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

Reprocessing of research reactor fuel the Dounreay option

Energy Technology Data Exchange (ETDEWEB)

Cartwright, P.

1997-08-01

Reprocessing is a proven process for the treatment of spent U/Al Research Reactor fuel. At Dounreay 12679 elements have been reprocessed during the past 30 years. For reactors converting to LEU fuel the uranium recovered in reprocessing can be blended down to less than 20% U{sub 235}, enrichment and be fabricated into new elements. For reactors already converted to LEU it is technically possible to reprocess spent silicide fuel to reduce the U{sub 235} burden and present to a repository only stable conditioned waste. The main waste stream from reprocessing which contains the Fission products is collected in underground storage tanks where it is kept for a period of at least five years before being converted to a stable solid form for return to the country of origin for subsequent storage/disposal. Discharges to the environment from reprocessing are low and are limited to the radioactive gases contained in the spent fuel and a low level liquid waste steam. Both of these discharges are independently monitored, and controlled within strict discharge limits set by the UK Government`s Scottish Office. Transportation of spent fuel to Dounreay has been undertaken using many routes from mainland Europe and has utilised over the past few years both chartered and scheduled vessel services. Several different transport containers have been handled and are currently licensed in the UK. This paper provides a short history of MTR reprocessing at Dounreay, and provides information to show reprocessing can satisfy the needs of MTR operators, showing that reprocessing is a valuable asset in non-proliferation terms, offers a complete solution and is environmentally acceptable.

Open problems in reprocessing of a molten salt reactor fuel

International Nuclear Information System (INIS)

Lelek, Vladimir; Vocka, Radim

2000-01-01

The study of fuel cycle in a molten salt reactor (MSR) needs deeper understanding of chemical methods used for reprocessing of spent nuclear fuel and preparation of MSR fuel, as well as of the methods employed for reprocessing of MSR fuel itself. Assuming that all the reprocessing is done on the basis of electrorefining, we formulate some open questions that should be answered before a flow sheet diagram of the reactor is designed. Most of the questions concern phenomena taking place in the vicinity of an electrode, which influence the efficiency of the reprocessing and sensibility of element separation. Answer to these questions would be an important step forward in reactor set out. (Authors)

Technology development of fast reactor fuel reprocessing technology in India

International Nuclear Information System (INIS)

Natarajan, R.; Raj, Baldev

2009-01-01

India is committed to the large scale induction of fast breeder reactors beginning with the construction of 500 MWe Prototype Fast Breeder Reactor, PFBR. Closed fuel cycle is a prerequisite for the success of the fast reactors to reduce the external dependence of the fuel. In the Indian context, spent fuel reprocessing, with as low as possible out of pile fissile inventory, is another important requirement for increasing the share in power generation through nuclear route as early as possible. The development of this complex technology is being carried out in four phases, the first phase being the developmental phase, in which major R and D issues are addressed, while the second phase is the design, construction and operation of a pilot plant, called CORAL (COmpact Reprocessing facility for Advanced fuels in Lead shielded cell. The third phase is the construction and operation of Demonstration of Fast Reactor Fuel Reprocessing Plant (DFRP) which will provide experience in fast reactor fuel reprocessing with high availability factors and plant throughput. The design, construction and operation of the commercial plant (FRP) for reprocessing of PFBR fuel is the fourth phase, which will provide the requisite confidence for the large scale induction of fast reactors

Status of power reactor fuel reprocessing in India

International Nuclear Information System (INIS)

Kansra, V.P.

1999-01-01

Spent fuel reprocessing in India started with the commissioning of the Trombay Plutonium Plant in 1964. This plant was intended for processing spent fuel from the 40 MWth research reactor CIRUS and recovering plutonium required for the research and development activities of the Indian Atomic Energy programme. India's nuclear energy programme aims at the recycle of plutonium in view of the limited national resources of natural uranium and abundant quantities of thorium. This is based on the approach which aims at separating the plutonium from the power reactor spent fuel, use it in the fast reactors to breed 233 U and utilise the 233 U generated to sustain a virtually endless source of power through thorium utilisation. The separated plutonium is also being utilised to fabricate MOX fuel for use in thermal reactors. Spent fuel treatment and extracting plutonium from it makes economic sense and a necessity for the Indian nuclear power programme. This paper describes the status and trends in the Indian programme for the reprocessing of power reactor fuels. The extraction of plutonium can also be seen as a far more positive approach to long-term waste management. The closed cycle approach visualised and pursued by the pioneers in the field is now steadily moving India towards the goal of a sustainable source of power through nuclear energy. The experience in building, operating and refurbishing the reprocessing facilities for uranium and thorium has resulted in acquiring the technological capability for designing, constructing, operating and maintaining reprocessing plants to match India's growing nuclear power programme. (author)

Power Reactor Thoria Reprocessing Facility (PRTRF), Trombay

International Nuclear Information System (INIS)

Dhami, P.S; Yadav, J.S; Agarwal, K.

2017-01-01

Exploitation of the abundant thorium resources to meet sustained energy demand forms the basis of the Indian nuclear energy programme. To gain reprocessing experience in thorium fuel cycle, thoria was irradiated in research reactor CIRUS in early sixties. Later in eighties, thoria bundles were used for initial flux flattening in some of the pressurized heavy water reactors (PHWRs). The research reactor irradiated thoria contained small content (∼ 2-3ppm) of "2"3"2U in "2"3"3U product, which did not pose any significant radiological problems during processing in Uranium Thorium Separation Facility (UTSF), Trombay. Thoria irradiated in PHWRs on discharge contained (∼ 0.5-1.5% "2"3"3U with significant "2"3"2U content (100-500 ppm) requiring special radiological attention. Based on the experience from UTSF, a new facility viz. Power Reactor Thoria Reprocessing Facility (PRTRF), Trombay was built which was hot commissioned in the year 2015

The search for advanced remote technology in fast reactor reprocessing

International Nuclear Information System (INIS)

Burch, W.D.; Herndon, J.N.; Stradley, J.G.

1990-01-01

Research and development in fast reactor reprocessing has been under way about 20 years in several countries throughout the world. During the past decade in France and the United Kingdom, active development programs have been carried out in breeder reprocessing. Actual fuels from their demonstration reactors have been reprocessed in small-scale facilities. Early US work in breeder reprocessing was carried out at the EBR-II facilities with the early metal fuels, and interest has renewed recently in metal fuels. A major, comprehensive program, focused on oxide fuels, has been carried out in the Consolidated Fuel Reprocessing Program (CFRP) at the Oak Ridge National Laboratory (ORNL) since 1974. Germany and Japan have also carried out development programs in breeder reprocessing, and Japan appears committed to major demonstration of breeder reactors and their fuel cycles. While much of the effort in all of these programs addressed process chemistry and process hardware, a significant element of many of these programs, particularly the CFRP, has been on advancements in facility concepts and remote maintenance features. This paper will focus principally on the search for improved facility concepts and better maintenance systems in the CFRP and, in turn, on how developments at ORNL have influenced the technology elsewhere

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Power Reactor Fuel Reprocessing Plant-2, Tarapur: a benchmark in Indian PHWR spent fuel reprocessing

International Nuclear Information System (INIS)

Pradhan, Sanjay; Dubey, K.; Qureshi, F.T.; Lokeswar, S.P.

2017-01-01

Power Reactor Fuel Reprocessing Plant-2 (PREFRE-2) is latest operating spent nuclear fuel reprocessing plant in India. This plant has improved design based on latest technology and feedback provided by the earlier plants. The design of PREFRE-2 plant is in five cycles of solvent extraction using TBP as extractant. The plant is commissioned in year 2011 after regulatory clearances

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.

Review of NRC Commission Papers on Regulatory Basis for Licensing and Regulating Reprocessing Facilities

Energy Technology Data Exchange (ETDEWEB)

Park, Jae Yeong; Shin, Hyeong Ki [KINS, Daejeon (Korea, Republic of)

2016-05-15

Spent nuclear fuel (SNF) accumulated in nuclear power plant has been a serious issue in most countries with operating nuclear power plants. Direct disposal of SNF could be a solution of the problem but many countries including the Republic of Korea have had a hard time selecting a site for high level waste repository because of low public acceptance. SNF recycling technologies consisting of reprocessing and transmutation have been developed so as to reduce the final volume of the disposed radioactive waste and to diminish the radiotoxicity of the waste. The Republic of Korea is now developing pyroprocessing and sodium-cooled fast reactor (SFR) technology to be used for the recycling of the wastes. KAERI has a plan to construct a pyroprocessing facility with a capacity of 30 tHM/y and a facility manufacturing TRU fuel for SFR by 2025. However, to license these facility and secure the safety, the current regulatory system related to SNF treatment needs to be improved and amended since the system has been developed focusing on facilities to examine irradiated nuclear materials. Status of reprocessing facility regulations developed by U.S.NRC was reviewed based on SECY papers. U.S.NRC has approved the development of a new rule referred to nationally as '10CFR Part 7x'. Existing 10CFR 50 and 70 has been evolved mainly for nuclear power plants and fuel cycle facilities whose radiological hazard is much lower than reprocessing plants respectively. U.S.NRC also derived many regulatory gaps including safety assessment methods, technical specification, general design criteria and waste classification and continue to develop the regulatory framework limited in scope to the resolution of Gap 5.

R and D of On-line Reprocessing Technology for Molten-Salt Reactor Systems

International Nuclear Information System (INIS)

Uhlir, Jan; Tulackova, Radka; Chuchvalcova Bimova, Karolina

2006-01-01

The Molten Salt Reactor (MSR) represents one of promising future nuclear reactor concept included in the Generation IV reactors family. The reactor can be operated as the thorium breeder or as the actinide transmuter. However, the future deployment of Molten-Salt Reactors will be significantly dependent on the successful mastering of advanced reprocessing technologies dedicated to their fuel cycle. Here the on-line reprocessing technology connected with the fuel circuit of MSR is of special importance because the reactor cannot be operated for a long run without the fuel salt clean-up. Generally, main MSR reprocessing technologies are pyrochemical, majority of them are fluoride technologies. The proposed flow-sheets of MSR on-line reprocessing are based on a combination of molten-salt / liquid metal extraction and electro-separation processes, which can be added to the gas extraction process already verified during the MSRE project in ORNL. The crucial separation method proposed for partitioning of actinides from fission products is based on successive Anodic dissolution and Cathodic deposition processes in molten fluoride media. (authors)

On-line reprocessing of a molten salt reactor: a simulation tool

International Nuclear Information System (INIS)

Simon, Nicole; Gastaldi, Olivier; Penit, Thomas; Cohin, Olivier; Campion, Pierre-Yves

2008-01-01

The molten salt reactor (MSR) is one of the concepts studied in the frame of GEN IV road-map. Due to the specific features of its liquid fuel, the reprocessing unit may be directly connected to the reactor. A modelling of this unit is presented. The final objective is to create a flexible computer reprocessing code which can use data from neutron calculations and can be coupled to a neutron code. Such a code allows the description of the whole behaviour of MSR, including, in a coupled manner, both the design of the core and the optimised reprocessing scheme effects. (authors)

French experience and prospects in the reprocessing of fast breeder reactor fuels

International Nuclear Information System (INIS)

Megy, J.

1983-06-01

Experience acquired in France in the field of reprocessing spent fuels from fast breeder reactors is recalled. Emphasis is put on characteristics and quantities of spent fuels reprocessed in La Hague and Marcoule facilities. Then reprocessing developments with the realisation of the new pilot plant TOR at Marcoule, new equipments and study of industrial reprocessing units are reviewed [fr

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

Reprocessing ability of high density fuels for research and test reactors

International Nuclear Information System (INIS)

Gay, A.; Belieres, M.

1997-01-01

The development of a new high density fuel is becoming a key issue for Research Reactors operators. Such a new fuel should be a Low Enrichment Uranium (LEU) fuel with a high density, to improve present in core performances. It must be compatible with the reprocessing in an industrial plant to provide a steady back-end solution. Within the framework of a work group CEA/CERCA/COGEMA on new fuel development for Research Reactors, COGEMA has performed an evaluation of the reprocessing ability of some fuel dispersants selected as good candidates. The results will allow US to classify these fuel dispersants from a reprocessing ability point of view. (author)

The integral fast reactor fuels reprocessing laboratory at Argonne National Laboratory, Illinois

International Nuclear Information System (INIS)

Wolson, R.D.; Tomczuk, Z.; Fischer, D.F.; Slawecki, M.A.; Miller, W.E.

1986-09-01

The processing of Integral Fast Reactor (IFR) metal fuel utilizes pyrochemical fuel reprocessing steps. These steps include separation of the fission products from uranium and plutonium by electrorefining in a fused salt, subsequent concentration of uranium and plutonium for reuse, removal, concentration, and packaging of the waste material. Approximately two years ago a facility became operational at Argonne National Laboratory-Illinois to establish the chemical feasibility of proposed reprocessing and consolidation processes. Sensitivity of the pyroprocessing melts to air oxidation necessitated operation in atmosphere-controlled enclosures. The Integral Fast Reactor Fuels Reprocessing Laboratory is described

Coupled study of the Molten Salt Fast Reactor core physics and its associated reprocessing unit

International Nuclear Information System (INIS)

Doligez, X.; Heuer, D.; Merle-Lucotte, E.; Allibert, M.; Ghetta, V.

2014-01-01

Highlights: • The limit on the reprocessing is due to the redox potential control. • Alkali and Earth-alkaline elements do not have to be extracted. • Criticality risks have to be studied in the reprocessing unit. • The neutronics properties are not sensitive to chemical data. • The reprocessing chemistry, from a pure numerical point of view, is an issue. - Abstract: Molten Salt Reactors (MSRs) are liquid-fuel reactors, in which the fuel is also the coolant and flows through the core. A particular configuration presented in this paper called the Molten Salt Fast Reactor consists in a Molten Salt Reactor with no moderator inside the core and a salt composition that leads to a fast neutron spectrum. Previous studies showed that this concept (previously called Thorium Molten Salt Reactor – Nonmoderated) has very promising characteristics. The liquid fuel implies a special reprocessing. Each day a small amount of the fuel salt is extracted from the core for on-site reprocessing. To study such a reactor, the materials evolution within the core has to be coupled to the reprocessing unit, since the latter cleans the salt quasi continuously and feeds the reactor. This paper details the issues associated to the numerical coupling of the core and the reprocessing. It presents how the chemistry is introduced inside the classical Bateman equation (evolution of nuclei within a neutron flux) in order to carry a numerical coupled study. To achieve this goal, the chemistry has to be modeled numerically and integrated to the equations of evolution. This paper presents how is it possible to describe the whole concept (reactor + reprocessing unit) by a system of equations that can be numerically solved. Our program is a connection between MCNP and a homemade evolution code called REM. Thanks to this tool; constraints on the fuel reprocessing were identified. Limits are specified to preserve the good neutronics properties of the MSFR. In this paper, we show that the limit

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)

R and D on fast reactor fuel reprocessing

International Nuclear Information System (INIS)

Subba Rao, R.V.; Vijaya Kumar, V.; Natarajan, R.

2012-01-01

Development of Fast Reactor Fuel Reprocessing technology, with low out of pile inventory, is carried out at the Indira Gandhi Centre for Atomic Research (IGCAR). Based on the successful R and D programme which addressed specific issues of fast reactor fuels, a pilot plant called CORAL was set up. This plant is operational since 2003 and several reprocessing campaigns with spent FBTR fuels of varying burnups have been carried out. Based on the valuable operating experience of CORAL, the design of demonstration fast reactor fuel reprocessing plant (DFRP) and the commercial reprocessing plant, FRP have been taken up. Concurrently R and D efforts are continuing for improving the process and equipment performance apart from reducing the waste volumes and the radiation exposures to the operating personnel. Some important R and D efforts are highlighted in the paper. Reducing the dissolution time is one of the vital area of investigation especially for the high plutonium bearing MOX fuels which are known to dissolve slowly. To address this as well as criticality issues, continuous dissolvers are being developed. Solvent extraction based process is employed for getting highly pure nuclear grade uranium and plutonium. In view of the lower cooling time the fission product activity in the spent fuel is higher, formulation of process flowsheet with reduced number of solvent extraction cycles to improve the decontamination of ruthenium and zirconium without the formation of second organic phase due to plutonium loading, is under investigation. Retention of plutonium in lean organic is another issue to be addressed as otherwise it would lead to further deterioration of the solvent on storage. Several reagents to effectively wash the lean solvent have been investigated and flowsheets have been formulated to recover the retained plutonium with minimum secondary wastes. Partitioning of uranium and plutonium is an important step and methods reported in the literature have inherent

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)

Reprocessing technology for present water reactor fuels

International Nuclear Information System (INIS)

McMurray, P.R.

1977-01-01

The basic Purex solvent extraction technology developed and applied in the U.S. in the 1950's provides a well-demonstrated and efficient process for recovering uranium and plutonium for fuel recycle and separating the wastes for further treatment and packaging. The technologies for confinement of radioactive effluents have been developed but have had limited utilization in the processing of commercial light water reactor fuels. Technologies for solidification and packaging of radioactive wastes have not yet been demonstrated but significant experience has been gained in laboratory and engineering scale experiments with simulated commercial reprocessing wastes and intermediate level wastes. Commercial scale experience with combined operations of all the required processes and equipment are needed to demonstrate reliable reprocessing centers

Benefit analysis of reprocessing and recycling light water reactor fuel

International Nuclear Information System (INIS)

1976-12-01

The macro-economic impact of reprocessing and recycling fuel for nuclear power reactors is examined, and the impact of reprocessing on the conservation of natural uranium resources is assessed. The LWR fuel recycle is compared with a throwaway cycle, and it is concluded that fuel recycle is favorable on the basis of economics, as well as being highly desirable from the standpoint of utilization of uranium resources

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

Probability of Criticality for MOX SNF

International Nuclear Information System (INIS)

P. Gottlieb

1999-01-01

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

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

Current Status of Spent Fast Reactor Fuel Reprocessing and Waste Treatment in Various Countries: United States of America

International Nuclear Information System (INIS)

2011-01-01

Due to the previous strategic US decision on treating SNF as waste and not pursuing the reprocessing option, development work for the FR fuel cycle was only performed in a few laboratories, although interest is now increasing again. ORNL together with ANL have been influential in promoting the wider use of centrifugal contactors (favoured due to the high fissile content and decay power of FR fuel materials), associated remote handling systems and hardware prototypes for most unit operations in the reprocessing conceptual designs in the context of their development of the Consolidated Fuel Reprocessing Program. There is limited experience with reprocessing tests on the Fast Flux Text Facility (FFTF) MOX fuel. ORNL has undertaken small tests on laboratory scale dissolution and solvent extraction of MOX fuel irradiated to 220 GW/t HM burnup at around 2 kg batch scale [180-186]. The initiative called the breeder reprocessing engineering test (BRET) was started in the 1980s with a focus on the developmental activity of the US DOE to demonstrate breeder fuel reprocessing technology while closing the fuel cycle for the FFTF. The process was supposed to be installed at the existing Fuels and Materials Examination Facility (FMEF) at the Hanford Site, Richland, Washington. The major objectives of BRET were to: - Develop and demonstrate reprocessing technology and systems for breeder fuel; - Close the fuel cycle for the FFTF; - Provide an integrated test of breeder reactor fuel cycle technology - reprocessing, safeguards and waste management. The quest for pyrochemical alternatives to aqueous reprocessing has been under way in the USA since the late 1950s. Approaches examined at various levels of development and for a variety of fuels include alloy melting, FP volatilization and adsorption, fluoride and chloride volatility methods, redox solvent extractions between liquid salt and metal phases, precipitation and fractional crystallization, and electrowinning and electro

Simulation tool of the on-line reprocessing unit of a molten salt reactor

International Nuclear Information System (INIS)

Simon, Nicole; Conocar, Olivier; Boussier, Hubert; Gastaldi, Olivier; Penit, Thomas; Walle, Eric; Huguet, Anne

2006-01-01

Molten salt reactor (MSR) is an interesting technology selected in the frame of the Generation IV forum. In the MSR, actinides are diluted in a molten salt which is both the fuel and the coolant. The ability of such a reactor is the reducing of the long-lived wastes due to high burn-up and the on-site simplified reprocessing directly connected with the core which preserve the salt properties necessary for its correct operation. Here is defined a flexible computer reprocessing code which can use data from neutronic calculations and can be coupled to a neutronic code. The code allow the description the whole behaviour of MSR, including, a coupled manner, both the design of the core and the optimised reprocessing scheme effects. (authors)

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)

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

International Nuclear Information System (INIS)

Clark, William D. Jr.

