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)

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

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

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

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

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)

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

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

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...

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)

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

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

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

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

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.

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)

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.

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.

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.

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

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

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

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

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

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

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

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

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)

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.

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)

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

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 ...

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.

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.

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.

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

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

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

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

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 Δsnf1Δsnf4 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...

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

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.

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

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.

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.

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...

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

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.

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.

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 ...

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)

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)

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

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

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)

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

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.

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.

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 Δsnf1Δsnf4 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

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.

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

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.

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

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.

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

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

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

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)

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)

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

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

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

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

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.

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

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)

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

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

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.

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.

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

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

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...

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.

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)

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.

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)

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...

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

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

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.

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)

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

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.

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

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....

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

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....

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

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

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.

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

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.

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

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

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

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.

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)

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

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

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

EQ6 Calculation for Chemical Degradation of Shippingport LWBR (TH/U Oxide) Spent Nuclear Fuel Waste Packages

International Nuclear Information System (INIS)

S. Arthur

2000-01-01

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 Shippingport Light Water Breeder Reactor (LWBR) (Ref. 1). The Shippingport LWBR SNF has been considered for disposal at the potential Yucca Mountain site. Because of the high content of fissile material in the SNF, the waste package (WP) design requires special consideration of the amount and placement of neutron absorbers and the possible loss of absorbers and SNF materials over geologic time. For some WPs, the outer shell corrosion-resistant material (CRM) and the corrosion-allowance inner shell may breach (Refs. 2 and 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 and neutron absorbers from the WP, further affecting the neutronics of the system. This study presents calculations of the long-term geochemical behavior of WPs containing a Shippingport LWBR SNF seed assembly, and high-level waste (HLW) glass canisters arranged according to the codisposal concept (Ref. 4). The specific study objectives were to determine: (1) The extent to which criticality control material, suggested for this WP design, will remain in the WP after corrosion/dissolution of the initial WP configuration (such that it can be effective in preventing criticality); (2) The extent to which fissile uranium and fertile thorium 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

EQ6 Calculation for Chemical Degradation of Shippingport LWBR (TH/U Oxide) Spent Nuclear Fuel Waste Packages

Energy Technology Data Exchange (ETDEWEB)

S. Arthur

2000-09-14

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 Shippingport Light Water Breeder Reactor (LWBR) (Ref. 1). The Shippingport LWBR SNF has been considered for disposal at the potential Yucca Mountain site. Because of the high content of fissile material in the SNF, the waste package (WP) design requires special consideration of the amount and placement of neutron absorbers and the possible loss of absorbers and SNF materials over geologic time. For some WPs, the outer shell corrosion-resistant material (CRM) and the corrosion-allowance inner shell may breach (Refs. 2 and 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 and neutron absorbers from the WP, further affecting the neutronics of the system. This study presents calculations of the long-term geochemical behavior of WPs containing a Shippingport LWBR SNF seed assembly, and high-level waste (HLW) glass canisters arranged according to the codisposal concept (Ref. 4). The specific study objectives were to determine: (1) The extent to which criticality control material, suggested for this WP design, will remain in the WP after corrosion/dissolution of the initial WP configuration (such that it can be effective in preventing criticality); (2) The extent to which fissile uranium and fertile thorium 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

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)

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

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)

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

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.

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.

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

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

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

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

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

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

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.

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.

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.

Characterization plan for Hanford spent nuclear fuel

International Nuclear Information System (INIS)

Abrefah, J.; Thornton, T.A.; Thomas, L.E.; Berting, F.M.; Marschman, S.C.

1994-12-01

Reprocessing of spent nuclear fuel (SNF) at the Hanford Site Plutonium-Uranium Extraction Plant (PUREX) was terminated in 1972. Since that time a significant quantity of N Reactor and Single-Pass Reactor SNF has been stored in the 100 Area K-East (KE) and K-West (KW) reactor basins. Approximately 80% of all US Department of Energy (DOE)-owned SNF resides at Hanford, the largest portion of which is in the water-filled KE and KW reactor basins. The basins were not designed for long-term storage of the SNF and it has become a priority to move the SNF to a more suitable location. As part of the project plan, SNF inventories will be chemically and physically characterized to provide information that will be used to resolve safety and technical issues for development of an environmentally benign and efficient extended interim storage and final disposition strategy for this defense production-reactor SNF

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

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...

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

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)

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...

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

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.

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

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.

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.

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

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

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...

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.

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

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

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.

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.

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

Issues relating to spent nuclear fuel storage on the Oak Ridge Reservation

International Nuclear Information System (INIS)

Klein, J.A.; Turner, D.W.

1994-01-01

Currently, about 2,800 metric tons of spent nuclear fuel (SNF) is stored in the US, 1,000 kg of SNF (or about 0.03% of the nation's total) are stored at the US Department of Energy (DOE) complex in Oak Ridge, Tennessee. However small the total quantity of material stored at Oak Ridge, some of the material is quite singular in character and, thus, poses unique management concerns. The various types of SNF stored at Oak Ridge will be discussed including: (1) High-Flux Isotope Reactor (HFIR) and future Advanced Neutron Source (ANS) fuels; (2) Material Testing Reactor (MTR) fuels, including Bulk Shielding Reactor (BSR) and Oak Ridge Research Reactor (ORR) fuels; (3) Molten Salt Reactor Experiment (MSRE) fuel; (4) Homogeneous Reactor Experiment (HRE) fuel; (5) Miscellaneous SNF stored in Oak Ridge National Laboratory's (ORNL's) Solid Waste Storage Areas (SWSAs); (6) SNF stored in the Y-12 Plant 9720-5 Warehouse including Health. Physics Reactor (HPRR), Space Nuclear Auxiliary Power (SNAP-) 10A, and DOE Demonstration Reactor fuels

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.

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

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)

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.

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.

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

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

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.

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

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

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

Third party liability of nuclear installation decommissioning with Russian nuclear submarines as an example: insurance versus technologies

International Nuclear Information System (INIS)

Gavrilov, S.D.; Derevyankin, A.A.; Khamyanov, L.P.; Kovalenko, V.N.; Kovalivich, O.M.; Smirnov, P.L.

2001-01-01

Third party and environment of civil liability damage caused by incidents at military nuclear installations, for instance at decommissioned NPS (nuclear powered submarines), may be divided into three main trends: -) Liability of NPS without high-enriched irradiated nuclear fuel (SNF) for its self-submersion (radiation incident); -) Liability of NPS with SNF aboard for its self-submersion (radiation incident); and -) Liability of floating NPS for its SNF discharge (nuclear accident). Without step-by-step transition from the Russian Federation guaranties to insurance and making allowance for liability limits according to the Vienna Convention approach, the sizes of the financial guarantee for the civil liability of the NPS owner (Russian state), in US dollars of 2000, are approximately assessed as the following: -) storing decommissioned NPS or a floating module without SNF - from 12 to 25 thousand dollars per year (per one submarine or module); -) storing decommissioned NPS with SNF inside reactors cores - from 25 to 40 thousand dollars per year; -) assembly-by-assembly removing SNF from reactors' core of decommissioned NPS - up to 1.5 million dollars for undamaged reactor per the discharging period; -) SNF removing within reactor using the filled in-space reactor's core by liquid-phased hardened or dispersed solid-phase materials from decommissioned NPS - from 30 to 50 thousand dollars for undamaged reactor per the discharging period. Both rates and sums for NPS with damaged reactors are to be estimated for the each damaged reactor and NPS at all. It is necessary to perform the measures reducing the risk of nuclear accidents of NPS with undamaged SNF and NPS with damaged reactors in possibly short time. It will allow not only to cut risks by ten times and more, but also to accumulate necessary insurance reserves faster. These measures can be partially or completely executed using the preventing measures reserves assigned to all decommissioned Russian NPS and

Third party liability of nuclear installation decommissioning with Russian nuclear submarines as an example: insurance versus technologies

Energy Technology Data Exchange (ETDEWEB)

Gavrilov, S.D. [PREKSAT Ltd., Moscow (Russian Federation); Derevyankin, A.A. [Reseaarch and Development Institute of Nuclear Power Engineering, Moscow (Russian Federation); Khamyanov, L.P. [All-Russian Research Institute on NPP Operation, Moscow (Russian Federation); Kovalenko, V.N. [Ministry for Nuclear Energy Of Russian, Moscow (Russian Federation); Kovalivich, O.M. [Research and Technological Center for Nuclear and Radiation Safety of Supervisory, Nuclear Energy State Commitee of Russia, Moscow (Russian Federation); Smirnov, P.L. [Nuclear Safety Institute of Russian Academy of Sciences, Moscow (Russian Federation)

2001-07-01

Third party and environment of civil liability damage caused by incidents at military nuclear installations, for instance at decommissioned NPS (nuclear powered submarines), may be divided into three main trends: -) Liability of NPS without high-enriched irradiated nuclear fuel (SNF) for its self-submersion (radiation incident); -) Liability of NPS with SNF aboard for its self-submersion (radiation incident); and -) Liability of floating NPS for its SNF discharge (nuclear accident). Without step-by-step transition from the Russian Federation guaranties to insurance and making allowance for liability limits according to the Vienna Convention approach, the sizes of the financial guarantee for the civil liability of the NPS owner (Russian state), in US dollars of 2000, are approximately assessed as the following: -) storing decommissioned NPS or a floating module without SNF - from 12 to 25 thousand dollars per year (per one submarine or module); -) storing decommissioned NPS with SNF inside reactors cores - from 25 to 40 thousand dollars per year; -) assembly-by-assembly removing SNF from reactors' core of decommissioned NPS - up to 1.5 million dollars for undamaged reactor per the discharging period; -) SNF removing within reactor using the filled in-space reactor's core by liquid-phased hardened or dispersed solid-phase materials from decommissioned NPS - from 30 to 50 thousand dollars for undamaged reactor per the discharging period. Both rates and sums for NPS with damaged reactors are to be estimated for the each damaged reactor and NPS at all. It is necessary to perform the measures reducing the risk of nuclear accidents of NPS with undamaged SNF and NPS with damaged reactors in possibly short time. It will allow not only to cut risks by ten times and more, but also to accumulate necessary insurance reserves faster. These measures can be partially or completely executed using the preventing measures reserves assigned to all decommissioned Russian NPS

Warm pre-stressing

International Nuclear Information System (INIS)

Hedner, G.

1983-01-01

Literature survey and critical evaluation of the phenomenon of warm pre-stressing (WPS) is presented. It is found that the cause of it is not clear and a calculated control is missing. The effect of irradiation is unknown, and the influence of WPS on the behaviour of reactor vessels is discussed. (G.B.)

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

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.

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

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)

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.

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.

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

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.

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.

Proposal of a dry storage installation in Angra NPP for spent nuclear fuel

International Nuclear Information System (INIS)

Romanato, Luiz S.; Rzyski, Barbara M.

2009-01-01

When nuclear fuel is removed from a power reactor core after the depletion of efficiency in generating energy is called Spent Nuclear Fuel (SNF). After its withdrawal from the reactor core, SNF is temporarily stored in pools usually at the same site of the reactor. During this time, short-living radioactive elements and generated heat undergo decay until levels that allow removing the SNF from the pool and sending it for reprocessing or a temporary storage whether any of its final destinations has not yet been defined. It can be loaded in casks and disposed during years in a dry storage installations until be sent to a reprocessing plant or deep repositories. Before any decision, reprocessing or disposal, the SNF needs to be safely and efficiently isolated in one of many types of storages that exist around the world. Worldwide, the amount of SNF increases annually and in the next years this amount will be higher as a consequence of new Nuclear Power Plants (NPP) construction. In Brazil, that is about to construct the Angra 3 nuclear power reactor, a project about the final destination of the SNF is not yet ready. This paper presents a proposal for a dry storage installation in the Angra NPP site since it can be an initial solution for the Brazilian's SNF, until a final decision is taken. (author)

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

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.

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

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.

Storage of spent nuclear fuel: the problem of spent nuclear fuel in Bulgaria

Energy Technology Data Exchange (ETDEWEB)

Boyadzhiev, Z; Vapirev, E [Kombinat Atomna Energetika, Kozloduj (Bulgaria)

1996-12-31

A review of existing technologies for wet and dry storage of spent nuclear fuel (SNF) and the reprocessing policies is presented. The problem of SNF in Bulgaria is arising from nonobservance of the obligation to return SNF back to the former Soviet Union as agreed in the construction contract. In November 1994 approximately 1800 fuel assemblies have been stored in away-from-reactor (AFR) facility and another 1060 in at-reactor (AR) pools. The national policy is to export SNF out of the country. The AFR facility has a limited capacity and it is designed only for WWER-440 fuel although work is going on to extend it in order to store WWER-1000 SNF. 14 refs.

Storage of spent nuclear fuel: the problem of spent nuclear fuel in Bulgaria

International Nuclear Information System (INIS)

Boyadzhiev, Z.; Vapirev, E.

1995-01-01

A review of existing technologies for wet and dry storage of spent nuclear fuel (SNF) and the reprocessing policies is presented. The problem of SNF in Bulgaria is arising from nonobservance of the obligation to return SNF back to the former Soviet Union as agreed in the construction contract. In November 1994 approximately 1800 fuel assemblies have been stored in away-from-reactor (AFR) facility and another 1060 in at-reactor (AR) pools. The national policy is to export SNF out of the country. The AFR facility has a limited capacity and it is designed only for WWER-440 fuel although work is going on to extend it in order to store WWER-1000 SNF. 14 refs

Status of the TRIGA shipments to the INEEL from Europe

International Nuclear Information System (INIS)

Mustin, T.; Stump, R.C.; Tyacke, M.J.

1997-01-01

This paper reports the activities underway by the US Department of Energy (DOE) for returning Training, Research, Isotope, General Atomics (TRIGA) spent nuclear fuel (SNF) from foreign research reactors (FRR) in four European countries to the Idaho National Engineering and Environmental Laboratory (INEEL). Those countries are Germany, Italy, Romania, and Slovenia. This is part of the ''Nuclear Weapons Nonproliferation Policy'' of returning research reactor SNF containing uranium enriched in the US. This paper describes the results of a pre-assessment trip in September, 1997, to these countries, including: history of the reactors and research being performed; inventory of TRIGA SNF; fuel types (stainless steel, aluminum, or Incoloy) and enrichments; and each country's plans for returning their TRIGA SNF to the INEEL

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.

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)

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)

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)

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

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.

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

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

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

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

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

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.

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.

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.

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)

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

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.

U.S. Commercial Spent Nuclear Fuel Assembly Characteristics - 1968-2013

Energy Technology Data Exchange (ETDEWEB)

Hu, Jianwei [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Peterson, Joshua L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bowman, Stephen M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-09-01

Activities related to management of spent nuclear fuel (SNF) are increasing in the US and many other countries. Over 240,000 SNF assemblies have been discharged from US commercial reactors since the late 1960s. The enrichment and burnup of SNF have changed significantly over the past 40 years, and fuel assembly designs have also evolved. Understanding the general characteristics of SNF helps regulators and other stakeholders form overall strategies towards the final disposal of US SNF. This report documents a survey of all US commercial SNF assemblies in the GC-859 database and provides reference SNF source terms (e.g., nuclide inventories, decay heat, and neutron/photon emission) at various cooling times up to 200 years after fuel discharge. This study reviews the distribution and evolution of fuel parameters of all SNF assemblies discharged over the past 40 years. Assemblies were categorized into three groups based on discharge year, and the median burnups and enrichments of each group were used to establish representative cases. An extended burnup case was created for boiling water reactor (BWR) fuels, and another was created for the pressurized water reactor (PWR) fuels. Two additional cases were developed to represent the eight mixed oxide (MOX) fuel assemblies in the database. Burnup calculations were performed for each representative case. Realistic parameters for fuel design and operations were used to model the SNF and to provide reference fuel characteristics representative of the current inventory. Burnup calculations were performed using the ORIGEN code, which is part of the SCALE nuclear modeling and simulation code system. Results include total activity, decay heat, photon emission, neutron flux, gamma heat, and plutonium content, as well as concentrations for 115 significant nuclides. These quantities are important in the design, regulation, and operations of SNF storage, transportation, and disposal systems.

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

Study of a brazilian cask and its installation for PWR spent nuclear fuel dry storage

International Nuclear Information System (INIS)

Romanato, Luiz Sergio

2009-01-01

Spent nuclear fuel (SNF) is removed from the nuclear reactor after the depletion on efficiency in generating energy. After the withdrawal from the reactor core, the SNF is temporarily stored in pools at the same site of the reactor. At this time, the generated heat and the short and medium lived radioactive elements decay to levels that allow removing SNF from the pool and sending it to temporary dry storage. In that phase, the fuel needs to be safely and efficiently stored, and then, it can be retrieved in a future, or can be disposed as radioactive waste. The amount of spent fuel increases annually and, in the next years, will still increase more, because of the construction of new nuclear plants. Today, the number of new facilities back up to levels of the 1970's, since it is greater than the amount of decommissioning in old installations. As no final decision on the back-end of the nuclear fuel cycle is foreseen in the near future in Brazil, either to recover the SNF or to consider it as radioactive waste, this material has to be isolated in some type of storage model existing around the world. In the present study it is shown that dry SNF storage is the best option. A national cask model for SNF as well these casks storage installation are proposed. It is a multidisciplinary study in which the engineering conceptual task was developed and may be applied to national SNF removed from the Brazilian power reactors, to be safely stored for a long time until the Brazilian authorities will decide about the site for final disposal. (author)

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)

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

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

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.

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

Savannah River Site, spent nuclear fuel management, draft environmental impact statement

International Nuclear Information System (INIS)

1998-12-01

The management of spent nuclear fuel (SNF) has been an integral part of the mission of the Savannah River Site (SRS) for more than 40 years. Until the early 1990s, SNF management consisted primarily of short-term onsite storage and reprocessing in the SRS chemical separation facilities to produce strategic nuclear materials. With the end of the Cold War, the US Department of Energy (DOE) decided to phase out reprocessing of SNF for the production of nuclear weapons materials. Therefore, the management strategy for this fuel has shifted from short-term storage and reprocessing for the recovery of highly-enriched uranium and transuranic isotopes to stabilization, when necessary, and interim storage pending final disposition that includes preparing aluminum-based SNF for placement in a geologic repository. In addition to the fuel already onsite, the SRS will receive SNF from foreign research reactors until 2009 and from domestic research reactors until, potentially, 2035. As a result, the safe and efficient management of SNF will continue to be an important SRS mission. This EIS evaluates the potential environmental impacts of DOE's proposed plans for management SNF assigned to SRS

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.

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.

CIRFT Data Update and Data Analyses for Spent Nuclear Fuel Vibration Reliability Study

Energy Technology Data Exchange (ETDEWEB)

Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wang, Hong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2018-01-01

The objective of this research is to collect experimental data on spent nuclear fuel (SNF) from pressurized water reactors (PWRs), including the H. B. Robinson Nuclear Power Station (HBR), Catawba Nuclear Station, North Anna Nuclear Power Station (NA), and the Limerick Nuclear Power Station (LMK) boiling water reactor (BWR). Data will be collected under simulated transportation environments using the cyclic integrated reversible-bending fatigue tester (CIRFT), an enabling hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL). These data will be used to support ongoing SNF modeling activities and to address regulatory issues associated with SNF transport.

Dry air oxidation kinetics of K-Basin spent nuclear fuel

International Nuclear Information System (INIS)

Abrefah, J.; Buchanan, H.C.; Gerry, W.M.; Gray, W.J.; Marschman, S.C.

1998-06-01

The safety and process analyses of the proposed Integrated Process Strategy (IPS) to move the N-Reactor spent nuclear fuel (SNF) stored at K-Basin to an interim storage facility require information about the oxidation behavior of the metallic uranium. Limited experiments have been performed on the oxidation reaction of SNF samples taken from an N-Reactor outer fuel element in various atmospheres. This report discusses studies on the oxidation behavior of SNF using two independent experimental systems: (1) a tube furnace with a flowing gas mixture of 2% oxygen/98% argon; and (2) a thermogravimetric system for dry air oxidation

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

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

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.

Analysis of back-end fuel-cycle issues in the ROK

Energy Technology Data Exchange (ETDEWEB)

Forrest, R.; Braun, C., E-mail: rforrest@gmail.com, E-mail: cbraun@stanford.edu [Stanford Univ., Center for International Security and Cooperation (CISAC), Stanford, CA (United States)

2014-07-01

We discuss mid and near term issues related to spent nuclear fuel (SNF) management in the Republic of Korea (ROK). Currently, 23 operational reactors in the ROK produce SNF that is stored mostly in pools adjacent to each reactor. It is expected that, due to the limited capacity of these pools, they will be saturated within the decade. The ROK may arrange for several additional years of pool storage via various mechanisms such as high density racking and intra-site SNF transhipment. However, these short term measures will quickly run their course, and the fundamental problem of SNF disposition will remain. We model the future of SNF pools of all reactors located on each of the reactor sites to verify declared saturation dates under constraints of current policy. Using this model we can examine the effect of current and planned mitigation measures in addressing short-term saturation. We then propose simple additional measures requiring changes in policy and calculate additional time such measures would afford. Projecting further into the future, we can then calculate saturation timelines under various fuel cycle scenarios and how such timelines would allow transition to future proposed closed cycles. Among other observations, we determine that even with the timely success of the ongoing pyroprocessing effort, interim SNF storage will be required. It follows that this would most logically be a single, centralized dry cask storage facility commonly called an Independent Spent Fuel Storage Installation (ISFSI). The purpose of this study is to inform a larger project between CISAC and the East Asia Institute (EAI) to examine important issues in nuclear security and energy in the ROK. (author)

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

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)

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

Preliminary Concept of Operations for the Spent Fuel Management System--WM2017

Energy Technology Data Exchange (ETDEWEB)

Cumberland, Riley M [ORNL; Adeniyi, Abiodun Idowu [ORNL; Howard, Rob L [ORNL; Joseph III, Robert Anthony [ORNL; Jarrell, Joshua J [ORNL; Nutt, Mark [Argonne National Laboratory (ANL)

2017-01-01

The Nuclear Fuels Storage and Transportation Planning Project (NFST) within the U.S. Department of Energy s Office of Nuclear Energy is tasked with identifying, planning, and conducting activities to lay the groundwork for developing interim storage and transportation capabilities in support of an integrated waste management system. The system will provide interim storage for commercial spent nuclear fuel (SNF) from reactor sites and deliver it to a repository. The system will also include multiple subsystems, potentially including; one or more interim storage facilities (ISF); one or more repositories; facilities to package and/or repackage SNF; and transportation systems. The project team is analyzing options for an integrated waste management system. To support analysis, the project team has developed a Concept of Operations document that describes both the potential integrated system and inter-dependencies between system components. The goal of this work is to aid systems analysts in the development of consistent models across the project, which involves multiple investigators. The Concept of Operations document will be updated periodically as new developments emerge. At a high level, SNF is expected to travel from reactors to a repository. SNF is first unloaded from reactors and placed in spent fuel pools for wet storage at utility sites. After the SNF has cooled enough to satisfy loading limits, it is placed in a container at reactor sites for storage and/or transportation. After transportation requirements are met, the SNF is transported to an ISF to store the SNF until a repository is developed or directly to a repository if available. While the high level operation of the system is straightforward, analysts must evaluate numerous alternative options. Alternative options include the number of ISFs (if any), ISF design, the stage at which SNF repackaging occurs (if any), repackaging technology, the types of containers used, repository design, component

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.

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)

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

Management of spent nuclear fuel on the Oak Ridge Reservation, Oak Ridge, Tennessee: Environmental assessment

International Nuclear Information System (INIS)

1996-02-01

On June 1, 1995, DOE issued a Record of Decision [60 Federal Register 28680] for the Department-wide management of spent nuclear fuel (SNF); regionalized storage of SNF by fuel type was selected as the preferred alternative. The proposed action evaluated in this environmental assessment is the management of SNF on the Oak Ridge Reservation (ORR) to implement this preferred alternative of regional storage. SNF would be retrieved from storage, transferred to a hot cell if segregation by fuel type and/or repackaging is required, loaded into casks, and shipped to off-site storage. The proposed action would also include construction and operation of a dry cask SNF storage facility on ORR, in case of inadequate SNF storage. Action is needed to enable DOE to continue operation of the High Flux Isotope Reactor, which generates SNF. This report addresses environmental impacts

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

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)

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.