1997-01-01

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

DOE SNF technology development necessary for final disposal

International Nuclear Information System (INIS)

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

1996-01-01

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

Conceptual design study on advanced aqueous reprocessing system for fast reactor fuel cycle

International Nuclear Information System (INIS)

Takata, Takeshi; Koma, Yoshikazu; Sato, Koji; Kamiya, Masayoshi; Shibata, Atsuhiro; Nomura, Kazunori; Ogino, Hideki; Koyama, Tomozo; Aose, Shin-ichi

2003-01-01

As a feasibility study on commercialized fast reactor cycle system, a conceptual design study is being progressed for the aqueous and pyrochemical processes from the viewpoint of economical competitiveness, efficient utilization of resources, decreasing environmental impact and proliferation resistance in Japan Nuclear Cycle Development Institute (JNC). In order to meet above-mentioned requirements, the survey on a range of reprocessing technologies and the evaluation of conceptual plant designs against targets for the future fast reactor cycle system have been implemented as the fist phase of the feasibility study. For an aqueous reprocessing process, modification of the conventional PUREX process (a solvent extraction process with purification of U/Pu, with nor recovery of minor actinides (MA)) and investigation of alternatives for the PUREX process has been carried out and design study of advanced aqueous reprocessing system and its alternatives has been conducted. The conceptual design of the advanced aqueous reprocessing system has been updated and evaluated by the latest R and D results of the key technologies such as crystallization, single-cycle extraction, centrifugal contactors, recovery of Am/Cm and waste processing. In this paper, the outline of the design study and the current status of development for advanced aqueous reprocessing system, NEXT process, are mentioned. (author)

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

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

Technical aspects of fuel reprocessing

International Nuclear Information System (INIS)

Groenier, W.S.

1982-02-01

The purpose of this paper is to present a brief description of fuel reprocessing and some present developments which show the reliability of nuclear energy as a long-term supply. The following topics are discussed: technical reasons for reprocessing; economic reasons for reprocessing; past experience; justification for advanced reprocessing R and D; technical aspects of current reprocessing development. The present developments are mainly directed at the reprocessing of breeder reactor fuels but there are also many applications to light-water reactor fuel reprocessing. These new developments involve totally remote operation, and maintenance. To demonstrate this advanced reprocessing concept, pilot-scale demonstration facilities are planned with commercial application occurring sometime after the year 2000

Fast breeder reactor fuel reprocessing R and D: technological development for a commercial plant

International Nuclear Information System (INIS)

Colas, J.; Saudray, D.; Coste, J.A.; Roux, J.P.; Jouan, A.

1987-01-01

The technological developments undertaken by the CEA are applied to a plant project of a 50 t/y capacity, having to reprocess in particular the SUPERPHENIX 1 reactor fuel. French experience on fast breeder reactor fuel reprocessing is presented, then the 50 t/y capacity plant project and the research and development installations. The R and D programs are described, concerning: head-end operations, solvent extractions, Pu02 conversion and storage, out-of-specification Pu02 redissolution, fission products solution vitrification, conditioning of stainless steel hulls by melting, development of remote operation equipments, study of corrosion and analytical problems

Economic evaluation of reprocessing and thermal reactor recycle

International Nuclear Information System (INIS)

Marshall, W.

This paper provides a summing up of the discussions on economic aspects in WG4. These discussions also took account of the strategic, ecological and public acceptance factors intimately involved. Tentative conclusions are put forward as a basis for discussion. Reprocessing may take place for reasons other than just strictly economic ones. The decisions facing various countries are rationalized by considering their reactions to the range of possible uranium prices and fast reactor costs in the future

Available Reprocessing and Recycling Services for Research Reactor Spent Nuclear Fuel

International Nuclear Information System (INIS)

2017-01-01

The high enriched uranium (HEU) take back programmes will soon have achieved their goals. When there are no longer HEU inventories at research reactors and no commerce in HEU for research reactors, the primary driver for the take back programmes will cease. However, research reactors will continue to operate in order to meet their various mission objectives. As a result, inventories of low enriched uranium spent nuclear fuel will continue to be created during the research reactors' lifetime and, therefore, there is a need to develop national final disposition routes. This publication is designed to address the issues of available reprocessing and recycling services for research reactor spent fuel and discusses the various back end management aspects of the research reactor fuel cycle.

Reprocessing techniques of LWR spent fuel for reutilization in hybrid systems and IV generation reactors

Energy Technology Data Exchange (ETDEWEB)

Aruquipa, Wilmer; Velasquez, Carlos E.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Costa, Antonella L. [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Barros, Graiciany de P. [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

2017-07-01

Since the era of nuclear technology begins, nuclear reactors have been produced spent fuel. This spent fuel contains material that could be recycle and reprocessed by different processes. All these processes aim to reduce the contribution to the final repository through the re-utilization of the nuclear material. Therefore, some new reprocessing options with non-proliferation characteristics have been proposed and the goal is to compare the different techniques used to maximize the effectiveness of the spent fuel utilization and to reduce the volume and long-term radiotoxicity of high-level waste by irradiation with neutron with high energy such as the ones created in hybrid reactors. In order to compare different recovery methods, the cross sections of fuels are calculated with de MCNP code, the first set consists of thorium-232 spiked with the reprocessed material and the second set in depleted uranium that containing 4.5% of U-235 spiked with the reprocessed material; These sets in turn are compared with the cross section of the UO{sub 2} in order to evaluate the efficiency of the reprocessed fuel as nuclear fuel. (author)

Current status and directions for fast reactor reprocessing

International Nuclear Information System (INIS)

Burch, W.D.

1983-01-01

The development of fast breeder reactors (FBRs) for commercial electric power production has been under way in several countries for more than 20 years. In the United States (US), as elsewhere, early work was focused on small reactors to prove the feasibility of concepts and later was followed by larger reactors to test engineering features and to develop fuel technology. Because of the perceived crisis in electrical generation expected late in this century, major efforts (including fuel cycle activities) were developed in the early 1970s to ensure the capability of developing and using this new form of nuclear power. However, because of the effects of the oil price rise and subsequent emphasis on conservation, and a slowdown of industrial growth, there has been a decline in such activities, particularly in the US, which was at one time (1970s) the world leader in reactor development. This paper provides a brief history of breeder reprocessing and describes the current status, with emphasis on US programs and glimpses into the future

Radiological considerations in the design of Reprocessing Uranium Plant (RUP) of Fast Reactor Fuel Cycle Facility (FRFCF), Kalpakkam

International Nuclear Information System (INIS)

Chandrasekaran, S.; Rajagopal, V.; Jose, M.T.; Venkatraman, B.

2012-01-01

A Fast Reactor Fuel Cycle Facility (FRFCF) being planned at Indira Gandhi Centre for Atomic Research, Kalpakkam is an integrated facility with head end and back end of fuel cycle plants co-located in a single place, to meet the refuelling needs of the prototype fast breeder reactor (PFBR). Reprocessed uranium oxide plant (RUP) is one such plant in FRFCF to built to meet annual requirements of UO 2 for fabrication of fuel sub-assemblies (FSAs) and radial blanket sub-assemblies (RSAs) for PFBR. RUP receives reprocessed uranium oxide powder (U 3 O 8 ) from fast reactor fuel reprocessing plant (FRP) of FRFCF. Unlike natural uranium oxide plant, RUP has to handle reprocessed uranium oxide which is likely to have residual fission products activity in addition to traces of plutonium. As the fuel used for PFBR is recycled within these plants, formation of higher actinides in the case of plutonium and formation of higher levels of 232 U in the uranium product would be a radiological problem to be reckoned with. The paper discussed the impact of handling of multi-recycled reprocessed uranium in RUP and the radiological considerations

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

Fast reactor fuel reprocessing development in the United States: an overview

International Nuclear Information System (INIS)

Groenier, W.S.; Burch, W.D.

1979-01-01

As a result of the reduced nuclear power demand and the growing concerns over the potential proliferation of sensitive nuclear materials, there has not been a necessity to make immediate decisions regarding near-term reprocessing and breeder reactor commercialization. Programs which formed the basic thrust of nuclear development in the early 1970's have already been adjusted: increased emphasis on problems of radioactive waste management; increased attention to nonproliferation objectives and subsequent reorientation of the overall fuel cycle and breeder programs; increased emphasis on a once-through light-water reactor technology; increased concern for a more detailed knowledge of the uranium resource base; reorientation of the uranium enrichment programs; and exploration of alternative fuel cycles (such as thorium) to minimize the use of plutonium. Nevertheless, major strategic decisions still loom over breeder commercialization, the breeder's requisite demand for reprocessing, and the future role of more proliferation-resistant nuclear technologies. The current program in the United States is organized to provide the necessary technology for the reprocessing of breeder fuels on a timetable that is consistent with the reactor development and demonstration program. Also addressed in this paper are the present day concerns of environmental protection, safety, nuclear material safeguards, and proliferation resistance. It is structured on the well-known Purex processing method but includes new efforts aimed at advanced and alternative fuels. At the present time, the program consists mainly of a generic effort that is planned to progress through an integrated equipment engineering demonstration to an eventual pilot-plant operation. Each of these facilities is viewed as a test bed for advanced and alternative processing steps to address the many significant technical and political issues. 16 figures

Remote maintenance in TOR fast reactor fuel reprocessing facility

International Nuclear Information System (INIS)

Eymery, R.; Constant, M.; Malterre, G.

1986-11-01

The TOR facility which is undergoing commissioning tests has a capacity of 5 T. HM/year which is enough for reprocessing all the Phenix fuel, with an excess capacity which is to be used for other fast reactors fuels. It is the result of enlargement and renovation of the old Marcoule pilot facility. A good load factor is expected through the use of equipment with increased reliability and easy maintenance. TOR will also be used to test new equipment developed for the large breeder fuel reprocessing plant presently in the design stage. The latter objective is specifically important for the parts of the plant involving mechanical equipment which are located in a new building: TOR 1. High reliability and flexibility will be obtained in this building thanks to the attention given to the integrated remote handling system [fr

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)

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

International Nuclear Information System (INIS)

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

2004-01-01

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

The measurement of neptunium in fast reactor fuel reprocessing

International Nuclear Information System (INIS)

Mair, M.A.; Savage, D.J.; Kyffin, T.W.

1986-02-01

Analytical techniques have been developed to measure neptunium in the feed, waste and product streams of a fast reactor fuel reprocessing plant. The estimated level of one microgram per milligram of plutonium in some solutions presented severe separation and measurement problems. An initial separation stage was essential, and both ion exchange and solvent extraction using thenoyltrifluoroacetone were studied. The redox chemistry of neptunium necessary to achieve good separation is considered. Spectrophotometry measurement of the stable neptunium/arsenazo III complex was selected for the final neptunium determination with additional analysis by radiometric methods. Incomplete recovery of neptunium during the separation stages necessitated yield measurements, using either neptunium-237 as an internal standard or the short lived gamma active neptunium-239 isotope as a tracer. The distribution of neptunium between the waste and product streams is discussed, in relation to the chemistry of neptunium in the reprocessing plant. (author)

Reprocessing decision

International Nuclear Information System (INIS)

Heising, C.D.

1978-01-01

The United States must decide whether to permit, delay, or prohibit the reprocessing and recycling of nuclear spent fuel. To permit reprocessing would allow recycle as early as 1985; to delay the decision for a later administration to deal with means spent fuel would mount up at nuclear reactor sites; to prohibit would eliminate recycling and mandate permanent storage. Bayesian decision analysis was used to examine reprocessing costs associated with risks and economic benefits. Three distinct categories of risk that are important in the nuclear fuel cycle are discussed. These are: health, environment, and safety risks; nuclear theft and sabotage; and nuclear weapons proliferation risks. Results are discussed from comparing nine routes to weapons-usuable mterial available to nonweapons states that desire a nuclear capability. These are: production reactor and military reporcessor; research reacotr and military reprocessor; power plant plus military reprocessor or commercial reprocessor; enrichment (centrifuge, gaseous diffusion, electromagnetic separation, or aerodynamic jet cascade); and accelerator. It was found that the commercial power reactor-commercial reprocessor route is comparatively unattractive to a nonweapons state. In summary, allowing nuclear fuel reprocessing to go forward in the United States can be expected to increase the costs to society by a maximum $360 million a year. This is approximately one-seventh of the expected benefit (reduced electricity bills) to be dderived by society from closing the fuel cycle. It appears that the permitting reprocessing now is logically preferable to delaying or prohibiting the technology, the author concludes

The French R and D programme for fast reactor fuel reprocessing

International Nuclear Information System (INIS)

Auchapt, P.; Bourgeois, M.; Calame-Longjean, A.; Miquel, P.; Sauteron, J.

1979-01-01

The process employed is the Purex process adapted to the specific case of fast breeder reactor fuels. The results achieved have demonstrated that the aqueous method can be applied to these fuels: nearly ten years of operation in the ATl workshop which reprocesses RAPSODIE fuels, and the good results obtained at the Marcoule pilot facility on large batches of fuel attest to this achievement. The CEA effort continues principally on extrapolation to industrial scale, thanks mainly to experiments conducted on industrial prototypes and to the launching of the TOR project, which will, as of 1984, allow reprocessing of FBR fuels on a significant scale, and which will provide extensive additional resources for R and D activities

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

Revisit of analytical methods for the process and plant control analyses during reprocessing of fast reactor fuels

International Nuclear Information System (INIS)

Subba Rao, R.V.

2016-01-01

CORAL (COmpact facility for Reprocessing of Advanced fuels in Lead cell) is an experimental facility for demonstrating the reprocessing of irradiated fast reactor fuels discharged from the Fast Breeder Test Reactor (FBTR). The objective of the reprocessing plant is to achieve nuclear grade plutonium and uranium oxides with minimum process waste volumes. The process flow sheet for the reprocessing of spent Fast Reactor Fuel consists of Transport of spent fuel, Chopping, Dissolution, Feed conditioning, Solvent Extraction cycle, Partitioning Cycle and Re-conversion of Plutonium nitrate and uranium nitrate to respective oxides. The efficiency and performance of the plant to achieve desired objective depends on the analyses of various species in the different steps adopted during reprocessing of fuels. The analytical requirements in the plant can be broadly classified as 1. Process control Analyses (Analyses which effect the performance of the plant- PCA); 2. Plant control Analyses (Analyses which indicates efficiency of the plant-PLCA); 3. Nuclear Material Accounting samples (Analyses which has bearing on nuclear material accounting in the plant - NUMAC) and Quality control Analyses (Quality of the input bulk chemicals as well as products - QCA). The analytical methods selected are based on the duration of analyses, precision and accuracies required for each type analytical requirement classified earlier. The process and plant control analyses requires lower precision and accuracies as compared to NUMAC analyses, which requires very high precision accuracy. The time taken for analyses should be as lower as possible for process and plant control analyses as compared to NUMAC analyses. The analytical methods required for determining U and Pu in process and plant samples from FRFR will be different as compared to samples from TRFR (Thermal Reactor Fuel Reprocessing) due to higher Pu to U ratio in FRFR as compared TRFR and they should be such that they can be easily

An assessment of KW Basin radionuclide activity when opening SNF canisters

International Nuclear Information System (INIS)

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

1995-01-01

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

Experimental research subject and renovation of chemical processing facility (CPF) for advanced fast reactor fuel reprocessing technology development

International Nuclear Information System (INIS)

Koyama, Tomozo; Shinozaki, Tadahiro; Nomura, Kazunori; Koma, Yoshikazu; Miyachi, Shigehiko; Ichige, Yoshiaki; Kobayashi, Tsuguyuki; Nemoto, Shin-ichi

2002-01-01

In order to enhance economical efficiency, environmental impact and nuclear nonproliferation resistance, the Advanced Reprocessing Technology, such as simplification and optimization of process, and applicability evaluation of the innovative technology that was not adopted up to now, has been developed for the reprocessing of the irradiated fuel taken out from a fast reactor. Renovation of the hot cell interior equipments, establishment and updating of glove boxes, installation of various analytical equipments, etc. in the Chemical Processing Facility (CPF) was done to utilize the CPF more positivity which is the center of the experimental field, where actual fuel can be used, for research and development towards establishment of the Advanced Reprocessing Technology development. The hot trials using the irradiated fuel pins of the experimental fast reactor 'JOYO' for studies on improved aqueous reprocessing technology, MA separation technology, dry process technology, etc. are scheduled to be carried out with these new equipments. (author)

The reprocessing of fast reactor fuels - the TOR project

International Nuclear Information System (INIS)

Calame-Longjean, A.; Le Bouhellec, J.; Schwob, Y.

1982-01-01

A description is given of development work on the proposed new French facility for the reprocessing of fast reactor fuel. This is the TOR facility (Traitement des Oxydes Rapides). Block diagrams give details of the TOR project as a whole and of the main line and R and D line of the TOR 1 facility which is a new works devoted to the head of the process. Modifications to existing plant which will form the TOR 2 and TOR 3 facilities are also described. (U.K.)

Catalogue and classification of technical safety rules for light-water reactors and reprocessing plants

International Nuclear Information System (INIS)

Bloser, M.; Fichtner, N.; Neider, R.

1975-08-01

This report on the cataloguing and classification of technical rules for land-based light-water reactors and reprocessing plants contains a list of classified rules. The reasons for the classification system used are given and discussed

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

Measurement and behaviour of technetium in fast reactor fuel reprocessing

International Nuclear Information System (INIS)

Ferguson, C.; Kyffin, T.W.

1986-02-01

A method is described for the spectrophotometric measurement of technetium in plant solutions from the reprocessing of fast reactor fuel. The technetium is selectively extracted using tri-iso-octylamine. After back extraction, thiocyanate is added, in the presence of tetrabutyl-ammonium hydroxide, to form the red hexa-thiocyanato anionic complex in a chloroform medium. The concentration of the technetium is then calculated from the spectrophotometric measurement of this complex. This method was applied to bulk samples, collected during a PFR fuel reprocessing campaign, to identify the main routes followed by technetium through the reprocessing plant. In order to understand the probable behaviour of technetium in the process plant streams, an investigation into the influence of plutonium IV nitrate on the extraction of Tc (VII) into 20%v/v tributyl phosphate/odourless kerosene solution from nitric acid solutions, was initiated. The results of this investigation, along with the known distribution coefficient for the extraction of the uranyl/technetium complex U0 2 (N0 3 )(Tc0 4 ).2TBP and the redox chemistry of technetium, are used to predict the probable behaviour of technetium in the process plant streams. This predicted behaviour is compared with the experimental results and reasonable agreement is obtained between experiment and theory, considering the history of the samples analysed. (author)

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

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

International Nuclear Information System (INIS)

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

1997-01-01

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

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

Pyrochemical reprocessing of molten salt fast reactor fuel: focus on the reductive extraction step

Directory of Open Access Journals (Sweden)

Rodrigues Davide

2015-12-01

Full Text Available The nuclear fuel reprocessing is a prerequisite for nuclear energy to be a clean and sustainable energy. In the case of the molten salt reactor containing a liquid fuel, pyrometallurgical way is an obvious way. The method for treatment of the liquid fuel is divided into two parts. In-situ injection of helium gas into the fuel leads to extract the gaseous fission products and a part of the noble metals. The second part of the reprocessing is performed by ‘batch’. It aims to recover the fissile material and to separate the minor actinides from fission products. The reprocessing involves several chemical steps based on redox and acido-basic properties of the various elements contained in the fuel salt. One challenge is to perform a selective extraction of actinides and lanthanides in spent liquid fuel. Extraction of actinides and lanthanides are successively performed by a reductive extraction in liquid bismuth pool containing metallic lithium as a reductive reagent. The objective of this paper is to give a description of the several steps of the reprocessing retained for the molten salt fast reactor (MSFR concept and to present the initial results obtained for the reductive extraction experiments realized in static conditions by contacting LiF-ThF4-UF4-NdF3 with a lab-made Bi-Li pool and for which extraction efficiencies of 0.7% for neodymium and 14.0% for uranium were measured. It was concluded that in static conditions, the extraction is governed by a kinetic limitation and not by the thermodynamic equilibrium.

Reprocessing in breeder fuel cycles

International Nuclear Information System (INIS)

Burch, W.D.; Groenier, W.S.

1982-01-01

Over the past decade, the United States has developed plans and carried out programs directed toward the demonstration of breeder fuel reprocessing in connection with the first breeder demonstration reactor. A renewed commitment to moving forward with the construction of the Clinch River Breeder Reactor (CRBR) has been made, with startup anticipated near the end of this decade. While plans for the CRBR and its associated fuel cycle are still being firmed up, the basic research and development programs required to carry out the demonstrations have continued. This paper updates the status of the reprocessing plans and programs. Policies call for breeder recycle to begin in the early to mid-1990's. Contents of this paper are: (1) evolving plans for breeder reprocessing (demonstration reprocessing plant, reprocessing head-end colocated at an existing facility); (2) relationship to LWR reprocessing; (3) integrated equipment test (IET) facility and related hardware development activities (mechanical considerations in shearing and dissolving, remote operations and maintenance demonstration phase of IET, integrated process demonstration phase of IET, separate component development activities); and (4) supporting process R and D

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)

Why reprocess

International Nuclear Information System (INIS)

Hagen, M.