Fuel fabrication and reprocessing for nuclear fuel cycle with inherent safety demands

Energy Technology Data Exchange (ETDEWEB)

Shadrin, Andrey Yurevich; Dvoeglazov, Konstantin Nikolaevich; Ivanov, Valentine Borisovich; Volk, Vladimir Ivanovich; Skupov, Mikhail Vladimirovich; Glushenkov, Alexey Evgenevich [Joint Stock Company ' ' The High Technological Research Institute of Inorganic Materials' ' , Moscow (Russian Federation); Troyanov, Vladimir Mihaylovich; Zherebtsov, Alexander Anatolievich [Innovation and Technology Center of Project ' ' PRORYV' ' , State Atomic Energy Corporation ' ' Rosatom' ' , Moscow (Russian Federation)

2015-06-01

The strategies adopted in Russia for a closed nuclear fuel cycle with fast reactors (FR), selection of fuel type and recycling technologies of spent nuclear fuel (SNF) are discussed. It is shown that one of the possible technological solutions for the closing of a fuel cycle could be the combination of pyroelectrochemical and hydrometallurgical methods of recycling of SNF. This combined scheme allows: recycling of SNF from FR with high burn-up and short cooling time; decreasing the volume of stored SNF and the amount of plutonium in a closed fuel cycle in FR; recycling of any type of SNF from FR; obtaining the high pure end uranium-plutonium-neptunium end-product for fuel refabrication using pellet technology.

Spent Fuel and Waste Management Activities for Cleanout of the 105 F Fuel Storage Basin at Hanford

International Nuclear Information System (INIS)

Morton, M. R.; Rodovsky, T. J.; Day, R. S.

2002-01-01

Clean-out of the F Reactor fuel storage basin (FSB) by the Environmental Restoration Contractor (ERC) is an element of the FSB decontamination and decommissioning and is required to complete interim safe storage (ISS) of the F Reactor. Following reactor shutdown and in preparation for a deactivation layaway action in 1970, the water level in the F Reactor FSB was reduced to approximately 0.6 m (2 ft) over the floor. Basin components and other miscellaneous items were left or placed in the FSB. The item placement was performed with a sense of finality, and no attempt was made to place the items in an orderly manner. The F Reactor FSB was then filled to grade level with 6 m (20 ft) of local surface material (essentially a fine sand). The reactor FSB backfill cleanout involves the potential removal of spent nuclear fuel (SNF) that may have been left in the basin unintentionally. Based on previous cleanout of four water-filled FSBs with similar designs (i.e., the B, C, D, and DR FSBs in the 1980s), it was estimated that up to five SNF elements could be discovered in the F Reactor FSB (1). In reality, a total of 10 SNF elements have been found in the first 25% of the F Reactor FSB excavation. This paper discusses the technical and programmatic challenges of performing this decommissioning effort with some of the controls needed for SNF management. The paper also highlights how many various technologies were married into a complete package to address the issue at hand and show how no one tool could be used to complete the job; but by combining the use of multiple tools, progress is being made

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.

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.

Spent nuclear fuel project product specification

International Nuclear Information System (INIS)

Pajunen, A.L.

1998-01-01

Product specifications are limits and controls established for each significant parameter that potentially affects safety and/or quality of the Spent Nuclear Fuel (SNF) packaged for transport to dry storage. The product specifications in this document cover the spent fuel packaged in MultiCanister Overpacks (MCOs) to be transported throughout the SNF Project. The SNF includes N Reactor fuel and single-pass reactor fuel. The FRS removes the SNF from the storage canisters, cleans it, and places it into baskets. The MCO loading system places the baskets into MCO/Cask assembly packages. These packages are then transferred to the Cold Vacuum Drying (CVD) Facility. After drying at the CVD Facility, the MCO cask packages are transferred to the Canister Storage Building (CSB), where the MCOs are removed from the casks, staged, inspected, sealed (by welding), and stored until a suitable permanent disposal option is implemented. The key criteria necessary to achieve these goals are documented in this specification

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

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.

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.

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

Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites

Energy Technology Data Exchange (ETDEWEB)

Maheras, Steven J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Best, Ralph E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Ross, Steven B. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Buxton, Kenneth A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); England, Jeffery L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McConnell, Paul E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Massaro, Lawrence M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jensen, Philip J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2015-09-30

A preliminary evaluation of removing spent nuclear fuel (SNF) from 13 shutdown nuclear power reactor sites was conducted. At these shutdown sites the nuclear power reactors have been permanently shut down and the sites have been decommissioned or are undergoing decommissioning. The shutdown sites were Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, Zion, Crystal River, Kewaunee, San Onofre, and Vermont Yankee. The evaluation was divided into four components: (1) characterization of the SNF and greater-than-Class C low-level radioactive waste (GTCC waste) inventory, (2) a description of the on-site infrastructure and conditions relevant to transportation of SNF and GTCC waste, (3) an evaluation of the near-site transportation infrastructure and experience relevant to shipping transportation casks containing SNF and GTCC waste, including identification of gaps in information, and (4) an evaluation of the actions necessary to prepare for and remove SNF and GTCC waste. Every site was found to have at least one off-site transportation mode option for removing its SNF and GTCC waste; some have multiple options. Experience removing large components during reactor decommissioning provided an important source of information used to identify the transportation mode options for the sites. Especially important in conducting the evaluation were site visits, through which information was obtained that would not have been available otherwise. Extensive photographs taken during the site visits proved to be particularly useful in documenting the current conditions at or near the sites. It is expected that additional site visits will be conducted to add to the information presented in the evaluation.

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

DSNF AND OTHER WASTE FORM DEGRADATION ABSTRACTION

International Nuclear Information System (INIS)

CUNNANE, J.

2004-01-01

Several hundred distinct types of DOE-owned spent nuclear fuel (DSNF) may potentially be disposed in the Yucca Mountain repository. These fuel types represent many more types than can be viably individually examined for their effect on the Total System Performance Assessment for the License Application (TSPA-LA). Additionally, for most of these fuel types, there is no known direct experimental test data for the degradation and dissolution of the waste form in repository groundwaters. The approach used in the TSPA-LA model is, therefore, to assess available information on each of 11 groups of DSNF, and to identify a model that can be used in the TSPA-LA model without differentiating between individual codisposal waste packages containing different DSNF types. The purpose of this report is to examine the available data and information concerning the dissolution kinetics of DSNF matrices for the purpose of abstracting a degradation model suitable for use in describing degradation of the DSNF inventory in the Total System Performance Assessment for the License Application. The data and information and associated degradation models were examined for the following types of DSNF: Group 1--Naval spent nuclear fuel; Group 2--Plutonium/uranium alloy (Fermi 1 SNF); Group 3--Plutonium/uranium carbide (Fast Flux Test Facility-Test Fuel Assembly SNF); Group 4--Mixed oxide and plutonium oxide (Fast Flux Test Facility-Demonstration Fuel Assembly/Fast Flux Test Facility-Test Demonstration Fuel Assembly SNF); Group 5--Thorium/uranium carbide (Fort St. Vrain SNF); Group 6--Thorium/uranium oxide (Shippingport light water breeder reactor SNF); Group 7--Uranium metal (N Reactor SNF); Group 8--Uranium oxide (Three Mile Island-2 core debris); Group 9--Aluminum-based SNF (Foreign Research Reactor SNF); Group 10--Miscellaneous Fuel; and Group 11--Uranium-zirconium hydride (Training Research Isotopes-General Atomics SNF). The analyses contained in this document provide an ''upper-limit'' (i

Microstructure and Physical Metallurgy in Melt-Dilute Treatment Technology for Al-Based Spent Nuclear Fuel

International Nuclear Information System (INIS)

Adams, T.M.; Peacock, H.B.

1998-09-01

Major challenges associated with the direct disposal of research reactor fuel in a repository include nonproliferation and criticality control, both of which may be a concern for HEU Al-SNF. Consideration must be given to the potential desirability and/or regulatory necessity of diluting the HEU SNF to below 20 percent enrichment. The probability for a criticality event as well as the issue of proliferation are greatly decreased by reducing the enrichment. The melt-dilute technology development program is focused on the development and implementation of a treatment technology for diluting HEU Al-SNF to LEU levels and qualifying this LEU Al-SNF form for geologic repository storage. An attribute of the melt-dilute process is its applicability to various types of FRR and DRR SNF

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.

Experience Of Using Metal-and-Concrete Cask TUK-108/1 For Storage And Transportation Of Spent Nuclear Fuel Of Decommissioned NPS

Energy Technology Data Exchange (ETDEWEB)

Barnes, E.; Dyer, R. [Environmental Protection Agency, Ronald Reagan Bldg. 3rd Floor 1200 Pennsylvania Av., NW Washington, D.C. 20024 (United States); Snipes, R. [Oak Ridge National Laboratories, VA (United States); Dolbenkov, V.G.; Guskov, V.D.; Korotkov, G.V. [Joint Stock Company ' KBSM' , 64 Lesnoy Av., St.Petersburg 194100 (Russian Federation); Makarchuk, T.F. [Joint Stock Company ' Atomstroyexport' , Potapovskiy str. 5, bld. 4, Moscow, 101990 (Russian Federation); Zakharchev, A.A. [State Corporation ' Rosatom' , 24-26 Ordinka St., Moscow, 100000 (Russian Federation)

2009-06-15

In past 10 years in Russia an intensive development of a new technology of management of spent nuclear fuel (SNF) has taken place. This technology is based on the concept of using a shielded cask which provides safety of its content (SNF) and meeting all other safety requirements to storage and transportation of SNF. Radiation protection against emission and non-propagation of activity outside the cask is ensured by the physical barriers such as all-metal or composite body, face work, inner structures to accommodate spent fuel assemblies (SFA), lids with sealing systems. Residual heat buildup is off-taken to the environment by natural way: emission and convection of surrounding air. The necessity in development of the cask technology of SNF management was conditioned by the situation at hand with defueling of Russian decommissioned nuclear-powered submarines (NPS) as the existed transport infrastructure and enterprises involved in fuel processing could not meet the demand for transportation and processing of SNF neither from reactors of all dismantled NPS, nor from reactors of NPS waiting for decommissioning. The US and Norway actively participated in the trilateral joint project with the Russian Federation aimed at creation of a cask prototype for interim storage and transportation of SNF of dismantled NPS. The 1.1 Project is a part of the Arctic Military Environmental Cooperation (AMEC) Program. In December 2000 the project was successfully completed by issuance of the certificate-permit for design and transportation of NP Submarine SNF. It was a first certified dual-purpose TUK from the MMC family. In these years 106 TUK-108/1 casks have been manufactured and supplied to PO Mayak, JSC CS Zvezda, JSC CS Zvezdochka and FSUE DalRAO. The storage pads for interim storage of TUK-108/1 have been built and currently are in operation on sites of SNF unloading from submarine reactors and SNF cask-loading such as JSC CS Zvezda, JSC CS Zvezdochka and FSUE DalRAO. In

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

Spent Nuclear Fuel Alternative Technology Risk Assessment

Energy Technology Data Exchange (ETDEWEB)

Perella, V.F.

1999-11-29

A Research Reactor Spent Nuclear Fuel Task Team (RRTT) was chartered by the Department of Energy (DOE) Office of Spent Fuel Management with the responsibility to recommend a course of action leading to a final technology selection for the interim management and ultimate disposition of the foreign and domestic aluminum-based research reactor spent nuclear fuel (SNF) under DOE''s jurisdiction. The RRTT evaluated eleven potential SNF management technologies and recommended that two technologies, direct co-disposal and an isotopic dilution alternative, either press and dilute or melt and dilute, be developed in parallel. Based upon that recommendation, the Westinghouse Savannah River Company (WSRC) organized the SNF Alternative Technology Program to further develop the direct co-disposal and melt and dilute technologies and provide a WSRC recommendation to DOE for a preferred SNF alternative management technology. A technology risk assessment was conducted as a first step in this recommendation process to determine if either, or both, of the technologies posed significant risks that would make them unsuitable for further development. This report provides the results of that technology risk assessment.

Spent Nuclear Fuel Alternative Technology Risk Assessment

International Nuclear Information System (INIS)

Perella, V.F.

1999-01-01

A Research Reactor Spent Nuclear Fuel Task Team (RRTT) was chartered by the Department of Energy (DOE) Office of Spent Fuel Management with the responsibility to recommend a course of action leading to a final technology selection for the interim management and ultimate disposition of the foreign and domestic aluminum-based research reactor spent nuclear fuel (SNF) under DOE''s jurisdiction. The RRTT evaluated eleven potential SNF management technologies and recommended that two technologies, direct co-disposal and an isotopic dilution alternative, either press and dilute or melt and dilute, be developed in parallel. Based upon that recommendation, the Westinghouse Savannah River Company (WSRC) organized the SNF Alternative Technology Program to further develop the direct co-disposal and melt and dilute technologies and provide a WSRC recommendation to DOE for a preferred SNF alternative management technology. A technology risk assessment was conducted as a first step in this recommendation process to determine if either, or both, of the technologies posed significant risks that would make them unsuitable for further development. This report provides the results of that technology risk assessment

Management of radioactive waste from reprocessing plants

International Nuclear Information System (INIS)

Kanwar Raj

2010-01-01

Reprocessing and recycling of both fissile and fertile components back into appropriate reactor systems is an integral part of three stage nuclear energy programme of India. Different steps involved in processing of spent nuclear fuel (SNF) are decladding, dissolution and recovery of fissile and fertile materials. Reprocessing of SNF is a complex process involving handling of large quantity of radioactive materials and processing chemicals. There are three reprocessing plants in operation in the country at Trombay, Tarapur and Kalpakkam. Out of these plants, Trombay reprocessing plant is engaged in reprocessing of SNF from research reactors and other two plants are processing of SNF from PHWRs. A facility is being built for reprocessing of thorium based spent fuel at BARC, Trombay based on the experience of pilot plant scale. Like other industrial activities of nuclear fuel cycle, fuel reprocessing facilities too generate various types of radioactive waste streams. These are generated in all the three physical forms namely solid, liquid and gas. These waste streams are primarily categorized on the basis of concentration of radionuclides, their half lives and toxicity. Management of these wastes aims at (a) recovery and recycle of useful materials, (b) concentration and confinement of radioactivity in inert and stable matrices, (c) minimization of final waste volume for disposal, (d) decontamination of effluents following ALARA principle and (e) minimization of radioactive discharge to the environment. The present paper outlines the salient features of management of different types of radioactive waste generated in reprocessing plants handling SNF from research reactors and PHWR

Russian RERTR program as a part of Joint US DOE-RF MINATOM collaboration on elimination of the threat connected to the use of HEU in research reactors

International Nuclear Information System (INIS)

Arkhangelsky, N.

2002-01-01

The Russian RERTR Program started at the end of 70's, the final goal of the program is to eliminate supplies of HEU in fuel elements and assemblies for foreign research reactors that were designed according to Russian projects. Basic directions of the work include: completion of the development of the fuel elements and assemblies on a basis of uranium dioxide; development of the fuel on a basis of U-Mo alloy; and development of pin type fuel elements. Fuel assemblies of WWR-M2 type with LEU were developed and qualified for using in foreign research reactors that use such type of fuel assemblies. These assemblies are ready for the supplying several operating foreign research reactors. There are more than 20 sites in Eastern European countries, former Soviet republics and another countries that have big amount of Russian origin HEU in fresh and spent fuel. The problem of the shipment of SNF from sites of research reactors is also very important for domestic Russian research reactors. More than ten years from its beginning the Russian RERTR program developed practically independently from the international RERTR program and only at the begin of 90's the Russian specialists started to contact with foreign scientists and the exchange of the scientific information has become more intensive. In September 1994, representatives of Minatom and DOE signed a protocol of intent to reduce an enrichment of uranium in research reactors. The main aspects of collaboration involve: Several domestic Russian research reactors such as WWR-M, IR-8 and others were investigated from the point of view of possibility of reducing of enrichment; financial support of the program from US DOE which is insufficient. The important part of international collaboration is the import of Russian origin spent and fresh fuel of research reactors to Russia. In August 2002 an impressive result of the Russian-American collaboration with support of IAEA and with the help and assistance of Yugoslavian side was

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

Issues related to EM management of DOE spent nuclear fuel

International Nuclear Information System (INIS)

Abbott, D.G.; Abashian, M.S.; Chakraborti, S.; Roberson, K.; Meloin, J.M.

1993-07-01

This document is a summary of the important issues involved in managing spent nuclear fuel (SNF) owned by the Department of Energy (DOE). Issues related to civilian SNF activities are not discussed. DOE-owned SNF is stored primarily at the Hanford Site, Idaho National Engineering Laboratory (INEL), Savannah River Site (SRS), Oak Ridge National Laboratory (ORNL), and West Valley Demonstration Project. Smaller quantities of SNF are stored at Brookhaven National Laboratory, Sandia National Laboratories, and Los Alamos National Laboratory (LANL). There is a wide variety of fuel types, including both low and high enrichment fuels from weapons production, DOE reactors, research and development programs, naval programs, and universities. Most fuel is stored in pools associated with reactor or reprocessing facilities. Smaller quantities are in dry storage. Physical conditions of the fuel range from excellent to poor or severely damaged. An issue is defined as an important question that must be answered or decision that must be made on a topic or subject relevant to achieving the complimentary objectives of (a) storing SNF in compliance with applicable regulations and orders until it can be disposed, and (b) safely disposing of DOE's SNF. The purpose of this document is to define the issues; no recommendations are made on resolutions. As DOE's national SNF management program is implemented, a system of issues identification, documentation, tracking, and resolution will be implemented. This document is an initial effort at issues identification. The first section of this document is an overview of issues that are common to several or all DOE facilities that manage SNF. The common issues are organized according to specific aspects of spent fuel management. This is followed by discussions of management issues that apply specifically to individual DOE facilities. The last section provides literature references

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

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

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

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

Spent Nuclear Fuel Alternative Technology Decision Analysis

International Nuclear Information System (INIS)

Shedrow, C.B.

1999-01-01

The Westinghouse Savannah River Company (WSRC) made a FY98 commitment to the Department of Energy (DOE) to recommend a technology for the disposal of aluminum-based spent nuclear fuel (SNF) at the Savannah River Site (SRS). The two technologies being considered, direct co-disposal and melt and dilute, had been previously selected from a group of eleven potential SNF management technologies by the Research Reactor Spent Nuclear Fuel Task Team chartered by the DOE''s Office of Spent Fuel Management. To meet this commitment, WSRC organized the SNF Alternative Technology Program to further develop the direct co-disposal and melt and dilute technologies and ultimately provide a WSRC recommendation to DOE on a preferred SNF alternative management technology

DSNF AND OTHER WASTE FORM DEGRADATION ABSTRACTION

Energy Technology Data Exchange (ETDEWEB)

J. CUNNANE

2004-11-19

Several hundred distinct types of DOE-owned spent nuclear fuel (DSNF) may potentially be disposed in the Yucca Mountain repository. These fuel types represent many more types than can be viably individually examined for their effect on the Total System Performance Assessment for the License Application (TSPA-LA). Additionally, for most of these fuel types, there is no known direct experimental test data for the degradation and dissolution of the waste form in repository groundwaters. The approach used in the TSPA-LA model is, therefore, to assess available information on each of 11 groups of DSNF, and to identify a model that can be used in the TSPA-LA model without differentiating between individual codisposal waste packages containing different DSNF types. The purpose of this report is to examine the available data and information concerning the dissolution kinetics of DSNF matrices for the purpose of abstracting a degradation model suitable for use in describing degradation of the DSNF inventory in the Total System Performance Assessment for the License Application. The data and information and associated degradation models were examined for the following types of DSNF: Group 1--Naval spent nuclear fuel; Group 2--Plutonium/uranium alloy (Fermi 1 SNF); Group 3--Plutonium/uranium carbide (Fast Flux Test Facility-Test Fuel Assembly SNF); Group 4--Mixed oxide and plutonium oxide (Fast Flux Test Facility-Demonstration Fuel Assembly/Fast Flux Test Facility-Test Demonstration Fuel Assembly SNF); Group 5--Thorium/uranium carbide (Fort St. Vrain SNF); Group 6--Thorium/uranium oxide (Shippingport light water breeder reactor SNF); Group 7--Uranium metal (N Reactor SNF); Group 8--Uranium oxide (Three Mile Island-2 core debris); Group 9--Aluminum-based SNF (Foreign Research Reactor SNF); Group 10--Miscellaneous Fuel; and Group 11--Uranium-zirconium hydride (Training Research Isotopes-General Atomics SNF). The analyses contained in this document provide an &apos

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

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

Waste Package Component Design Methodology Report

International Nuclear Information System (INIS)

D.C. Mecham

2004-01-01

requirements of the YMP. Four waste package configurations have been selected to illustrate the application of the methodology during the licensing process. These four configurations are the 21-pressurized water reactor absorber plate waste package (21-PWRAP), the 44-boiling water reactor waste package (44-BWR), the 5 defense high-level radioactive waste (HLW) DOE spent nuclear fuel (SNF) codisposal short waste package (5-DHLWDOE SNF Short), and the naval canistered SNF long waste package (Naval SNF Long). Design work for the other six waste packages will be completed at a later date using the same design methodology. These include the 24-boiling water reactor waste package (24-BWR), the 21-pressurized water reactor control rod waste package (21-PWRCR), the 12-pressurized water reactor waste package (12-PWR), the 5 defense HLW DOE SNF codisposal long waste package (5-DHLWDOE SNF Long), the 2 defense HLW DOE SNF codisposal waste package (2-MC012-DHLW), and the naval canistered SNF short waste package (Naval SNF Short). This report is only part of the complete design description. Other reports related to the design include the design reports, the waste package system description documents, manufacturing specifications, and numerous documents for the many detailed calculations. The relationships between this report and other design documents are shown in Figure 1

Waste Package Component Design Methodology Report

Energy Technology Data Exchange (ETDEWEB)

D.C. Mecham

2004-07-12

and operational requirements of the YMP. Four waste package configurations have been selected to illustrate the application of the methodology during the licensing process. These four configurations are the 21-pressurized water reactor absorber plate waste package (21-PWRAP), the 44-boiling water reactor waste package (44-BWR), the 5 defense high-level radioactive waste (HLW) DOE spent nuclear fuel (SNF) codisposal short waste package (5-DHLWDOE SNF Short), and the naval canistered SNF long waste package (Naval SNF Long). Design work for the other six waste packages will be completed at a later date using the same design methodology. These include the 24-boiling water reactor waste package (24-BWR), the 21-pressurized water reactor control rod waste package (21-PWRCR), the 12-pressurized water reactor waste package (12-PWR), the 5 defense HLW DOE SNF codisposal long waste package (5-DHLWDOE SNF Long), the 2 defense HLW DOE SNF codisposal waste package (2-MC012-DHLW), and the naval canistered SNF short waste package (Naval SNF Short). This report is only part of the complete design description. Other reports related to the design include the design reports, the waste package system description documents, manufacturing specifications, and numerous documents for the many detailed calculations. The relationships between this report and other design documents are shown in Figure 1.

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.

Spent Nuclear Fuel Alternative Technology Decision Analysis

Energy Technology Data Exchange (ETDEWEB)

Shedrow, C.B.

1999-11-29

The Westinghouse Savannah River Company (WSRC) made a FY98 commitment to the Department of Energy (DOE) to recommend a technology for the disposal of aluminum-based spent nuclear fuel (SNF) at the Savannah River Site (SRS). The two technologies being considered, direct co-disposal and melt and dilute, had been previously selected from a group of eleven potential SNF management technologies by the Research Reactor Spent Nuclear Fuel Task Team chartered by the DOE''s Office of Spent Fuel Management. To meet this commitment, WSRC organized the SNF Alternative Technology Program to further develop the direct co-disposal and melt and dilute technologies and ultimately provide a WSRC recommendation to DOE on a preferred SNF alternative management technology.

Market driven strategy for acquisition of waste acceptance and transportation services for commercial spent fuel in the United States

International Nuclear Information System (INIS)

Lemeshewky, W.; Macaluso, C.; Smith, P.; Teer, B.