1977-01-01

The problem of whether to reprocess spent nuclear fuel elements has been studied already in the early days of the commercial utilization of nuclear power and has been answered positively. This also, and in particular, applies to the United States. Under the new American nuclear policy reprocessing is rejected only for reasons of non-proliferation. Although these are valid reasons, the effectiveness of a ban on reprocessing, as fas as the non-profileration of nuclear weapons is concerned, is not accepted worldwide because the necessary knowledge either already exists in many countries or can be obtained. Only if there had been a realistic chance to prevent the proliferation of nuclear weapons, also the other industrialized countries would have seconded the policy of the United States. A country like the Federal Republic of Germany, with a substantial long-term nuclear power program based initially on light water reactors, subsequently on advanced reactor systems, cannot do without a complete nuclear fuel cycle. This reasoning is outlined in the light of economic and radioecological aspects. Extensive experience on a technical scale is available in the reprocessing sector. The technical problems associated with this activity have been solved in principle and have largely been demonstrated to function in practice. (orig.) [de

Pyro-electrochemical reprocessing of irradiated MOX fast reactor fuel, testing of the reprocessing process with direct MOX fuel production

Energy Technology Data Exchange (ETDEWEB)

Kormilitzyn, M.V.; Vavilov, S.K.; Bychkov, A.V.; Skiba, O.V.; Chistyakov, V.M.; Tselichshev, I.V

2000-07-01

One of the advanced technologies for fast reactor fuel recycle is pyro-electrochemical molten salt technology. In 1998 we began to study the next phase of the irradiated oxide fuel reprocessing new process MOX {yields} MOX. This process involves the following steps: - Dissolution of irradiated fuel in molten alkaline metal chlorides, - Purification of melt from fission products that are co-deposited with uranium and plutonium oxides, - Electrochemical co-deposition of uranium and plutonium oxides under the controlled cathode potential, - Production of granulated MOX (crushing,salt separation and sizing), and - Purification of melt from fission products by phosphate precipitation. In 1998 a series of experiments were prepared and carried out in order to validate this process. It was shown that the proposed reprocessing flowsheet of irradiated MOX fuel verified the feasibility of its decontamination from most of its fission products (rare earths, cesium) and minor-actinides (americium, curium)

Pyro-electrochemical reprocessing of irradiated MOX fast reactor fuel, testing of the reprocessing process with direct MOX fuel production

International Nuclear Information System (INIS)

Kormilitzyn, M.V.; Vavilov, S.K.; Bychkov, A.V.; Skiba, O.V.; Chistyakov, V.M.; Tselichshev, I.V.

2000-01-01

One of the advanced technologies for fast reactor fuel recycle is pyro-electrochemical molten salt technology. In 1998 we began to study the next phase of the irradiated oxide fuel reprocessing new process MOX → MOX. This process involves the following steps: - Dissolution of irradiated fuel in molten alkaline metal chlorides, - Purification of melt from fission products that are co-deposited with uranium and plutonium oxides, - Electrochemical co-deposition of uranium and plutonium oxides under the controlled cathode potential, - Production of granulated MOX (crushing,salt separation and sizing), and - Purification of melt from fission products by phosphate precipitation. In 1998 a series of experiments were prepared and carried out in order to validate this process. It was shown that the proposed reprocessing flowsheet of irradiated MOX fuel verified the feasibility of its decontamination from most of its fission products (rare earths, cesium) and minor-actinides (americium, curium)

High temperature gas-cooled reactor (HTGR) graphite pebble fuel: Review of technologies for reprocessing

Energy Technology Data Exchange (ETDEWEB)

Mcwilliams, A. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2015-09-08

This report reviews literature on reprocessing high temperature gas-cooled reactor graphite fuel components. A basic review of the various fuel components used in the pebble bed type reactors is provided along with a survey of synthesis methods for the fabrication of the fuel components. Several disposal options are considered for the graphite pebble fuel elements including the storage of intact pebbles, volume reduction by separating the graphite from fuel kernels, and complete processing of the pebbles for waste storage. Existing methods for graphite removal are presented and generally consist of mechanical separation techniques such as crushing and grinding chemical techniques through the use of acid digestion and oxidation. Potential methods for reprocessing the graphite pebbles include improvements to existing methods and novel technologies that have not previously been investigated for nuclear graphite waste applications. The best overall method will be dependent on the desired final waste form and needs to factor in the technical efficiency, political concerns, cost, and implementation.

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

Prospects of Using Reprocessed Uranium in CANDU Reactors, in the U.S. GNEP Program

International Nuclear Information System (INIS)

Ellis, Ronald James

2007-01-01

Current Global Nuclear Energy Partnership (GNEP) plans envision reprocessing spent fuel (SF) with view to minimizing high-level waste (HLW) repository use and recovering actinides (U, Np, Pu, Am, and Cm) for transmutation in reactors as fuel and targets. The reprocessed uranium (RU), however, is to be disposed of. This paper presents a limited-scope analysis of possible reuse of RU in CANDU (Canada Deuterium Uranium) Reactors, within the context of the US GNEP program. Other papers on this topic submitted to this conference discuss the possibility of RU reuse in light-water reactors (LWRs) (with enrichment) and offer an independent economic analysis of RU reuse. A representative RU uranium 'vector', from reprocessed spent LWR fuel, comprises 98.538 wt% 238U, 0.46 wt% 236 U, 0.986 wt% 235 U, and 0.006 wt% 234 U. After multiple recyclings, the concentration of 234 U can approach 0.02 wt%. The presence of 234 U and 236 U in RU reduces the reactivity and fuel lifetime (exit burnup), which is particularly an issue in LWRs. While in PWR analyses, the burnup penalty caused by the concentration of 236 U in RU needs to be offset by additional 235 U enrichment in the amount of ∼25% to 30% of the weight percentage of the 236 U; however, the effect in CANDU is much smaller. Furthermore, since the 235 U content in RU exceeds that of natural uranium, CANDU offers the advantageous option of uranium recycling without reenrichment. The exit burnup of CANDU RU-derived fuel is considerably larger than that for natural uranium-fueled scenario, despite the presence of 234 U and 236 U.

Cryotechnics used for activity retention in nuclear reactors and in the reprocessing of nuclear fuels

International Nuclear Information System (INIS)

Cavalieri, L.; Gutowski, H.; Patzelt, A.; Schroeder, E.

The use of low temperature technology for cleaning waste gas is demonstrated by four examples from nuclear technology: 1. Treatment of the gaseous fission products and oxyhydrogen gas components (catalytic H 2 oxidation, cold drying, Kr/Xe delay by adsorption) in the primary circuit of a 1300 MWe light water reactor. 2. Cleaning of the helium cooling gas of a high temperature reactor (delay adsorption for short-lived Kr/Xe isotopes, CO/H 2 oxidation CO 2 /H 2 O adsorption and finally low temperature adsorption of Kr/Xe/N 2 and N 2 recondensation). 3. Reprocessing of heavy water moderator by separation of the tritium with low temperature rectification and 4. Cleaning of solvent waste gas of a reprocessing plant (precleaning with various stages of process for separating aerosols, oil, Kr/Xe/N 2 , H 2 , NOsub(x) and low temperature rectification of krypton). (RB) [de

Consolidated fuel reprocessing program

Science.gov (United States)

1985-04-01

A survey of electrochemical methods applications in fuel reprocessing was completed. A dummy fuel assembly shroud was cut using the remotely operated laser disassembly equipment. Operations and engineering efforts have continued to correct equipment operating, software, and procedural problems experienced during the previous uranium compaigns. Fuel cycle options were examined for the liquid metal reactor fuel cycle. In high temperature gas cooled reactor spent fuel studies, preconceptual designs were completed for the concrete storage cask and open field drywell storage concept. These and other tasks operating under the consolidated fuel reprocessing program are examined.

The economic influence of reprocessing strategy in the early stages of a commercial breeder programme

International Nuclear Information System (INIS)

Pounder, F.

1982-01-01

The effect on reprocessing cost of constructing fast reactors in nuclear parks is examined and compared with carrying out reprocessing for a range of installation programmes of fast reactor in central reprocessing facilities. Consideration is also given to the economics of storing irradiated fuel to improve the load factor of reprocessing plants and to reprocessing both thermal reactor and fast reactor fuel in a common plant. (author)

Statement on the Consolidated Fuel Reprocessing Program

International Nuclear Information System (INIS)

Trauger, D.B.

1984-01-01

Oak Ridge National Laboratory has chosen the following objectives for future reprocessing plant design: reduced radiation exposure to workers; minimal environmental impact; improved plant operation and maintenance; improved accountability; no plutonium diversion; and reduced overall capital and operating cost. These objectives lead to a plant with totally remote operation. The Breeder Reactor Engineering Test (BRET) has been designed to perform a key role in demonstrating advanced reprocessing technology. It has been scheduled to be available to reprocess spent fuel from the Fast Flux Test Facility. The principal features of the Consolidated Fuel Reprocessing Program and of the BRET facility are appropriate for all reactor types

Reprocessing: experience and future outlooks

International Nuclear Information System (INIS)

Rapin, M.

1981-01-01

It is shown that reprocessing is the best way to cope with irradiated fuels since it provides an optimized waste conditioning for long term storage, the possibility to recycle fissile material and the reduction of Pu diversion risk. The reprocessing constraints are discussed from political, technical, safety, public acceptance, and economical points of view. The French reprocessing programme (thermal reactor fuel fast breeder fuels) is presented together with a short review of the reprocessing experience and outlooks out of France [fr

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

Reprocessing of ''fast'' fuel in France

International Nuclear Information System (INIS)

Sauteron, J.; Bourgeois, M.; Le Bouhellec, J.; Miquel, P.

1976-05-01

The results of laboratory studies as well as pilot testing (AT-I La Hague, Marcoule, Fontenay-aux-Roses) in reprocessing of fast breeder reactor fuels are described. The paper covers all steps: head end, aqueous and fluoride volatility processes, and waste treatment. In conclusion, it is demonstrated why it is still too early to define a strategy of industrial reprocessing for this reactor type

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.

Fuel reprocessing experience in India: Technological and economic considerations

International Nuclear Information System (INIS)

Prasad, A.N.; Kumar, S.V.

1983-01-01

The approach to the reprocessing of irradiated fuel from power reactors in India is conditioned by the non-availability of highly enriched uranium with the consequent need for plutonium for the fast-reactor programme. With this in view, the fuel reprocessing programme in India is developing in stages matching the nuclear power programme. The first plant was set up in Trombay to reprocess the metallic uranium fuel from the research reactor CIRUS. The experience gained in the construction and operation of this plant, and in its subsequent decommissioning and reconstruction, has not only provided the know-how for the design of subsequent plants but has indicated the fruitful areas of research and development for efficient utilization of limited resources. The Trombay plant also handled successfully, on a pilot scale, the reprocessing of irradiated thorium fuel to separate uranium-233. The second plant at Tarapur has been built for reprocessing spent fuels from the power reactors at Tarapur (BWR) and Rajasthan (PHWR). The third plant, at present under design, will reprocess the spent fuels from the power reactors (PHWR) and the Fast Breeder Test Reactor (FBTR) located at Kalpakkam. Through the above approach experience has been acquired which will be useful in the design and construction of even larger plants which will become necessary in the future as the nuclear power programme grows. The strategies considered for the sizing and siting of reprocessing plants extend from the idea of small plants, located at nuclear power station sites, to a large-size central plant, located at an independent site, serving many stations. The paper discusses briefly the experience in reprocessing uranium and thorium fuels and also in decommissioning. An attempt is made to outline the technological and economic aspects which are relevant under different circumstances and which influence the size and siting of the fuel reprocessing plants and the expected lead times for construction

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

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

Operational experiences in radiation protection in fast reactor fuel reprocessing facility

International Nuclear Information System (INIS)

Meenakshisundaram, V.; Rajagopal, V.; Santhanam, R.; Baskar, S.; Madhusoodanan, U.; Chandrasekaran, S.; Balasundar, S.; Suresh, K.; Ajoy, K.C.; Dhanasekaran, A.; Akila, R.; Indira, R.

2008-01-01

The Compact Reprocessing facility for Advanced fuels in Lead cells (CORAL), situated at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam is a pilot plant to reprocess the mixed carbide fuel, for the first time in the world. Reprocessing of fuel with varying burn-ups up to 155 G Wd/t, irradiated at Fast Breeder Test Reactor (FBTR), has been successfully carried out at CORAL. Providing radiological surveillance in a fuel reprocessing facility itself is a challenging task, considering the dynamic status of the sources and the proximity of the operator with the radioactive material and it is more so in a fast reactor fuel reprocessing facility due to handling of higher burn-up fuels associated with radiation fields and elevated levels of fissile material content from the point of view of criticality hazard. A very detailed radiation protection program is in place at CORAL. This includes, among others, monitoring the release of 85 Kr and other fission products and actinides, if any, through stack on a continuous basis to comply with the regulatory limits and management of disposal of different types of radioactive wastes. Providing radiological surveillance during the operations such as fuel transport, chopping and dissolution and extraction cycle was without any major difficulty, as these were carried out in well-shielded and high integrity lead cells. Enforcement of exposure control assumes more importance during the analysis of process samples and re-conversion operations due to the presence of fission product impurities and also since the operations were done in glove boxes and fume hoods. Although the radiation fields encountered in process area were marginally higher, due to the enforcement of strict administrative controls, the annual exposure to the radiation workers was well within the regulatory limit. As the facility is being used as test bed for validation of prototype equipment, periodic inspection and maintenance of components such as centrifuge

Development, experience and innovation in reprocessing

International Nuclear Information System (INIS)

Delange, M.

1985-01-01

The author describes landmarks in the development of the reprocessing industry in France and then presents objectives for the future (extension of reprocessing of fuel from breeder reactors) together with the technological resources deployed to attain them [fr

Indian experience in fuel reprocessing

International Nuclear Information System (INIS)

Prasad, A.N.; Kumar, S.V.

1977-01-01

Plant scale experience in fuel reprocessing in India was started with the successful design, execution and commissioning of the Trombay plant in 1964 to reprocess aluminium clad metallic uranium fuel from the 40 MWt research reactor. The plant has helped in generating expertise and trained manpower for future reprocessing plants. With the Trombay experience, a larger plant of capacity 100 tonnes U/year to reprocess spent oxide fuels from the Tarapur (BWR) and Rajasthan (PHWR) power reactors has been built at Tarapur which is undergoing precommissioning trial runs. Some of the details of this plant are dealt with in this paper. In view of the highly corrosive chemical attack the equipment and piping are subjected to in a fuel reprocessing plant, some of them require replacement during their service if the plant life has to be extended. This calls for extensive decontamination for bringing the radiation levels low enough to establish direct accesss to such equipment. For making modifications in the plant to extend its life and also to enable expansion of capacity, the Trombay plant has been successfully decontaminated and partially decommissioned. Some aspects of thi decontamination campaign are presented in this paper

Environmental evaluation of reprocessing

International Nuclear Information System (INIS)

1979-01-01

This paper addresses two specific points. (a) The means by which it is established that reprocessing is carried out within the basic standards for radiological protection set by the ICRP. (b) A summary of the products, wastes and effluents of reprocessing together with the energy and water resources required. It is concluded that reprocessing of spent thermal reactor fuel can be undertaken whilst conforming to the basic standards set by ICRP. For domestic reasons of public acceptability some countries adopt very strict limits. Any attempt at comparisons between limits set by individual countries could lead to misunderstandings if account is not taken of these additional factors which may in turn influence the cost of reprocessing

Reprocessing RTR fuel in the La Hague plants

Energy Technology Data Exchange (ETDEWEB)

Thomasson, J. [Cogema, 78 - Velizy Villacoublay (France); Drain, F.; David, A. [SGN, 78 - Saint Quentin en Yveline (France)

2001-07-01

Starting in 2006, research reactors operators will be fully responsible for their research and testing reactors spent fuel back-end management. It appears that the only solution for this management is treatment-conditioning, which could be done at the La Hague reprocessing complex in France. The fissile material can be separated in the reprocessing plants and the final waste can be encapsulated in a matrix adapted to its potential hazards. RTR reprocessing at La Hague would require some modifications, since the plant had been primarily designed to reprocess fuel from light water reactors. Many provisions have been taken at the plant design stage, however, and the modifications would be feasible even during active operations, as was done from 1993 to 1995 when a new liquid waste management was implemented, and when one of the two vitrification facilities was improved. To achieve RTR back-end management, COGEMA and its partners are also conducting R and D to define a new generation of LEU fuel with performance characteristics approximating those of HEU fuel. This new-generation fuel would be easier to reprocess. (author)

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

Recycling of reprocessed uranium

International Nuclear Information System (INIS)

Randl, R.P.

1987-01-01

Since nuclear power was first exploited in the Federal Republic of Germany, the philosophy underlying the strategy of the nuclear fuel cycle has been to make optimum use of the resource potential of recovered uranium and plutonium within a closed fuel cycle. Apart from the weighty argument of reprocessing being an important step in the treatment and disposal of radioactive wastes, permitting their optimum ecological conditioning after the reprocessing step and subsequent storage underground, another argument that, no doubt, carried weight was the possibility of reducing the demand of power plants for natural uranium. In recent years, strategies of recycling have emerged for reprocessed uranium. If that energy potential, too, is to be exploited by thermal recycling, it is appropriate to choose a slightly different method of recycling from the one for plutonium. While the first generation of reprocessed uranium fuel recycled in the reactor cuts down natural uranium requirement by some 15%, the recycling of a second generation of reprocessed, once more enriched uranium fuel helps only to save a further three per cent of natural uranium. Uranium of the second generation already carries uranium-232 isotope, causing production disturbances, and uranium-236 isotope, causing disturbances of the neutron balance in the reactor, in such amounts as to make further fabrication of uranium fuel elements inexpedient, even after mixing with natural uranium feed. (orig./UA) [de

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

Fuel salt reprocessing influence on the MSFR behavior and on its associated reprocessing unit

International Nuclear Information System (INIS)

Doligez, X.

2010-10-01

In order to face with the growing of the energy demand, the nuclear industry has to reach the fourth generation technology. Among those concept, molten salt reactor, and especially the fast neutron spectrum configuration, seems very promising: indeed breeding is achievable while the feedback coefficient are still negative. However, the reprocessing salt scheme is not totally set down yet. A lot of uncertainties remain on chemical properties of the salt. Thanks to numerical simulation we studied the behavior of the molten Salt Fast Reactor coupled to a nominal reprocessing unit. We are now able to determine heat transfer and radiation in each elementary step of the unit and, by this way determine those that need special study for radioprotection. We also studied which elements are fundamental to extract for the reactor operation. Finally, we present a sensibility analysis of the chemical uncertainties to few relevant properties of the reactor behavior. (author)

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

Reprocessing of LEU silicide fuel at Dounreay

International Nuclear Information System (INIS)

Cartwright, P.

1996-01-01

UKAEA have recently reprocessed two LEU silicide fuel elements in their MTR fuel reprocessing plant at Dounreay. The reprocessing was undertaken to demonstrate UKAEA's commitment to the world-wide research reactor communities future needs. Reprocessing of LEU silicide fuel is seen as a waste treatment process, resulting in the production of a liquid feed suitable for conditioning in a stable form of disposal. The uranium product from the reprocessing can be used as a blending feed with the HEU to produce LEU for use in the MTR cycle. (author)

Spent fuel reprocessing options

International Nuclear Information System (INIS)

2008-08-01

The objective of this publication is to provide an update on the latest developments in nuclear reprocessing technologies in the light of new developments on the global nuclear scene. The background information on spent fuel reprocessing is provided in Section One. Substantial global growth of nuclear electricity generation is expected to occur during this century, in response to environmental issues and to assure the sustainability of the electrical energy supply in both industrial and less-developed countries. This growth carries with it an increasing responsibility to ensure that nuclear fuel cycle technologies are used only for peaceful purposes. In Section Two, an overview of the options for spent fuel reprocessing and their level of development are provided. A number of options exist for the treatment of spent fuel. Some, including those that avoid separation of a pure plutonium stream, are at an advanced level of technological maturity. These could be deployed in the next generation of industrial-scale reprocessing plants, while others (such as dry methods) are at a pilot scale, laboratory scale or conceptual stage of development. In Section Three, research and development in support of advanced reprocessing options is described. Next-generation spent fuel reprocessing plants are likely to be based on aqueous extraction processes that can be designed to a country specific set of spent fuel partitioning criteria for recycling of fissile materials to advanced light water reactors or fast spectrum reactors. The physical design of these plants must incorporate effective means for materials accountancy, safeguards and physical protection. Section four deals with issues and challenges related to spent fuel reprocessing. The spent fuel reprocessing options assessment of economics, proliferation resistance, and environmental impact are discussed. The importance of public acceptance for a reprocessing strategy is discussed. A review of modelling tools to support the

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...