1998-05-01

The Department of Energy has the responsibility for the shipment of spent nuclear fuel (SNF) from commercial reactors to a Federal facility for storage and/or disposal. DOE has developed a strategy for a market driven approach for the acquisition of transportation services and equipment which will maximize the participation of private industry. To implement this strategy, DOE is planning to issue a Request for Proposal (RFP) for the provision of the required services and equipment to accept SNF from the utilities and transport the SNF to a Federal facility. The paper discusses this strategy and describes the RFP

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

Standard guide for characterization of spent nuclear fuel in support of geologic repository disposal

CERN Document Server

American Society for Testing and Materials. Philadelphia

2009-01-01

1.1 This guide provides guidance for the types and extent of testing that would be involved in characterizing the physical and chemical nature of spent nuclear fuel (SNF) in support of its interim storage, transport, and disposal in a geologic repository. This guide applies primarily to commercial light water reactor (LWR) spent fuel and spent fuel from weapons production, although the individual tests/analyses may be used as applicable to other spent fuels such as those from research and test reactors. The testing is designed to provide information that supports the design, safety analysis, and performance assessment of a geologic repository for the ultimate disposal of the SNF. 1.2 The testing described includes characterization of such physical attributes as physical appearance, weight, density, shape/geometry, degree, and type of SNF cladding damage. The testing described also includes the measurement/examination of such chemical attributes as radionuclide content, microstructure, and corrosion product c...

ATR Spent Fuel Options Study

International Nuclear Information System (INIS)

Connolly, Michael James; Bean, Thomas E.; Brower, Jeffrey O.; Luke, Dale E.; Patterson, M. W.; Robb, Alan K.; Sindelar, Robert; Smith, Rebecca E.; Tonc, Vincent F.; Tripp, Julia L.; Winston, Philip L.

2017-01-01

The Advanced Test Reactor (ATR) is a materials and fuels test nuclear reactor that performs irradiation services for the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), Naval Reactors, the National Nuclear Security Administration (NNSA), and other research programs. ATR achieved initial criticality in 1967 and is expected to operate in support of needed missions until the year 2050 or beyond. It is anticipated that ATR will generate approximately 105 spent nuclear fuel (SNF) elements per year through the year 2050. Idaho National Laboratory (INL) currently stores 2,008 ATR SNF elements in dry storage, 976 in wet storage, and expects to have 1,000 elements in wet storage before January 2017. A capability gap exists at INL for long-term (greater than the year 2050) management, in compliance with the Idaho Settlement Agreement (ISA), of ATR SNF until a monitored retrievable geological repository is open. INL has significant wet and dry storage capabilities that are owned by the DOE Office of Environmental Management (EM) and operated and managed by Fluor Idaho, which include the Idaho Nuclear Technology and Engineering Center's (INTEC's) CPP-666, CPP-749, and CPP-603. In addition, INL has other capabilities owned by DOE-NE and operated and managed by Battelle Energy Alliance, LLC (BEA), which are located at the Materials and Fuel Complex (MFC). Additional storage capabilities are located on the INL Site at the Naval Reactors Facility (NRF). Current INL SNF management planning, as defined in the Fluor Idaho contract, shows INTEC dry fuel storage, which is currently used for ATR SNF, will be nearly full after transfer of an additional 1,000 ATR SNF from wet storage. DOE-NE tasked BEA with identifying and analyzing options that have the potential to fulfill this capability gap. BEA assembled a team comprised of SNF management experts from Fluor Idaho, Savannah River Site (SRS), INL/BEA, and the MITRE Corp with an objective of developing and analyzing

ATR Spent Fuel Options Study

Energy Technology Data Exchange (ETDEWEB)

Connolly, Michael James [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bean, Thomas E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Brower, Jeffrey O. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Luke, Dale E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Patterson, M. W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Robb, Alan K. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sindelar, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Rebecca E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Tonc, Vincent F. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Tripp, Julia L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Winston, Philip L. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2017-01-01

The Advanced Test Reactor (ATR) is a materials and fuels test nuclear reactor that performs irradiation services for the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), Naval Reactors, the National Nuclear Security Administration (NNSA), and other research programs. ATR achieved initial criticality in 1967 and is expected to operate in support of needed missions until the year 2050 or beyond. It is anticipated that ATR will generate approximately 105 spent nuclear fuel (SNF) elements per year through the year 2050. Idaho National Laboratory (INL) currently stores 2,008 ATR SNF elements in dry storage, 976 in wet storage, and expects to have 1,000 elements in wet storage before January 2017. A capability gap exists at INL for long-term (greater than the year 2050) management, in compliance with the Idaho Settlement Agreement (ISA), of ATR SNF until a monitored retrievable geological repository is open. INL has significant wet and dry storage capabilities that are owned by the DOE Office of Environmental Management (EM) and operated and managed by Fluor Idaho, which include the Idaho Nuclear Technology and Engineering Center’s (INTEC’s) CPP-666, CPP-749, and CPP-603. In addition, INL has other capabilities owned by DOE-NE and operated and managed by Battelle Energy Alliance, LLC (BEA), which are located at the Materials and Fuel Complex (MFC). Additional storage capabilities are located on the INL Site at the Naval Reactors Facility (NRF). Current INL SNF management planning, as defined in the Fluor Idaho contract, shows INTEC dry fuel storage, which is currently used for ATR SNF, will be nearly full after transfer of an additional 1,000 ATR SNF from wet storage. DOE-NE tasked BEA with identifying and analyzing options that have the potential to fulfill this capability gap. BEA assembled a team comprised of SNF management experts from Fluor Idaho, Savannah River Site (SRS), INL/BEA, and the MITRE Corp with an objective of developing and analyzing

Managing Spent Nuclear Fuel at the Idaho National Laboratory

Energy Technology Data Exchange (ETDEWEB)

Thomas Hill; Denzel L. Fillmore

2005-10-01

The Idaho National Laboratory (INL) has a large inventory of diverse types of spent nuclear fuel (SNF). This legacy derives from the history of the INL as the National Reactor Testing Station, and from its mission to recover HEU from SNF and to test and examine SNF after irradiation. The INL also has a large diversity of SNF storage facilities, some 50 years old. SNF at INL has many forms—from intact assemblies down to metallurgical mounts, and some fuel has been wet stored for over 40 years. SNF is stored bare or in metal cans under water, or dry in vaults, caissons or casks. Inspection shows varying corrosion and degradation of the SNF and its storage cans. SNF has been stored in 10 different facilities: 5 pools, one cask storage pad, one vault, two generations of caisson facilities, and one modular Independent Spent Fuel Storage Installation (ISFSI). The pools range in age from 40 years old to the most modern in the US Department of Energy (DOE) complex. The near-term objective is to move SNF from older pools to interim dry storage, allowing shutdown and decommissioning of the older facilities. This move involves drying methods that are dependent on fuel type. The long-term objective is to have INL SNF in safe dry storage and ready to be shipped to the National Repository. The unique features of the INL SNF requires special treatments and packaging to meet the proposed repository acceptance criteria and SNF will be repackaged in standardized canisters for shipment and disposal in the National Repository. Disposal will use the standardized canisters that can be co-disposed with High Level Waste glass logs to limit the total fissile material in a repository waste package. The DOE standardized canister also simplifies the repository handling of the multitude of DOE SNF sizes and shapes.

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

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.

SMILE: experimental results of the WP4 PTS large scale test performed on a component in terms of cracked cylinder involving warm pre-stress

International Nuclear Information System (INIS)

Kerkhof, K.; Bezdikian, G.; Moinereau, D.; Dahl, A; Wadier, Y.; Gilles, P.; Keim, E.; Chapuliot, S.; Taylor, N.; Lidbury, D.; Sharples, J.; Budden, P.; Siegele, D.; Nagel, G.; Bass, R.; Emond, D.

2005-01-01

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 thick walled 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). No crack initiation occurred during cooling (thermal shock loading) although the loading path crossed the fracture toughness curve in the transition region. The benefit of the WPS-effect by final re-loading up to fracture in the lower shelf region, was shown clearly. The corresponding fracture load during reloading in the lower shelf region was significantly higher than the crack initiation values of the original material in the lower shelf region. The post test fractographic evaluation showed that the fracture mode was predominantly cleavage fracture also with some secondary cracks emanating from major crack. (authors)

Hanford spent fuel inventory baseline

International Nuclear Information System (INIS)

Bergsman, K.H.

1994-01-01

This document compiles technical data on irradiated fuel stored at the Hanford Site in support of the Hanford SNF Management Environmental Impact Statement. Fuel included is from the Defense Production Reactors (N Reactor and the single-pass reactors; B, C, D, DR, F, H, KE and KW), the Hanford Fast Flux Test Facility Reactor, the Shipping port Pressurized Water Reactor, and small amounts of miscellaneous fuel from several commercial, research, and experimental reactors

INEL integrated spent nuclear fuel consolidation task team report

International Nuclear Information System (INIS)

Henry, R.N.; Clark, J.H.; Chipman, N.A.

1994-01-01

This document describes a draft plan and schedule to consolidate spent nuclear fuel (SNF) and special nuclear material (SNW) from aging storage facilities throughout the Idaho National Engineering Laboratory (INEL) to the Idaho Chemical Processing Plant (ICPP) in a safe, cost-effective, and expedient manner. A fully integrated and resource-loaded schedule was developed to achieve consolidation as soon as possible. All of the INEL SNF and SNM management task, projects, and related activities from fiscal year 1994 to the end of the consolidation period are logic-tied and integrated with each other. The schedule and plan are presented to initiate discussion of their implementation, which is expected to generate alternate concepts that can be evaluated using the methodology described in this report. Three perturbations to consolidating SNF as soon as possible are also explored. If the schedule is executed as proposed, the new and on-going consolidation activities will require about 6 years to complete and about $25.3M of additional funding. Reduced annual operating costs are expected to recover the additional investment in about 6.4 years. The total consolidation program as proposed will cost about $66.8M and require about 6 years to recover via reduced operating costs from retired SNF/SNM storage facilities. Detailed schedules and cost estimates for the Test Reactor Area Materials Test Reactor canal transfers are included as an example of the level of detail that is typical of the entire schedule (see Appendix D). The remaining work packages for each of the INEL SNF consolidation transfers are summarized in this document. Detailed cost and resource information is available upon request for any of the SNF consolidation transfers

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),...

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.

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

A report on the transport of MTR-type spent fuel assemblies of the Philippine Research Reactor (PRR-1)

International Nuclear Information System (INIS)

Yoshisaki, Magno B.; Leopando, Leonardo S.

1999-03-01

Fifty one (51) fuel assemblies of mixed enrichment from the Philippine Research Reactor (PRR-1), consisting of 50 spent and 1 fresh, were shipped to the United States last 14 March 1999 under the U.S. Return of Foreign Research Reactor (FRR) fuel policy. The shipment was in line with the U.S. initiative to implement its Record of Decision (ROD) which took effect on 13 May 1996 to accept and manage all FRR uranium fuel of U.S. origin and enriched in the United States. The shipment program would last10 years, ending midnight of 13 May 2006. The ROD provided a 3 year extension period within which to accept FRR spent nuclear fuel (SNF) withdrawn from reactors after 2006. The U.S. policy gave priority to the NPT significance of high enriched U, as the prime target of the return of FRR policy. Classified as a developing country, the Philippines, through the PNRI, signed a contract with the U.S. Department of Energy for the cost-free shipment of PRR-1 spent fuel to the United States. Spent fuel loading and transport operations to the port area lasted seven (7) days, from 8 to 14 March 1999. (Author)

Vacuum Drying Tests for Storage of Aluminum Spent Nuclear Fuel

International Nuclear Information System (INIS)

Chen, K.F.; Large, W.S.; Sindelar, R.L.

1998-05-01

A total inventory of up to approximately 32,000 aluminum-based spent nuclear fuel (Al SNF) assemblies are expected to be shipped to Savannah River Site (SRS) from domestic and foreign research reactors over the next several decades. Treatment technologies are being developed as alternatives to processing for the ultimate disposition of Al SNF in the geologic repository. One technology, called Direct/Co-disposal of Al SNF, would place the SNF into a canister ready for disposal in a waste package, with or without canisters containing high-level radioactive waste glass logs, in the repository. The Al SNF would be transferred from wet storage and would need to be dried in the Al SNF canister. The moisture content inside the Al SNF canister is limited to avoid excessive Al SNF corrosion and hydrogen buildup during interim storage before disposal. A vacuum drying process was proposed to dry the Al SNF in a canister. There are two major concerns for the vacuum drying process. One is water inside the canister could become frozen during the vacuum drying process and the other one is the detection of dryness inside the canister. To vacuum dry an irradiated fuel in a heavily shielded canister, it would be very difficult to open the lid to inspect the dryness during the vacuum drying operation. A vacuum drying test program using a mock SNF assembly was conducted to demonstrate feasibility of drying the Al SNF in a canister. These tests also served as a check-out of the drying apparatus for future tests in which irradiated fuel would be loaded into a canister under water followed by drying for storage

An improved characterization method for international accountancy measurements of fresh and irradiated mixed oxide (MOX) fuel: helping achieve continual monitoring and safeguards through the fuel cycle

Energy Technology Data Exchange (ETDEWEB)

Evans, Louise G [Los Alamos National Laboratory; Croft, Stephen [Los Alamos National Laboratory; Swinhoe, Martyn T [Los Alamos National Laboratory; Tobin, S. J. [Los Alamos National Laboratory; Menlove, H. O. [Los Alamos National Laboratory; Schear, M. A. [Los Alamos National Laboratory; Worrall, Andrew [U.K. NNL

2011-01-13

Nuclear fuel accountancy measurements are conducted at several points through the nuclear fuel cycle to ensure continuity of knowledge (CofK) of special nuclear material (SNM). Non-destructive assay (NDA) measurements are performed on fresh fuel (prior to irradiation in a reactor) and spent nuclear fuel (SNF) post-irradiation. We have developed a fuel assembly characterization system, based on the novel concept of 'neutron fingerprinting' with multiplicity signatures to ensure detailed CofK of nuclear fuel through the entire fuel cycle. The neutron fingerprint in this case is determined by the measurement of the various correlated neutron signatures, specific to fuel isotopic composition, and therefore offers greater sensitivity to variations in fissile content among fuel assemblies than other techniques such as gross neutron counting. This neutron fingerprint could be measured at the point of fuel dispatch (e.g. from a fuel fabrication plant prior to irradiation, or from a reactor site post-irradiation), monitored during transportation of the fuel assembly, and measured at a subsequent receiving site (e.g. at the reactor site prior to irradiation, or reprocessing facility post-irradiation); this would confirm that no unexpected changes to the fuel composition or amount have taken place during transportation and/ or reactor operations. Changes may indicate an attempt to divert material for example. Here, we present the current state of the practice of fuel measurements for both fresh mixed oxide (MOX) fuel and SNF (both MOX and uranium dioxide). This is presented in the framework of international safeguards perspectives from the US and UK. We also postulate as to how the neutron fingerprinting concept could lead to improved fuel characterization (both fresh MOX and SNF) resulting in: (a) assured CofK of fuel across the nuclear fuel cycle, (b) improved detection of SNM diversion, and (c) greater confidence in safeguards of SNF transportation.

The Decay of Communism: Managing Spent Nuclear Fuel in the Soviet Union, 1937-1991

International Nuclear Information System (INIS)

Hoegselius, Per

2010-09-01

The historical evolution of spent nuclear fuel (SNF) decision-making in Western Europe and North America is already fairly well-known. For the former socialist countries of Eastern Europe, and in particular the Soviet Union, we know less. There have recently been several good studies of Soviet nuclear power history (e.g. Schmid 2004, 2006, Josephson 2005), but none of them has gone into any depth when it comes to SNF, but rather focused on nuclear power reactors, public acceptance, the role of the media, etc. There are also several good overviews available that problematize the radioactive legacy of the Soviet Union, including the SNF and waste issue, but these studies do not address the historical dynamics and evolution of SNF management over a longer period of time; in other words, they fail to explain how and why the present state of affairs have actually come into being. The aim of this paper is to provide historical insight into the dynamics of SNF decision-making in the Soviet Union, from the origins of nuclear engineering in the 1930s to the collapse of the country in 1991. The nuclear fuel system can be described as a large technical system with a variety of interrelated components. The system is 'large' both because it involves key links between geographically disperse activities, and because it involves a variety of technologies, organizations and people that influence the dynamics and evolution of the system. Soviet SNF history is of particular interest in this context, with a nuclear fuel system that was the most complex in the world. The USSR was a pioneer within nuclear power and developed a variety of reactor designs and technologies for uranium mining, conversion and enrichment, as well as for transport, treatment, storage and reprocessing of spent nuclear fuel. It explored both military and civil uses of the atom, and an enormous amount of people and organizations were involved in realizing highly ambitious nuclear programmes. The USSR is of

CIRFT Data Update and Data Analyses for Spent Nuclear Fuel Vibration Reliability Study

Energy Technology Data Exchange (ETDEWEB)

Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wang, Hong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-08-01

The objective of this research is to collect experimental data on spent nuclear fuel (SNF) from pressurized water reactors (PWRs), including the H. B. Robinson Nuclear Power Station (HBR), Catawba Nuclear Station, North Anna Nuclear Power Station (NA), and the Limerick Nuclear Power Station (LMK) boiling water reactor (BWR).

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.

Cost analysis of spent nuclear fuel management

International Nuclear Information System (INIS)

Robertson, D.L.M.; Ford, L.M.

1993-01-01

The Department of Energy Civilian Radioactive Waste Management System (CRWMS) is chartered to develop a waste management system for the safe disposal of spent nuclear fuel (SNF) from the 131 nuclear power reactors in the United States and a certain amount of high level waste (HLW) from reprocessing operations. The current schedule is to begin accepting SNF in 1998 for storage at a Monitored Retrievable Storage (MRS) facility. Subsequently, beginning in 2010, the system is scheduled to begin accepting SNF at a permanent geologic repository in 2010 and HLW in 2015. At this time, a MRS site has not been selected. Yucca Mountain, Nevada is currently being evaluated as the candidate site for the repository for permanent geologic disposal of SNF. All SNF, with the possible exception of the SNF from the western reactors, is currently planned to be shipped to or through the MRS site en route to the repository. The repository will operate in an acceptance and performance confirmation phase for a 50 year period beginning in 2010 with an additional nine year closure and five year decontamination and decommissioning period. The MRS has a statutory maximum capacity of 15,000 Metric Tons Uranium (MTU), with a further restriction that it may not store more than 10,000 MTU until the repository begins accepting waste. The repository is currently scheduled to store 63,000 MTU of SNF and an additional 7,000 MTU equivalent of HLW for a total capacity of 70,000 MTU. The amended act specified the MRS storage limits and identified Yucca Mountain as the only site to be characterized. Also, an Office of the Nuclear Waste Negotiator was established to secure a voluntary host site for the MRS. The MRS, the repository, and all waste containers/casks will go through a Nuclear Regulatory Commission licensing process much like the licensing process for a nuclear power plant. Environmental assessments and impact statements will be prepared for both the MRS and repository

Advantages on dry interim storage for spent nuclear fuel

International Nuclear Information System (INIS)

Romanato, L.S.; Rzyski, B.M.

2006-01-01

When the nuclear fuel lose its ability to efficiently create energy it is removed from the core reactor and moved to a storage unit waiting for a final destination. Generally, the spent nuclear fuel (SNF) remains inside concrete basins with water within the reactors facility for the radioactive activity decay. Water cools the generated heat and shields radioactivity emissions. After some period of time in water basins the SNF can be sent to a definitive deposition in a geological repository and handled as radioactive waste or to reprocessing installations, or still wait for a future solution. Meanwhile, SNF remains stored for a period of time in dry or wet installations, depending on the method adopted by the nuclear power plant or other plans of the country. In many SNF wet storage sites the capacity can be fulfilled very quickly. If so, additional area or other alternative storage system should be given. There are many options to provide capacity increase in the wet storage area, but dry storages are worldwide preferred since it reduces corrosion concerns. In the wet storage the temperature and water purity should be constantly controlled whereas in the dry storage the SNF stands protected in specially designed canisters. Dry interim storages are practical and approved in many countries especially that have the 'wait and see' philosophy (wait to see new technologies development). This paper shows the advantages of dry interim storages sites in comparison with the wet ones and the nowadays problems as terrorism. (Author)

Advantages on dry interim storage for spent nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Romanato, L.S. [Centro Tecnologico da Marinha em Sao Paulo, Av. Professor Lineu Prestes 2468, 05508-900 Sao Paulo (Brazil); Rzyski, B.M. [IPEN/ CNEN-SP, 05508-000 Sao Paulo (Brazil)]. e-mail: romanato@ctmsp.mar.mil.br

2006-07-01

When the nuclear fuel lose its ability to efficiently create energy it is removed from the core reactor and moved to a storage unit waiting for a final destination. Generally, the spent nuclear fuel (SNF) remains inside concrete basins with water within the reactors facility for the radioactive activity decay. Water cools the generated heat and shields radioactivity emissions. After some period of time in water basins the SNF can be sent to a definitive deposition in a geological repository and handled as radioactive waste or to reprocessing installations, or still wait for a future solution. Meanwhile, SNF remains stored for a period of time in dry or wet installations, depending on the method adopted by the nuclear power plant or other plans of the country. In many SNF wet storage sites the capacity can be fulfilled very quickly. If so, additional area or other alternative storage system should be given. There are many options to provide capacity increase in the wet storage area, but dry storages are worldwide preferred since it reduces corrosion concerns. In the wet storage the temperature and water purity should be constantly controlled whereas in the dry storage the SNF stands protected in specially designed canisters. Dry interim storages are practical and approved in many countries especially that have the 'wait and see' philosophy (wait to see new technologies development). This paper shows the advantages of dry interim storages sites in comparison with the wet ones and the nowadays problems as terrorism. (Author)

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...

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.

Spent fuel measurements. passive neutron albedo reactivity (PNAR) and photon signatures

Energy Technology Data Exchange (ETDEWEB)

Eigenbrodt, Julia [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Menlove, Howard Olsen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-03-29

The International Atomic Energy Agency’s (IAEA) safeguards technical objective is the timely detection of a diversion of a significant quantity of nuclear material from peaceful activities to the manufacture of nuclear weapons or of other nuclear explosive devices or for purposes unknown, and deterrence of such diversion by the risk of early detection. An important IAEA task towards meeting this objective is the ability to accurately and reliably measure spent nuclear fuel (SNF) to verify reactor operating parameters and verify that the fuel has not been removed from reactors or SNF storage facilities. This dissertation analyzes a method to improve the state-of-the-art of nuclear material safeguards measurements using two combined measurement techniques: passive neutron albedo reactivity (PNAR) and passive spectral photon measurements.

DIissolution of low enriched uranium from the experimental breeder reactor-II fuel stored at the Idaho National Laboratory

Energy Technology Data Exchange (ETDEWEB)

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

2017-06-28

The Idaho National Laboratory (INL) is actively engaged in the development of electrochemical processing technology for the treatment of fast reactor fuels using irradiated fuel from the Experimental Breeder Reactor-II (EBR-II) as the primary test material. The research and development (R&D) activities generate a low enriched uranium (LEU) metal product from the electrorefining of the EBR-II fuel and the subsequent consolidation and removal of chloride salts by the cathode processor. The LEU metal ingots from past R&D activities are currently stored at INL awaiting disposition. One potential disposition pathway is the shipment of the ingots to the Savannah River Site (SRS) for dissolution in H-Canyon. Carbon steel cans containing the LEU metal would be loaded into reusable charging bundles in the H-Canyon Crane Maintenance Area and charged to the 6.4D or 6.1D dissolver. The LEU dissolution would be accomplished as the final charge in a dissolver batch (following the dissolution of multiple charges of spent nuclear fuel (SNF)). The solution would then be purified and the ²³⁵U enrichment downblended to allow use of the U in commercial reactor fuel. To support this potential disposition path, the Savannah River National Laboratory (SRNL) developed a dissolution flowsheet for the LEU using samples of the material received from INL.

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)

Creep Analysis of Aluminum-Based Spent Nuclear Fuel in Repository Storage

International Nuclear Information System (INIS)

Gong, C.; Lam, P.S.; Sindelar, R.L.