Italian experience with pilot reprocessing plants

International Nuclear Information System (INIS)

Cao, S.; Dworschak, H.; Rolandi, G.; Simonetta, R.

1977-01-01

Problems and difficulties recently experienced in the reprocessing technology of high burnup power reactor fuel elements have shown the importance of pilot plant experiments to optimize the separation processes and to test advanced equipment on a representative scale. The CNEN Eurex plant, in Saluggia (Vercelli), with a 50 kg/d thruput, in operation since '71, has completed several reprocessing campaigns on MTR type fuel elements. Two different chemical flowsheets based respectively on TBP and tertiary amines were thoroughly tested and compared: a concise comparative evaluation of the results obtained with the two schemes is given. Extensive modifications have then been introduced (namely a new headend cell equipped with a shear) to make the plant suitable to reprocess power reactor fuels. The experimental program of the plant includes a joint CNEN-AECL reprocessing experiment on CANDU (Pickering) type fuel elements to demonstrate a two cycle, amine based recovery of the plutonium. Later, a stock of high burnup fuel elements from the PWR Trino power station will be reprocessed to recover Pu and U with a Purex type flowsheet. ITREC, the second CNEN experimental reprocessing plant located at Trisaia Nuclear Center (Matera), started active operation two years ago. In the first campaign Th-U mixed oxide fuel elements irradiated in the Elk River reactor were processed. Results of this experiment are reported. ITREC special design features confer a high degree of versability to the plant allowing for substantial equipment modification under remote control conditions. For this reason the plant will be principally devoted in the near future to advanced equipment testing. Along this line high speed centrifugal contactor of a new type developed in Poland will be tested in the plant in the frame of a joint experiment between CNEN and the Polish AEC. Later on the plant program will include experimental campaign on fast reactor fuels; a detailed study on this program is in

Nuclear Fuel Reprocessing

International Nuclear Information System (INIS)

Simpson, Michael F.; Law, Jack D.

2010-01-01

This is a submission for the Encyclopedia of Sustainable Technology on the subject of Reprocessing Spent Nuclear Fuel. Nuclear reprocessing is the chemical treatment of spent fuel involving separation of its various constituents. Principally, it is used to recover useful actinides from the spent fuel. Radioactive waste that cannot be re-used is separated into streams for consolidation into waste forms. The first known application of nuclear reprocessing was within the Manhattan Project to recover material for nuclear weapons. Currently, reprocessing has a peaceful application in the nuclear fuel cycle. A variety of chemical methods have been proposed and demonstrated for reprocessing of nuclear fuel. The two most widely investigated and implemented methods are generally referred to as aqueous reprocessing and pyroprocessing. Each of these technologies is described in detail in Section 3 with numerous references to published articles. Reprocessing of nuclear fuel as part of a fuel cycle can be used both to recover fissionable actinides and to stabilize radioactive fission products into durable waste forms. It can also be used as part of a breeder reactor fuel cycle that could result in a 14-fold or higher increase in energy utilization per unit of natural uranium. Reprocessing can also impact the need for geologic repositories for spent fuel. The volume of waste that needs to be sent to such a repository can be reduced by first subjecting the spent fuel to reprocessing. The extent to which volume reduction can occur is currently under study by the United States Department of Energy via research at various national laboratories and universities. Reprocessing can also separate fissile and non-fissile radioactive elements for transmutation.

Fuel reprocessing/fabrication interface

International Nuclear Information System (INIS)

Benistan, G.; Blanchon, T.; Galimberti, M.; Mignot, E.

1987-01-01

EDF has conducted a major research, development and experimental programme concerning the recycling of plutonium and reprocessed uranium in pressurized water reactors, in collaboration with its major partners in the nuclear fuel cycle industry. Studies already conducted have demonstrated the technical and economic advantages of this recycling, as also its feasibility with due observance of the safety and reliability criteria constantly applied throughout the industrial development of the nuclear power sector in France. Data feedback from actual experience will make it possible to control the specific technical characteristics of MOX and reprocessed uranium fuels to a higher degree, as also management, viewed from the economic standpoint, of irradiated fuels and materials recovered from reprocessing. The next step will be to examine the reprocessing of MOX for reprocessed uranium fuels, either for secondary recycling in the PWR units, or, looking further ahead, in the fast breeders or later generation PWR units, after a storage period of a few years

Trends in fuel reprocessing safety research

International Nuclear Information System (INIS)

Tsujino, Takeshi

1981-01-01

With the operation of a fuel reprocessing plant in the Power Reactor and Nuclear Fuel Development Corporation (PNC) and the plan for a second fuel reprocessing plant, the research on fuel reprocessing safety, along with the reprocessing technology itself, has become increasingly important. As compared with the case of LWR power plants, the safety research in this field still lags behind. In the safety of fuel reprocessing, there are the aspects of keeping radiation exposure as low as possible in both personnel and local people, the high reliability of the plant operation and the securing of public safety in accidents. Safety research is then required to establish the safety standards and to raise the rate of plant operation associated with safety. The following matters are described: basic ideas for the safety design, safety features in fuel reprocessing, safety guideline and standards, and safety research for fuel reprocessing. (J.P.N.)

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

Technical study report on reprocessing system

International Nuclear Information System (INIS)

Tanaka, Hiroshi; Kasai, Yoshimitsu; Kawamura, Fumio

2000-07-01

The feasibility studies on commercialized fast breeder reactor (FBR) recycle system, in which reactor system and recycle technologies are related on consideration, are performed considering the attainable perspectives for the followings: ensuring safety, economic competitiveness to future LWRs, efficient utilization of resources, reduction of environmental burden, and enhancement of nuclear non-proliferation. In the studies of Reprocessing system of FBR, evaluation was made for the non-aqueous reprocessing of pyrochemical and fluoride volatility process and also for the aqueous reprocessing with a view to streamlining. As a result, it was estimated that each system has prospects of coming into practical use in terms of technique. In the economical efficiency assessment, it was estimated to have economic competitiveness to future LWRs. And the technical research items of each system are picked out. Hereafter, more detail design study will be performed for each system. (author)

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

Direction of reprocessing technology development based on 30 years operation of Tokai reprocessing plant

International Nuclear Information System (INIS)

Nomura, S; Tanaka, T.; Ohshima, H.

2006-01-01

Full text: Full text: Recent global interest focuses the possibility of recycling of spent fuel with advanced fast reactor fuel cycle system. Goal of closed fuel cycle is to achieve the maximum use of uranium resources and minimum disposal of waste by multi recycle of TRU as a competitive nuclear energy system. The future reprocessing and fuel fabrication system should be synchronized completely with the advanced reactor system and waste treatment and disposal back-end system to complete closed fuel cycle. To realize such system, current reprocessing system should be changed to handle Pu-U-Minor Actinide with more reductions in the cost and less waste volume, as well as an inherent proliferation resistance. For the successful industrialization of advanced reprocessing technology, it is necessary to combine three key elements of R and D efforts, engineering base demonstration and experiences of plant operation. Tokai Reprocessing Facilities licensed a maximum capacity of 0.7tHM/day began a hot operation in 1977 and reprocessed l,100tHM U02 spent fuel and 20tHM ATR-MOX with a continuous technological improvements under IAEA full scope safeguards. With 30 years experience, candidate of key technologies proposed for realizing the next advanced reprocessing are as follows: 1) Simplified co-extraction process of Pu-Np-U by using multistage centrifugal extractors in stead of pulsed columns; 2) Corrosion free components in acid condition by using corrosion resistant refractory alloys and ceramics; 3) Co-conversion technology to MA containing MOX powder by micro-wave heating method for a short process for MA containing MOX pellets fabrication; 4) Advanced verification of high level radioactive liquid waste combining separation technology of TRU and LLFP elements; 5) Advanced chemical analysis and monitoring system for TRU elements in a plant. These advanced reprocessing technologies will be applied mainly to reprocess the LWR spent fuel accumulated past and future

MOX fuel reprocessing and recycling

International Nuclear Information System (INIS)

Guillet, J.L.

1990-01-01

This paper is devoted to the reprocessing of MOX fuel in UP2-800 plant at La Hague, and to the MOX successive reprocessing and recycling. 1. MOX fuel reprocessing. In a first step, the necessary modifications in UP2-800 to reprocess MOX fuel are set out. Early in the UP2-800 project, actions have been taken to reprocess MOX fuel without penalty. They consist in measures regarding: Dissolution; Radiological shieldings; Nuclear instrumentation; Criticality. 2. Mox successive reprocessing and recycling. The plutonium recycling in the LWR is now a reality and, as said before, the MOX fuel reprocessing is possible in UP2-800 plant at La Hague. The following actions in this field consist in verifying the MOX successive reprocessing and recycling possibilities. After irradiation, the fissile plutonium content of irradiated MOX fuel is decreased and, in this case, the re-use of plutonium in the LWR need an important increase of initial Pu enrichment inconsistent with the Safety reactor constraints. Cogema opted for reprocessing irradiated MOX fuel in dilution with the standard UO2 fuel in appropriate proportions (1 MOX for 4 UO2 fuel for instance) in order to save a fissile plutonium content compatible with MOX successive recycling (at least 3 recyclings) in LWR. (author). 2 figs

Power Reactor Fuel Reprocessing Plant-1: a stepping stone in Indian PHWR spent fuel reprocessing

International Nuclear Information System (INIS)

Pradhan, Sanjay; Dubey, K.; Qureshi, F.T.; Lokeswar, S.P.

2017-01-01

India has low reserves of uranium and high reserves of thorium. In order to optimize resource utilization India has adopted a closed fuel cycle to ensure long-term energy security. The optimum resource utilization is feasible only by adopting reprocessing, conditioning and recycle options. It is very much imperative to view spent fuel as a vital resource material and not a waste to be disposed off. Thus, spent nuclear fuel reprocessing forms an integral part of the Indian Nuclear Energy Programme. Aqueous reprocessing based on PUREX technology is in use for more than 50 years and has reached a matured status

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

Operating experience in reprocessing

International Nuclear Information System (INIS)

Schueller, W.

1983-01-01

Since 1953, reprocessing has accumulated 180 years of operating experience in ten plants, six of them with 41 years of operation in reprocessing oxide fuel from light water reactors. After abortive, premature attempts at what is called commercial reprocessing, which had been oriented towards the market value of recoverable uranium and plutonium, non-military reprocessing technologies have proved their technical feasibility, since 1966 on a pilot scale and since 1976 on an industrial scale. Reprocessing experience obtained on uranium metal fuel with low and medium burnups can now certainly be extrapolated to oxide fuel with high burnup and from pilot plants to industrial scale plants using the same technologies. The perspectives of waste management of the nuclear power plants operated in the Federal Republic of Germany should be viewed realistically. The technical problems still to be solved are in a balanced relationship to the benefit arising to the national economy out of nuclear power generation and can be solved in time, provided there are clearcut political boundary conditions. (orig.) [de

Alternate extractants to tributyl phosphate for reactor fuel reprocessing

International Nuclear Information System (INIS)

Crouse, D.J.; Arnold, W.D.; Hurst, F.J.

1983-01-01

Both tri(n-hexyl) phosphate (THP) and tri(2-ethylhexyl) phosphate (TEHP) have some important potential process advantages over TBP for reactor fuel reprocessing. These include negligible aqueous phase solubility and less tendency toward third phase and crud formation. The alkyl chain branching of TEHP makes it much more stable to chemical degradation than TBP and probably also accounts for its much weaker ruthenium extraction. The higher uranium and plutonium extraction power of THP and TEHP allows higher solvent loadings in extraction but makes them somewhat more difficult to strip. The phase separation properties of 1.09 M solutions of THP and TEHP are inferior to those of 1.09 M TBP (30 vol %) but are favorable at lower concentrations. Use of more dilute THP and TEHP solutions is recommended for this reason and to obtain a better balance of extraction power in the extraction versus stripping steps

Reprocessing RTR fuel in the La Hague plants

International Nuclear Information System (INIS)

Thomasson, J.; Drain, F.; David, A.

2001-01-01

Starting in 2006, research reactors operators will be fully responsible for the back-end management of their spent fuel. It appears that the only solution for this management is treatment-conditioning, which could be done at the La Hague reprocessing complex in France. The fissile material can be separated in the reprocessing plants and the final waste can be encapsulated in a matrix adapted to its potential hazards. RTR reprocessing at La Hague would require some modifications, since the plant had been primarily designed to reprocess fuel from light water reactors. Many provisions have been taken at the plant design stage, however, and the modifications would be feasible even during active operations, as was done from 1993 to 1995 when a new liquid waste management was implemented, and when one of the two vitrification facilities was improved. To achieve RTR back-end management, COGEMA and its partners are also conducting R and D to define a new generation of LEU fuel with performance characteristics approximating those of HEU fuel. This new-generation fuel would be easier to reprocess. (author)

Reprocessing RTR fuel in the La Hague plants

Energy Technology Data Exchange (ETDEWEB)

Thomasson, J. [Cogema, F-78140 Velizy (France); Drain, F.; David, A. [SGN, F-78182 Saint Quentin en Yvelines (France)

2001-07-01

Starting in 2006, research reactors operators will be fully responsible for the back-end management of their spent fuel. It appears that the only solution for this management is treatment-conditioning, which could be done at the La Hague reprocessing complex in France. The fissile material can be separated in the reprocessing plants and the final waste can be encapsulated in a matrix adapted to its potential hazards. RTR reprocessing at La Hague would require some modifications, since the plant had been primarily designed to reprocess fuel from light water reactors. Many provisions have been taken at the plant design stage, however, and the modifications would be feasible even during active operations, as was done from 1993 to 1995 when a new liquid waste management was implemented, and when one of the two vitrification facilities was improved. To achieve RTR back-end management, COGEMA and its partners are also conducting R and D to define a new generation of LEU fuel with performance characteristics approximating those of HEU fuel. This new-generation fuel would be easier to reprocess. (author)

Spent fuel reprocessing past experience and future prospects

International Nuclear Information System (INIS)

Megy, J.

1983-09-01

A large experience has been gathered from the early fifties till now in the field of spent fuel reprocessing. As the main efforts in the world have been made for developping the reactors and the fuel fabrication industry to feed them, the spent fuel reprocessing activities came later and have not yet reached the industrial maturity existing to day for plants such as PWRs. But in the principal nuclear countries spent fuel reprocessing is to day considered as a necessity with two simultaneous targets: 1. Recovering the valuable materials, uranium and plutonium. 2. Conditionning the radioactive wastes to ensure safe definitive storage. The paper reviews the main steps: 1. Reprocessing for thermal reactor fuels: large plants are already operating or in construction, but in parallel a large effort of R and D is still under way for improvements. 2. The development of fast breeder plants implies associated fuel reprocessing facilities: pilot plants have demonstrated the closing of the cycle. The main difficulties encountered will be examined and particularly the importance of taking into account the problems of effluents processing and wastes storage [fr

Light water reactor fuel reprocessing and recycling

International Nuclear Information System (INIS)

1977-07-01

This document was originally intended to provide the basis for an environmental impact statement to assist ERDA in making decisions with respect to possible LWR fuel reprocessing and recycling programs. Since the Administration has recently made a decision to indefinitely defer reprocessing, this environmental impact statement is no longer needed. Nevertheless, this document is issued as a report to assist the public in its consideration of nuclear power issues. The statement compares the various alternatives for the LWR fuel cycle. Costs and environmental effects are compared. Safeguards for plutonium from sabotage and theft are analyzed

Reprocessing

International Nuclear Information System (INIS)

Couture, J.; Rougeau, J.-P.

1987-01-01

The course of development of a comprehensive nuclear power industry has its own pace which implies the timely progressive and consistent mastery of each industrial step. In the nuclear fuel it is not surprising that the back-end services have lastly reached the industrial stage. In France, we have now fully completed the industrial demonstration of the closed fuel cycle. Our experience covers all necessary steps : transportation of spent fuel, storage, reprocessing, waste conditioning, recovered uranium recycling, plutonium recycling in thermal MOX fuels, plutonium-based fuel for FBR. While FBR development is a long term target, recycling of fissile materials in present LWR reactors appears to be a source of noticable savings. In the meantime rational management of waste material is the key for increased safety and better environment protection. Reprocessing activity is certainly the major achievement of the back-end strategy. The proven efficiency of this technique as it is implemented at La Hague facility gives the full assurance of a smooth operation of the under completion UP3 unit. The base-load management system which applies during the first ten years of its operation will make possible a noticable reduction of the commercial price for reprocessing services by the end of the century. Industrial maturity being confirmed, economic maturity is now the outstanding merit of the reprocessing and recycling strategy. It is a permanent challenge, to which the response is definitely positive in the sense of reducing the nuclear KWh production cost. (author)

Critical experiment needs and plans of the consolidated fuel reprocessing program

International Nuclear Information System (INIS)

Primm, R.T.

1984-01-01

An integral part of the United States Department of Energy (DOE) plan for the development of breeder reactors is the development of the capability for fuel reprocessing. The Consolidated Fuel Reprocessing Program (CFRP) was established by the DOE to identify and conduct research and development activities in this area. The DOE is currently proposing that a capability to reprocess fast reactor fuel be established in the Fuels and Materials Examination Facility at the Hanford Engineering Development Laboratory. This capability would include conversion of plutonium nitrate to plutonium oxide. The reprocessing line is designated the Breeder Reprocessing Engineering Test (BRET). Criticality safety remains an important critetion in the design of the BRET. The different steps in the reprocessing are reviewed and areas where additional critical experiments are needed have been indentified as also areas where revision or clarification of existing criticality safety standards are desirable

Nuclear fuel reprocessing in the UK

International Nuclear Information System (INIS)

Allardice, R.; Harris, D.; Mills, A.

1983-01-01

Nuclear fuel reprocessing has been carried out on an industrial scale in the United Kingdom since 1952. Two large reprocessing plants have been constructed and operated at Windscale, Cumbria and two smaller specialized plants have been constructed and operated at Dounreay, Northern Scotland. At the present time, the second of the two Windscale plants is operating, and Government permission has been given for a third reprocessing plant to be built on that site. At Dounreay, one of the plants is operating in its original form, whilst the second is now operating in a modified form, reprocessing fuel from the prototype fast reactor. This chapter describes the development of nuclear fuel reprocessing in the UK, commencing with the research carried out in Canada immediately after the Second World War. A general explanation of the techniques of nuclear fuel reprocessing and of the equipment used is given. This is followed by a detailed description of the plants and processes installed and operated in the UK

Nuclear fuel reprocessing in the UK

International Nuclear Information System (INIS)

Allardice, R.H.; Harris, D.W.; Mills, A.

1983-01-01

Nuclear fuel reprocessing has been carried out on an industrial scale in the United Kingdom since 1952. Two large reprocessing plants have been constructed and operated at Windscale, Cumbria and two smaller specialized plants have been constructed and operated at Dounreay, Northern Scotland. At the present time, the second of the two Windscale plants is operating, and Government permission has been given for a third reprocessing plant to be built on that site. At Dounreay, one of the plants is operating in its original form, whilst the second is now operating in a modified form, reprocessing fuel from the prototype fast reactor. This chapter describes the development of nuclear fuel reprocessing in the UK, commencing with the research carried out in Canada immediately after the Second World War. A general explanation of the techniques of nuclear fuel reprocessing and of the equipment used is given. This is followed by a detailed description of the plants and processes installed and operated in the UK. (author)

Reprocessing decision: a study in policymaking under uncertainty

International Nuclear Information System (INIS)

Heising, C.D.

1978-01-01

The U.S. reprocessing decision is examined in this thesis. Decision analysis is applied to develop a rational framework for the assessment of policy alternatives. Benefits and costs for each alternative are evaluated and compared in dollar terms to determine the optimal decision. A fuel cycle simulation model is constructed to assess the economic value of reprocessing light water reactor (LWR) spent fuel and recycling plutonium. In addition, a dynamic fuel substitution model is used to estimate the economic effects of the reprocessing decision's influence on the introduction date of the liquid metal fast breeder reactor (LMFBR). Risks estimated in dollar terms for comparison with the economic values include those related to health, the environment and safety, nuclear theft and sabotage, and nuclear proliferation

Fast reactor fuel reprocessing plant D1206: disassembly cave window 4 replacement

International Nuclear Information System (INIS)

Sutherland, H.G.; Beckitt, S.; Potts, A.B.

1996-01-01

At UKAEA's fast reactor reprocessing plant at Dounreay, the containment glass on the zinc bromide cave viewing window tank failed after 13 years active use. External shielding was fitted and the window tank subsequently drained to make it safe. Fuel cropping operations carried out behind the window were resited to enable cave work to continue whilst a project team made arrangements and plans to replace the damaged window. Because of the complexity of the task and high (alpha, beta, gamma and neutron) radiation levels in excess of 500 Sv/hr a rehearsal facility was built to develop the remote handling techniques to be employed in the task. (UK)

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

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)

Processes for the control of 14CO2 during reprocessing

International Nuclear Information System (INIS)

Notz, K.J.; Holladay, D.W.; Forsberg, C.W.; Haag, G.L.