1998-07-01

Aluminum-clad, aluminum-based spent nuclear fuels (Al SNF) from foreign and domestic research reactors are being consolidated at the Savannah River Site (SRS). These fuels are planned to be put into dry storage followed by disposal in the federal repository. Temperature conditions in storage and disposal systems due to nuclear decay heat sources will promote creep information of the fuel elements. Excessive deformation of the Al SNF will cause gross distortion (slump) of the fuels and may cause gross cladding rupture

The Decay of Communism: Managing Spent Nuclear Fuel in the Soviet Union, 1937-1991

Energy Technology Data Exchange (ETDEWEB)

Hoegselius, Per (History of Science and Technology, Royal Inst. of Technology, Stockholm (Sweden)), e-mail: perho@kth.se

2010-09-15

The historical evolution of spent nuclear fuel (SNF) decision-making in Western Europe and North America is already fairly well-known. For the former socialist countries of Eastern Europe, and in particular the Soviet Union, we know less. There have recently been several good studies of Soviet nuclear power history (e.g. Schmid 2004, 2006, Josephson 2005), but none of them has gone into any depth when it comes to SNF, but rather focused on nuclear power reactors, public acceptance, the role of the media, etc. There are also several good overviews available that problematize the radioactive legacy of the Soviet Union, including the SNF and waste issue, but these studies do not address the historical dynamics and evolution of SNF management over a longer period of time; in other words, they fail to explain how and why the present state of affairs have actually come into being. The aim of this paper is to provide historical insight into the dynamics of SNF decision-making in the Soviet Union, from the origins of nuclear engineering in the 1930s to the collapse of the country in 1991. The nuclear fuel system can be described as a large technical system with a variety of interrelated components. The system is 'large' both because it involves key links between geographically disperse activities, and because it involves a variety of technologies, organizations and people that influence the dynamics and evolution of the system. Soviet SNF history is of particular interest in this context, with a nuclear fuel system that was the most complex in the world. The USSR was a pioneer within nuclear power and developed a variety of reactor designs and technologies for uranium mining, conversion and enrichment, as well as for transport, treatment, storage and reprocessing of spent nuclear fuel. It explored both military and civil uses of the atom, and an enormous amount of people and organizations were involved in realizing highly ambitious nuclear programmes. The USSR is

Enhanced styrene recovery from waste polystyrene pyrolysis using response surface methodology coupled with Box-Behnken design.

Science.gov (United States)

Mo, Yu; Zhao, Lei; Wang, Zhonghui; Chen, Chia-Lung; Tan, Giin-Yu Amy; Wang, Jing-Yuan

2014-04-01

A work applied response surface methodology coupled with Box-Behnken design (RSM-BBD) has been developed to enhance styrene recovery from waste polystyrene (WPS) through pyrolysis. The relationship between styrene yield and three selected operating parameters (i.e., temperature, heating rate, and carrier gas flow rate) was investigated. A second order polynomial equation was successfully built to describe the process and predict styrene yield under the study conditions. The factors identified as statistically significant to styrene production were: temperature, with a quadratic effect; heating rate, with a linear effect; carrier gas flow rate, with a quadratic effect; interaction between temperature and carrier gas flow rate; and interaction between heating rate and carrier gas flow rate. The optimum conditions for the current system were determined to be at a temperature range of 470-505°C, a heating rate of 40°C/min, and a carrier gas flow rate range of 115-140mL/min. Under such conditions, 64.52% WPS was recovered as styrene, which was 12% more than the highest reported yield for reactors of similar size. It is concluded that RSM-BBD is an effective approach for yield optimization of styrene recovery from WPS pyrolysis. Copyright © 2014 Elsevier Ltd. All rights reserved.

Solubility of uranium in liquid gallium, indium and their alloys

International Nuclear Information System (INIS)

Volkovich, Vladimir A.; Maltsev, Dmitry S.; Yamschikov, Leonid F.; Osipenko, Alexander G.; Kormilitsyn, Mikhail V.

2014-01-01

Pyrochemical reprocessing of spent nuclear fuels (SNF) employing molten salts and liquid metals as working media is considered as a possible alternative to the existing liquid extraction (PUREX) processes. Liquid salts and metals allow reprocessing highly irradiated high burn-up fuels with short cooling times, including the fuels of fast neutron reactors. Pyrochemical technology opens a way to practical realization of short closed fuel cycle. Liquid low-melting metals are immiscible with molten salts and can be effectively used for separation (or selective extraction) of SNF components dissolved in fused salts. Binary or ternary alloys of eutectic compositions can be employed to lower the melting point of the metallic phase. However, the information on SNF components behaviour and properties in ternary liquid metal alloys is very scarce

Potential benefits and impacts on the CRWMS transportation system of filling spent fuel shipping casks with depleted uranium silicate glass

International Nuclear Information System (INIS)

Pope, R.B.; Forsberg, C.W.; DeHart, M.D.; Childs, K.W.; Tang, J.S.

1996-01-01

A new technology, the Depleted Uranium Silicate COntainer Fill System (DUSCOFS), is proposed to improve the performance and reduce the uncertainties of geological disposal of spent nuclear fuel (SNF), thus reducing both radionuclide release rates from the waste package and the potential for repository nuclear criticality events. DUSCOFS may also provide benefits for SNF storage and transport if it is loaded into the container early in the waste management cycle. Assessments have been made of the benefits to be derived by placing depleted uranium silicate (DUS) glass into SNF containers for enhancing repository performance assessment and controlling criticality over geologic times in the repository. Also, the performance, benefits, and impacts which can be derived if the SNF is loaded into a multi-purpose canister with DUS glass at a reactor site have been assessed. The DUSCOFS concept and the benefits to the waste management cycle of implementing DUSCOFS early in the cycle are discussed in this paper

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.

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.

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

Short-Term Nose-Only Water-Pipe (Shisha Smoking Exposure Accelerates Coagulation and Causes Cardiac Inflammation and Oxidative Stress in Mice

Directory of Open Access Journals (Sweden)

Abderrahim Nemmar

2015-01-01

Full Text Available Background/Aim: Water-pipe smoking (WPS has acquired worldwide popularity, and is disseminating particularly rapidly in Europe and North America. However, little is known about the short-term cardiovascular effects of WPS. Methods: Presently, we assessed the short-term cardiovascular effects of nose-only exposure to mainstream WPS in BALB/c mice for 30 min/day for 5 consecutive days. Control mice were exposed to air. At the end of the exposure period, several cardiovascular endpoints were measured. Results: WPS did not affect the number of leukocytes and the plasma concentrations of C-reactive protein, tumor necrosis factor α (TNFα and interleukin-6 (IL-6. Likewise, plasma levels of lipid peroxidation (LPO, reduced glutathione (GSH and catalase were not affected by WPS. By contrast, WPS aggravated in vivo thrombosis by shortening the thrombotic occlusion time in pial arterioles and venules. The number of circulating platelets was reduced by WPS suggesting the occurrence of platelet aggregation in vivo. Elevated concentrations of fibrinogen and plasminogen activator inhibitor-1 were seen after the exposure to WPS. Blood samples taken from mice exposed to WPS and exposed to adenosine diphosphate showed more platelet aggregation. The heart concentrations of IL-6 and TNFα were augmented by WPS. Likewise, heart levels of LPO, reactive oxygen species and the antioxidants catalase and GSH were increased by WPS. However, the systolic blood pressure and heart rate were not affected by WPS. Conclusion: It can be concluded that short-term exposure to WPS exerts procoagulatory effects and induce cardiac inflammation and oxidative stress. At the time point investigated, there was no evidence for blood inflammation or oxidative stress.

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

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.

Overview of the spent nuclear fuel project at Hanford

International Nuclear Information System (INIS)

Daily, J.L.

1995-02-01

The Spent Nuclear Fuel Project's mission at Hanford is to open-quotes Provide safe, economic and environmentally sound management of Hanford spent nuclear fuel in a manner which stages it to final disposition.close quotes The inventory of spent nuclear fuel (SNF) at the Hanford Site covers a wide variety of fuel types (production reactor to space reactor) in many facilities (reactor fuel basins to hot cells) at locations all over the Site. The 2,129 metric tons of Hanford SNF represents about 80% of the total US Department of Energy (DOE) inventory. About 98.5% of the Hanford SNF is 2,100 metric tons of metallic uranium production reactor fuel currently stored in the 1950s vintage K Basins in the 100 Area. This fuel has been slowly corroding, generating sludge and contaminating the basin water. This condition, coupled with aging facilities with seismic vulnerabilities, has been identified by several groups, including stakeholders, as being one of the most urgent safety and environmental concerns at the Hanford Site. As a direct result of these concerns, the Spent Nuclear Fuel Project was recently formed to address spent fuel issues at Hanford. The Project has developed the K Basins Path Forward to remove fuel from the basins and place it in dry interim storage. Alternatives that addressed the requirements were developed and analyzed. The result is a two-phased approach allowing the early removal of fuel from the K Basins followed by its stabilization and interim storage consistent with the national program

Burn-up credit criticality safety benchmark phase VII - UO2 fuel: study of spent fuel compositions for long-term disposal

International Nuclear Information System (INIS)

2012-01-01

After spent nuclear fuel (SNF) is discharged from a nuclear reactor, fuel composition and reactivity continue to vary as a function of time due to the decay of unstable nuclides. Accurate predictions of the concentrations of long-lived radionuclides in SNF, which represent a significant potential hazard to human beings and to the environment over a very long period, are particularly necessary for radiological dose assessments. This report assesses the ability of existing computer codes and associated nuclear data to predict isotopic compositions and their corresponding neutron multiplication factor (k eff ) values for pressurised-water-reactor (PWR) UO 2 fuel at 50 GWd/MTU burn-up in a generic spent fuel cask configuration. Fuel decay compositions and k eff values have been calculated for 30 post-irradiation time steps out to one million years

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

Regulation at nuclear fuel cycle

International Nuclear Information System (INIS)

2002-01-01

This bulletin contains information about activities of the Nuclear Regulatory Authority of the Slovak Republic (UJD). In this leaflet the role of the UJD in regulation at nuclear fuel cycle is presented. The Nuclear Fuel Cycle (NFC) is a complex of activities linked with production of nuclear fuel for nuclear reactors as a source of energy used for production of electricity and heat, and of activities linked with spent nuclear fuel handling. Activities linked with nuclear fuel (NF) production, known as the Front-End of Nuclear Fuel Cycle, include (production of nuclear fuel from uranium as the most frequently used element). After discharging spent nuclear fuel (SNF) from nuclear reactor the activities follow linked with its storage, reprocessing and disposal known as the Back-End of Nuclear Fuel Cycle. Individual activity, which penetrates throughout the NFC, is transport of nuclear materials various forms during NF production and transport of NF and SNF. Nuclear reactors are installed in the Slovak Republic only in commercial nuclear power plants and the NFC is of the open type is imported from abroad and SNF is long-term supposed without reprocessing. The main mission of the area of NFC is supervision over: - assurance of nuclear safety throughout all NFC activities; - observance of provisions of the Treaty on Non-Proliferation of Nuclear Weapons during nuclear material handling; with an aim to prevent leakage of radioactive substances into environment (including deliberated danage of NFC sensitive facilities and misuse of nuclear materials to production of nuclear weapons. The UJD carries out this mission through: - assessment of safety documentation submitted by operators of nuclear installations at which nuclear material, NF and SNF is handled; - inspections concentrated on assurance of compliance of real conditions in NFC, i.e. storage and transport of NF and SNF; storage, transport and disposal of wastes from processing of SNF; with assumptions of the safety

Naval Waste Package Design Report

International Nuclear Information System (INIS)

M.M. Lewis

2004-01-01

A design methodology for the waste packages and ancillary components, viz., the emplacement pallets and drip shields, has been developed to provide designs that satisfy the safety and operational requirements of the Yucca Mountain Project. This methodology is described in the ''Waste Package Design Methodology Report'' Mecham 2004 [DIRS 166168]. To demonstrate the practicability of this design methodology, four waste package design configurations have been selected to illustrate the application of the methodology. These four design configurations are the 21-pressurized water reactor (PWR) Absorber Plate waste package, the 44-boiling water reactor (BWR) waste package, the 5-defense high-level waste (DHLW)/United States (U.S.) Department of Energy (DOE) spent nuclear fuel (SNF) Co-disposal Short waste package, and the Naval Canistered SNF Long waste package. Also included in this demonstration is the emplacement pallet and continuous drip shield. The purpose of this report is to document how that design methodology has been applied to the waste package design configurations intended to accommodate naval canistered SNF. This demonstrates that the design methodology can be applied successfully to this waste package design configuration and support the License Application for construction of the repository

Annex D 200 Area Interim Storage Area Final Safety Analysis Report Volume 5 (FSAR) (Section 1 and 2)

International Nuclear Information System (INIS)

CARRELL, R.D.

2003-01-01

The 200 Area Interim Storage Area (200 Area ISA) at the Hanford Site provides for the interim storage of non-defense reactor spent nuclear fuel (SNF) housed in aboveground dry cask storage systems. The 200 Area ISA is a relatively simple facility consisting of a boundary fence with gates, perimeter lighting, and concrete and gravel pads on which to place the dry storage casks. The fence supports safeguards and security and establishes a radiation protection buffer zone. The 200 Area ISA is nominally 200,000 ft 2 and is located west of the Canister Storage Building (CSB). Interim storage at the 200 Area ISA is intended for a period of up to 40 years until the materials are shipped offsite to a disposal facility. This Final Safety Analysis Report (FSAR) does not address removal from storage or shipment from the 200 Area ISA. Three different SNF types contained in three different dry cask storage systems are to be stored at the 200 Area ISA, as follows: (1) Fast Flux Test Facility (FFTF) Fuel--Fifty-three interim storage casks (ISC), each holding a core component container (CCC), will be used to store the FFTF SNF currently in the 400 Area. (2) Neutron Radiography Facility (NRF)TRIGA--One Rad-Vault container stores two DOT-6M 3 containers and six NRF TRIGA casks. (3) Commercial Light Water Reactor Fuel--Six International Standards Organization (ISO) containers, each holding a NAC-1 cask with an inner commercial light water reactor (LWR) canister, are used for storing commercial LWR SNF from the 300 Area. An aboveground dry cask storage location is necessary for the spent fuel because the current storage facilities are being shut down and deactivated. The spent fuel is being transferred to interim storage because there is no permanent repository storage currently available

Calculation Method for the Projection of Future Spent Nuclear Fuel Discharges

International Nuclear Information System (INIS)

B. McLeod

2002-01-01

This report describes the calculation method developed for the projection of future utility spent nuclear fuel (SNF) discharges in regard to their timing, quantity, burnup, and initial enrichment. This projection method complements the utility-supplied RW-859 data on historic discharges and short-term projections of SNF discharges by providing long-term projections that complete the total life cycle of discharges for each of the current U.S. nuclear power reactors. The method was initially developed in mid-1999 to update the SNF discharge projection associated with the 1995 RW-859 utility survey (CRWMS M and O 1996). and was further developed as described in Rev. 00 of this report (CRWMS M and O 2001a). Primary input to the projection of SNF discharges is the utility projection of the next five discharges from each nuclear unit, which is provided via the revised final version of the Energy Information Administration (EIA) 1998 RW-859 utility survey (EIA 2000a). The projection calculation method is implemented via a set of Excel 97 spreadsheets. These calculations provide the interface between receipt of the utility five-discharge projections that are provided in the RW-859 survey, and the delivery of projected life-cycle SNF discharge quantities and characteristics in the format requisite for performing logistics analysis to support design of the Civilian Radioactive Waste Management System (CRWMS). Calculation method improvements described in this report include the addition of a reactor-specific maximum enrichment-based discharge burnup limit. This limit is the consequence of the enrichment limit, currently 5 percent. which is imposed as a Nuclear Regulatory Commission (NRC) license condition on nuclear fuel fabrication plants. In addition, the calculation method now includes the capability for projecting future nuclear plant power upratings, consistent with many such recent plant uprates and the prospect of additional future uprates. Finally. this report

DOE-owned spent nuclear fuel strategic plan. Revision 1

International Nuclear Information System (INIS)

1996-09-01

The Department of Energy (DOE) is responsible for safely and efficiently managing DOE-owned spent nuclear fuel (SNF) and SNF returned to the US from foreign research reactors (FRR). The fuel will be treated where necessary, packaged suitable for repository disposal where practicable, and placed in interim dry storage. These actions will remove remaining vulnerabilities, make as much spent fuel as possible ready for ultimate disposition, and substantially reduce the cost of continued storage. The goal is to complete these actions in 10 years. This SNF Strategic Plan updates the mission, vision, objectives, and strategies for the management of DOE-owned SNF articulated by the SNF Strategic Plan issued in December 1994. The plan describes the remaining issues facing the EM SNF Program, lays out strategies for addressing these issues, and identifies success criteria by which program progress is measured. The objectives and strategies in this plan are consistent with the following Em principles described by the Assistance Secretary in his June 1996 initiative to establish a 10-year time horizon for achieving most program objectives: eliminate and manage the most serious risks; reduce mortgage and support costs to free up funds for further risk reduction; protect worker health and safety; reduce generation of wastes; create a collaborative relationship between DOE and its regulators and stakeholders; focus technology development on cost and risk reduction; and strengthen management and financial control

The Yucca Mountain Repository - Too Little, Too Late

International Nuclear Information System (INIS)

Eriksson, L.G.; Pentz, D.L.

2009-01-01

In 2008, the U.S. Department of Energy (US DOE) announced that the nation's first (and only pursued) deep geological disposal system (repository) for 70,000 metric tonnes of spent nuclear fuel (SNF) and other high-level radioactive waste (HLW) at the Yucca Mountain (YM) site in Nevada would: 1. Not be able to accommodate the projected stockpile of utility-generated SNF beyond 2010. 2. Open no earlier than in 2020, i.e., more than 22 years behind the statutory-mandated opening date. In the meantime, the US DOE is legally obligated to compensate the utilities from January 31, 1998, until it takes title to the utilities' SNF. In 2005 when the YM SNF repository was projected to open in 2010, the utilities estimated that, depending upon how close to 2010 the YM repository opened, the 'breach-of-contract' compensation could be in the range of between 100 billion and 300 billion U.S. dollars ($300 B), which would exceed the 2008 projected life-cycle cost of $96 B for the YM repository. It thus seems appropriate to look beyond the YM repository and call upon the U.S. Congress to promptly act and open new avenues allowing the US DOE to more timely and cost-effectively take title to both existing and pending SNF the current fleet of 104 reactors will generate through the next 60 years. Options for SNF arising from an additional 50 reactors should also be provided. Based on our more than 60 years of combined involvement in nuclear waste management in the USA and abroad, we submit the following industrial-scale-proven, repository-related, nuclear-waste-management and disposition solutions for prompt Congressional consideration and action: 1. An increase in the disposal capacity (and perhaps mission) of the YM repository. 2. Prompt establishment of at least one large federal monitored retrievable storage (MRS) facility for utility-generated SNF. 3. Continued research in reprocessing options of existing and pending SNF with defined milestones. 4. Resurrection of a second

The Planning, Licensing, Modifications, and Use of a Russian Vessel for Shipping Spent Nuclear Fuel by Sea in Support of the DOE RRRFR Program

International Nuclear Information System (INIS)

Tyacke, Michael; Bolshinsky, Igor; Tomczak, Wlodzimierz; Naletov, Sergey; Pichugin, Oleg

2001-01-01

The Russian Research Reactor Fuel Return (RRRFR) Program, under the U.S. Department of Energy's Global Threat Reduction Initiative, began returning Russian-supplied high-enriched uranium (HEU) spent nuclear fuel (SNF), stored at Russian-designed research reactors throughout the world, to Russia in January 2006. During the first years of making HEU SNF shipments, it became clear that the modes of transportation needed to be expanded from highway and railroad to include sea and air to meet the extremely aggressive commitment of completing the first series of shipments by the end of 2010. The first shipment using sea transport was made in October 2008 and used a non-Russian flagged vessel. The Russian government reluctantly allowed a one-time use of the foreign-owned vessel into their highly secured seaport, with the understanding that any future shipments would be made using a vessel owned and operated by a Russian company. ASPOL-Baltic of St. Petersburg, Russia, owns and operates a small fleet of vessels and has a history of shipping nuclear materials. ASPOL-Baltic's vessels were licensed for shipping nuclear materials; however, they were not licensed to transport SNF materials. After a thorough review of ASPOL Baltic's capabilities and detailed negotiations, it was agreed that a contract would be let with ASPOL-Baltic to license and refit their MCL Trader vessel for hauling SNF in support of the RRRFR Program. This effort was funded through a contract between the RRRFR Program, Idaho National Laboratory, and Radioactive Waste Management Plant of Swierk, Poland. This paper discusses planning, Russian and international maritime regulations and requirements, Russian authorities reviews and approvals, licensing, design, and modifications made to the vessel in preparation for SNF shipments. A brief summary of actual shipments using this vessel, experiences, and lessons learned also are described.

The Planning, Licensing, Modifications, and Use of a Russian Vessel for Shipping Spent Nuclear Fuel by Sea in Support of the DOE RRRFR Program

Energy Technology Data Exchange (ETDEWEB)

Michael Tyacke; Dr. Igor Bolshinsky; Wlodzimierz Tomczak; Sergey Naletov; Oleg Pichugin

2001-10-01

The Russian Research Reactor Fuel Return (RRRFR) Program, under the U.S. Department of Energy’s Global Threat Reduction Initiative, began returning Russian-supplied high-enriched uranium (HEU) spent nuclear fuel (SNF), stored at Russian-designed research reactors throughout the world, to Russia in January 2006. During the first years of making HEU SNF shipments, it became clear that the modes of transportation needed to be expanded from highway and railroad to include sea and air to meet the extremely aggressive commitment of completing the first series of shipments by the end of 2010. The first shipment using sea transport was made in October 2008 and used a non-Russian flagged vessel. The Russian government reluctantly allowed a one-time use of the foreign-owned vessel into their highly secured seaport, with the understanding that any future shipments would be made using a vessel owned and operated by a Russian company. ASPOL-Baltic of St. Petersburg, Russia, owns and operates a small fleet of vessels and has a history of shipping nuclear materials. ASPOL-Baltic’s vessels were licensed for shipping nuclear materials; however, they were not licensed to transport SNF materials. After a thorough review of ASPOL Baltic’s capabilities and detailed negotiations, it was agreed that a contract would be let with ASPOL-Baltic to license and refit their MCL Trader vessel for hauling SNF in support of the RRRFR Program. This effort was funded through a contract between the RRRFR Program, Idaho National Laboratory, and Radioactive Waste Management Plant of Swierk, Poland. This paper discusses planning, Russian and international maritime regulations and requirements, Russian authorities’ reviews and approvals, licensing, design, and modifications made to the vessel in preparation for SNF shipments. A brief summary of actual shipments using this vessel, experiences, and lessons learned also are described.

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.

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).

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.

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.

Direct conversion of surplus fissile materials, spent nuclear fuel, and other materials to high-level-waste glass

International Nuclear Information System (INIS)

Forsberg, C.W.; Elam, K.R.

1995-01-01

With the end of the cold war the United States, Russia, and other countries have excess plutonium and other materials from the reductions in inventories of nuclear weapons. The United States Academy of Sciences (NAS) has recommended that these surplus fissile materials (SFMs) be processed so they are no more accessible than plutonium in spent nuclear fuel (SNF). This spent fuel standard, if adopted worldwide, would prevent rapid recovery of SFMs for the manufacture of nuclear weapons. The NAS recommended investigation of three sets of options for disposition of SFMs while meeting the spent fuel standard: (1) incorporate SFMs with highly radioactive materials and dispose of as waste, (2) partly burn the SFMs in reactors with conversion of the SFMs to SNF for disposal, and (3) dispose of the SFMs in deep boreholes. The US Government is investigating these options for SFM disposition. A new method for the disposition of SFMs is described herein: the simultaneous conversion of SFMs, SNF, and other highly radioactive materials into high-level-waste (HLW) glass. The SFMs include plutonium, neptinium, americium, and 233 U. The primary SFM is plutonium. The preferred SNF is degraded SNF, which may require processing before it can be accepted by a geological repository for disposal

Review of consequences of uranium hydride formation in N-Reactor fuel elements stored in the K-Basins

Energy Technology Data Exchange (ETDEWEB)

Weber, J.W.

1994-09-28

The 105-K Basins on the Hanford site are used to store uranium fuel elements and assemblies irradiated in and discharged from N Reactor. The storage cylinders in KW Basin are known to have some broken N reactor fuel elements in which the exposed uranium is slowly reacting chemically with water in the cylinder. The products of these reactions are uranium oxide, hydrogen, and potentially some uranium hydride. The purpose of this report is to document the results f the latest review of potential, but highly unlikely accidents postulated to occur as closed cylinders containing N reactor fuel assemblies are opened under water in the KW basin and as a fuel assembly is raised from the basin in a shipping cask for transportation to the 327 Building for examination as part of the SNF Characterization Program. The postulated accidents reviews in this report are considered to bound all potential releases of radioactivity and hydrogen. These postulated accidents are: (1) opening and refill of a cylinder containing significant amounts of hydrogen and uranium hydride; and (2) draining of the single element can be used to keep the fuel element submerged in water after the cask containing the can and element is lifted from the KW Basin. Analysis shows the release of radioactivity to the site boundary is significantly less than that allowed by the K Basin Safety Evaluation. Analysis further shows there would be no damage to the K Basin structure nor would there be injury to personnel for credible events.

SOURCE OF BURNUP VALUES FOR COMMERCIAL SPENT NUCLEAR FUEL ASSEMBLIES

International Nuclear Information System (INIS)

BSC

2004-01-01

Waste packages are loaded with commercial spent nuclear fuel (SNF) that satisfies the minimum burnup requirements of a criticality loading curve. The burnup value assigned by the originating nuclear utility to each SNF assembly (assigned burnup) is used to load waste packages in compliance with a criticality loading curve. The burnup provided by a nuclear utility has uncertainties, so conservative calculation methods are used to characterize those uncertainties for incorporation into the criticality loading curves. Procedural safety controls ensure that the correct assembly is loaded into each waste package to prevent a misload that could create a condition affecting the safety margins. Probabilistic analyses show that procedural safety controls can minimize the chance of a misload but can not completely eliminate the possibility. Physical measurements of burnup with instrumentation in the surface facility are not necessary due to the conservative calculation methods used to produce the criticality loading curves. The reactor records assigned burnup of a commercial SNF assembly contains about two percent uncertainty, which is increased to five-percent to ensure conservatism. This five-percent uncertainty is accommodated by adjusting the criticality loading curve. Also, the record keeping methods of nuclear utilities are not uniform and the level of detail required by the NRC has varied over the last several decades. Thus, some SNF assemblies may have assigned burnups that are averages for a batch of assemblies with similar characteristics. Utilities typically have access to more detailed core-follow records that allow the batch average burnup to be changed to an assembly specific burnup. Alternatively, an additional safety margin is incorporated into the criticality loading curve to accommodate SNF assemblies with batch average burnups or greater uncertainties due to the methodology used by the nuclear utility. The utility records provide the assembly identifier

On the market of wind with hydro-pumped storage systems in autonomous Greek islands

International Nuclear Information System (INIS)

Caralis, G.; Zervos, A.; Rados, K.

2010-01-01

In autonomous islands, the wind penetration is restricted due to technical reasons related with the safe operation of the electrical systems. The combined use of wind energy with pumped storage (WPS) is considered as a mean to exploit the abundant wind potential, increase the wind installed capacity and substitute conventional peak supply. In this paper, the experience gained from the analysis of WPS in three specific islands is used towards the estimation of the WPS market in autonomous Greek islands. Parameterized diagrams and a methodology towards the pre-dimensioning and initial design of the WPS are proposed and used towards the estimation of the market in autonomous Greek islands. The objective is to make an initial general prefeasibility study of WPS prospects in the autonomous Greek islands. Results show that there is a significant market for WPS in Greece and the development cost of WPS is competitive to the fuel cost of local power stations in autonomous islands. (author)

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

Improved thermal source term generation capability for use in performance assessment and system studies

International Nuclear Information System (INIS)

King, J.; Rhodes, C.

1994-01-01

This paper describes work performed by the Civilian Radioactive Waste Management System (CRWMS) Management and Operating (M ampersand O) Contractor to improve spent nuclear fuel (SNF) waste stream characterization for system studies. It discusses how these new capabilities may be exploited for thermal source term generation for use in repository performance assessment modeling. SNF historical discharges have been exhaustively tracked, and significant effort has gone into capturing, verifying, and electronically managing spent fuel inventory data. Future discharge projections are produced annually by the Energy Information Administration (EIA) using sophisticated computer models. The output of these models is coupled with annually updated SNF historical discharges to produce what is referred to as the open-quotes reactor database.close quotes This database and related data are published in a variety of ways including on magnetic media for consistent use by analysts or other interested parties

Evaluation of Codisposal Viability for TH/U Carbide (Fort Saint Vrain HTGR) DOE-Owned Fuel

International Nuclear Information System (INIS)

Radulescu, H.

2001-01-01

There are more than 250 forms of US Department of Energy (DOE)-owned spent nuclear fuel (SNF). Due to the variety of the spent nuclear fuel, the National Spent Nuclear Fuel Program has designated nine representative fuel groups for disposal criticality analyses based on fuel matrix, primary fissile isotope, and enrichment. The Fort Saint Vrain reactor (FSVR) SNF has been designated as the representative fuel for the Th/U carbide fuel group. The FSVR SNF consists of small particles (spheres of the order of 0.5-mm diameter) of thorium carbide or thorium and high-enriched uranium carbide mixture, coated with multiple, thin layers of pyrolytic carbon and silicon carbide, which serve as miniature pressure vessels to contain fission products and the U/Th carbide matrix. The coated particles are bound in a carbonized matrix, which forms fuel rods or ''compacts'' that are loaded into large hexagonal graphite prisms. The graphite prisms (or blocks) are the physical forms that are handled in reactor loading and unloading operations, and which will be loaded into the DOE standardized SNF canisters. The results of the analyses performed will be used to develop waste acceptance criteria. The items that are important to criticality control are identified based on the analysis needs and result sensitivities. Prior to acceptance to fuel from the Th/U carbide fuel group for disposal, the important items for the fuel types that are being considered for disposal under the Th/U carbide fuel group must be demonstrated to satisfy the conditions determined in this report

Evaluation of Codisposal Viability for TH/U Carbide (Fort Saint Vrain HTGR) DOE-Owned Fuel

Energy Technology Data Exchange (ETDEWEB)

H. radulescu

2001-09-28

There are more than 250 forms of US Department of Energy (DOE)-owned spent nuclear fuel (SNF). Due to the variety of the spent nuclear fuel, the National Spent Nuclear Fuel Program has designated nine representative fuel groups for disposal criticality analyses based on fuel matrix, primary fissile isotope, and enrichment. The Fort Saint Vrain reactor (FSVR) SNF has been designated as the representative fuel for the Th/U carbide fuel group. The FSVR SNF consists of small particles (spheres of the order of 0.5-mm diameter) of thorium carbide or thorium and high-enriched uranium carbide mixture, coated with multiple, thin layers of pyrolytic carbon and silicon carbide, which serve as miniature pressure vessels to contain fission products and the U/Th carbide matrix. The coated particles are bound in a carbonized matrix, which forms fuel rods or ''compacts'' that are loaded into large hexagonal graphite prisms. The graphite prisms (or blocks) are the physical forms that are handled in reactor loading and unloading operations, and which will be loaded into the DOE standardized SNF canisters. The results of the analyses performed will be used to develop waste acceptance criteria. The items that are important to criticality control are identified based on the analysis needs and result sensitivities. Prior to acceptance to fuel from the Th/U carbide fuel group for disposal, the important items for the fuel types that are being considered for disposal under the Th/U carbide fuel group must be demonstrated to satisfy the conditions determined in this report.

Equilibrium transuranic management scheme for diverse fuel cycle analysis

International Nuclear Information System (INIS)

Haas, Jason; Lee, John C.

2008-01-01

A key issue cited in the U.S. Department of Energy's report to Congress (2003) on the research path for the Advanced Fuel Cycle Initiative (AFCI) is an accurate estimation of life cycle costs for the construction, operation, decontamination and decommissioning of all fuel cycle facilities. In this report we discuss the methodology and validation of a fuel cycle model based on equilibrium operation. We apply our model to a diverse set of advanced reactors and fuel types in order to determine the most effective transmuting system while simultaneously minimizing fuel cycle costs. Our analysis shows that a nearly instant equilibrium modeling of fuel cycle scenarios can accurately approximate the detailed complex dynamic models developed by national laboratories. Our analysis also shows that the cost of transmuting Spent Nuclear Fuel (SNF) from a UO 2 fueled Pressurized Water Reactor (PWR) is minimized by utilizing the thorium cycle in sodium cooled fast reactors and is near the cost for long term repository storage of SNF at Yucca Mountain. (authors)

Treatment of High-Level Waste Arising from Pyrochemical Processes

International Nuclear Information System (INIS)

Lizin, A.A.; Kormilitsyn, M.V.; Osipenko, A.G.; Tomilin, S.V.; Lavrinovich, Yu.G.

2013-01-01

JSC “SSC RIAR” has been performing research and development activities in support of closed fuel cycle of fast reactor since the middle of 1960s. Fuel cycle involves fabrication and reprocessing of spent nuclear fuel (SNF) using pyrochemical methods of reprocessing in molten alkali metal chlorides. At present pyrochemical methods of SNF reprocessing in molten chlorides has reached such a level in their development that makes it possible to compare their competitiveness with classic aqueous methods. Their comparative advantage lies in high safety, compactness, high protectability as to nonproliferation of nuclear materials, and reduction of high level waste volume

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.

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.

MANAGING SPENT NUCLEAR FUEL WASTES AT THE IDAHO NATIONAL LABORATORY

Energy Technology Data Exchange (ETDEWEB)

Hill, Thomas J

2005-09-01

The Idaho National Engineering Laboratory (INL) has a large inventory of diverse types of spent nuclear fuel (SNF). This legacy is in part due to the history of the INL as the National Reactor Testing Station, in part to its mission to recover highly enriched uranium from SNF and in part to it’s mission to test and examine SNF after irradiation. The INL also has a large diversity of SNF storage facility, some dating back 50 years in the site history. The success of the INL SNF program is measured by its ability to: 1) achieve safe existing storage, 2) continue to receive SNF from other locations, both foreign and domestic, 3) repackage SNF from wet storage to interim dry storage, and 4) prepare the SNF for dispositioning in a federal repository. Because of the diversity in the SNF and the facilities at the INL, the INL is addressing almost very condition that may exist in the SNF world. Many of solutions developed by the INL are applicable to other SNF storage sites as they develop their management strategy. The SNF being managed by the INL are in a variety of conditions, from intact assemblies to individual rods or plates to powders, rubble, and metallurgical mounts. Some of the fuel has been in wet storage for over forty years. The fuel is stored bare, or in metal cans and either wet under water or dry in vaults, caissons or casks. Inspections have shown varying degrees of corrosion and degradation of the fuel and the storage cans. Some of the fuel has been recanned under water, and the conditions of the fuel inside the second or third can are unknown. The fuel has been stored in one of 10 different facilities: five wet pools and one casks storage pad, one vault, two generations of caisson facilities, and one modular Independent Spent Fuel Storage Installation (ISFSI). The wet pools range from forty years old to the most modern pool in the US Department of Energy (DOE) complex. The near-term objective is moving the fuel in the older wet storage facilities to

Annex D-200 Area Interim Storage Area Final Safety Analysis Report [FSAR] [Section 1 and 2

International Nuclear Information System (INIS)

CARRELL, R.D.

2002-01-01

The 200 Area Interim Storage Area (200 Area ISA) at the Hanford Site provides for the interim storage of non-defense reactor spent nuclear fuel (SNF) housed in aboveground dry cask storage systems. The 200 Area ISA is a relatively simple facility consisting of a boundary fence with gates, perimeter lighting, and concrete and gravel pads on which to place the dry storage casks. The fence supports safeguards and security and establishes a radiation protection buffer zone. The 200 Area ISA is nominally 200,000 ft 2 and is located west of the Canister Storage Building (CSB). Interim storage at the 200 Area ISA is intended for a period of up to 40 years until the materials are shipped off-site to a disposal facility. This Final Safety Analysis Report (FSAR) does not address removal from storage or shipment from the 200 Area ISA. Three different SNF types contained in three different dry cask storage systems are to be stored at the 200 Area ISA, as follows: (1) Fast Flux Test Facility Fuel--Fifty-three interim storage casks (ISC), each holding a core component container (CCC), will be used to store the Fast Flux Test Facility (FFTF) SNF currently in the 400 Area. (2) Neutron Radiography Facility (NRF) TRIGA'--One Rad-Vault' container will store two DOT-6M3 containers and six NRF TRIGA casks currently stored in the 400 Area. (3) Commercial Light Water Reactor Fuel--Six International Standards Organization (ISO) containers, each holding a NAC-I cask4 with an inner commercial light water reactor (LWR) canister, will be used for commercial LWR SNF from the 300 Area. An aboveground dry cask storage location is necessary for the spent fuel because the current storage facilities are being shut down and deactivated. The spent fuel is being transferred to interim storage because there is no permanent repository storage currently available

Annex D-200 Area Interim Storage Area Final Safety Analysis Report [FSAR] [Section 1 & 2

Energy Technology Data Exchange (ETDEWEB)

CARRELL, R D

2002-07-16

The 200 Area Interim Storage Area (200 Area ISA) at the Hanford Site provides for the interim storage of non-defense reactor spent nuclear fuel (SNF) housed in aboveground dry cask storage systems. The 200 Area ISA is a relatively simple facility consisting of a boundary fence with gates, perimeter lighting, and concrete and gravel pads on which to place the dry storage casks. The fence supports safeguards and security and establishes a radiation protection buffer zone. The 200 Area ISA is nominally 200,000 ft{sup 2} and is located west of the Canister Storage Building (CSB). Interim storage at the 200 Area ISA is intended for a period of up to 40 years until the materials are shipped off-site to a disposal facility. This Final Safety Analysis Report (FSAR) does not address removal from storage or shipment from the 200 Area ISA. Three different SNF types contained in three different dry cask storage systems are to be stored at the 200 Area ISA, as follows: (1) Fast Flux Test Facility Fuel--Fifty-three interim storage casks (ISC), each holding a core component container (CCC), will be used to store the Fast Flux Test Facility (FFTF) SNF currently in the 400 Area. (2) Neutron Radiography Facility (NRF) TRIGA'--One Rad-Vault' container will store two DOT-6M3 containers and six NRF TRIGA casks currently stored in the 400 Area. (3) Commercial Light Water Reactor Fuel--Six International Standards Organization (ISO) containers, each holding a NAC-I cask4 with an inner commercial light water reactor (LWR) canister, will be used for commercial LWR SNF from the 300 Area. An aboveground dry cask storage location is necessary for the spent fuel because the current storage facilities are being shut down and deactivated. The spent fuel is being transferred to interim storage because there is no permanent repository storage currently available.

Nevada commercial spent nuclear fuel transportation experience

International Nuclear Information System (INIS)

1991-09-01

The purpose of this report is to present an historic overview of commercial reactor spent nuclear fuel (SNF) shipments that have occurred in the state of Nevada, and to review the accident and incident experience for this type of shipments. Results show that between 1964 and 1990, 309 truck shipments covering approximately 40,000 miles moved through Nevada; this level of activity places Nevada tenth among the states in the number of truck shipments of SNF. For the same period, 15 rail shipments moving through the State covered approximately 6,500 miles, making Nevada 20th among the states in terms of number of rail shipments. None of these shipments had an accident or an incident associated with them. Because the data for Nevada are so limited, national data on SNF transportation and the safety of truck and rail transportation in general were also assessed

Management Of Hanford KW Basin Knockout Pot Sludge As Spent Nuclear Fuel

International Nuclear Information System (INIS)

Raymond, R. E.; Evans, K. M.

2012-01-01

CH2M Hill Plateau Remediation Company (CHPRC) and AREVA Federal Services, LLC (AFS) have been working collaboratively to develop and deploy technologies to remove, transport, and interim store remote-handled sludge from the 10S-K West Reactor Fuel Storage Basin on the U.S. Department of Energy (DOE) Hanford Site near Richland, WA, USA. Two disposal paths exist for the different types of sludge found in the K West (KW) Basin. One path is to be managed as Spent Nuclear Fuel (SNF) with eventual disposal at an SNF at a yet to be licensed repository. The second path will be disposed as remote-handled transuranic (RH-TRU) waste at the Waste Isolation Pilot Plant (WIPP) in Carlsbad, NM. This paper describes the systems developed and executed by the Knockout Pot (KOP) Disposition Subproject for processing and interim storage of the sludge managed as SNF, (i.e., KOP material)

Actinide-only and full burn-up credit in criticality assessment of RBMK-1500 spent nuclear fuel storage cask using axial burn-up profile

Energy Technology Data Exchange (ETDEWEB)

Barkauskas, V., E-mail: vytenis.barkauskas@ftmc.lt; Plukiene, R., E-mail: rita.plukiene@ftmc.lt; Plukis, A., E-mail: arturas.plukis@ftmc.lt

2016-10-15

Highlights: • RBMK-1500 fuel burn-up impact on k{sub eff} in the SNF cask was calculated using SCALE 6.1. • Positive end effect was noticed at certain burn-up for the RBMK-1500 spent nuclear fuel. • The non-uniform uranium depletion is responsible for the end effect in RBMK-1500 SNF. • k{sub eff} in the SNF cask does not exceed a value of 0.95 which is set in the safety requirements. - Abstract: Safe long-term storage of spent nuclear fuel (SNF) is one of the main issues in the field of nuclear safety. Burn-up credit application in criticality analysis of SNF reduces conservatism of usually used fresh fuel assumption and implies a positive economic impact for the SNF storage. Criticality calculations of spent nuclear fuel in the CONSTOR® RBMK-1500/M2 cask were performed using pre-generated ORIGEN-ARP spent nuclear fuel composition libraries, and the results of the RBMK-1500 burn-up credit impact on the effective neutron multiplication factor (k{sub eff}) have been obtained and are presented in the paper. SCALE 6.1 code package with the STARBUCKS burn-up credit evaluation tool was used for modeling. Pre-generated ARP (Automatic Rapid Processing) crosssection libraries based on ENDF/B-VII cross section library were used for fast burn-up inventory modeling. Different conditions in the SNF cask were modeled: 2.0% and 2.8% initial enrichment fuel of various burn-up and water density inside cavities of the SNF cask. The fuel composition for the criticality analysis was chosen taking into account main actinides and most important fission products used in burn-up calculations. A significant positive end effect is noticed from 15 GWd/tU burn-up for 2.8% enrichment fuel and from 9 GWd/tU for 2.0% enrichment fuel applying the actinide-only approach. The obtained results may be applied in further evaluations of the RBMK type reactor SNF storage as well as help to optimize the SNF storage volume inside the CONSTOR® RBMK-1500/M2 cask without compromising criticality

Actinide-only and full burn-up credit in criticality assessment of RBMK-1500 spent nuclear fuel storage cask using axial burn-up profile

International Nuclear Information System (INIS)

Barkauskas, V.; Plukiene, R.; Plukis, A.

2016-01-01

Highlights: • RBMK-1500 fuel burn-up impact on k_e_f_f in the SNF cask was calculated using SCALE 6.1. • Positive end effect was noticed at certain burn-up for the RBMK-1500 spent nuclear fuel. • The non-uniform uranium depletion is responsible for the end effect in RBMK-1500 SNF. • k_e_f_f in the SNF cask does not exceed a value of 0.95 which is set in the safety requirements. - Abstract: Safe long-term storage of spent nuclear fuel (SNF) is one of the main issues in the field of nuclear safety. Burn-up credit application in criticality analysis of SNF reduces conservatism of usually used fresh fuel assumption and implies a positive economic impact for the SNF storage. Criticality calculations of spent nuclear fuel in the CONSTOR® RBMK-1500/M2 cask were performed using pre-generated ORIGEN-ARP spent nuclear fuel composition libraries, and the results of the RBMK-1500 burn-up credit impact on the effective neutron multiplication factor (k_e_f_f) have been obtained and are presented in the paper. SCALE 6.1 code package with the STARBUCKS burn-up credit evaluation tool was used for modeling. Pre-generated ARP (Automatic Rapid Processing) crosssection libraries based on ENDF/B-VII cross section library were used for fast burn-up inventory modeling. Different conditions in the SNF cask were modeled: 2.0% and 2.8% initial enrichment fuel of various burn-up and water density inside cavities of the SNF cask. The fuel composition for the criticality analysis was chosen taking into account main actinides and most important fission products used in burn-up calculations. A significant positive end effect is noticed from 15 GWd/tU burn-up for 2.8% enrichment fuel and from 9 GWd/tU for 2.0% enrichment fuel applying the actinide-only approach. The obtained results may be applied in further evaluations of the RBMK type reactor SNF storage as well as help to optimize the SNF storage volume inside the CONSTOR® RBMK-1500/M2 cask without compromising criticality safety.

76 FR 35137 - Vulnerability and Threat Information for Facilities Storing Spent Nuclear Fuel and High-Level...

Science.gov (United States)

2011-06-16

... High-Level Radioactive Waste AGENCY: U.S. Nuclear Regulatory Commission. ACTION: Public meeting... Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater Than Class C Waste,'' and 73... Spent Nuclear Fuel (SNF) and High-Level Radioactive Waste (HLW) storage facilities. The draft regulatory...

The Question of Queue: Implications for -Best Practice- in Cross-country Transport of Commercial Spent Nuclear Fuel

International Nuclear Information System (INIS)

Williams, J.M.

2009-01-01

The 'Standard Contract' authorized by the Nuclear Waste Policy Act of 1982 (Section 302(a)) provides that priority for acceptance of spent nuclear fuel (SNF) shall be based on the date of its discharge from civilian nuclear reactors. Through 2007, about 2,100 discharges of about 58,000 metric tons have created a priority ranking (or 'queue') for US DOE spent fuel acceptance and transport. Since 1982, consideration of the task of large-scale, cross-country SNF transport (by the National Academies and others) has led to several recommendations for 'best practice' in such an unprecedented campaign. Many of these recommendations, however, are inconsistent with the acceptance priority established by the Standard Contract, and in fact cannot be implemented under its provisions. This paper considers the SNF acceptance rankings established by the Standard Contract, and the barrier these place on best practice cross-country transport of the nation's inventory of SNF. Using a series of case studies, the paper explores the challenge of best practice transport from selected shipment origins under current arrangements. The case studies support preliminary conclusions regarding the inconsistency between best practice SNF transport and the Standard Contract acceptance queue, with reference to particular origins sites and their utility owners. The paper concludes with a suggestion for resolving the inconsistencies, and recommended next steps in the inquiry. (authors)

Exercise Training Mitigates Water Pipe Smoke Exposure-Induced Pulmonary Impairment via Inhibiting NF-κB and Activating Nrf2 Signalling Pathways

Directory of Open Access Journals (Sweden)

Abderrahim Nemmar

2018-01-01

Full Text Available Water pipe smoking is a tobacco smoking method commonly used in Eastern countries and is gaining popularity in Europe and North America, in particular among adolescents and young adults. Several clinical and experimental studies have reported that exposure to water pipe smoke (WPS induces lung inflammation and impairment of pulmonary function. However, the mechanisms of such effects are not understood, as are data on the possible palliative effect of exercise training. The present study evaluated the effects of regular aerobic exercise training (treadmill: 5 days/week, 40 min/day on subchronic exposure to WPS (30 minutes/day, 5 days/week for 2 months. C57BL/6 mice were exposed to air or WPS with or without exercise training. Airway resistance measured using forced oscillation technique was significantly and dose-dependently increased in the WPS-exposed group when compared with the air-exposed one. Exercise training significantly prevented the effect of WPS on airway resistance. Histologically, the lungs of WPS-exposed mice had focal moderate interstitial inflammatory cell infiltration consisting of neutrophil polymorphs, plasma cells, and lymphocytes. There was a mild increase in intra-alveolar macrophages and a focal damage to alveolar septae in some foci. Exercise training significantly alleviated these effects and also decreased the WPS-induced increase of tumor necrosis factor α and interleukin 6 concentrations and attenuated the increase of 8-isoprostane in lung homogenates. Likewise, the lung DNA damage induced by WPS was significantly inhibited by exercise training. Moreover, exercise training inhibited nuclear factor kappa-B (NF-κB expression induced by WPS and increased that of nuclear factor erythroid 2-related factor 2 (Nrf2. Our findings suggest that exercise training significantly mitigated WPS-induced increase in airway resistance, inflammation, oxidative stress, and DNA damage via mechanisms that include inhibiting NF-κB and

Sampling and analysis plan for the preoperational environmental survey of the spent nuclear fuel project facilities

International Nuclear Information System (INIS)

MITCHELL, R.M.