1980-01-01

The fixation of 14 CO 2 may be required at some future time because of the significant fractional contribution of 14 C, via the ingestion pathway, to the total population dose from the nuclear fuel cycle, even though the actual quantity of this dose is very small when compared to natural background. The work described here was done in support of fuel reprocessing development, of both graphite fuel (HTGRs) and metal-clad fuel (LWRs and LMFBRs), and was directed to the control of 14 CO 2 released during reprocessing operations. However, portions of this work are also applicable to the control of 14 CO 2 released during reactor operation. The work described falls in three major areas: (1) The application of liquid-slurry fixation with Ca(OH) 2 , which converts the CO 2 to CaCO 3 , carried out after treatment of the CO 2 -containing stream to remove other gaseous radioactive components, mainly 85 Kr. This approach is primarily for application to HTGR fuel reprocessing. (2) The above process for CO 2 fixation, but used ahead of Kr removal, and followed by a molecular sieve process to take out the 85 Kr. This approach was developed for use with HTGR reprocessing, but certain aspects also have application to metal-clad fuel reprocessing and to reactor operation. (3) The use of solid Ba(OH) 2 hydrate reacting directly with the gaseous phase. This process is generally applicable to both reprocessing and to reactor operation

Economic evaluation of reprocessing

International Nuclear Information System (INIS)

This paper, which also appears as an Appendix to the Final Working Group 4 report, considers the economics of the four basic options available in nuclear programmes namely: the once-through cycle; reprocessing with uranium recycle and plutonium storage; reprocessing with both uranium and plutonium recycle; and the fast reactor. These options are represented by four separate areas on a ''phase diagram'' showing the relationship between relative generating costs and uranium ore price. The basic algebra defining each component of electricity cost is given for each option. The diagram can take different forms depending upon the relative magnitudes of the costs of reprocessing and MOX fuel fabrication and whether the once-through fuel cycle is acceptable or not on grounds other than strictly economic, i.e. environmental grounds. The shortcomings of this form of presentation are also identified

Study on safety of crystallization method applied to dissolver solution in fast breeder reactor reprocessing

International Nuclear Information System (INIS)

Okuno, Hiroshi; Fujine, Yukio; Asakura, Toshihide; Murazaki, Minoru; Koyama, Tomozo; Sakakibara, Tetsuro; Shibata, Atsuhiro

1999-03-01

The crystallization method is proposed to apply for recovery of uranium from dissolution liquid, enabling to reduce handling materials in later stages of reprocessing used fast breeder reactor (FBR) fuels. This report studies possible safety problems accompanied by the proposed method. Crystallization process was first defined in the whole reprocessing process, and the quantity and the kind of treated fuel were specified. Possible problems, such as criticality, shielding, fire/explosion, and confinement, were then investigated; and the events that might induce accidental incidents were discussed. Criticality, above all the incidents, was further studied by considering exampled criticality control of the crystallization process. For crystallization equipment, in particular, evaluation models were set up in normal and accidental operation conditions. Related data were selected out from the nuclear criticality safety handbooks. The theoretical densities of plutonium nitrates, which give basic and important information, were estimated in this report based on the crystal structure data. The criticality limit of crystallization equipment was calculated based on the above information. (author)

History and current status of nuclear fuel reprocessing technology

International Nuclear Information System (INIS)

Funasaka, Hideyuki; Nagai, Toshihisa; Washiya, Tadahiro

2008-01-01

History and present state of fast breeder reactor was reviewed in series. As a history and current status of nuclear fuel reprocessing technology, this ninth lecture presented the progress of the FBR fuel reprocessing technology and advanced reprocessing processes. FBR fuel reprocessing technology had been developed to construct the reprocessing equipment test facilities (RETF) based on PUREX process technologies. With economics, reduction of environmental burdens and proliferation resistance taken into consideration, advanced aqueous method for nuclear fuel cycle activities has been promoted as the government's basic policy. Innovative technologies on mechanical disassembly, continuous rotary dissolver, crystallizer, solvent extraction and actinides recovery have been mainly studied. (T. Tanaka)

Present status of foreign reprocessing technology

International Nuclear Information System (INIS)

Otagaki, Takao; Ishikawa, Yasusi; Mori, Jyunichi

2000-03-01

In considering extensively and evaluating advanced nuclear fuel recycle technologies then selecting credible one among those technology options and establishing practicable plan of future fast reactor fuel recycle technology, it is important to investigate foreign reprocessing information extensively and minutely as much as possible then to know trends of reprocessing technology development in the world and present technology level of each country. This report is intending to present information of the status and the technology of operating, constructing and closed foreign reprocessing facilities in the world, including, mixed oxide (MOX) fuel reprocessing technology. The conceptual study of 'Foreign Reprocessing Technology Database' was also performed in order to add or revise the information easily. The eight countries, France, The U.K., Russia, The U.S., Germany, Belgium, India and China, were studied regarding outline of the facilities, operation status, future plan, technical information of process flow sheet, primary components, maintenance system etc, construction and operating costs, accidents or troubles, decommissioning status. (author)

Economic evaluation of reprocessing

International Nuclear Information System (INIS)

1978-11-01

This discussion paper first identifies the main factors which influence the economic assessment of reprocessing. It proposes the use of a diagram - the so-called ''phase diagram'' - which plots the fast reactor premium against the price of uranium. The diagram delineates areas where the once-through fuel cycle, thermal recycle and fast reactor will be the preferred choice from micro-economic considerations. The paper then goes on to consider the circumstances under which a country may or may not wish to introduce thermal recycle or fast reactors. Finally, a procedure for further discussion on economic considerations with WG4 is proposed

Safeguards for reprocessing and enrichment plants

International Nuclear Information System (INIS)

1977-01-01

Agency safeguards are entering a new phase with the coming under active safeguards for the first time of reprocessing plants in several regions of the world. This is taking place at a time when not only the safeguards aspect itself is coming under international scrutiny, but also at a time when the necessity of reprocessing plants is being called into question. Attracting less attention at the moment, but potentially of equal significance, are the enrichment plants that soon will be coming under Agency safeguards. It is not unreasonable in view of the present controversies to ask what is the significance of these reprocessing and enrichment plants, what are the problems concerning safeguards that appear to have given rise to the controversies, and how these problems are to be solved. The question of significance is an easy one to answer. The output of these plants is material which some people consider can be used directly for military purposes, whereas the output from other plants, for instance, reactors, would require long and extensive processing before it could be used for military purposes. Like most short answers, this one is an over-simplification which requires some elaboration to make it strictly accurate. For example, the material output of a power reactor is in the form of irradiate assemblies containing plutonium which is potentially of military use if the irradiation had been within a certain range. However, to utilize this plutonium under clandestine conditions, the highly radioactive material would have to be secretly transported to a reprocessing plant and there would have to be simultaneous falsification of the reactor material accounts and the plant records. Such falsification would be difficult to conceal. The total time required to obtain usable plutonium would be many months. Diversion of material from a uranium fabrication plant making fuel for power reactors would be easier physically but strategically it would be of little value. The

Pilot studies of an extraction process for reprocessing of spent fuel from fast reactors: Hardware and process details of extractor selection

International Nuclear Information System (INIS)

Anisimov, V.I.; Pavlovich, V.B.; Smetanin, E.Ya.; Glazunov, N.V.; Shklyar, L.I.; Dubrovskii, V.G.; Serov, A.V.; Zakharkin, B.S.; Konorchenko, V.D.; Korotkov, I.A.; Neumoev, N.V.; Renard, E.V.

1992-01-01

While acknowledging the bold and persistent efforts of U.S. and Russian specialists to develop the concept of pyrochemical reprocessing of spent nuclear fuel from fast reactors on remote-controlled equipment for removal of actinides from the fission products one should recognize that the tasks of reprocessing such fuel can be handled only by using water-extraction technology, especially since the known Purex process continues to be improved to the point that a single-cycle scheme may be developed. This article presents results of pilot studies conducted in hot cells using multistage extractors in continuous counterflow operation; data on various extractor types used in reprocessing spent mixed oxide nuclear fuel; advantages and disadvantages of centrifugal and pulsed column extractor; comparison of column-type and centrifugal extractors; and extraction process

Current liquid metal cooled fast reactor concepts: use of the dry reprocess fuel

International Nuclear Information System (INIS)

Park, Jee Won; Jeong, C. J.; Yang, M. S.

2003-03-01

Recent Liquid metal cooled Fast Reactor (LFR) concepts are reviewed for investigating the potential usability of the Dry Reprocess Fuel (DRF). The LFRs have been categorized into two different types: the sodium cooled and the lead cooled systems. In each category, overall design and engineering concepts are collected which includes those of S-PRISM, AFR300, STAR, ENHS and more. Specially, the nuclear fuel types which can be used in these LFRs, have been summarized and their thermal, physical and neutronic characteristics are tabulated. This study does not suggest the best-matching LFR for the DRF, but shows good possibility that the DRF fuel can be used in future LFRs

Current liquid metal cooled fast reactor concepts: use of the dry reprocess fuel

Energy Technology Data Exchange (ETDEWEB)

Park, Jee Won; Jeong, C. J.; Yang, M. S

2003-03-01

Recent Liquid metal cooled Fast Reactor (LFR) concepts are reviewed for investigating the potential usability of the Dry Reprocess Fuel (DRF). The LFRs have been categorized into two different types: the sodium cooled and the lead cooled systems. In each category, overall design and engineering concepts are collected which includes those of S-PRISM, AFR300, STAR, ENHS and more. Specially, the nuclear fuel types which can be used in these LFRs, have been summarized and their thermal, physical and neutronic characteristics are tabulated. This study does not suggest the best-matching LFR for the DRF, but shows good possibility that the DRF fuel can be used in future LFRs.

Nuclear fuel reprocessing expansion strategies

International Nuclear Information System (INIS)

Gallagher, J.M.

1975-01-01

A description is given of an effort to apply the techniques of operations research and energy system modeling to the problem of determination of cost-effective strategies for capacity expansion of the domestic nuclear fuel reprocessing industry for the 1975 to 2000 time period. The research also determines cost disadvantages associated with alternative strategies that may be attractive for political, social, or ecological reasons. The sensitivity of results to changes in cost assumptions was investigated at some length. Reactor fuel types covered by the analysis include the Light Water Reactor (LWR), High-Temperature Gas-Cooled Reactor (HTGR), and the Fast Breeder Reactor (FBR)

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

Fuel reprocessing at a loss to prove its justification

International Nuclear Information System (INIS)

Traube, K.

1986-01-01

Commercial utilization of nuclear energy is possible with or without fuel reprocessing of spent fuel elements. Demands on terminal storage are about equal in both cases. There is no reason - excluding the military one - to decide in favour of fuel reprocessing instead of direct terminal storage, for neither does fuel reprocessing offer advantages in regard of the safety of nuclear waste disposal, nor is it necessary to produce plutonium for the breeder reactor. Fuel reprocessing is analyzed considering those changed aspects with a view to scarcer uranium resources, juridical motives, and what is termed the development deficit. (DG) [de

Improved annular centrifugal contactor for solvent extraction reprocessing of nuclear reactor fuel

International Nuclear Information System (INIS)

Bernstein, G.J.; Leonard, R.A.; Ziegler, A.A.; Steindler, M.J.

1978-01-01

An improved annular centrifugal contactor has been developed for solvent extraction reprocessing of spent nuclear reactor fuel. The design is an extension of a contactor developed several years ago at Argonne National Laboratory. Its distinguishing features are high throughput, high stage efficiency and the ability to handle a broad range of aqueous-to-organic phase flow ratios and density ratios. Direct coupling of the mixing and separating rotor to a motorized spindle simplifies the design and makes the contactor particularly suitable for remote maintenance. A unit that is critically safe by geometry is under test and a larger unit is being fabricated. Multi-stage miniature contactors operating on the annular mixing principle are being used for laboratory flow sheet studies. 8 figures

Reprocessing of spent plasma

International Nuclear Information System (INIS)

Pierini, G.

1981-01-01

This invention relates to a process for removing helium and other impurities from a mixture containing deuterium and tritium, a deuterium/tritium mixture when purified in accordance with such a process and, more particularly, to a process for the reprocessing of spent plasma removed from a thermofusion reactor. (U.K.)

Possibilities of tritium removal from waste waters of pressurized water reactors and fuel reprocessing plants

International Nuclear Information System (INIS)

Ribnikar, S.V.; Pupezin, J.D.

1975-01-01

Starting from parameters known for heavy water production processes, a parallel was made with separation of tritium from water. The quantity in common is the total cascade flow. The most efficient processes appear to be hydrogen sulfide, water exchange, hydrogen- and water distillation. Prospects of application of new processes are discussed briefly. Problems concerning detritiation of pressurized water reactors and large fuel reprocessing plants are analyzed. Detritiation of the former should not present problems. With the latter, economical detritiation can be achieved only after some plant flow patterns are changed. (U.S.)

Fuel reprocessing data validation using the isotope correlation technique

International Nuclear Information System (INIS)

Persiani, P.J.; Bucher, R.G.; Pond, R.B.; Cornella, R.J.

1990-01-01

The Isotope Correlation Technique (ICT), in conjunction with the gravimetric (Pu/U ratio) method for mass determination, provides an independent verification of the input accountancy at the dissolver or accountancy stage of the reprocessing plant. The Isotope Correlation Technique has been applied to many classes of domestic and international reactor systems (light-water, heavy-water, and graphite reactors) operating in a variety of modes (power, research, and production reactors), and for a variety of reprocessing fuel cycle management strategies. Analysis of reprocessing operations data based on isotopic correlations derived for assemblies in a PWR environment and fuel management scheme, yielded differences between the measurement-derived and ICT-derived plutonium mass determinations of (- 0.02 ± 0.23)% for the measured U-235 and (+ 0.50 ± 0.31)% for the measured Pu-239, for a core campaign. The ICT analyses has been implemented for the plutonium isotopics in a depleted uranium assembly in a heavy-water, enriched uranium system and for the uranium isotopes in the fuel assemblies in light-water, highly-enriched systems

DESIGN ANALYSIS FOR THE NAVAL SNF WASTE PACKAGE

International Nuclear Information System (INIS)

T.L. Mitchell

2000-01-01

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

Reprocessing of spent nuclear fuel

International Nuclear Information System (INIS)

Kidd, S.

2008-01-01

The closed fuel cycle is the most sustainable approach for nuclear energy, as it reduces recourse to natural uranium resources and optimises waste management. The advantages and disadvantages of used nuclear fuel reprocessing have been debated since the dawn of the nuclear era. There is a range of issues involved, notably the sound management of wastes, the conservation of resources, economics, hazards of radioactive materials and potential proliferation of nuclear weapons. In recent years, the reprocessing advocates win, demonstrated by the apparent change in position of the USA under the Global Nuclear Energy Partnership (GNEP) program. A great deal of reprocessing has been going on since the fourties, originally for military purposes, to recover plutonium for weapons. So far, some 80000 tonnes of used fuel from commercial power reactors has been reprocessed. The article indicates the reprocessing activities and plants in the United Kigdom, France, India, Russia and USA. The aspect of plutonium that raises the ire of nuclear opponents is its alleged proliferation risk. Opponents of the use of MOX fuels state that such fuels represent a proliferation risk because the plutonium in the fuel is said to be 'weapon-use-able'. The reprocessing of used fuel should not give rise to any particular public concern and offers a number of potential benefits in terms of optimising both the use of natural resources and waste management.

Cost and availability of gadolinium for nuclear fuel reprocessing plants

International Nuclear Information System (INIS)

Klepper, O.H.

1985-06-01

Gadolinium is currently planned for use as a soluble neutron poison in nuclear fuel reprocessing plants to prevent criticality of solutions of spent fuel. Gadolinium is relatively rare and expensive. The present study was undertaken therefore to estimate whether this material is likely to be available in quantities sufficient for fuel reprocessing and at reasonable prices. It was found that gadolinium, one of 16 rare earth elements, appears in the marketplace as a by-product and that its present supply is a function of the production rate of other more prevalent rare earths. The potential demand for gadolinium in a fuel reprocessing facility serving a future fast reactor industry amounts to only a small fraction of the supply. At the present rate of consumption, domestic supplies of rare earths containing gadolinium are adequate to meet national needs (including fuel reprocessing) for over 100 years. With access to foreign sources, US demands can be met well beyond the 21st century. It is concluded therefore that the supply of gadolinium will quite likely be more than adequate for reprocessing spent fuel for the early generation of fast reactors. The current price of 99.99% pure gadolinium oxide lies in the range $50/lb to $65/lb (1984 dollars). By the year 2020, in time for reprocessing spent fuel from an early generation of large fast reactors, the corresponding values are expected to lie in the $60/lb to $75/lb (1984 dollars) price range. This increase is modest and its economic impact on nuclear fuel reprocessing would be minor. The economic potential for recovering gadolinium from the wastes of nuclear fuel reprocessing plants (which use gadolinium neutron poison) was also investigated. The cost of recycled gadolinium was estimated at over twelve times the cost of fresh gadolinium, and thus recycle using current recovery technology is not economical. 15 refs., 4 figs., 11 tabs

The molten salt reactors (MSR) pyro chemistry and fuel cycle for innovative nuclear systems

International Nuclear Information System (INIS)

Brossard, Ph.; Garzenne, C.; Mouney, H.

2002-01-01

In the frame of the studies on next generation nuclear systems, and especially for the molten salt reactors and for the integrated fuel cycle (as IFR), the fuel cycle constraints must be taken into account in the preliminary studies of the system to improve the cycle and reactor optimisation. Among the purposes for next generation nuclear systems, sustainability and waste (radio-toxicity and mass) management are important goals. These goals imply reprocessing and recycling strategies. The objectives of this workshop are to present and to share the different strategies and scenarios, the needs based on these scenarios, the experimental facilities available today or in the future and their capabilities, the needs for demonstration. It aims at: identifying the needs for fuel cycle based on solid fuel or liquid fuel, and especially, the on-line reprocessing or clean up for the molten salt reactors; assessing the state-of-the-art on the pyro-chemistry applied to solid fuel and to present the research activities; assessing the state-of-the-art on liquid fuels (or others), and to present the research activities; expressing the R and D programs for pyro-chemistry, molten salt, and also to propose innovative processes; and proposing some joint activities in the frame of GEDEON and PRACTIS programs. This document brings together the transparencies of 18 contributions dealing with: scenario studies with AMSTER concept (Scenarios, MSR, breeders (Th) and burners); fuel cycle for innovative systems; current reprocessing of spent nuclear fuel (SNF) in molten salts (review of pyro-chemistry processes (non nuclear and nuclear)); high temperature NMR spectroscopies in molten salts; reductive extraction of An from molten fluorides (salt - liquid metal extraction); electrochemistry characterisation; characterisation with physical methods - extraction coefficient and kinetics; electrolytic extraction; dissolution-precipitation of plutonium in the eutectic LiCl-KCl (dissolution and

Economic evaluation of reprocessing

International Nuclear Information System (INIS)

1979-02-01

This paper presents a progress report of work undertaken relevant to the economic evaluation of reprocessing. It sets out the assumptions to be made for the preparation of the economic ''phase diagram'' - a plot of fast reactor premium against uranium (U 3 O 8 ) price. The paper discusses the assumptions to be made in respect of present worth methodology, LWR fuel logistics, U 3 O 8 price, enrichment tails, plutonium values, fast reactor premium and proposes a set of reference costs to be used for the preparation of the phase diagram

Problem statement: international safeguards for a light-water reactor fuels reprocessing plant

International Nuclear Information System (INIS)

Shipley, J.P.; Hakkila, E.A.; Dietz, R.J.; Cameron, C.P.; Bleck, M.E.; Darby, J.L.

1979-03-01

This report considers the problem of developing international safeguards for a light-water reactor (LWR) fuel reprocessing/conversion facility that combines the Purex process with conversion of plutonium nitrate to the oxide by means of plutonium (III) oxalate precipitation and calcination. Current international safeguards systems are based on the complementary concepts of materials accounting and containment and surveillance, which are designed to detect covert, national diversion of nuclear material. This report discusses the possible diversion threats and some types of countermeasures, and it represents the first stage in providing integrated international safeguards system concepts that make optimum use of available resources. The development of design methodology to address this problem will constitute a significant portion of the subsequent effort. Additionally, future technology development requirements are identified. 8 figures, 1 table

RIAR experimental base development concept 1. Multi-purpose pyrochemical complex for experimental justification of innovative closed fuel cycle technologies

Energy Technology Data Exchange (ETDEWEB)

Bychkov, A.V.; Kormilitsyn, M.V. [Research Institute of Atomic Reactors, Dimitrovgrad-10, Ulyanovsk region, 433510 (Russian Federation)

2009-06-15

The principles of closed FC arrangement on the basis of non-aqueous methods allow the development of production addressing two tasks simultaneously: production of fresh fuel and reprocessing of irradiated fuel, that makes it possible to achieve the industrial level of implementation of closed FC of fast reactors of new generation in a series variant of standardized process modules on the basis of innovative pyrochemical high-effective compact technologies. For the purpose of experimental justification of innovative closed FC technologies at the RIAR site, the existing experimental base is being updated and a multi-purpose pyrochemical complex is developed: - Experimental complex of pyrochemical molten salt facilities to reprocess all types of spent fuel (MOX, nitride, metallic, IMF) of fast reactors of new generation (BN-800, MBIR, BREST). - Experimental complex of facilities to master a gas-fluoride technology of reprocessing intractable fuel, research reactors fuel and thermal SNF. - Transition of the existing facility of pyro-electrochemical production of MOX fuel into the mode of reprocessing of the BN-800 MOX SNF. - Renovation of the facilities for production of fuel elements from experimental, re-fabricated, innovative and high-active fuel - a complex of heavy and glove boxes - to produce experimental fuel elements and targets with MAs on the basis of oxides (vibro and pellets), mixed nitrides, metal alloys and inert matrices in heavy boxes. - Upgrading of the complex for mastering and demonstration of the processes for radioactive waste management and spent fuel pyrochemical reprocessing. The report covers main concept and design solutions, plans and schedule of the program for development of pyrochemical complex for experimental justification of innovative closed FC technologies. (authors)

Nuclear energy without reprocessing: a future made possible by energy conservation

International Nuclear Information System (INIS)

Feiveson, Harold

1987-01-01

The achievements of energy conservation and the promise of future progress have rendered economically unattractive for many decades the separation of plutonium from spent reactor fuel (reprocessing) and its recycle into breeder or light water reactors. Despite this, reprocessing and recycle are already underway in Europe and Japan. If unchecked, these activities, by the end of the century, will lead to an annual flow in routine commerce of tens of thousands of kilograms of separated plutonium, a circumstance which would pose a grave threat to international security. However, continued stagnation of nuclear power programs in much of the world, due significantly to greatly reduced growth rates in electricity demand, will underscore the economic folly of reprocessing and recycle and may eventually encourage their abandonment. (author)

Noble Gas Measurement and Analysis Technique for Monitoring Reprocessing Facilities

International Nuclear Information System (INIS)

William S. Charlton

1999-01-01

An environmental monitoring technique using analysis of stable noble gas isotopic ratios on-stack at a reprocessing facility was developed. This technique integrates existing technologies to strengthen safeguards at reprocessing facilities. The isotopic ratios are measured using a mass spectrometry system and are compared to a database of calculated isotopic ratios using a Bayesian data analysis method to determine specific fuel parameters (e.g., burnup, fuel type, fuel age, etc.). These inferred parameters can be used by investigators to verify operator declarations. A user-friendly software application (named NOVA) was developed for the application of this technique. NOVA included a Visual Basic user interface coupling a Bayesian data analysis procedure to a reactor physics database (calculated using the Monteburns 3.01 code system). The integrated system (mass spectrometry, reactor modeling, and data analysis) was validated using on-stack measurements during the reprocessing of target fuel from a U.S. production reactor and gas samples from the processing of EBR-II fast breeder reactor driver fuel. These measurements led to an inferred burnup that matched the declared burnup with sufficient accuracy and consistency for most safeguards applications. The NOVA code was also tested using numerous light water reactor measurements from the literature. NOVA was capable of accurately determining spent fuel type, burnup, and fuel age for these experimental results. Work should continue to demonstrate the robustness of this system for production, power, and research reactor fuels

Reprocessing free nuclear fuel production via fusion fission hybrids

Energy Technology Data Exchange (ETDEWEB)

Kotschenreuther, Mike, E-mail: mtk@mail.utexas.edu [Intitute for Fusion Studies, University of Texas at Austin (United States); Valanju, Prashant; Mahajan, Swadesh [Intitute for Fusion Studies, University of Texas at Austin (United States)

2012-05-15

Fusion fission hybrids, driven by a copious source of fusion neutrons can open qualitatively 'new' cycles for transmuting nuclear fertile material into fissile fuel. A totally reprocessing-free (ReFree) Th{sup 232}-U{sup 233} conversion fuel cycle is presented. Virgin fertile fuel rods are exposed to neutrons in the hybrid, and burned in a traditional light water reactor, without ever violating the integrity of the fuel rods. Throughout this cycle (during breeding in the hybrid, transport, as well as burning of the fissile fuel in a water reactor) the fissile fuel remains a part of a bulky, countable, ThO{sub 2} matrix in cladding, protected by the radiation field of all fission products. This highly proliferation-resistant mode of fuel production, as distinct from a reprocessing dominated path via fast breeder reactors (FBR), can bring great acceptability to the enterprise of nuclear fuel production, and insure that scarcity of naturally available U{sup 235} fuel does not throttle expansion of nuclear energy. It also provides a reprocessing free path to energy security for many countries. Ideas and innovations responsible for the creation of a high intensity neutron source are also presented.

Reprocessing free nuclear fuel production via fusion fission hybrids

International Nuclear Information System (INIS)

Kotschenreuther, Mike; Valanju, Prashant; Mahajan, Swadesh

2012-01-01

Fusion fission hybrids, driven by a copious source of fusion neutrons can open qualitatively “new” cycles for transmuting nuclear fertile material into fissile fuel. A totally reprocessing-free (ReFree) Th 232 –U 233 conversion fuel cycle is presented. Virgin fertile fuel rods are exposed to neutrons in the hybrid, and burned in a traditional light water reactor, without ever violating the integrity of the fuel rods. Throughout this cycle (during breeding in the hybrid, transport, as well as burning of the fissile fuel in a water reactor) the fissile fuel remains a part of a bulky, countable, ThO 2 matrix in cladding, protected by the radiation field of all fission products. This highly proliferation-resistant mode of fuel production, as distinct from a reprocessing dominated path via fast breeder reactors (FBR), can bring great acceptability to the enterprise of nuclear fuel production, and insure that scarcity of naturally available U 235 fuel does not throttle expansion of nuclear energy. It also provides a reprocessing free path to energy security for many countries. Ideas and innovations responsible for the creation of a high intensity neutron source are also presented.

Development of pulsed plate columns for fast reactor fuel reprocessing

International Nuclear Information System (INIS)

Jenkins, J.A.; Logsdail, D.H.; Lyall, E.; Myers, P.E.; Partridge, B.A.

1987-01-01

The UK Atomic Energy Authority has undertaken a development programme on solvent extraction equipment for reprocessing fast reactor fuels. As part of this programme a solvent extraction pilot plant has been built at Harwell in which a variety of flowsheet conditions can be simulated using the system uranyl nitrate/nitric acid (UN/HNO 3 ) - 20% tri-n-butyl phosphate in odourless kerosene (TBP/OK). The main purpose of present pilot plant operations is to study the performance of pulsed plate columns, with the following specific objectives: to measure the volumetric throughput capacity of the columns, - to study the effect of scale-up of column diameter on U mass transfer performance, - to provide hydraulic and mass transfer data for a dynamic simulation model of pulsed column operation, - to develop and test instruments and ancillary equipment. This poster describes the pilot plant and is illustrated by experimental data, with particular reference to an external settler for controlling the removal of aqueous phase from columns operated with the aqueous phase dispersed

Advances in reprocessing

International Nuclear Information System (INIS)

Giraud, J.P.; Guais, J.C.

1993-01-01

In a comprehensive nuclear energy program based on Light Water Reactor, closing the nuclear fuel cycle by reprocessing the spent fuel and recycling the recovered fissile materials is a key activity which is now fully mastered at the industrial level. In France a large, modern commercial reprocessing plant called UP3 is operating at La Hague since 18 months in excellent conditions regarding products quality, plant availability, safety and waste management. At the same time, industrial capacities for plutonium recycling by MOX fuel fabrication are under operation and larger units are in construction in France and in Europe. Our customers, the utilities which are engaged in a complete closed fuel cycle in Japan, in Germany, Switzerland, Belgium, the Netherlands, and in France, are having a comprehensive industrial system available for their spent fuel management. Three main objectives are being met by this system: (1) saving natural resources by recycling energetic material: plutonium and uranium; (2) solving the waste management question by a segregating the waste according to their characteristics for a proper conditioning, in particular with vitrification for HLW; and (3) preparing the future developments of nuclear power generation with advanced reactors, and best Pu use, and keeping open progresses in long lived waste processing and disposal

Physical and economical aspects of Pu multiple recycling on the basis of REMIX reprocessing technology in thermal reactors

International Nuclear Information System (INIS)

Teplov, P.S.; Alekseev, P.N.; Bobrov, E.A.; Chibinyaev, A.V.

2016-01-01

The basic strategy of Russian nuclear energy is propagation of a closed fuel cycle on the basis of fast breeder and thermal reactors, as well as the solution of the spent nuclear fuel accumulation and resource problems. The 3 variants of multiple Pu and U recycling in Russian pressurized water reactor concept reactors on the basis of Regenerated Mixture of U, Pu oxides (REMIX) reprocessing technology are considered in this work. The REMIX fuel is fabricated from an un-separated mixture of uranium and plutonium obtained during spent fuel reprocessing with further makeup by enriched natural U or reactor grade Pu. This makes it possible to recycle several times the total amount of Pu obtained from the spent fuel. The main difference in Pu recycling is the concept of 100% or partial fuel loading of the core. The third variant is heterogeneous composition of enriched uranium and uranium-plutonium mixed oxide fuel pins in one fuel assembly. It should be noted that all fuel assemblies with Pu require the involvement of expensive technologies during manufacturing. These 3 variants of the full core loadings can be balanced on zero Pu accumulation in the cycle. The various physical and economical aspects of Pu and U multiple recycling in selected variants are presented in the given work. The better results in natural uranium savings can be achieved for the standard REMIX(UOX) approach. The usage of regenerated materials in thermal power reactors gives not more than 30% saving of natural uranium consumption. It is important to note that the use of REMIX technology in the closed fuel cycle is more expensive than the open fuel cycle with direct spent fuel disposal. It is also important to see that uranium multiple recycling in the REMIX fuel form or using the re-enrichment process leads to the degradation of the uranium isotope composition. The "2"3"6U and "2"3"2U concentrations in the fuel are increasing and the regenerated uranium treatment becomes more complicated

Reprocessing in Sweden: History and perspective

International Nuclear Information System (INIS)

Hultgren, Aa.; Oesterlund, C.G.

1990-10-01

Against the background of nuclear power development and installation in Sweden an overview is presented of the parallel domestic development of the reprocessing of spent nuclear fuel. The original selection of the natural uranium - heavy water reactor in the 1950s included spent fuel reprocessing and recycle, and process and plant studies were performed to that end. The switch to light water reactors in the 1960s did not change the planning to recycle; however, the participation in the Eurochemic undertaking, and the delay in the nuclear programme stopped further domestic development work. A number of governmental committee investigations in the 1970s on the radioactive waste issue and, above all, the decision to phase out nuclear power by 2010, after a referendum following the TMI-accident, finally resulted in a decision to plan only for direct disposal of spent nuclear fuel. This policy still prevails. (42 refs.)

Advantages of co-located spent fuel reprocessing, repository and underground reactor facilities

International Nuclear Information System (INIS)

Mahar, James M.; Kunze, Jay F.; Wes Myers, Carl; Loveland, Ryan

2007-01-01

The purpose of this work is to extend the discussion of potential advantages of the underground nuclear park (UNP) concept by making specific concept design and cost estimate comparisons for both present Generation III types of reactors and for some of the modular Gen IV or the GNEP modular concept. For the present Gen III types, we propose co-locating reprocessing and (re)fabrication facilities along with disposal facilities in the underground park. The goal is to determine the site costs and facility construction costs of such a complex which incorporates the advantages of a closed fuel cycle, nuclear waste repository, and ultimate decommissioning activities all within the UNP. Modular power generation units are also well-suited for placement underground and have the added advantage of construction using current and future tunnel boring machine technology. (authors)

Optimal reactor strategy for commercializing fast breeder reactors

International Nuclear Information System (INIS)

Yamaji, Kenji; Nagano, Koji

1988-01-01

In this paper, a fuel cycle optimization model developed for analyzing the condition of selecting fast breeder reactors in the optimal reactor strategy is described. By dividing the period of planning, 1966-2055, into nine ten-year periods, the model was formulated as a compact linear programming model. With the model, the best mix of reactor types as well as the optimal timing of reprocessing spent fuel from LWRs to minimize the total cost were found. The results of the analysis are summarized as follows. Fast breeder reactors could be introduced in the optimal strategy when they can economically compete with LWRs with 30 year storage of spent fuel. In order that fast breeder reactors monopolize the new reactor market after the achievement of their technical availability, their capital cost should be less than 0.9 times as much as that of LWRs. When a certain amount of reprocessing commitment is assumed, the condition of employing fast breeder reactors in the optimal strategy is mitigated. In the optimal strategy, reprocessing is done just to meet plutonium demand, and the storage of spent fuel is selected to adjust the mismatch of plutonium production and utilization. The price hike of uranium ore facilitates the commercial adoption of fast breeder reactors. (Kako, I.)

Problems of nuclear fuel reprocessing in Japan

International Nuclear Information System (INIS)

Tanaka, Naojiro

1974-01-01

The reprocessing capacity of the plant No. 1 of Power Reactor and Nuclear Fuel Development Corporation, which is scheduled to start operation in fiscal year 1975, will be insufficient after fiscal year 1978 for the estimated demand for reprocessing based on Japanese nuclear energy development program. Taking into consideration the results examined by JAIF's study team to Europe and the U.S., it is necessary that Japan builds 2nd reprocessing plant. But there will be a gap from 1978 to 1984 during which Japan must rely on overseas reprocessing services. The establishment of a reprocessing system is a task of national scale, and there are many problems to be solved before it can be done. These include the problems of site and environment, the problem of treatment and disposal of radioactive wastes, the raising of huge required funds and so on. Therefore, even if a private enterprise is allowed to undertake the task, it will be impossible to achieve the aim without the cooperation and assistance of the government. (Wakatsuki, Y.)

Molten salt breeder reactor

International Nuclear Information System (INIS)

1977-01-01

MSBR Study Group formed in October 1974 has studied molten salt breeder reactor and its various aspects. Usage of a molten salt fuel, extremely interesting as reactor chemistry, is a great feature to MSBR; there is no need for separate fuel making, reprocessing, waste storage facilities. The group studied the following, and these results are presented: molten salt technology, molten salt fuel chemistry and reprocessing, reactor characteristics, economy, reactor structural materials, etc. (Mori, K.)

Reprocessing input data validation

International Nuclear Information System (INIS)

Persiani, P.J.; Bucher, R.G.; Pond, R.B.; Cornella, R.J.

1990-01-01

The Isotope Correlation Technique (ICT), in conjunction with the gravimetric (Pu/U ratio) method for mass determination, provides an independent verification of the input accountancy at the dissolver or accountancy stage of the reprocessing plant. The Isotope Correlation Technique has been applied to many classes of domestic and international reactor systems (light-water, heavy-water, graphite, and liquid-metal) operating in a variety of modes (power, research, production, and breeder), and for a variety of reprocessing fuel cycle management strategies. Analysis of reprocessing operations data based on isotopic correlations derived for assemblies in a PWR environment and fuel management scheme, yielded differences between the measurement-derived and ICT-derived plutonium mass determinations of (-0.02 ± 0.23)% for the measured U-235 and (+0.50 ± 0.31)% for the measured Pu-239, for a core campaign. The ICT analyses has been implemented for the plutonium isotopics in a depleted uranium assembly in a heavy-water, enriched uranium system and for the uranium isotopes in the fuel assemblies in light-water, highly-enriched systems. 7 refs., 5 figs., 4 tabs

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.

Management of reprocessed uranium. Current status and future prospects

International Nuclear Information System (INIS)

2007-02-01

There is worldwide interest in developing advanced and innovative technologies for nuclear fuel cycles, minimizing waste and environmental impacts. As of the beginning of 2003, about 171000 tonnes heavy metal spent nuclear fuel is in storage, while smaller amounts have been reprocessed. In several countries, including France, India, Japan and the Russian Federation, spent fuel has been viewed as a national energy resource. Some countries hold reprocessed uranium as the result of their commercial reprocessing service contracts for reprocessing the spent fuel of others. Reprocessed uranium has a potential value for recycling either directly or after appropriate treatment. This report analyses the existing options, approaches and developments in the management of reprocessed uranium. It includes the technical issues involved in managing reprocessed uranium which are RepU arisings, storage, chemical conversion, re-enrichment, fuel fabrication, transport, reactor irradiation, subsequent reprocessing and disposal options, as well as assessment of holistic environmental impacts. The objective of this document is to overview the information on the current status and future trends in the management of RepU and to identify major issues to be considered for future projects

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.

Reprocessing flowsheet and material balance for MEU spent fuel

International Nuclear Information System (INIS)

Abraham, L.

1978-10-01

In response to nonproliferation concerns, the high-temperature gas-cooled reactor (HTGR) Fuel Recycle Development Program is investigating the processing requirements for a denatured medium-enriched uranium--thorium (MEU/Th) fuel cycle. Prior work emphasized the processing requirements for a high-enriched uranium--thorium (HEU/Th) fuel cycle. This report presents reprocessing flowsheets for an HTGR/MEU fuel recycle base case. Material balance data have been calculated for reprocessing of spent MEU and recycle fuels in the HTGR Recycle Reference Facility (HRRF). Flowsheet and mass flow effects in MEU-cycle reprocessing are discussed in comparison with prior HEU-cycle flowsheets

Development of fluoride reprocessing technology for molten salt transmutation reactor systems in the Czech Republic

International Nuclear Information System (INIS)

Uhlir, J.; Hosnedl, P.; Matal, O.

2000-01-01

At present, the transmutation of spent nuclear fuel is considered a prospective alternative conception with respect to the current conception based on the non-reprocessed spent fuel disposal into a deep geological repository. The Czech research and development programme in the area of partitioning is directed primarily on the development of the fuel cycle technology for the accelerator - driven subcritical reactor with a liquid fuel based on fluoride melts. The final objective of the research programme is the development of pyrochemical technologies suitable for a continuous or semi-continuous separation process which would allow practically perfect utilization of the transmutation potentialities of the reactor system. The present research is directed particularly on the development of suitable fluoride separation methods the target of which is the removal of the uranium component from spent nuclear fuel and on the research of the electro-separation procedures and further on the development of appropriate construction materials and equipment for the technology of fluoride salt melts. (authors)

Application of electrochemical techniques in fuel reprocessing- an overview

Energy Technology Data Exchange (ETDEWEB)

Rao, M K; Bajpai, D D; Singh, R K [Power Reactor Fuel Reprocessing Plant, Tarapur (India)

1994-06-01

The operating experience and development work over the past several years have considerably improved the wet chemical fuel reprocessing PUREX process and have brought the reprocessing to a stage where it is ready to adopt the introduction of electrochemical technology. Electrochemical processes offer advantages like simplification of reprocessing operation, improved performance of the plant and reduction in waste volume. At Power Reactor Fuel Reprocessing plant, Tarapur, work on development and application of electrochemical processes has been carried out in stages. To achieve plant scale application of these developments, a new electrochemical cycle is being added to PUREX process at PREFRE. This paper describes the electrochemical and membrane cell development activities carried out at PREFRE and their current status. (author). 5 refs., 4 tabs.

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

Feasibility study for adapting ITREC plant to reprocessing LMFBR fuels

International Nuclear Information System (INIS)

Moccia, A.; Rolandi, G.

1976-05-01

The report evaluates the feasibility of adapting ITREC plant to the reprocessing LMFBR fuels, with the double purpose of: 1) recovering valuable Pu contained in these fuels and recycling it to the fabrication plant; 2) trying, on a pilot scale, the chemical process technology to be applied in a future industrial plant for reprocessing the fuel elements discharged from fast breeder power reactors

The refurbishment of the D1206 fuel reprocessing plant

International Nuclear Information System (INIS)

Bailey, G.

1988-01-01

The term decommissioning can be applied not only to reactors but to any nuclear plant, laboratory, building or part of a building that may have been associated with radioactive material and needs to be restored to clean conditions. In this case the decommissioning and reconstruction of the Dounreay Fast Reactor fuel reprocessing plant, so that plutonium oxide could be reprocessed as well as enriched uranium fuel, is described. The work included improving containment and shielding, building a new head-end treatment cave for the more complex and larger fuel elements, improving the ventilation and constructing a new dissolver. In this paper the breakdown cave and dissolver cell are described and compared and the work done explained. (U.K.)

Head-end reprocessing equipment remote maintenance demonstration

International Nuclear Information System (INIS)

Evans, J.H.; Metz, C.F. III.