1999-01-01

This sampling and analysis plan will support the preoperational environmental monitoring for construction, development, and operation of the Spent Nuclear Fuel (SNF) Project facilities, which have been designed for the conditioning and storage of spent nuclear fuels; particularly the fuel elements associated with the operation of N-Reactor. The SNF consists principally of irradiated metallic uranium, and therefore includes plutonium and mixed fission products. The primary effort will consist of removing the SNF from the storage basins in K East and K West Areas, placing in multicanister overpacks, vacuum drying, conditioning, and subsequent dry vault storage in the 200 East Area. The primary purpose and need for this action is to reduce the risks to public health and safety and to the environment. Specifically these include prevention of the release of radioactive materials into the air or to the soil surrounding the K Basins, prevention of the potential migration of radionuclides through the soil column to the nearby Columbia River, reduction of occupational radiation exposure, and elimination of the risks to the public and to workers from the deterioration of SNF in the K Basins

Rethinking the economics of centralized spent fuel storage

International Nuclear Information System (INIS)

Wood, T.W.; Short, S.M.; Dippold, D.G.; Rod, S.R.; Williams, J.W.

1991-04-01

The technology for extended storage of spent nuclear fuel (SNF), either at-reactor or in a centralized facility such as a monitored retrievable storage (MRS) facility, is well-developed and proven from an engineering and safety perspective. The question of whether spent fuel should await its final geologic disposal while at a reactor site or in an MRS facility is essentially an economic one. While intuition and previous results suggest that centralized storage will be more economical than at-reactor storage beyond some break-even quantity of SNF, the incremental costs of pool storage at-reactor are close to zero as long as pool capacity is generally available. Thus, if economics is the prime motivator, the quantity of spent fuel required to warrant centralized storage could be quite large. The economics of centralizing the storage of spent fuel at a single site, as opposed to continued storage at over 100 reactor sites, has been the subject of several recent analyses. Most of these analyses involved calculating the benefits of an MRS facility with a pre-defined MRS operating scenario. This paper reverses this approach to economic analysis of the MRS by seeking the optimal MRS operating scenario (in terms of the parameters listed above) implied by the economic incentives arising from the relative costs of at-reactor storage and centralized storage. This approach treats an MRS as a possible storage location that will be used according to its economic value in system operation. 5 refs., 5 figs

Review of Technical Studies in the United States in Support of Burnup Credit Regulatory Guidance

International Nuclear Information System (INIS)

Wagner, John C.; Parks, Cecil V.; Mueller, Don; Gauld, Ian C.

2010-01-01

Taking credit for the reduction in reactivity associated with fuel depletion can enable more cost-effective, higher-density storage, transport, disposal, and reprocessing of spent nuclear fuel (SNF) while maintaining sufficient subcritical margin to establish an adequate safety basis. Consequently, there continues to be considerable interest in the United States (U.S.), as well as internationally, in the increased use of burnup credit in SNF operations, particularly related to storage, transport, and disposal of commercial SNF. This interest has motivated numerous technical studies related to the application of burnup credit, both domestically and internationally, as well as the design of SNF storage, transport and disposal systems that rely on burnup credit for maintaining subcriticality. Responding to industry requests and needs, the U.S. Nuclear Regulatory Commission (NRC) initiated a burnup credit research program in 1999, with support from the Oak Ridge National Laboratory (ORNL), to develop regulatory guidance and the supporting technical bases for allowing and expanding the use of burnup credit in pressurized-water reactor SNF storage and transport applications. Although this NRC research program has not been continuous since its inception, considerable progress has been achieved in many key areas in terms of increased understanding of relevant phenomena and issues, availability of relevant information and data, and subsequently updated regulatory guidance for expanded use of burnup credit. This paper reviews technical studies performed by ORNL for the U.S. NRC burnup credit research program. Examples of topics include reactivity effects associated with reactor operating characteristics, fuel assembly characteristics, burnable absorbers, control rods, spatial burnup distributions, cooling time, and assembly misloading; methods and data for validation of isotopic composition predictions; methods and data for validation of criticality calculations; and

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

Spent Nuclear Fuel Project FY 1996 Multi-Year Program Plan WBS No. 1.4.1, Revision 1

International Nuclear Information System (INIS)

1995-09-01

This document describes the Spent Nuclear Fuel (SNF) Project portion of the Hanford Strategic Plan for the Hanford Reservation in Richland, Washington. The SNF Project was established to evaluate and integrate the urgent risks associated with N-reactor fuel currently stored at the Hanford site in the K Basins, and to manage the transfer and disposition of other spent nuclear fuels currently stored on the Hanford site. An evaluation of alternatives for the expedited removal of spent fuels from the K Basin area was performed. Based on this study, a Recommended Path Forward for the K Basins was developed and proposed to the U.S. DOE

Spent Nuclear Fuel Project FY 1996 Multi-Year Program Plan WBS No. 1.4.1, Revision 1

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-09-01

This document describes the Spent Nuclear Fuel (SNF) Project portion of the Hanford Strategic Plan for the Hanford Reservation in Richland, Washington. The SNF Project was established to evaluate and integrate the urgent risks associated with N-reactor fuel currently stored at the Hanford site in the K Basins, and to manage the transfer and disposition of other spent nuclear fuels currently stored on the Hanford site. An evaluation of alternatives for the expedited removal of spent fuels from the K Basin area was performed. Based on this study, a Recommended Path Forward for the K Basins was developed and proposed to the U.S. DOE.

Extraction and antioxidant activities assay of polysaccharides from white Hyacinth bean and promoting-growth to probiotics

Directory of Open Access Journals (Sweden)

Ni LEI

2016-12-01

Full Text Available The extraction parameters of water-soluble polysaccharides (WPs from white hyacinth bean were optimized using single-factor and orthogonal experiment. The antioxidant activities of WPs were presented by assaying three different radicals, 2,2-diphenyl-1-picrylhydrazy radical (DPPH radical, hydroxyl radical and superoxide radical scavenging ability. In addition, the effects of WPs obtained on the growth of three probiotic strains (Lactobacillus acidophilus LA5, Bifidobacterium bifidum BB01 and Lactobacillus bulgaricus LB6 were also determined by measuring the OD and pH value of culture medium. According to the results, the optimum extraction parameters were as follows: the ratio of water to material was 50, extraction time was 2h and the extraction temperature was 95°C. The yield of WPs reached 1.15±0.07% under this condition. In addition, the WPs had different scavenging ability on three radicals (hydroxyl > DPPH > superoxide. And the WPs could promote the growth of LA5, BB01 and LB6.

Determination of uncertainties of PWR spent fuel radionuclide inventory based on real operational history data

Energy Technology Data Exchange (ETDEWEB)

Fast, Ivan; Bosbach, Dirk [Institute of Energy- and Climate Research, Nuclear Waste Management and Reactor Safety Research, IEK-6, Forschungszentrum, Julich GmbH, (Germany); Aksyutina, Yuliya; Tietze-Jaensch, Holger [German Product Control Office for Radioactive Waste (PKS), Institute of Energy- and Climate Research, Nuclear Waste Management and Reactor Safety Research, IEK-6, Forschungszentrum Julich GmbH, (Germany)

2015-07-01

A requisite for the official approval of the safe final disposal of SNF is a comprehensive specification and declaration of the nuclear inventory in SNF by the waste supplier. In the verification process both the values of the radionuclide (RN) activities and their uncertainties are required. Burn-up (BU) calculations based on typical and generic reactor operational parameters do not encompass any possible uncertainties observed in real reactor operations. At the same time, the details of the irradiation history are often not well known, which complicates the assessment of declared RN inventories. Here, we have compiled a set of burnup calculations accounting for the operational history of 339 published or anonymized real PWR fuel assemblies (FA). These histories were used as a basis for a 'SRP analysis', to provide information about the range of the values of the associated secondary reactor parameters (SRP's). Hence, we can calculate the realistic variation or spectrum of RN inventories. SCALE 6.1 has been employed for the burn-up calculations. The results have been validated using experimental data from the online database - SFCOMPO-1 and -2. (authors)

Determination of uncertainties of PWR spent fuel radionuclide inventory based on real operational history data

International Nuclear Information System (INIS)

Fast, Ivan; Bosbach, Dirk; Aksyutina, Yuliya; Tietze-Jaensch, Holger

2015-01-01

A requisite for the official approval of the safe final disposal of SNF is a comprehensive specification and declaration of the nuclear inventory in SNF by the waste supplier. In the verification process both the values of the radionuclide (RN) activities and their uncertainties are required. Burn-up (BU) calculations based on typical and generic reactor operational parameters do not encompass any possible uncertainties observed in real reactor operations. At the same time, the details of the irradiation history are often not well known, which complicates the assessment of declared RN inventories. Here, we have compiled a set of burnup calculations accounting for the operational history of 339 published or anonymized real PWR fuel assemblies (FA). These histories were used as a basis for a 'SRP analysis', to provide information about the range of the values of the associated secondary reactor parameters (SRP's). Hence, we can calculate the realistic variation or spectrum of RN inventories. SCALE 6.1 has been employed for the burn-up calculations. The results have been validated using experimental data from the online database - SFCOMPO-1 and -2. (authors)

Extending the foreign spent fuel acceptance program: Policy and implementation issues

International Nuclear Information System (INIS)

Lyman, Edwin S.

2005-01-01

The May 2006 expiration date of the Foreign Research Reactor Spent Nuclear Fuel (FRR SNF) Acceptance Program is fast approaching. In April 2004, Energy Secretary Spencer Abraham instructed the Energy Department to 'initiate actions necessary to extend .... the fuel acceptance deadline'. However, extending the deadline may not be a simple task. The limits on the original program resulted from a delicate negotiation among many stakeholders. Any proposal to increase the duration and scope of the program will have to be considered in the context of DOE's failure since 1996 to develop viable treatment, packaging and long-term disposal options for FRR SNF. It is also unclear whether accepting additional low-enriched uranium FRR SNF can be justified on security grounds. This paper will propose criteria for acceptance of spent fuel under an extension that are intended to minimize controversy and ensure consistency with a threat-based prioritization of homeland security expenditures. (author)

Nuclear fuels

International Nuclear Information System (INIS)

Gangwani, Saloni; Chakrabortty, Sumita

2011-01-01

Nuclear fuel is a material that can be consumed to derive nuclear energy, by analogy to chemical fuel that is burned for energy. Nuclear fuels are the most dense sources of energy available. Nuclear fuel in a nuclear fuel cycle can refer to the fuel itself, or to physical objects (for example bundles composed of fuel rods) composed of the fuel material, mixed with structural, neutron moderating, or neutron reflecting materials. Long-lived radioactive waste from the back end of the fuel cycle is especially relevant when designing a complete waste management plan for SNF. When looking at long-term radioactive decay, the actinides in the SNF have a significant influence due to their characteristically long half-lives. Depending on what a nuclear reactor is fueled with, the actinide composition in the SNF will be different. The following paper will also include the uses. advancements, advantages, disadvantages, various processes and behavior of nuclear fuels

Standard guide for evaluation of materials used in extended service of interim spent nuclear fuel dry storage systems

CERN Document Server

American Society for Testing and Materials. Philadelphia

2010-01-01

1.1 Part of the total inventory of commercial spent nuclear fuel (SNF) is stored in dry cask storage systems (DCSS) under licenses granted by the U.S. Nuclear Regulatory Commission (NRC). The purpose of this guide is to provide information to assist in supporting the renewal of these licenses, safely and without removal of the SNF from its licensed confinement, for periods beyond those governed by the term of the original license. This guide provides information on materials behavior under conditions that may be important to safety evaluations for the extended service of the renewal period. This guide is written for DCSS containing light water reactor (LWR) fuel that is clad in zirconium alloy material and stored in accordance with the Code of Federal Regulations (CFR), at an independent spent-fuel storage installation (ISFSI). The components of an ISFSI, addressed in this document, include the commercial SNF, canister, cask, and all parts of the storage installation including the ISFSI pad. The language of t...

Analysis of scenarios of the inclusion of fast reactors in the nuclear power of Russia in the context of sustainable development with the use of the INPRO methodology

International Nuclear Information System (INIS)

Usanov, V.I.; Kagramanyan, V.S.; Kalashnikov, A.G.; Korobeinikov, V.V.; Korobitsyn, V.E.; Moseyev, A.L.; Poplavskaya, E.V.

2013-01-01

Conclusions: • The two-component NES of VVER and BN reactors can meet some critical challenges of the present nuclear industry and provide a substantial contribution to enhancing sustainability of a national NP: – basically to solve up to 2050 the problem of the VVER SNF accumulation by using Pu from VVER in MOX fuel for BN reactors; – to ensure management of Pu from VVER to reduce it by 2070 to operational reserve and thus to enhance the NES proliferation resistance; – to save natural U and SWU and thus to facilitate U supply and enrichment capacities for planed deployment of VVERs in Russia and abroad. • Implementation of these opportunities might be a substance of the first phase of the NFC closure • While some INPRO indicators have shown remarkable advantages of the NES with BNs comparing to the present system, some issues in economics and NFC technologies have not got convincing answers. • These challenges along with a crucial safety issues are addressed in the Federal target programmes on transition to a CNFC with advanced FRs which are currently run in Russia

Long-Term Dry Storage of High Burn-Up Spent Pressurized Water Reactor (PWR) Fuel in TAD (Transportation, Aging, and Disposal) Containers

International Nuclear Information System (INIS)

Hwang, Yong Soo

2008-12-01

A TAD canister, in conjunction with specially-designed over-packs can accomplish the functions of transportation, aging, and disposal (TAD) in the management of spent nuclear fuel (SNF). Industrial dry cask systems currently available for SNF are licensed for storage-only or for dual-purpose (i.e., storage and transportation). By extending the function to include the indefinite storage and perhaps, eventual geologic disposal, the TAD canister would have to be designed to enhance, among others, corrosion resistance, thermal stability, and criticality-safety control. This investigative paper introduces the use of these advanced iron-based, corrosion-resistant materials for SNF transportation, aging, and disposal.The objective of this investigative project is to explore the interest that KAERI would research and develop its specific SAM coating materials for the TAD canisters to satisfy the requirements of corrosion-resistance, thermal stability, and criticality-controls for long-term dry storage of high burn-up spent PWR fuel

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

Enriching the Web Processing Service

Science.gov (United States)

Wosniok, Christoph; Bensmann, Felix; Wössner, Roman; Kohlus, Jörn; Roosmann, Rainer; Heidmann, Carsten; Lehfeldt, Rainer

2014-05-01

The OGC Web Processing Service (WPS) provides a standard for implementing geospatial processes in service-oriented networks. In its current version 1.0.0 it allocates the operations GetCapabilities, DescribeProcess and Execute, which can be used to offer custom processes based on single or multiple sub-processes. A large range of ready to use fine granular, fundamental geospatial processes have been developed by the GIS-community in the past. However, modern use cases or whole workflow processes demand specifications of lifecycle management and service orchestration. Orchestrating smaller sub-processes is a task towards interoperability; a comprehensive documentation by using appropriate metadata is also required. Though different approaches were tested in the past, developing complex WPS applications still requires programming skills, knowledge about software libraries in use and a lot of effort for integration. Our toolset RichWPS aims at providing a better overall experience by setting up two major components. The RichWPS ModelBuilder enables the graphics-aided design of workflow processes based on existing local and distributed processes and geospatial services. Once tested by the RichWPS Server, a composition can be deployed for production use on the RichWPS Server. The ModelBuilder obtains necessary processes and services from a directory service, the RichWPS semantic proxy. It manages the lifecycle and is able to visualize results and debugging-information. One aim will be to generate reproducible results; the workflow should be documented by metadata that can be integrated in Spatial Data Infrastructures. The RichWPS Server provides a set of interfaces to the ModelBuilder for, among others, testing composed workflow sequences, estimating their performance and to publish them as common processes. Therefore the server is oriented towards the upcoming WPS 2.0 standard and its ability to transactionally deploy and undeploy processes making use of a WPS

COMPLETION OF THE FIRST INTEGRATED SPENT NUCLEAR FUEL TRANSSHIPMENT/INTERIM STORAGE FACILITY IN NW RUSSIA

International Nuclear Information System (INIS)

Dyer, R.S.; Barnes, E.; Snipes, R.L.; Hoeibraaten, S.; Gran, H.C.; Foshaug, E.; Godunov, V.

2003-01-01

Northwest and Far East Russia contain large quantities of unsecured spent nuclear fuel (SNF) from decommissioned submarines that potentially threaten the fragile environments of the surrounding Arctic and North Pacific regions. The majority of the SNF from the Russian Navy, including that from decommissioned nuclear submarines, is currently stored in on-shore and floating storage facilities. Some of the SNF is damaged and stored in an unstable condition. Existing Russian transport infrastructure and reprocessing facilities cannot meet the requirements for moving and reprocessing this amount of fuel. Additional interim storage capacity is required. Most of the existing storage facilities being used in Northwest Russia do not meet health and safety, and physical security requirements. The United States and Norway are currently providing assistance to the Russian Federation (RF) in developing systems for managing these wastes. If these wastes are not properly managed, they could release significant concentrations of radioactivity to these sensitive environments and could become serious global environmental and physical security issues. There are currently three closely-linked trilateral cooperative projects: development of a prototype dual-purpose transport and storage cask for SNF, a cask transshipment interim storage facility, and a fuel drying and cask de-watering system. The prototype cask has been fabricated, successfully tested, and certified. Serial production is now underway in Russia. In addition, the U.S. and Russia are working together to improve the management strategy for nuclear submarine reactor compartments after SNF removal

Accelerator-driven system design concept for disposing of spent nuclear fuels

International Nuclear Information System (INIS)

Gohar, Y.; Cao, Y.; Kellogg, R.; Merzari, E.

2015-01-01

At present, the US SNF (Spent Nuclear Fuel) inventory is growing by about 2,000 metric tonnes (MT) per year from the current operating nuclear power plants to reach about 70,000 MT by 2015. This SNF inventory contains about 1% transuranics (700 MT), which has about 115 MT of minor actinides. Accelerator-driven systems utilising proton accelerators with neutron spallation targets and subcritical blankets can be utilised for transmuting these transuranics, simultaneously generating carbon free energy, and significantly reducing the capacity of the required geological repository storage facility for the spent nuclear fuels. A fraction of the SNF plutonium can be used as a MOX fuel in the current/future thermal power reactors and as a starting fuel for future fast power reactors. The uranium of the spent nuclear fuel can be recycled for use in future nuclear power plants. This paper shows that only four to five accelerator-driven systems operating for less than 33 full power years can dispose of the US SNF inventory expected by 2015. In addition, a significant fraction of the long-lived fission products will be transmuted at the same time. Each system consists of a proton accelerator with a neutron spallation target and a subcritical assembly. The accelerator beam parameters are 1 GeV protons and 25 MW beam power, which produce 3 GWt in the subcritical assembly. A liquid metal (lead or lead-bismuth eutectic) spallation target is selected because of design advantages. This target is located at the centre of the subcritical assembly to maximise the utilisation of spallation neutrons. Because of the high power density in the target material, the target has its own coolant loop, which is independent of the subcritical assembly coolant loop. Mobile fuel forms with transuranic materials without uranium are considered in this work with liquid lead or lead-bismuth eutectic as fuel carrier

Rethinking the economics of centralized spent fuel storage

International Nuclear Information System (INIS)

Wood, T.W.; Short, S.M.; Dippold, D.G.; Rod, S.R.; Williams, J.W.

1991-01-01

The technology for extended storage of spent nuclear fuel (SNF), either at-reactor or in a centralized facility such as a monitored retrievable storage (MRS) facility, is well-developed and proven from an engineering and safety perspective. The question of whether spent fuel should await its final geologic disposal while at a reactor site or in an MRS facility is essentially an economic one. While intuition and previous results suggest that centralized storage will be more economical than at-reactor storage beyond some break-even quantity of SNF, the incremental costs of pool storage at-reactor are close to zero as long as pool capacity is generally available. Thus, if economics is the prime motivator, the quantity of spent fuel required to warrant centralized storage could be quite large. The economics of centralizing the storage of spent fuel at a single site, as opposed to continued storage at over 100 reactor sites, has been the subject of several recent analyses. Most of these analyses involved calculating the benefits of an MRS facility (in terms of avoided utility costs) with a pre-defined MRS operating scenario (e.g., spent fuel acceptance schedule, storage capacity, and typical storage cycle). While these analyses provided some insight into the economic justification for an MRS facility, even the most favorable scenarios resulted in net costs of hundreds of millions of dollars when evaluated on a discounted cash flow basis

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

Quantitative and radiological assessment of PYRO-SFR closed fuel cycle against direct disposal of spent nuclear fuel from Korean nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Khan, Muhammad Minhaj; Lee, Suhong; Cheong, Jaehak; Whang, Jooho [Kyunghee University, Seoul (Korea, Republic of)

2016-12-15

Calculations have been made to estimate the generation of SNF including Plutonium (Pu) and minor actinides (MAs) from PWR and PHWR NPPs by using the IAEA code Nuclear Fuel Cycle Simulation System (NFCSS) for the period of 2016, 2030 as 12 more NPPs will be added by the year 2029. In order to find the optimize option for SNF management through sustainable use of nuclear energy system and reduce the associated radiological risk, 4 scenarios to burnout the Pu and MAs are analyzed. Estimation for the amount of SNF including major radionuclides has made for the year of 2016, 2030, 2089 and 2109 with an approximate amount of 148,19.65MT, 248,34.6 MT, 41572.23 MT and 61272.24 MT of SNF will be generated respectively. Radioactivity and radiotoxicity is calculated in order to access the radiological risk in terms of ingestion. Four Transmutation strategies is predicted to make Korean NPPs sustainable by incorporating KALIMER-600 (Burner) reactor in Korean nuclear fleet with an approximate transmutation rate of 99.77%, 96.57%, 95.34% and 97.58% for MA only (scenario-1), (MA & Pu) scenario-2, scenario-3 (MA only till the year 2125) and scenario-4 MA & Pu till the year 2148) respectively.

Spent nuclear fuel recycling with plasma reduction and etching

Science.gov (United States)

Kim, Yong Ho

2012-06-05

A method of extracting uranium from spent nuclear fuel (SNF) particles is disclosed. Spent nuclear fuel (SNF) (containing oxides of uranium, oxides of fission products (FP) and oxides of transuranic (TRU) elements (including plutonium)) are subjected to a hydrogen plasma and a fluorine plasma. The hydrogen plasma reduces the uranium and plutonium oxides from their oxide state. The fluorine plasma etches the SNF metals to form UF6 and PuF4. During subjection of the SNF particles to the fluorine plasma, the temperature is maintained in the range of 1200-2000 deg K to: a) allow any PuF6 (gas) that is formed to decompose back to PuF4 (solid), and b) to maintain stability of the UF6. Uranium (in the form of gaseous UF6) is easily extracted and separated from the plutonium (in the form of solid PuF4). The use of plasmas instead of high temperature reactors or flames mitigates the high temperature corrosive atmosphere and the production of PuF6 (as a final product). Use of plasmas provide faster reaction rates, greater control over the individual electron and ion temperatures, and allow the use of CF4 or NF3 as the fluorine sources instead of F2 or HF.

Effects of Uncertainty and Spatial Variability on Seepage into Drifts in the Yucca Mountain Total system Performance Assessment Model

International Nuclear Information System (INIS)

Kalinich, D. A.; Wilson, M. L.

2001-01-01

Seepage into the repository drifts is an important factor in total-system performance. Uncertainty and spatial variability are considered in the seepage calculations. The base-case results show 13.6% of the waste packages (WPs) have seepage. For 5th percentile uncertainty, 4.5% of the WPs have seepage and the seepage flow decreased by a factor of 2. For 95th percentile uncertainty, 21.5% of the WPs have seepage and the seepage flow increased by a factor of 2. Ignoring spatial variability resulted in seepage on 100% of the WPs, with a factor of 3 increase in the seepage flow

State of the art of the welding method for sealing spent nuclear fuel canister made of copper. Part 2 - EBW

International Nuclear Information System (INIS)

Salonen, T.

2014-05-01

This report consist the results of the development of the electron beam welding (EBW) method for sealing spent nuclear fuel (SNF) disposal canister. This report has been used as background material for selection of the sealing method for the SNF canister. Report contains the state of the art knowledge of the EBW method and research and development (R and D) results done by Posiva. Relevant R and D results of EB-welds done by SKB are also reviewed in this report. Requirements set for the welding and weld are present. These requirements are based on the long term safety and also some part of requirements are set by other processes like non-destructive testing (NDT) and manufacturing processes of components. Initial state of the weld is described in this report. Initial state has significant effect on the long term safety issues like corrosion resistance and creep ductility. Also short and long term mechanical properties as well as corrosion properties are described. Microstructure and residual stresses of the weld is represented in this report. Report consists also imperfections of the weld and statistical analysis of the evaluation of the probability of the largest defect size on the weld. Results of corrosion and creep tests of EB-welds are reviewed in this report. EBW process and machine are described. Preliminary designing of the EBW-machine has been done including component handling equipments. Preliminary welding procedure specification (pWPS) has drawn up and qualification of the personnel is described briefly. In-line process and quality control system including seam tracking system is implemented in modern EBW machine. Also NDT methods for inspection of the weld are described in this report. Concerning the results from the research and development work it can be concluded that EB welding method is suitable method for sealing SNF canister. Weld material fulfils requirements set by the long term safety. The welding system is robust and reliable and it is based

Nuclear Waste Fund management

International Nuclear Information System (INIS)

Rosselli, R.