1989-01-01

Prototype equipment for reprocessing breeder reactor nuclear fuel was installed in the Remote Operation and Maintenance Demonstration (ROMD) area of the Consolidated Fuel Reprocessing Program (CFRP) facility at the Oak Ridge National Laboratory (ORNL) in order to evaluate the design of this equipment in a cold mock-up of a remotely maintained hot cell. This equipment included the Remote Disassembly System (RDS) and the Remote Shear System (RSS). These systems were disassembled and reassembled remotely by using the extensive remote handling systems that are installed in this simulated hot-cell environment. 5 refs., 5 figs

PYRO, a system for modeling fuel reprocessing

International Nuclear Information System (INIS)

Ackerman, J.P.

1989-01-01

Compact, on-site fuel reprocessing and waste management for the Integral Fast Reactor are based on the pyrochemical reprocessing of metal fuel. In that process, uranium and plutonium in spent fuel are separated from fission products in an electrorefiner using liquid cadmium and molten salt solvents. Quantitative estimates of the distribution of the chemical elements among the metal and salt phases are essential for development of both individual pyrochemical process steps and the complete process. This paper describes the PYRO system of programs used to generate reliable mass flows and compositions

The regulations concerning the reprocessing business of spent fuels

International Nuclear Information System (INIS)

1981-01-01

This rule is stipulated under the provisions of reprocessing business in the law concerning regulation of nuclear raw materials, nuclear fuel materials and nuclear reactors and to execute them. Basic terms are defined, such as exposure radiation dose, cumulative dose, control area, security area, surrounding monitoring area, worker, radioactive waste and facility for discharging into the sea. The application for the designation for reprocessing business under the law shall include the maximum reprocessing capacities per day and per year of each kind of spent fuel, to be reprocessed and the location, structure and equipment of reprocessing facilities as specified in the regulation. Records shall be made in each works or enterprise on the inspection, operation and maintenance of reprocessing facilities, radiation control, accidents and weather, and kept for particular periods respectively. Reprocessing enterprisers shall set up control area, security area and surrounding monitoring area to restrict entrance, etc. Specified measures shall be taken by these enterprisers concerning the exposure radiation doses of workers. Reprocessing facilities shall be inspected and examined more than once a day. The regular self-inspection and operation of reprocessing facilities, the transport and storage of nuclear fuel materials, the disposal of radioactive wastes in works or enterprises where reprocessing facilities are located, and security rules are defined in detail, respectively. (Okada, K.)

Safeguarding a future industrial reprocessing plant

International Nuclear Information System (INIS)

1978-11-01

This paper is submitted to Working Group 5, Sub-Group B for information. It is being submitted to Working Group 4 for discussion at their meeting in January 1979 and shows that by a combination of accountancy, surveillance and containment a reliable safeguards system can be designed for the reprocessing of fuels of the BWR and PWR type. Its arguments can, in general terms, be applied to plants for reprocessing LMFBR fuels, with due allowance for future advances which should improve our overall knowledge of the reliability of safeguards systems. In the reprocessing of fast reactor (LMFBR) fuels, as compared with LWR fuels, the main differences are the higher plutonium concentration and lower heavy metal throughput in the early stages of the reprocessing operations. At later stages in the process (after plutonium/uranium separation) the plants could be similar and have similar safeguarding problems. Plants for reprocessing LMFBR on a commercial scale will not be in operation for a number of years. In these plants greater attention may have to be paid to safeguards at the early stages, especially to waste/raffinate streams, than in the PWR/BWR reprocessing plant. The actual balance between containment, surveillance and accountancy adopted will depend on the status of the technology of safeguards and reprocessing. It can be anticipated that improvements to measurement systems will be made which may allow greater reliance on actual measurement. Treatment and recycle of solid wastes will advance and could therefore lead to improvements in accountancy in, for example, the ''head-end''

Study on remain actinides recovery in pyro reprocessing

International Nuclear Information System (INIS)

Suharto, Bambang

1996-01-01

The spent fuel reprocessing by dry process called pyro reprocessing have been studied. Most of U, Pu and MA (minor actinides) from the spent fuel will be recovered and be fed back to the reactor as new fuel. Accumulation of remain actinides will be separated by extraction process with liquid cadmium solvent. The research was conducted by computer simulation to calculate the stage number required. The calculation's results showed on the 20 stages extractor more than 99% actinides can be separated. (author)

The economics of reprocessing versus direct disposal of spent nuclear fuel

International Nuclear Information System (INIS)

Bunn, M.; Holdren, J.P.; Fetter, S.; Zwaan, B. van der

2007-01-01

The economics of reprocessing versus direct disposal of spent nuclear fuel are assessed. The break-even uranium price at which reprocessing spent nuclear fuel from existing light water reactors (LWRs) and recycling the resulting plutonium and uranium in LWRs would become economic is estimated for a wide range of reprocessing prices and other fuel cycle costs and parameters. The contribution of each fuel cycle option to the cost of electricity is also estimated. A similar analysis is performed for the breakeven uranium price at which deploying fast neutron reactors (FRs) would become competitive compared with a once-through fuel cycle in LWRs, for a range of differences in capital cost between LWRs and FRs. Available information about reprocessing prices and various other fuel cycle costs and input parameters are reviewed, as well as the quantities of uranium likely to be recoverable worldwide at a range of different possible future prices. It is concluded that the once-through fuel cycle is likely to remain significantly cheaper than reprocessing and recycling in either LWRs or FRs for at least the next 50 years. Finally, there is a discussion of how scarce and expensive repository space would have to become before separation and transmutation would be economically attractive. (author)

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 regulations concerning the reprocessing business of spent fuels

International Nuclear Information System (INIS)

1980-01-01

The office ordinance is established under the provisions related to reprocessing businesses of the law concerning regulation of nuclear raw materials, nuclear fuel materials and reactors, to enforce the provisions. The basic terms are defined, such as exposure radiation dose; accumulated dose; controlled area; maintenance area; surrounding watch area; employee; radioactive waste; the facilities for discharge to sea. An application for the designation of reprocessing businesses shall be filed, listing the following matters: the maximum daily and yearly reprocessing capacities for each kind of spent fuel; the location and general structure of reprocessing facilities; the structures of buildings; the structure and equipments of main reprocessing facilities, the storage facilities for products and the disposal facilities for radioactive wastes; the equipments of measuring and control system facilities and radiation control facilities, etc. Records shall be made on the inspection of reprocessing facilities, radiation control, operation, maintenance, the accidents of reprocessing facilities and weather, and kept for the period from one to ten years, respectively. Any person engaging in reprocessing businesses shall set up control, maintenance and surrounding watch areas, and take specified measures to restrict the entrance of persons. The measures to be taken against exposure radiation dose, the inspection, regular independent examination and operation of reprocessing facilities and other related matters are stipulated in detail. (Okada, K.)

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...

Industrial experience of irradiated nuclear fuel reprocessing

International Nuclear Information System (INIS)

Delange, M.

1981-01-01

At the moment and during the next following years, France and La Hague plant particularly, own the greatest amount of industrial experience in the field of reprocessing, since this experience is referred to three types of reactors, either broadly spread all through the world (GCR and LWR) or ready to be greatly developed in the next future (FBR). Then, the description of processes and technologies used now in France, and the examination of the results obtained, on the production or on the security points of view, are a good approach of the actual industrial experience in the field of spent fuel reprocessing. (author)

Legal basis of Russian origin irradiated WWER nuclear fuel import to the Russian Federation

International Nuclear Information System (INIS)

Kanashov, B.; Dorofeev, A.; Komarov, S.; Smirnov, V.; Kolupaev, D.; Kriger, O.

2008-01-01

In the process of spent nuclear fuel (SNF) returning from Armenia, Bulgaria, Hungary, Slovakia, Ukraine, Finland and the Czech Republic to the Russian Federation the following issues have to be considered: 1) Does the legal opportunity of SNF import to the RF exist? 2) Does the technical opportunity for SNF acceptance at reprocessing or disposal facility exist? 3) What are the basic conditions for SNF import? 4) What are the basic conditions for return or retaining of reprocessing products including RAW? The first issue is a legal one and has to be resolved within the framework of federal laws, RF government regulations and international agreements. The second issue is normative-technical. It is regulated by documents of Rostechnadzor (Federal agency on ecological, technological and nuclear supervision), federal norms and regulations in the field of atomic energy usage, industry standards, and in case they are absent, by technical specifications for SNF supply. The last two issues are resolved in the process of drafting foreign trade contracts on SNF import. Generally, Russian regulatory framework is developed enough to regulate SNF import and handling, even in most complicated cases. Nevertheless, when foreign trade contracts on SNF import being drafted there may be disputed regarding both SNF import and RAW return. This report concerns the RF legal and regulatory basis on terms and conditions of SNF import, interim storage, reprocessing and reprocessing products handling in the RF. (authors)

Policy in France regarding the back-end of the fuel cycle reprocessing/recycling route

International Nuclear Information System (INIS)

Gloaguen, A.; Lenail, B.

1991-01-01

The decision taken in early 1970s to base the French power policy on the use of pressurized water reactors also included the strategy for the back end of the nuclear fuel cycle based on reprocessing, waste conditioning for the final disposal in the most suitable form in terms of safety and plutonium recycling to fast breeder reactors. Twenty years have elapsed, and substantial development and investment have been made. New evidences have emerged especially regarding breeder development, and the initial choice has been proved to be sound. EDF and COGEMA, the French utility and fuel cycle companies, respectively, are working together in order to take the best advantage of past efforts. The good behavior of MOX fuel in EDF reactors and the excellent start of the UP3 reprocessing plant of La Hague, which was completed and commissioned in August, 1990, made EDF and COGEMA extremely confident for future decision. The French choice made in favor of fuel reprocessing the history of fuel reprocessing in France, the policy concerning the back end of nuclear fuel cycle of EDF, and the present consideration and circumstances on this matter are reported. (K.I.)

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.

Long-term effects of a shortsighted us political decision banning commercial waste reprocessing

International Nuclear Information System (INIS)

Winterle, B.T.

2003-01-01

In 1977, US President Jimmy Carter ordered a drastic change in United States nuclear policy, banning all commercial reprocessing activities, and cancelling plans and funding for fast-breeding reactors. It was a political decision, influenced by immediate world events, and intended to be temporary. Contemporary documents and speeches given by Carter are examined in depth to determine the reasoning behind this dramatic change in US nuclear policy and the response by other countries. President Ronald Reagan rescinded these bans, but general mistrust by the US nuclear industry of further changes in political policy prevented further commercial reprocessing activities and construction of fast-breeding reactors in the US. The increase in the number of operating US fission reactors along with the lack of reprocessing activities caused concern among both Reagan and members of Congress. The resulting legislation set the stage for the proposed High-Level-Waste Repository at Yucca Mountain, Nevada. construction of the proposed repository had faced many serious setbacks due to several factors. The uncertainty regarding the project's timely completion has drastic repercussions for the involved parties. The conclusion of the author is that further evaluation into commercial reprocessing activities should be undertaken before final approval of the proposed repository is granted. (author)

The regulations concerning the reprocessing business of spent fuels

International Nuclear Information System (INIS)

1979-01-01

The regulations are defined under provisions concerning the reprocessing business in the law for the regulations of nuclear source materials, nuclear fuel materials and reactors. The basic concepts and terms are explained, such as: exposure dose; accumulative dose; controlled area; safeguarded area; inspected surrounding area; employee; radioactive waste and marine discharging facilities. Any person who gets permission for design of reprocessing facilities and method of the construction shall file an application, listing name and address of the person and the works or the place of enterprise where reprocessing facilities are to be set up, design of such facilities and method of the construction, in and out-put chart of nuclear fuel materials in reprocessing course, etc. Records shall be made and kept for particularly periods in each works or enterprise on inspection of reprocessing facilities, control of dose, operation, maintenance, accident of reprocessing facilities and weather. Detailed prescriptions are settled on entrance limitation to controlled area, exposure dose, inspection and check, regular independent examination and operation of reprocessing facilities, transportation in the works or the enterprise, storage, disposal, safeguard and measures in dangerous situations, etc. Reports shall be filed on exposure dose of employees and other specified matters in the forms attached and in the case otherwise defined. (Okada, K.)

Nitrate conversion and supercritical fluid extraction of UO2-CeO2 solid solution prepared by an electrolytic reduction-coprecipitation method

International Nuclear Information System (INIS)

Zhu, L.Y.; Duan, W.H.; Wen, M.F.; Xu, J.M.; Zhu, Y.J.

2014-01-01

A low-waste technology for the reprocessing of spent nuclear fuel (SNF) has been developed recently, which involves the conversion of actinide and lanthanide oxides with liquid N 2 O 4 into their nitrates followed by supercritical fluid extraction of the nitrates. The possibility of the reprocessing of SNF from high-temperature gas-cooled reactors (HTGRs) with nitrate conversion and supercritical fluid extraction is a current area of research in China. Here, a UO 2 -CeO 2 solid solution was prepared as a surrogate for a UO 2 -PuO 2 solid solution, and the recovery of U and Ce from the UO 2 -CeO 2 solid solution with liquid N 2 O 4 and supercritical CO 2 containing tri-n-butyl phosphate (TBP) was investigated. The UO 2 -CeO 2 solid solution prepared by electrolytic reduction-coprecipitation method had square plate microstructures. The solid solution after heat treatment was completely converted into nitrates with liquid N 2 O 4 . The XRD pattern of the nitrates was similar to that of UO 2 (NO 3 ) 2 . 3H 2 O. After 120 min of online extraction at 25 MPa and 50 , 99.98% of the U and 98.74% of the Ce were recovered from the nitrates with supercritical CO 2 containing TBP. The results suggest a promising potential technology for the reprocessing of SNF from HTGRs. (orig.)

Status and prospects for reprocessing

International Nuclear Information System (INIS)

Rossney, G.K.

1977-01-01

Following the formation of United Reprocessors (U.R.G.) in 1976 by British Nuclear Fuels Limited (B.N.F.L.) in the United Kingdom, the Commissariat a l'Energie Atomique (C.E.A.) in France and K.E.W.A. Kernbrennstoff-Wiederaufarbeitungs-Gesellschaft MBH (K.E.W.A.) in Germany, collaboration is now well established for the marketing of their reprocessing services for irradiated oxide fuel from thermal reactors. In addition collaboration in the continued evolution of the technology has progressed and an extensive research and development programme has been established, the results of which are exchanged between the shareholders. During 1976 the U.K. Government has given approval to B.N.F.L. to sign further contracts with foreign customers, subject to certain conditions. In France, the fuel cycle activities of the C.E.A. have been vested in a new company (Compagnie Generale Des Matieres Nucleaires (C.O.G.E.M.A.)) and their La Hague plant has commenced reprocessing operations on irradiated oxide fuel. In Germany, an agreement has been signed between K.E.W.A. and P.W.K. for the pre-project study for the proposed German plant. Against this background this paper reviews the present status of reprocessing by the shareholders of U.R.G. and the prospects for reprocessing

Environmental survey of the reprocessing and waste management portions of the LWR fuel cycle: a task force report

Energy Technology Data Exchange (ETDEWEB)

Bishop, W.P.; Miraglia, F.J. Jr. (eds.)

1976-10-01

This Supplement deals with the reprocessing and waste management portions of the nuclear fuel cycle for uranium-fueled reactors. The scope of the report is limited to the illumination of fuel reprocessing and waste management activities, and examination of the environmental impacts caused by these activities on a per-reactor basis. The approach is to select one realistic reprocessing and waste management system and to treat it in enough depth to illuminate the issues involved, the technology available, and the relationships of these to the nuclear fuel cycle in general and its environmental impacts.

Environmental survey of the reprocessing and waste management portions of the LWR fuel cycle: a task force report

International Nuclear Information System (INIS)

Bishop, W.P.; Miraglia, F.J. Jr.

1976-10-01

This Supplement deals with the reprocessing and waste management portions of the nuclear fuel cycle for uranium-fueled reactors. The scope of the report is limited to the illumination of fuel reprocessing and waste management activities, and examination of the environmental impacts caused by these activities on a per-reactor basis. The approach is to select one realistic reprocessing and waste management system and to treat it in enough depth to illuminate the issues involved, the technology available, and the relationships of these to the nuclear fuel cycle in general and its environmental impacts

Economic evaluation of reprocessing. Indicative UK position

International Nuclear Information System (INIS)

1979-05-01

This paper, which also appears as an Appendix to the final Working Group 4 report, forms part of the overall economic assessment of reprocessing. The indicative national position and illustrative ''phase diagram'' for the United Kingdom is presented. Under conditions pertaining in the United Kingdom the diagram suggests that: if uranium prices rise rapidly the fast reactor would become economic in the decade 1990-2000, if uranium prices rise more slowly, the fast reactor would become economic in the decade 2000-2010

Breeder reactor fuel reprocessing

International Nuclear Information System (INIS)

Trauger, D.B.

1983-01-01

The time cycle for breeder reactor development and deployment is longer than the planning horizons for most private industry and governments. The potential advantage and possible desperate need for widely deployed breeder reactors in the future seems to dictate that suitable long-term development and deployment programs be established to provide an adequate base of technology and in time to meet the need. The problems of failing to do so and being confronted with a major requirement for nuclear energy could result in very serious economic and social disruption. The cost of maintaining the needed program, although substantial, is certainly modest compared with the potential problems which could ensue should we fail to proceed

Pilot and pilot-commercial plants for reprocessing spent fuels of FBR type reactors

International Nuclear Information System (INIS)

Shaldaev, V.S.; Sokolova, I.D.

1988-01-01

A review of modern state of investigations on the FBR mixed oxide uranium-plutonium fuel reprocessing abroad is given. Great Britain and France occupy the leading place in this field, operating pilot plants of 5 tons a year capacity. Technology of spent fuel reprocessing and specific features of certain stages of the technological process are considered. Projects of pilot and pilot-commercial plants of Great Britain, France, Japan, USA are described. Economic problems of the FBR fuel reprocessing are touched upon

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

Internal dose evaluation from actinide intakes during nuclear power reactor spent fuel reprocessing

International Nuclear Information System (INIS)

Pawar, S.K.; Kumar, Ranjeet; Gamre, Rupali; Purohit, R.G.

2011-01-01

Full text: Indian PHWR reactors are using natural uranium as fuel. After use they are discharged from the core and send for fuel reprocessing to extract the unused uranium and plutonium. Plutonium and other actinides are formed by activation of 238 U with neutrons and subsequent decay. During reprocessing of the spent fuel, major long lived actinides (Pu, Am and U) may become radiological safety hazard. Actinides intakes are more probable during declading and chopping of spent fuel. During routine plant operation in reprocessing, exposure to Pu is a major concern along with Am and U in working environment due to its higher radiological hazard and occupational workers are likely to get exposed to plutonium, Americium and Uranium mostly through inhalation. Internally deposited Pu-isotopes, Am-isotope and U-isotopes are estimated using techniques such as lung counting (in-vivo) and urine and faecal bioassay (in-vitro). Evaluation of internal dose of actinides is dependent upon urinary excreted activity. To estimate the internally deposited Pu, U and Am at an intake level of about one ALI (ICRP-78, 1997) of occupational workers, urine bioassay is the preferred technique due to high detection sensitivity, ease of sample handling and economical method. A small and measurable fraction of internally deposited Pu, Am and U are excreted through urine whose content is dependent on time of inhalation, quantity and type of chemical form of inhaled material (S and M class). A standardized radiochemical analysis method for separation and estimation of Pu, Am and U is used to evaluate the urinary excreted activity and internal dose. Several measurements techniques are employed for the estimation of plutonium, Americium and Uranium for example, Alpha Spectrometry, Gamma Spectrometry, Neutron Activation Analysis, Mass Spectrometry and Fission Track Analysis. The radiochemical separation followed by alpha counting and/or spectrometry is chosen due to its ease of handling and

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...

Assessment of proliferation resistances of aqueous reprocessing techniques using the TOPS methodology

International Nuclear Information System (INIS)

Åberg Lindell, M.; Grape, S.; Håkansson, A.; Jacobsson Svärd, S.