1984-01-01

The Nuclear Waste Policy Act of 1982 (NWPA) established two separate special bank accounts: the Nuclear Waste Fund (NWF) was established to finance all of the Federal Government activities associated with the disposal of High-Level Waste (HLW) or Spent Nuclear Fuel (SNF). The Interim Storage Fund (ISF) is the financial mechanism for the provision of Federal Interim Storage capacity, not to exceed 1900 metric tons of SNF at civilian power reactors. The management of these funds is discussed. Since the two funds are identical in features and the ISF has not yet been activated, the author's remarks are confined to the Nuclear Waste Fund. Three points discussed include legislative features, current status, and planned activities

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...

Type of female sex worker and other risk factors of syphilis

Directory of Open Access Journals (Sweden)

Roselinda Roselinda

2016-03-01

Full Text Available AbstrakLatar belakang: Sifilis termasuk salah satu dari penyakit infeksi menular seksual kronis disebabkankuman Treponema pallidum yang dapat menyebabkan kecacatan pada penderita dan anak yang dilahirkan.Tujuan analisis ini adalah untuk melihat hubungan tipe dan lama kerja sebagai Wanita Pekerja Seks(WPS dengan kejadian penyakit sifilis di 7 kota di Indonesia.Metode: Data dari Survei WPS menggunakan kuesioner terstruktur di 7 kota (Kupang, Samarinda,Pontianak, Yogyakarta, Timika, Makassar dan Tangerang di Indonesia tahun 2007, desain potong lintangdan responden dipilih secara cluster random sampling dari WPS langsung dan tidak langsung yangmemenuhi kriteria definisi operasional. Diagnosa Sifilis ditegakkan dengan pemeriksaan laboratoriumRapid Plasma Reagent (RPR dan Treponema pallidum Haemaglutination Assay (TPHA.Hasil: Sebanyak 1750 responden ikut dalam penelitian dan 12.2% terindikasi sifilis. Kota Makassar mempunyaiprevalensi yang tertinggi sebesar 55,2%. WPS yang berlokasi di luar pulau Jawa memiliki risiko terinfeksi sifilis3,16 kali lebih tinggi dibandingkan dengan WPS yang berlokasi dipulau Jawa [risiko relatif suaian (RRa = 3,16 ;P= 0,000]. Tipe WPS tidak langsung memiliki risiko 46% lebih banyak untuk terinfeksi sifilis dibandingkan denganWPS langsung (RRa = 1,46 ; 95%; P = 0,002, sedangkan WPS yang mencari pengobatan ke dokter memiliki risiko58% lebih besar dibandingkan yang berobat ke sarana kesehatan langsung (RRa = 1,58; P = 0.006.Kesimpulan: Lokasi WPS yang berada di luar pulau Jawa, tipe WPS tidak langsung memiliki risiko lebihtinggi untuk terinfeksi penyakit sifilis. Wanita pekerja seks yang mencari pengobatan ke dokter menyebabkanpenyakit terindikasi lebih tinggi dibandingkan jika berobat ke sarana pelayanan kesehatan lainnya. (HealthScience Journal of Indonesia 2015;6:132-6Kata kunci: sifilis, wanita pekerja seks, Indonesia AbstractBackground: Syphilis is one of the chronicle sexual transmission diseases which is caused by

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...

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

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.)

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.

Current state of nuclear fuel cycles in nuclear engineering and trends in their development according to the environmental safety requirements

Science.gov (United States)

Vislov, I. S.; Pischulin, V. P.; Kladiev, S. N.; Slobodyan, S. M.

2016-08-01

The state and trends in the development of nuclear fuel cycles in nuclear engineering, taking into account the ecological aspects of using nuclear power plants, are considered. An analysis of advantages and disadvantages of nuclear engineering, compared with thermal engineering based on organic fuel types, was carried out. Spent nuclear fuel (SNF) reprocessing is an important task in the nuclear industry, since fuel unloaded from modern reactors of any type contains a large amount of radioactive elements that are harmful to the environment. On the other hand, the newly generated isotopes of uranium and plutonium should be reused to fabricate new nuclear fuel. The spent nuclear fuel also includes other types of fission products. Conditions for SNF handling are determined by ecological and economic factors. When choosing a certain handling method, one should assess these factors at all stages of its implementation. There are two main methods of SNF handling: open nuclear fuel cycle, with spent nuclear fuel assemblies (NFAs) that are held in storage facilities with their consequent disposal, and closed nuclear fuel cycle, with separation of uranium and plutonium, their purification from fission products, and use for producing new fuel batches. The development of effective closed fuel cycles using mixed uranium-plutonium fuel can provide a successful development of the nuclear industry only under the conditions of implementation of novel effective technological treatment processes that meet strict requirements of environmental safety and reliability of process equipment being applied. The diversity of technological processes is determined by different types of NFA devices and construction materials being used, as well as by the composition that depends on nuclear fuel components and operational conditions for assemblies in the nuclear power reactor. This work provides an overview of technological processes of SNF treatment and methods of handling of nuclear fuel

Effects of Burnable Absorbers on PWR Spent Nuclear Fuel

International Nuclear Information System (INIS)

O'Leary, P.M.; Pitts, M.L.

2000-01-01

Burnup credit is an ongoing issue in designing and licensing transportation and storage casks for spent nuclear fuel (SNF). To address this issue, in July 1999, the U.S. Nuclear Regulatory Commission (NRC), Spent Fuel Project Office, issued Interim Staff Guidance-8 (ISG-8), Revision 1 allowing limited burnup credit for pressurized water reactor (PWR) spent nuclear fuel (SNF) to be used in transport and storage casks. However, one of the key limitations for a licensing basis analysis as stipulated in ISG-8, Revision 1 is that ''burnup credit is restricted to intact fuel assemblies that have not used burnable absorbers''. Because many PWR fuel designs have incorporated burnable-absorber rods for more than twenty years, this restriction places an unnecessary burden on the commercial nuclear power industry. This paper summarizes the effects of in-reactor irradiation on the isotopic inventory of PWR fuels containing different types of integral burnable absorbers (BAs). The work presented is illustrative and intended to represent typical magnitudes of the reactivity effects from depleting PWR fuel with different types of burnable absorbers

On a possibility to ground a reliable and safe disposing of a spent nuclear fuel from nuclear reactors RBMK in deep boreholes

International Nuclear Information System (INIS)

Kedrovskij, O.L.

1998-01-01

In order to isolate a spent nuclear fuel (SNF), it is proposed to dispose it, after 30 years keeping on the day surface, in boreholes of up to 4 km depth and 350-1020 mm diameter drilled in low permeable platform basement crystalline rocks, that allows one to localize SNF radionuclides till their full decay. It is shown that the method requires relatively low investments and enables the volume of a burial being increased during wastes income. Along with consideration and assessment of hydrodynamic, geological and hydrogeological parameters of a rock massive and rocks preferable for the new method, engineering solutions for the borehole design and hermetizing structure and assessments of the technogeneous influence on the environment are given. The questions are also considered of a possible shortening of the terms of keeping SNF in a surface storage before their burying, that enables one to decrease expenses for surface storage constructions and to promote the cleaning of the sphere of human habitation from the most active and dangerous wastes

Geochemistry Model Validation Report: Material Degradation and Release Model

Energy Technology Data Exchange (ETDEWEB)

H. Stockman

2001-09-28

The purpose of this Analysis and Modeling Report (AMR) is to validate the Material Degradation and Release (MDR) model that predicts degradation and release of radionuclides from a degrading waste package (WP) in the potential monitored geologic repository at Yucca Mountain. This AMR is prepared according to ''Technical Work Plan for: Waste Package Design Description for LA'' (Ref. 17). The intended use of the MDR model is to estimate the long-term geochemical behavior of waste packages (WPs) containing U. S . Department of Energy (DOE) Spent Nuclear Fuel (SNF) codisposed with High Level Waste (HLW) glass, commercial SNF, and Immobilized Plutonium Ceramic (Pu-ceramic) codisposed with HLW glass. The model is intended to predict (1) the extent to which criticality control material, such as gadolinium (Gd), will remain in the WP after corrosion of the initial WP, (2) the extent to which fissile Pu and uranium (U) will be carried out of the degraded WP by infiltrating water, and (3) the chemical composition and amounts of minerals and other solids left in the WP. The results of the model are intended for use in criticality calculations. The scope of the model validation report is to (1) describe the MDR model, and (2) compare the modeling results with experimental studies. A test case based on a degrading Pu-ceramic WP is provided to help explain the model. This model does not directly feed the assessment of system performance. The output from this model is used by several other models, such as the configuration generator, criticality, and criticality consequence models, prior to the evaluation of system performance. This document has been prepared according to AP-3.10Q, ''Analyses and Models'' (Ref. 2), and prepared in accordance with the technical work plan (Ref. 17).

Geochemistry Model Validation Report: Material Degradation and Release Model

International Nuclear Information System (INIS)

Stockman, H.

2001-01-01

The purpose of this Analysis and Modeling Report (AMR) is to validate the Material Degradation and Release (MDR) model that predicts degradation and release of radionuclides from a degrading waste package (WP) in the potential monitored geologic repository at Yucca Mountain. This AMR is prepared according to ''Technical Work Plan for: Waste Package Design Description for LA'' (Ref. 17). The intended use of the MDR model is to estimate the long-term geochemical behavior of waste packages (WPs) containing U. S . Department of Energy (DOE) Spent Nuclear Fuel (SNF) codisposed with High Level Waste (HLW) glass, commercial SNF, and Immobilized Plutonium Ceramic (Pu-ceramic) codisposed with HLW glass. The model is intended to predict (1) the extent to which criticality control material, such as gadolinium (Gd), will remain in the WP after corrosion of the initial WP, (2) the extent to which fissile Pu and uranium (U) will be carried out of the degraded WP by infiltrating water, and (3) the chemical composition and amounts of minerals and other solids left in the WP. The results of the model are intended for use in criticality calculations. The scope of the model validation report is to (1) describe the MDR model, and (2) compare the modeling results with experimental studies. A test case based on a degrading Pu-ceramic WP is provided to help explain the model. This model does not directly feed the assessment of system performance. The output from this model is used by several other models, such as the configuration generator, criticality, and criticality consequence models, prior to the evaluation of system performance. This document has been prepared according to AP-3.10Q, ''Analyses and Models'' (Ref. 2), and prepared in accordance with the technical work plan (Ref. 17)

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.

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.)

Licensing Air and Transboundary Shipments of Spent Nuclear Fuel

International Nuclear Information System (INIS)

Komarov, S.V.; Budu, M.E.; Derganov, D.V.; Savina, O.A.; Bolshinsky, I.M.; Moses, S.D.; Biro, L.

2016-01-01

Since 1996 the IAEA TS-R-1 regulation included new requirements applicable to transport of fissile materials by air. The later 2005 and 2009 editions confirmed the validity of those provisions. Despite the fact that the IAEA TS-R-1 allows for air shipments of SNF in Type B and Type C packages, the examples of such shipments are not abundant. Nuclear regulatory bodies and transport safety experts are cautious about air shipments of SNF. Why so? What are the risks? What are the alternatives? In this new regulatory framework, in 2009, two air shipments in Type B packages of Research Reactor (RR) Spent Nuclear Fuel (SNF) from Romania and Libya were performed under the U.S. DOE/NNSA RRRFR Program. The first licensing process of such shipment brought up many questions about package and shipment safety from the licensing experts' side and so the scope of analyses exceeded the requirements of IAEA. Under the thorough supervision of Rosatom and witnessed by DOE and CNCAN, all questions were answered by various strength analyses and risk evaluations. But the progress achieved didn't stop here. In 2010-2011, an energy absorption container (EAC) with titanium spheres as absorbers based on the SKODA VPVR/M cask was designed as the first Type C package in the world destined for RR SNF, currently under approval process. At the same time, intense preparations for the safe removal of the Russian-origin damaged RR SNF from Serbia, Vinca were in progress. The big amount of SNF and its rapidly worsening condition imposed as requirements to organize only one shipment as fast as possible, i.e. using at the maximum extent the entire experience available from other SNF shipments. The long route, several transit countries and means of transport, two different casks, new European regulations and many other issues resulted for the Serbian shipment in one of the most complex SNF shipments’ licensing exercise. This paper shows how the international regulatory framework ensures the

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

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.)

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

Cask size and weight reduction through the use of depleted uranium dioxide-concrete material

International Nuclear Information System (INIS)

Lobach, S.Yu.; Haire, J.M.

2007-01-01

Newly developed depleted uranium (DU) composite materials enable fabrication of spent nuclear fuel (SNF) transport and storage casks that are smaller and lighter in weight than casks made with conventional materials. One such material is DU dioxide (DUO2)-concrete, so-called DUCRETE TM . This paper examines the radiation shielding efficiency of DUCRETE as compared with that of a conventional concrete cask that holds 32 pressurized-water-reactor SNF assemblies. In this analysis, conventional concrete shielding material is replaced with DUCRETE. The thickness of the DUCRETE shielding is adjusted to give the same radiation surface dose, 200 mrem/h (2 mSv/hr), as the conventional concrete cask. It was found that the concrete shielding thickness decreased from 71 to 20 cm and that the cask radial cross-section shielding area was reduced approx 50 %. The weight was reduced approx 21 %, from 154 to approx 127 tons. Should one choose to add an extra outer ring of SNF assemblies, the number of such assemblies would increase from 32 to 52. In this case, the outside cask diameter would still decrease, from 169 to 137 cm. However, the weight would increase somewhat from 156 to 177 tons. Neutron cask surface dose is only approx 10 % of the gamma dose. These reduced sizes and weights will significantly influence the design of next-generation SNF casks

Conceptual Design of Low Fusion Power Hybrid System for Waste Transmutation

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

DRUP (Direct Reuse of Used PWR) fuel has same process with DUPIC (Direct Use of spent PWR fuel Into CANDU reactor). There are 2 big benefits by using DRUP fuel in Hybrid system. One is fissile production during operating period. Required power is decreased by fissile production from DRUP fuel. When the fusion power is reduced, integrity of structure materials is not significantly weakened due to reduction of 14.1MeV high energy neutrons. In addition, required amount of tritium for self-sufficiency TBR (Tritium Breeding Ratio ≥ 1.1) is decreased. Therefore, it is possible to further loading the SNF as much as the amount of lithium decreased. It is effective in transmutation. The other one is that DRUP fuel is also SNF. Therefore, using DRUP fuel is reusing of SNF, as a result it makes reduction of SNF from PWR. However, thermal neutron system is suitable for using DRUP fuel compared to fast neutron system. Therefore, transmutation zone designed (U-TRU)Zr fuel and fissile production zone designed DRUP fuel are separated in this study. In this paper, using DRUP fuel for low fusion power in hybrid system is suggested. Fusion power is decreased by using DRUP fuel. As a result, TBR is satisfied design condition despite of using natural lithium. In addition, not only (U-TRU)Zr fuel but also DRUP fuel are transmuted.

MCO Monitoring activity description

International Nuclear Information System (INIS)

SEXTON, R.A.

1998-01-01

Spent Nuclear Fuel remaining from Hanford's N-Reactor operations in the 1970s has been stored under water in the K-Reactor Basins. This fuel will be repackaged, dried and stored in a new facility in the 200E Area. The safety basis for this process of retrieval, drying, and interim storage of the spent fuel has been established. The monitoring of MCOS in dry storage is a currently identified issue in the SNF Project. This plan outlines the key elements of the proposed monitoring activity. Other fuel stored in the K-Reactor Basins, including SPR fuel, will have other monitoring considerations and is not addressed by this activity description

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.)

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

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.

Nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

This patent describes an improved liquid metal nuclear reactor construction comprising: (a) a nuclear reactor core having a bottom platform support structure; (b) a reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core; (c) a containment structure surrounding the reactor vessel and having a sidewall spaced outwardly from the reactor vessel side wall and having a base mat spaced below the reactor vessel bottom end wall; (d) a central small diameter post anchored to the containment structure base mat and extending upwardly to the reactor vessel to axially fix the bottom end wall of the reactor vessel and provide a center column support for the lower end of the reactor core; (e) annular support structure disposed in the reactor vessel on the bottom end wall and extending about the lower end of the core; (f) structural support means disposed between the containment structure base mat and bottom end of the reactor vessel wall and cooperating for supporting the reactor vessel at its bottom end wall on the containment structure base mat to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event; (g) a bed of insulating material disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall; freely expand radially from the central post as it heats up while providing continuous support thereof; (h) a deck supported upon the wall of the containment vessel above the top open end of the reactor vessel; and (i) extendible and retractable coupling means extending between the deck and the top open end of the reactor vessel and flexibly and sealably interconnecting the reactor vessel at its top end to the deck

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

Health issues in the Arab American community. Tobacco use patterns among high school students: do Arab American youth differ?

Science.gov (United States)

Weglicki, Linda S; Templin, Thomas; Hammad, Adnan; Jamil, Hikmet; Abou-Mediene, Sharifa; Farroukh, Mona; Rice, Virginia Hill

2007-01-01

To determine tobacco use rates (cigarette, water pipe smoking [WPS] or narghile) in Arab American compared to non-Arab youth. A convenience sample of 2,782 14- to 18-year-old high school students from a midwest community completed a 21-item tobacco use history survey. Seventy-one percent of the participants were ArA. Grades 9 through 12 were equally represented. Results included 'ever tried cigarettes [narghile]' (20%, 39%); 'smoked cigarettes [narghile] in the past 30 days' (7%, 22%); and 'regular smoking [narghile]' (3%, 15%) for ArA and non-Arab youths, respectively. Each was significantly related to grade and ethnicity. WPS for ArA and non-Arab youths was (38%, 21%); (17%, 11%); and (7%, 5%) for 'ever used,' 'used in the past 30 days,' and 'regular use,' respectively. Grade, ethnicity, and sex were significantly related to WPS. Cigarette smoking rates for non-Arab youth were lower than current national youth smoking rates but significantly higher than ArA youth. Rates for ArA youth were much lower than current national reported data. Rates of WPS for US youth, regardless of race or ethnicity, are not known. Findings from this study indicate that both ArA and non-Arab youth are experimenting and using WPS regularly. These results underscore the importance of assessing novel forms of tobacco use, particularly WPS, a growing phenomenon among US youth.

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.)

The Impact of Operating Parameters and Correlated Parameters for Extended BWR Burnup Credit

Energy Technology Data Exchange (ETDEWEB)

Ade, Brian J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Marshall, William B. J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ilas, Germina [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bowman, Stephen M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-06-01

Applicants for certificates of compliance for spent nuclear fuel (SNF) transportation and dry storage systems perform analyses to demonstrate that these systems are adequately subcritical per the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Parts 71 and 72. For pressurized water reactor (PWR) SNF, these analyses may credit the reduction in assembly reactivity caused by depletion of fissile nuclides and buildup of neutron-absorbing nuclides during power operation. This credit for reactivity reduction during depletion is commonly referred to as burnup credit (BUC). US Nuclear Regulatory Commission (NRC) staff review BUC analyses according to the guidance in the Division of Spent Fuel Storage and Transportation Interim Staff Guidance (ISG) 8, Revision 3, Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transportation and Storage Casks.

System design description for the whole element furnace testing system

International Nuclear Information System (INIS)

Ritter, G.A.; Marschman, S.C.; MacFarlan, P.J.; King, D.A.

1998-05-01

This document provides a detailed description of the Hanford Spent Nuclear Fuel (SNF) Whole Element Furnace Testing System located in the Postirradiation Testing Laboratory G-Cell (327 Building). Equipment specifications, system schematics, general operating modes, maintenance and calibration requirements, and other supporting information are provided in this document. This system was developed for performing cold vacuum drying and hot vacuum drying testing of whole N-Reactor fuel elements, which were sampled from the 105-K East and K West Basins. The proposed drying processes are intended to allow dry storage of the SNF for long periods of time. The furnace testing system is used to evaluate these processes by simulating drying sequences with a single fuel element and measuring key system parameters such as internal pressures, temperatures, moisture levels, and off-gas composition

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.)

The CAREM reactor and present currents in reactor design

International Nuclear Information System (INIS)

Ordonez, J.P.

1990-01-01

INVAP has been working on the CAREM project since 1983. It concerns a very low power reactor for electrical energy generation. The design of the reactor and the basic criteria used were described in 1984. Since then, a series of designs have been presented for reactors which are similar to CAREM regarding the solutions presented to reduce the chance of major nuclear accidents. These designs have been grouped under different names: Advanced Reactors, Second Generation Reactors, Inherently Safe Reactors, or even, Revolutionary Reactors. Every reactor fabrication firm has, at least, one project which can be placed in this category. Presently, there are two main currents of Reactor Design; Evolutionary and Revolutionary. The present work discusses characteristics of these two types of reactors, some revolutionary designs and common criteria to both types. After, these criteria are compared with CAREM reactor design. (Author) [es

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

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

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)

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

Reactor science and technology: operation and control of reactors

International Nuclear Information System (INIS)

Qiu Junlong

1994-01-01

This article is a collection of short reports on reactor operation and research in China in 1991. The operation of and research activities linked with the Heavy Water Research Reactor, Swimming Pool Reactor and Miniature Neutron Source Reactor are briefly surveyed. A number of papers then follow on the developing strategies in Chinese fast breeder reactor technology including the conceptual design of an experimental fast reactor (FFR), theoretical studies of FFR thermo-hydraulics and a design for an immersed sodium flowmeter. Reactor physics studies cover a range of topics including several related to work on zero power reactors. The section on reactor safety analysis is concerned largely with the assessment of established, and the presentation of new, computer codes for use in PWR safety calculations. Experimental and theoretical studies of fuels and reactor materials for FBRs, PWRs, BWRs and fusion reactors are described. A final miscellaneous section covers Mo-Tc isotope production in the swimming pool reactor, convective heat transfer in tubes and diffusion of tritium through plastic/aluminium composite films and Li 2 SiO 3 . (UK)

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.)

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)

Review of oxidation rates of DOE spent nuclear fuel : Part 1 : nuclear fuel

International Nuclear Information System (INIS)

Hilton, B.A.

2000-01-01

The long-term performance of Department of Energy (DOE) spent nuclear fuel (SNF) in a mined geologic disposal system depends highly on fuel oxidation and subsequent radionuclide release. The oxidation rates of nuclear fuels are reviewed in this two-volume report to provide a baseline for comparison with release rate data and technical rationale for predicting general corrosion behavior of DOE SNF. The oxidation rates of nuclear fuels in the DOE SNF inventory were organized according to metallic, Part 1, and non-metallic, Part 2, spent nuclear fuels. This Part 1 of the report reviews the oxidation behavior of three fuel types prototypic of metallic fuel in the DOE SNF inventory: uranium metal, uranium alloys and aluminum-based dispersion fuels. The oxidation rates of these fuels were evaluated in oxygen, water vapor, and water. The water data were limited to pure water corrosion as this represents baseline corrosion kinetics. Since the oxidation processes and kinetics discussed in this report are limited to pure water, they are not directly applicable to corrosion rates of SNF in water chemistry that is significantly different (such as may occur in the repository). Linear kinetics adequately described the oxidation rates of metallic fuels in long-term corrosion. Temperature dependent oxidation rates were determined by linear regression analysis of the literature data. As expected the reaction rates of metallic fuels dramatically increase with temperature. The uranium metal and metal alloys have stronger temperature dependence than the aluminum dispersion fuels. The uranium metal/water reaction exhibited the highest oxidation rate of the metallic fuel types and environments that were reviewed. Consequently, the corrosion properties of all DOE SNF may be conservatively modeled as uranium metal, which is representative of spent N-Reactor fuel. The reaction rate in anoxic, saturated water vapor was essentially the same as the water reaction rate. The long-term intrinsic

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.

Roles of plasma neutron source reactor in development of fusion reactor engineering: Comparison with fission reactor engineering

International Nuclear Information System (INIS)

Hirayama, Shoichi; Kawabe, Takaya

1995-01-01

The history of development of fusion power reactor has come to a turning point, where the main research target is now shifting from the plasma heating and confinement physics toward the burning plasma physics and reactor engineering. Although the development of fusion reactor system is the first time for human beings, engineers have experience of development of fission power reactor. The common feature between them is that both are plants used for the generation of nuclear reactions for the production of energy, nucleon, and radiation on an industrial scale. By studying the history of the development of the fission reactor, one can find the existence of experimental neutron reactors including irradiation facilities for fission reactor materials. These research neutron reactors played very important roles in the development of fission power reactors. When one considers the strategy of development of fusion power reactors from the points of fusion reactor engineering, one finds that the fusion neutron source corresponds to the neutron reactor in fission reactor development. In this paper, the authors discuss the roles of the plasma-based neutron source reactors in the development of fusion reactor engineering, by comparing it with the neutron reactors in the history of fission power development, and make proposals for the strategy of the fusion reactor development. 21 refs., 6 figs

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)

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

Comparison between TRU burning reactors and commercial fast reactor

International Nuclear Information System (INIS)

Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

2001-03-01

Research and development for stabilizing or shortening the radioactive wastes including in spent nuclear fuel are widely conducted in view point of reducing the environmental impact. Especially it is effective way to irradiate and transmute long-lived TRU by fast reactors. Two types of loading way were previously proposed. The former is loading relatively small amount of TRU in all commercial fast reactors and the latter is loading large amount of TRU in a few TRU burning reactors. This study has been intended to contribute to the feasibility studies on commercialized fast reactor cycle system. The transmutation and nuclear characteristics of TRU burning reactors were evaluated and compared with those of conventional transmutation system using commercial type fast reactor based upon the investigation of technical information about TRU burning reactors. Major results are summarized as follows. (1) Investigation of technical information about TRU burning reactors. Based on published reports and papers, technical information about TRU burning reactor concepts transmutation system using convectional commercial type fast reactors were investigated. Transmutation and nuclear characteristics or R and D issue were investigated based on these results. Homogeneously loading of about 5 wt% MAs on core fuels in the conventional commercial type fast reactor may not cause significant impact on the nuclear core characteristics. Transmutation of MAs being produced in about five fast reactors generating the same output is feasible. The helium cooled MA burning fast reactor core concept propose by JAERI attains criticality using particle type nitride fuels which contain more than 60 wt% MA. This reactor could transmute MAs being produced in more than ten 1000 MWe-LWRs. Ultra-long life core concepts attaining more than 30 years operation without refueling by utilizing MA's nuclear characteristics as burnable absorber and fertile nuclides were proposed. Those were pointed out that

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.)

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

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

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.

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

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.)

Performance Specification Shippinpark Pressurized Water Reactor Fuel Drying and Canister Inerting System for PWR Core 2 Blanket Fuel Assemblies Stored within Shippingport Spent Fuel Canisters

International Nuclear Information System (INIS)

JOHNSON, D.M.

2000-01-01

This specification establishes the performance requirements and basic design requirements imposed on the fuel drying and canister inerting system for Shippingport Pressurized Water Reactor (PWR) Core 2 blanket fuel assemblies (BFAs) stored within Shippingport spent fuel (SSFCs) canisters (fuel drying and canister inerting system). This fuel drying and canister inerting system is a component of the U.S. Department of Energy, Richland Operations Office (RL) Spent Nuclear Fuels Project at the Hanford Site. The fuel drying and canister inerting system provides for removing water and establishing an inert environment for Shippingport PWR Core 2 BFAs stored within SSFCs. A policy established by the U.S. Department of Energy (DOE) states that new SNF facilities (this is interpreted to include structures, systems and components) shall achieve nuclear safety equivalence to comparable U.S. Nuclear Regulatory Commission (NRC)-licensed facilities. This will be accomplished in part by applying appropriate NRC requirements for comparable NRC-licensed facilities to the fuel drying and canister inerting system, in addition to applicable DOE regulations and orders

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.

Research reactor standards and their impact on the TRIGA reactor community

International Nuclear Information System (INIS)

Richards, W.J.

1980-01-01

The American Nuclear Society has established a standards committee devoted to writing standards for research reactors. This committee was formed in 1971 and has since that time written over 15 standards that cover all aspects of research reactor operation. The committee has representation from virtually every group concerned with research reactors and their operation. This organization includes University reactors, National laboratory reactors, Nuclear Regulatory commission, Department of Energy and private nuclear companies and insurers. Since its beginning the committee has developed standards in the following areas: Standard for the development of technical specifications for research reactors; Quality control for plate-type uranium-aluminium fuel elements; Records and reports for research reactors; Selection and training of personnel for research reactors; Review of experiments for research reactors; Research reactor site evaluation; Quality assurance program requirements for research reactors; Decommissioning of research reactors; Radiological control at research reactor facilities; Design objectives for and monitoring of systems controlling research reactor effluents; Physical security for research reactor facilities; Criteria for the reactor safety systems of research reactors; Emergency planning for research reactors; Fire protection program requirements for research reactors; Standard for administrative controls for research reactors. Besides writing the above standards, the committee is very active in using communications with the nuclear regulatory commission on proposed rules or positions which will affect the research reactor community

Waste Package Lifting Calculation

International Nuclear Information System (INIS)

H. Marr

2000-01-01

The objective of this calculation is to evaluate the structural response of the waste package during the horizontal and vertical lifting operations in order to support the waste package lifting feature design. The scope of this calculation includes the evaluation of the 21 PWR UCF (pressurized water reactor uncanistered fuel) waste package, naval waste package, 5 DHLW/DOE SNF (defense high-level waste/Department of Energy spent nuclear fuel)--short waste package, and 44 BWR (boiling water reactor) UCF waste package. Procedure AP-3.12Q, Revision 0, ICN 0, calculations, is used to develop and document this calculation

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

Improved nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding liquid metal coolant and housing the core within the pool. A generally cylindrical concrete containment structure surrounds the reactor vessel and a central support pedestal is anchored to the containment structure base mat and supports the bottom wall of the reactor vessel and the reactor core. The periphery of the reactor vessel bore is supported by an annular structure which allows thermal expansion but not seismic motion of the vessel, and a bed of thermally insulating material uniformly supports the vessel base whilst allowing expansion thereof. A guard ring prevents lateral seismic motion of the upper end of the reactor vessel. The periphery of the core is supported by an annular structure supported by the vessel base and keyed to the vessel wall so as to be able to expand but not undergo seismic motion. A deck is supported on the containment structure above the reactor vessel open top by annular bellows, the deck carrying the reactor control rods such that heating of the reactor vessel results in upward expansion against the control rods. (author)

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)

Development of Reactor Console Simulator for PUSPATI TRIGA Reactor

International Nuclear Information System (INIS)

Mohd Idris Taib; Izhar Abu Hussin; Mohd Khairulezwan Abdul Manan; Nufarhana Ayuni Joha; Mohd Sabri Minhat

2012-01-01

The Reactor Console Simulator will be an interactive tool for operator training and teaching of PUSPATI TRIGA Reactor. Behaviour and characteristic for reactor console and reactor itself can be evaluated and understand. This Simulator will be used as complement for actual present reactor console. Implementation of man-machine interface is using computer screens, keyboard and mouse. Multiple screens are used to match the physical of present reactor console. LabVIEW software are using for user interface and mathematical calculation. Polynomial equation based on control rods calibration data as well as operation parameters record was used to calculate the estimated reactor console parameters. (author)

ALARA Principle Application for Loading Spent Nuclear Fuel Assemblies from Nuclear Research Reactor WR-S Mergal-Bucharest Romania into Transportation Casks

International Nuclear Information System (INIS)

Dragusin, M.

2009-01-01

Safety implementation of Spent Nuclear Fuels Assemblies (SNFA) handling procedures at the WR-S reactor site is ensured by technical perfection and reliability of equipment, monitoring of its condition, qualification and discipline of personnel as well as organization and execution of work complied with requirements of regulatory documents, process procedures, guidance and manuals. The personnel training for execution loading of SNF FAs is other important aspect for radiation protection and safely activities. Estimations carried out using Micro Shield software show that maximal dose rate upon working site when loading four FAs into basket of cask will not exceed 1.7 and 956;Sv/h, excluding natural radiation. Radiation Safety Analyses estimates for loading 70 SNFA in 18 transportation casks are: maximal individual dose: 4274.7 and 956;Sv, maximal expected collective dose persons: 17 031.2 man and 956;Sv. By application ALARA principle with technical and administrative measures the loading process developed in the following conditions: maximal individual dose: 68 and 956;Sv, the collective dose persons: 732 man and 956;Sv. The work will presented the technical measures and procedures applied in loading process.

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

Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

International Nuclear Information System (INIS)

Greneche, D.

2014-01-01

This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

Control of reactor coolant flow path during reactor decay heat removal

International Nuclear Information System (INIS)

Hunsbedt, A.N.

1988-01-01

This patent describes a sodium cooled reactor of the type having a reactor hot pool, a slightly lower pressure reactor cold pool and a reactor vessel liner defining a reactor vessel liner flow gap separating the hot pool and the cold pool along the reactor vessel sidewalls and wherein the normal sodium circuit in the reactor includes main sodium reactor coolant pumps having a suction on the lower pressure sodium cold pool and an outlet to a reactor core; the reactor core for heating the sodium and discharging the sodium to the reactor hot pool; a heat exchanger for receiving sodium from the hot pool, and removing heat from the sodium and discharging the sodium to the lower pressure cold pool; the improvement across the reactor vessel liner comprising: a jet pump having a venturi installed across the reactor vessel liner, the jet pump having a lower inlet from the reactor vessel cold pool across the reactor vessel liner and an upper outlet to the reactor vessel hot pool

The research reactors their contribution to the reactors physics

International Nuclear Information System (INIS)

Barral, J.C.; Zaetta, A.; Johner, J.; Mathoniere, G.

2000-01-01

The 19 october 2000, the french society of nuclear energy organized a day on the research reactors. This associated report of the technical session, reactors physics, is presented in two parts. The first part deals with the annual meeting and groups general papers on the pressurized water reactors, the fast neutrons reactors and the fusion reactors industry. The second part presents more technical papers about the research programs, critical models, irradiation reactors (OSIRIS and Jules Horowitz) and computing tools. (A.L.B.)

Reactor noise analysis of experimental fast reactor 'JOYO'

International Nuclear Information System (INIS)

Ohtani, Hideji; Yamamoto, Hisashi

1980-01-01

As a part of dynamics tests in experimental fast reactor ''JOYO'', reactor noise tests were carried out. The reactor noise analysis techniques are effective for study of plant characteristics by determining fluctuations of process signals (neutron signal, reactor inlet temperature signals, etc.), which are able to be measured without disturbances for reactor operations. The aims of reactor noise tests were to confirm that no unstable phenomenon exists in ''JOYO'' and to gain initial data of the plant for reference of the future data. Data for the reactor noise tests treated in this paper were obtained at 50 MW power level. Fluctuations of process signals were amplified and recorded on analogue tapes. The analysis was performed using noise code (NOISA) of digital computer, with which statistical values of ASPD (auto power spectral density), CPSD (cross power spectral density), and CF (coherence function) were calculated. The primary points of the results are as follows. 1. RMS value of neutron signal at 50 MW power level is about 0.03 MW. This neutron fluctuation is not disturbing reactor operations. 2. The fluctuations of A loop reactor inlet temperatures (T sub(AI)) are larger than the fluctuations of B loop reactor inlet temperature (T sub(BI)). For this reason, the major driving force of neutron fluctuations seems to be the fluctuations of T sub(AI). 3. Core and blanket subassemblies can be divided into two halves (A and B region), with respect to the spacial motion of temperature in the reactor core. A or B region means the region in which sodium temperature fluctuations in subassembly are significantly affected by T sub(AI) or T sub(BI), respectively. This phenomenon seems to be due to the lack of mixing of A and B loop sodium in lower plenum of reactor vessel. (author)

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

Reactor water spontaneous circulation structure in reactor pressure vessel

International Nuclear Information System (INIS)

Takahashi, Kazumi

1998-01-01

The gap between the inner wall of a reactor pressure vessel of a BWR type reactor and a reactor core shroud forms a down comer in which reactor water flows downwardly. A feedwater jacket to which feedwater at low temperature is supplied is disposed at the outer circumference of the pressure vessel just below a gas/water separator. The reactor water at the outer circumferential portion just below the air/water separator is cooled by the feedwater jacket, and the feedwater after cooling is supplied to the feedwater entrance disposed below the feedwater jacket by way of a feedwater introduction line to supply the feedwater to the lower portion of the down comer. This can cool the reactor water in the down comer to increase the reactor water density in the down comer thereby forming strong downward flows and promote the recycling of the reactor water as a whole. With such procedures, the reactor water can be recycled stably only by the difference of the specific gravity of the reactor water without using an internal pump. In addition, the increase of the height of the pressure vessel can be suppressed. (I.N.)

Building an Elastic Parallel OGC Web Processing Service on a Cloud-Based Cluster: A Case Study of Remote Sensing Data Processing Service

Directory of Open Access Journals (Sweden)

Xicheng Tan

2015-10-01

Full Text Available Since the Open Geospatial Consortium (OGC proposed the geospatial Web Processing Service (WPS, standard OGC Web Service (OWS-based geospatial processing has become the major type of distributed geospatial application. However, improving the performance and sustainability of the distributed geospatial applications has become the dominant challenge for OWSs. This paper presents the construction of an elastic parallel OGC WPS service on a cloud-based cluster and the designs of a high-performance, cloud-based WPS service architecture, the scalability scheme of the cloud, and the algorithm of the elastic parallel geoprocessing. Experiments of the remote sensing data processing service demonstrate that our proposed method can provide a higher-performance WPS service that uses less computing resources. Our proposed method can also help institutions reduce hardware costs, raise the rate of hardware usage, and conserve energy, which is important in building green and sustainable geospatial services or applications.

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.)

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

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.)

RB Research nuclear reactor RB reactor, Annual report for 2000

International Nuclear Information System (INIS)

Milosevic, M.

2000-12-01

Report on RB reactor operation during 2000 contains 3 parts. Part one contains a brief description of reactor operation and reactor components, relevant dosimetry data and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level-meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization with a comprehensive list of publications resulting from experiments done at the RB reactor. It contains data about reactor operation during previous 14 years, i.e. from 1986 - 2000

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.)

A Metagenomic Survey of Serpentinites and Nearby Soils in Taiwan

Science.gov (United States)

Li, K. Y.; Hsu, Y. W.; Chen, Y. W.; Huang, T. Y.; Shih, Y. J.; Chen, J. S.; Hsu, B. M.

2016-12-01

The serpentinite of Taiwan is originated from the subduction zone of the Eurasian plate and the Philippine Sea plate. Many small bodies of serpentinite are scattered around the lands of the East Rift Valley, which are also one of the major agricultural areas in Taiwan. Since microbial communities play a role both on weathering process and soil recovery, uncovering the microbial compositions in serpentinites and surrounding soils may help people to understand the roles of microorganisms on serpentinites during the nature weathering process. In this study, microorganisms growing on the surface of serpentinites, in the surrounding soil, and agriculture soils that are miles of horizontal distance away from serpentinite were collected. Next generation sequencing (NGS) was carried out to examine the metagenomics of uncultured microbial community in these samples. The metagenomics were further clustered into operational taxonomic units (OTUs) to analyze relative abundance, heatmap of OTUs, and principal coordinates analysis (PCoA). Our data revealed the different types of geographic material had their own distinct structures of microbial community. In serpentinites, the heatmaps based on the phylogenetic pattern showed that the OTUs distributions were similar in phyla of Bacteroidetes, Cyanobacteria, Proteobacteria, Verrucomicrobia, and WPS-1/WPS-2. On the other hand, the heatmaps of phylogenetic pattern of agriculture soils showed that the OTUs distributions in phyla of Chloroflexi, Acidobacteria, Actinobacteria, WPS-1/WPS-2, and Proteobacteria were similar. In soil nearby the serpentinite, some clusters of OTUs in phyla of Bacteroidetes, Cyanobacteria, and WPS-1/WPS-2 have disappeared. Our data provided evidence regarding kinetic evolutions of microbial communities in different geographic materials.

Influence of the Number of Predicted Words on Text Input Speed in Participants With Cervical Spinal Cord Injury.

Science.gov (United States)

Pouplin, Samuel; Roche, Nicolas; Vaugier, Isabelle; Jacob, Antoine; Figere, Marjorie; Pottier, Sandra; Antoine, Jean-Yves; Bensmail, Djamel

2016-02-01

To determine whether the number of words displayed in the word prediction software (WPS) list affects text input speed (TIS) in people with cervical spinal cord injury (SCI), and whether any influence is dependent on the level of the lesion. A cross-sectional trial. A rehabilitation center. Persons with cervical SCI (N=45). Lesion level was high (C4 and C5, American Spinal Injury Association [ASIA] grade A or B) for 15 participants (high-lesion group) and low (between C6 and C8, ASIA grade A or B) for 30 participants (low-lesion group). TIS was evaluated during four 10-minute copying tasks: (1) without WPS (Without); (2) with a display of 3 predicted words (3Words); (3) with a display of 6 predicted words (6Words); and (4) with a display of 8 predicted words (8Words). During the 4 copying tasks, TIS was measured objectively (characters per minute, number of errors) and subjectively through subject report (fatigue, perception of speed, cognitive load, satisfaction). For participants with low-cervical SCI, TIS without WPS was faster than with WPS, regardless of the number of words displayed (Pwords displayed in a word prediction list on TIS; however, perception of TIS differed according to lesion level. For persons with low-cervical SCI, a small number of words should be displayed, or WPS should not be used at all. For persons with high-cervical SCI, a larger number of words displayed increases the comfort of use of WPS. Copyright © 2016 American Congress of Rehabilitation Medicine. Published by Elsevier Inc. All rights reserved.

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.

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

Validation of the TRACE code for the system dynamic simulations of the molten salt reactor experiment and the preliminary study on the dual fluid molten salt reactor

International Nuclear Information System (INIS)

He, Xun

2016-01-01

Molten Salt Reactor (MSR), which was confirmed as one of the six Generation IV reactor types by the GIF (Generation IV International Forum in 2008), recently draws a lot of attention all around the world. Due to the application of liquid fuels the MSR can be regarded as the most special one among those six GEN-IV reactor types in a sense. A unique advantage of using liquid nuclear fuel lies in that the core melting accident can be thoroughly eliminated. Besides, a molten salt reactor can have several fuel options, for instance, the fuel can be based on "2"3"5U, "2"3"2Th-"2"3"3U, "2"3"8U-"2"3"9Pu cycle or even the spent nuclear fuel (SNF), so the reactor can be operated as a breeder or as an actinides burner both with fast, thermal or epi-thermal neutron spectrum and hence, it has excellent features of the fuel sustainability and for the non-proliferation. Furthermore, the lower operating pressure not only means a lower risk of the explosion as well as the radioactive leakage but also implies that the reactor vessel and its components can be lightweight, thus lowering the cost of equipments. So far there is no commercial MSR being operated. However, the MSR concept and its technical validation dates back to the 1960s to 1970s, when the scientists and engineers from ORNL (Oak Ridge National Laboratory) in the United States managed to build and run the world's first civilian molten salt reactor called MSRE (Molten Salt Reactor Experiment). The MSRE was an experimental liquid-fueled reactor with 10 MW thermal output using "4LiF-BeF_2-ZrF_4-UF_4 as the fuel also as the coolant itself. The MSRE is usually taken as a very important reference case for many current researches to validate their codes and simulations. Without exception it works also as a benchmark for this thesis. The current thesis actually consists of two main parts. The first part is about the validation of the current code for the old MSRE concept, while the second one is about the demonstration of a new

Validation of the TRACE code for the system dynamic simulations of the molten salt reactor experiment and the preliminary study on the dual fluid molten salt reactor

Energy Technology Data Exchange (ETDEWEB)

He, Xun

2016-06-14

Molten Salt Reactor (MSR), which was confirmed as one of the six Generation IV reactor types by the GIF (Generation IV International Forum in 2008), recently draws a lot of attention all around the world. Due to the application of liquid fuels the MSR can be regarded as the most special one among those six GEN-IV reactor types in a sense. A unique advantage of using liquid nuclear fuel lies in that the core melting accident can be thoroughly eliminated. Besides, a molten salt reactor can have several fuel options, for instance, the fuel can be based on {sup 235}U, {sup 232}Th-{sup 233}U, {sup 238}U-{sup 239}Pu cycle or even the spent nuclear fuel (SNF), so the reactor can be operated as a breeder or as an actinides burner both with fast, thermal or epi-thermal neutron spectrum and hence, it has excellent features of the fuel sustainability and for the non-proliferation. Furthermore, the lower operating pressure not only means a lower risk of the explosion as well as the radioactive leakage but also implies that the reactor vessel and its components can be lightweight, thus lowering the cost of equipments. So far there is no commercial MSR being operated. However, the MSR concept and its technical validation dates back to the 1960s to 1970s, when the scientists and engineers from ORNL (Oak Ridge National Laboratory) in the United States managed to build and run the world's first civilian molten salt reactor called MSRE (Molten Salt Reactor Experiment). The MSRE was an experimental liquid-fueled reactor with 10 MW thermal output using {sup 4}LiF-BeF{sub 2}-ZrF{sub 4}-UF{sub 4} as the fuel also as the coolant itself. The MSRE is usually taken as a very important reference case for many current researches to validate their codes and simulations. Without exception it works also as a benchmark for this thesis. The current thesis actually consists of two main parts. The first part is about the validation of the current code for the old MSRE concept, while the second

Radiation exposures associated with shipments of foreign research reactor spent nuclear fuel

International Nuclear Information System (INIS)

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

1999-01-01

In accordance with the Record of Decision on a Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel (ROD) (DOE, 1996a), the U.S. Department of Energy (DOE) is implementing a 13-year program under which DOE accepts foreign research reactor spent nuclear fuel (SNF) containing uranium that was enriched in the United States. The ROD required that DOE take several steps to ensure low environmental and health impacts resulting from the implementation of the program. These efforts mainly focus on transportation related activities that the analysis of potential environmental impacts in the Environmental Impact Statement on a Proposed Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel (EIS) (DOE, 1996b) identified as having the potential for exceeding current radiation protection guidelines. Consequently, DOE issued a Mitigation Action Plan to reduce the likelihood of potential adverse environmental impacts associated with the policy established in the ROD. As shown in the EIS, incident-free radiation exposures to members of the ship's crew, port workers, and ground transportation personnel due to shipments of spent nuclear fuel from foreign research reactors are expected to be below the radiation exposure limit of 100 mrem (1 mSv) per year established to protect the general public. However, the analysis in the EIS demonstrated that port and transportation workers could conceivably receive a cumulative radiation dose above the limit established for the general public if, for example, they are involved in multiple shipments within one year or if the radiation levels outside the casks are at the maximum allowable regulatory limit (10 mrem/hr [0.1 mSv/h] at 2 meters from the surface of the cask). With the program successfully underway, DOE has collected information from the shipments in accordance with the Mitigation Action Plan. The information to date has demonstrated that the analysis in

The Influence of Seed Rate and Fertilizer Type on Growth of Tridax procumbens in Subhumid Nigeria

Directory of Open Access Journals (Sweden)

Onifade, OS.

2001-01-01

Full Text Available Forage production of Tridax procumbens was examined from three seed rates (20, 40 and 60 kg/ha, four types of fertilizer [(unfertilized control, single superphosphate (SSP, 18 % P, NPK (20 : 10 : 10, and calcium ammonium nitrate (CAN, 25 % N] and four harvest stages (6, 9, 12 and 15 weeks post sowing, WPS. Fertilizer rates used were 20 kg/ha P for (SSP and 50 kg/ha N for each of NPK and CAN. Plant height (cm, density (plants/m2 and spread (cm were measured at 3, 6, 9, 12 and 15 WPS. Sampling for yield estimation in each plot was done at 6, 9, 12 and 15 WPS. Mean plant height with respect to seed rate, fertilizer type and growth stage significantly (P 2000 kg DM/ha. Application of SSP, NPK and CAN gave more total DM yields than the unfertilized control plot respectively by about 69, 75 and 75 %. Total DM yield varied from 1024 kg DM/ha when T. procumbens was harvested at an earlier stage (6 WPS to 2130 kg DM/ha at a later harvest (12 WPS. The main CP of leaf was significantly (P <0.05 affected by ail the treatments imposed.