2013-01-01

Highlights: • Proliferation resistances of three possible LFR fuel cycles are assessed. • The TOPS methodology has been chosen for the PR assessment. • Reactor operation, reprocessing and fuel fabrication are examined. • Purex, Ganex, and a combination of Purex, Diamex and Sanex, are compared. • The safeguards analysis speaks in favor of Ganex as opposed to the Purex process. - Abstract: The aim of this study is to assess and compare the proliferation resistances (PR) of three possible Generation IV lead-cooled fast reactor fuel cycles, involving the reprocessing techniques Purex, Ganex and a combination of Purex, Diamex and Sanex, respectively. The examined fuel cycle stages are reactor operation, reprocessing and fuel fabrication. The TOPS methodology has been chosen for the PR assessment, and the only threat studied is the case where a technically advanced state diverts nuclear material covertly. According to the TOPS methodology, the facilities have been divided into segments, here roughly representing the different forms of nuclear material occurring in each examined fuel cycle stage. For each segment, various proliferation barriers have been assessed. The results make it possible to pinpoint where the facilities can be improved. The results show that the proliferation resistance of a fuel cycle involving recycling of minor actinides is higher than for the traditional Purex reprocessing cycle. Furthermore, for the purpose of nuclear safeguards, group actinide extraction should be preferred over reprocessing options where pure plutonium streams occur. This is due to the fact that a solution containing minor actinides is less attractive to a proliferator than a pure Pu solution. Thus, the safeguards analysis speaks in favor of Ganex as opposed to the Purex process

Reprocessing technology of liquid metal cooled fast breeder reactor fuel

International Nuclear Information System (INIS)

Baetsle, L.H.; Broothaerts, J.; Heylen, P.R.; Eschrich, H.; Geel, J. van

1974-11-01

All the important aspects of LMFBR fuel reprocessing are critically reviewed in this report. Storage and transportation techniques using sodium, inert gas, lead, molten salts and organic coolants are comparatively discussed in connection with cooling time and de-activation techniques. Decladding and fuel disaggregation of UO 2 -PuO 2 fuel are reviewed according to the present state of R and D in the main nuclear powers. Strong emphasis is put on on voloxidation, mechanical pulverization and molten salt disaggregation in connection with volatilization of gaseous fission products. Release of fission gases and the resulting off-gas treatment are discussed in connection with cooling time, burn up and dissagregation techniques. The review is limited to tritium, iodine xenon-krypton and radioactive airborne particulates. Dissolution, solvent extraction and plutonium purification problems specifically connected to LMFBR fuel are reviewed with emphasis on the differences between LWR and fast fuel reprocessing. Finally the categories of wastes produced by reprocessing are analysed according to their origin in the plant and their alpha emitters content. The suitable waste treatment techniques are discussed in connection with the nature of the wastes and the ultimate disposal technique. (author)

Hybrid KED/XRF measurement of minor actinides in reprocessing plants

International Nuclear Information System (INIS)

Hsue, S.T.; Collins, M.L.

1996-01-01

Minor actinides have received considerable attention recently in the nuclear power industry. Because of their potential value as recycle fuels in thermal and breeder reactors, reprocessing plants may have an economic incentive to extract Np, Am, and Cm from their waste streams. This report discusses the technique of hybrid densitometry and its potential to measure Np and Am in reprocessing plants. Precision estimates are made for the hybrid analysis of Np and Am in two types of dissolver solutions

Study on the possibility of supercritical fluid extraction for reprocessing of spent nuclear fuel from high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Duan Wuhua; Zhu Liyang; Zhu Yongjun; Xu Jingming

2011-01-01

International interest in high temperature gas-cooled reactor (HTGR) has been increasing in recent years. It is important to study on reprocessing of spent nuclear fuel from HTGR for recovery of nuclear resource and reduction of nuclear waste. Treatment of UO 2 pellets for preparing fuel elements of the 10 MW high temperature gas-cooled reactor (HTR-10) using supercritical fluid extraction was investigated. UO 2 pellets are difficult to be directly dissolved and extracted with TBP-HNO 3 complex in supercritical CO 2 (SC-CO 2 ), and the extraction efficiency is only about 7% under experimental conditions. UO 2 pellets are also difficult to be converted completely into nitrate with N 2 O 4 . When UO 2 pellets break spontaneously into U 3 O 8 powders with particle size below 100 μm under O 2 flow and 600degc, the extraction efficiency of U 3 O 8 powders with TBP-HNO 3 complex in SC-CO 2 can reach more than 98%. U 3 O 8 powders are easy to be completely converted into nitrate with N 2 O 4 . The extraction efficiency of the nitrate product with TBP in SC-CO 2 can reach more than 99%. So it has a potential prospect that application of supercritical fluid extraction in reprocessing of spent nuclear fuel from HTGR. (author)

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

Used mixed oxide fuel reprocessing at RT-1 plant

Energy Technology Data Exchange (ETDEWEB)

Kolupaev, D.; Logunov, M.; Mashkin, A.; Bugrov, K.; Korchenkin, K. [FSUE PA ' Mayak' , 30, Lenins str, Ozersk, 460065 (Russian Federation); Shadrin, A.; Dvoeglazov, K. [ITCP ' PRORYV' , 2/8 Malaya Krasmoselskay str, Moscow, 107140 (Russian Federation)

2016-07-01

Reprocessing of the mixed uranium-plutonium spent nuclear fuel of the BN-600 reactor was performed at the RT-1 plant twice, in 2012 and 2014. In total, 8 fuel assemblies with a burn-up from 73 to 89 GW day/t and the cooling time from 17 to 21 years were reprocessed. The reprocessing included the stages of dissolution, clarification, extraction separation of U and Pu with purification from the fission products, refining of uranium and plutonium at the relevant refining cycles. Dissolution of the fuel composition of MOX used nuclear fuel (UNF) in nitric acid solutions in the presence of fluoride ion has occurred with the full transfer of actinides into solution. Due to the high content of Pu extraction separation of U and Pu was carried out on a nuclear-safe equipment designed for the reprocessing of highly enriched U spent nuclear fuel and Pu refining. Technological processes of extraction, separation and refining of actinides proceeded without deviations from the normal mode. The output flow of the extraction outlets in their compositions corresponded to the regulatory norms and remained at the level of the compositions of the streams resulting from the reprocessing of fuel types typical for the RT-1 plant. No increased losses of Pu into waste have been registered during the reprocessing of BN-600 MOX UNF an compare with VVER-440 uranium UNF reprocessing. (authors)

Reprocessing business in the UK

Energy Technology Data Exchange (ETDEWEB)

Wilkinson, W L

1985-01-01

The development of the process for separating uranium, plutonium and fission products from irradiated fuel began in Britain in late 1940s, and the first separation plant was operated at Sellafield in 1952. This plant was operated very well for more than 12 years with the overall availability over 95%. The second separation plant to meet the needs of the growing nuclear power program became operational in 1964. This plant has been extremely successful, but the significant improvement was made to extend the operating life of the key items. In mid 1970s, by the introduction of uranium oxide fuel reactors, significant reprocessing capacity became to be required. Therefore, it was decided to embark upon the development of a thermal oxide reprocessing plant (THORP) to complement the existing facilities at Sellafield. The THORP is a very large complex of plants. The first duty for the THORP is to reprocess 6,000 t U of oxide fuel in 10 years. But the plant is designed for the life of 25 years. The plant has the capacity of 1200 tes/year. The scope covered by the THORP, the plant processes and the wastes produced from the THORP are described. (Kako, I.).

Remote maintenance ''lessons learned'' on prototypical reprocessing equipment

International Nuclear Information System (INIS)

Kring, C.T.; Schrock, S.L.

1990-01-01

Hardware representative of essentially every major equipment item necessary for reprocessing breeder reactor nuclear fuel has been installed and tested for remote maintainability. This testing took place in a cold mock-up of a remotely maintained hot cell operated by the Consolidated Fuel Reprocessing Program (CFRP) within the Fuel Recycle Division at Oak Ridge National Laboratory (ORNL). The reprocessing equipment tested included a Disassembly System, a Shear System, a Dissolver System, an Automated Sampler System, removable Equipment Racks on which various chemical process equipment items were mounted, and an advanced servomanipulator (ASM). These equipment items were disassembled and reassembled remotely by using the remote handling systems that are available within the cold mock-up area. This paper summarizes the ''lessons learned'' as a result of the numerous maintenance activities associated with each of these equipment items. 4 refs., 3 figs., 1 tab

Development of centrifugal contactor for FBR fuel reprocessing

International Nuclear Information System (INIS)

Washiya, Tadahiro; Takeuchi, Masayuki; Suganuma, Takashi; Aose, Shinichi; Ogino, Hideki

2003-01-01

In the Feasibility Study on Commercialized Fast Reactor Cycle Systems, the aqueous reprocessing technology is nominated as a candidate for future reprocessing system, which supposes to apply a centrifugal contactor in the extraction process. For the reprocessing plant, the centrifugal contactor has great advantages such as reducing solvent degradation, improving of equipment utilization rate, compact designing of equipment layout and critical safety domination. From these advantages, the centrifugal contactor is crucial equipment in the aqueous reprocessing process. Since 1985, JNC has been developing the centrifugal contactor. The single unit development has been accomplished and basic characteristics such as extraction performance, fluidic performance and remote maintenance performance have been determined. A durability test has been conducted for high longevity, with consideration given to the nitric acid mist and estimation of the equipment lifetime. System test equipment with centrifugal contactors of engineering scale was installed, and uranium test was conducted. Up to now, a standard flow sheet test in the extraction process and mal-operation test assuming the one stage shutdown condition have been performed. (author)

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.

Composite reprocessing of spent nuclear fuel - is a way to low waste nuclear power

International Nuclear Information System (INIS)

Kosyakov, Valentin

2005-01-01

Further development of nuclear power in many respects depend on the solution of the problems connected to high level radioactive wastes (HLRW), containing highly toxic long-lived radionuclides. Long-term controlled storage of HLRW manages expensively and any advanced technology of reprocessing of spent nuclear fuel (SNF), besides recovery of the basic products, should be aimed at the reduction of this waste amount. However, the existing SNF reprocessing technology, using PUREX - process, is aimed only at extraction of uranium and plutonium, considering the remaining fraction (other transuranium elements and all fission products) as HLRW. In this work an attempt is made to give quantitative and qualitative characteristics to the isotopes and the elements which are included in the composition of HLRW after 15-years storage. Depending on the radiation properties of the isotopes included, these elements were divided into three categories: 1. The elements represented by only stable isotopes; 2. The elements represented by mostly low radioactive isotopes; 3. The elements represented by highly toxic long-lived radionuclides. As a result it appeared, that the weight percentage of the elements of the first, the second and the third categories in HLRW was: 60, 25 and 15% respectively. It means, that the amount of the real HLRW to be disposed in a deep geological repository could be reduced at least by a factor of 6, if to recover completely only 7 most dangerous elements (Sr, J, Cs, Sm, Np, Am and Cm) from the solution remaining after extraction of uranium and plutonium. Then it is meaningful to recover the elements of the first category from the remaining mix. As the main part of this fraction is represented by rare earth elements and noble metals, which can easily find many useful application. (author)

Research and development of pyro-reprocessing and its world status

International Nuclear Information System (INIS)

Inoue, Tadashi

2005-01-01

The next generation fuel cycle requires a strong resistance of nuclear proliferation and lightening the environmental burden as well as safety and economic advantage. The pyro-reprocessing technology satisfies such kinds of requirements. Central Research Institute of Electric Power Industry, CRIEPI, has been involving the development of metal fuel cycle integrated pyro-reprocessing with metal-electrorefining and metal fuel fast reactor since 1986. The study on pyro-processing technology of spent MOX fuel from LWR has been also started. Based on the fast that metal-electrorefining does not produce pure plutonium but transuranium elements, irradiation experiment of metal fuel with minor actinides is carried out by use of Phenix Fast Reactor in France. This article reports an overview of pyro-reprocessing and the present status of its research and development. The R and D activity proceeds to the process verification by use of genuine material and the development of engineering model of the process after finishing the verification of elemental technology. Irradiation study of metal fuel will be started by use of JOYO Fast Reactor as well as Phenix Fast Reactor. The target at 2015 is to finish the irradiation programs by both reactors and to demonstrate the pyro-process flow and related technologies by use of irradiated material. After finishing this stage, we expect to be technically feasible to design a pyro-process facility with a throughput of several tones of spent fuels. While R and D on pyro-technology has started initially in the U.S. and followed by CRIEPI, the several activities, currently, are followed in European and Asian nations. The engineering installation of electrochemical reduction successfully achieved by uranium test with 20 kg/batch and the construction of hot cell for handling a 20 kg/batch spent fuel finished in the Korean Atomic Energy Research Institute, KAERI. China has started R and D on metal fuel fast reactor and pyro-reprocessing as a

How power is generated in a nuclear reactor

International Nuclear Information System (INIS)

Swaminathan, V.

1978-01-01

Power generation by nuclear fission as a result of chain reaction caused by neutrons interacting with fissile material such as 235 U, 233 U and 239 Pu is explained. Electric power production by reactor is schematically illustrated. Materials used in thermal reactor and breeder reactor are compared. Fuel reprocessing and disposal of radioactive waste coming from reprocessing plant is briefly described. Nuclear activities in India are reviewed. Four heavy water plants and two power reactors are under construction and will be operative in the near future. Two power reactors are already in operation. Nuclear Fuel Complex at Hyderabad supplies fuel element to the reactors. Fuel reprocessing and waste management facility has been set up at Tarapur. Bhabha Atomic Research Centre at Bombay and Reactor Research Centre at Kalpakkam near Madras are engaged in applied and basic research in nuclear science and engineering. (B.G.W.)

Reprocessed uranium recycling: the ideal and the facts

International Nuclear Information System (INIS)

Comte, D.

1998-01-01

Commercial reprocessing of Light Water Reactor (LWR) spent fuel provides Reprocessed Uranium (RepU) and plutonium, both products containing a large amount of energy. As the gap, already quite large, between uranium consumption and production threatens to grow, these materials show today a strong strategic benefit, constituting a substantial and reliable source of supply for nuclear fuel manufacturing. RepU, which represents about 96% of recycled materials, can be used in all major types of nuclear power reactor currently in operation. This paper focuses on RepU recycling in LWRs, the feasibility of which is technically well established world-wide, and industrially demonstrated through experience gained within the COGEMA Group. Contrary to a commonly accepted assumption, which probably derives from natural uranium prices remaining for more than a decade at a low level which cannot be sustained into the future, recycling RepU can also bring strong economic benefits to the utilities. It generates savings on uranium procurement that exceed expenditures associated with storing and processing RepU. Thus, the use of this product is an attractive option. The strategic and economic benefits of recycling RepU will become compelling in the near future. Anticipating these needs, the COGEMA Group has developed capabilities to offer utilities the services covering all steps from reprocessing of spent fuel to fuel fabrication using RepU. (author)

Equipment specifications for an electrochemical fuel reprocessing plant

International Nuclear Information System (INIS)

Hemphill, Kevin P.

2010-01-01

Electrochemical reprocessing is a technique used to chemically separate and dissolve the components of spent nuclear fuel, in order to produce new metal fuel. There are several different variations to electrochemical reprocessing. These variations are accounted for by both the production of different types of spent nuclear fuel, as well as different states and organizations doing research in the field. For this electrochemical reprocessing plant, the spent fuel will be in the metallurgical form, a product of fast breeder reactors, which are used in many nuclear power plants. The equipment line for this process is divided into two main categories, the fuel refining equipment and the fuel fabrication equipment. The fuel refining equipment is responsible for separating out the plutonium and uranium together, while getting rid of the minor transuranic elements and fission products. The fuel fabrication equipment will then convert this plutonium and uranium mixture into readily usable metal fuel.

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)

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

Analyzing the reprocessing decision: plutonium recycle and nuclear proliferation. Final report

International Nuclear Information System (INIS)

Heising, C.D.; Connolly, T.J.

1978-11-01

The United States decision to defer indefinitely the reprocessing of spent nuclear fuel is examined in this thesis. Bayesian decision analysis is applied to develop a rational framework for the assessment of alternatives. Benefits and costs for each alternative are evaluated and compared in dollar terms to determine the optimal decision. A fuel cycle simulation model is constructed to assess the economic value of reprocessing light water reactor (LWR) spent fuel and recycling plutonium. In addition, a dynamic fuel substitution model is used to estimate the economic effects of the reprocessing decision's influence on the introduction date of the liquid metal fast breeder reactor (LMFBR). The analysis of benefits and costs is extended to include the social costs due to technological risks, such as accident risk, nuclear theft and/or sabotage, and international nuclear proliferation. These social costs are expressed in dollar terms for comparison with the conventional economic values. Results of the analysis indicate that the domestic social costs are less than the economic benefits by more than three orders of magnitude, and that the permit option dominates those of delay or prohibit. An examination of proliferation risk indicates a factor of approximately 7 between cost-benefits. Thus, on the basis of this analysis, it appears that to permit reprocessing is optimal over delaying or prohibiting the technology

Treatment of spent fuels from research reactors and reactor development programs in Germany

International Nuclear Information System (INIS)

Closs, K.D.

1999-01-01

Quite a great number of different types of spent fuel from research reactors and development programs exists in Germany. The general policy is to send back to the USA as long as possible fuel from MTRs and TRIGAs of USA origin. An option is reprocessing in Great Britain or France. This option is pursued as long as reprocessing and reuse of the recovered material is economically justifiable. For those fuels which cannot be returned to the USA or which will not be reprocessed, a domestic back-up solution of spent fuel management has been developed in Germany, compatible with the management of spent fuel from power reactors. It consists in dry storage in special casks and, later on, direct disposal. Preliminary results from experimental R and D investigations with research reactor fuel and experience from LWR fuel lead to the conclusion that the direct disposal option even for research reactor fuel or exotic fuel does not impose major technical difficulties for the German waste management and disposal concept. (author)

Status of French reactors

International Nuclear Information System (INIS)

Ballagny, A.

1997-01-01

The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm 3 . The OSIRIS reactor has already been converted to LEU. It will use U 3 Si 2 as soon as its present stock of UO 2 fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (except if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU

Idaho Chemical Processing Plant Spent Fuel and Waste Management Technology Development Program Plan

International Nuclear Information System (INIS)

1993-09-01

The Department of Energy (DOE) has received spent nuclear fuel (SNF) at the Idaho Chemical Processing Plant (ICPP) for interim storage and reprocessing since 1953. Reprocessing of SNF has resulted in an existing inventory of 1.5 million gallons of radioactive sodium-bearing liquid waste and 3800 cubic meters (m 3 ) of calcine, in addition to the 768 metric tons (MT) of SNF and various other fuel materials in inventory. To date, the major activity of the ICPP has been the reprocessing of SNF to recover fissile uranium; however, recent changes in world events have diminished the demand to recover and recycle this material. As a result, DOE has discontinued reprocessing SNF for uranium recovery, making the need to properly manage and dispose of these and future materials a high priority. In accordance with the Nuclear Waste Policy Act (NWPA) of 1982, as amended, disposal of SNF and high-level waste (HLW) is planned for a geological repository. Preparation of SNF, HLW, and other radioactive wastes for disposal may include mechanical, physical, and/or chemical processes. This plan outlines the program strategy of the ICPP Spent Fuel and Waste Management Technology Development Program (SF ampersand WMTDP) to develop and demonstrate the technology required to ensure that SNF and radioactive waste will properly stored and prepared for final disposal. Program elements in support of acceptable interim storage and waste minimization include: developing and implementing improved radioactive waste treatment technologies; identifying and implementing enhanced decontamination and decommissioning techniques; developing radioactive scrap metal (RSM) recycle capabilities; and developing and implementing improved technologies for the interim storage of SNF

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

Reprocessing of AHWR spent-fuel: Challenges and strategies

International Nuclear Information System (INIS)

Kant, S.

2005-01-01

Reprocessing of advanced heavy water reactor (AHWR) spent-fuel involves separation of Th, 233 U and Pu, from the fission products and from one another. A proper combination of Purex and Thorex processes is required. The technology development for a reprocessing facility is extremely complex owing to high fissile content, high levels of irradiation, presence high of levels of 232 U, difficulty in thoria dissolution, presence of thorium as the major constituent, problems due to third phase formation with Th, etc. It demands for development of suitable dissolution, solvent extraction, criticality control, U-Pu partitioning, and other equipments and/or techniques. Process modelling, simulation and optimisation are crucial in predicting behaviour of equipments/cycles, and in arriving at safe and optimum flowsheet. A significant success in this field has been achieved. This paper describes the reprocessing aspects pertaining to AHWR spent-fuel, indicating the major technological challenges, strategies to be followed and development requirements. A schematic flowsheet is proposed for Th- 233 U-Pu separation. (author)

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)

Noble gas atmospheric monitoring at reprocessing facilities

International Nuclear Information System (INIS)

Nakhleh, C.W.; Perry, R.T. Jr.; Poths, J.; Stanbro, W.D.; Wilson, W.B.; Fearey, B.L.

1997-01-01

The discovery in Iraq after the Gulf War of the existence of a large clandestine nuclear-weapon program has led to an across-the-board international effort, dubbed Programme 93+2, to improve the effectiveness and efficiency of International Atomic Energy Agency (IAEA) safeguards. One particularly significant potential change is the introduction of environmental monitoring (EM) techniques as an adjunct to traditional safeguards methods. Monitoring of stable noble gas (Kr, Xe) isotopic abundances at reprocessing plant stacks appears to be able to yield information on the burnup and type of the fuel being processed. To estimate the size of these signals, model calculations of the production of stable Kr, Xe nuclides in reactor fuel and the subsequent dilution of these nuclides in the plant stack are carried out for two case studies: reprocessing of PWR fuel with a burnup of 35 GWd/tU, and reprocessing of CAND fuel with a burnup of 1 GWd/tU. For each case, a maximum-likelihood analysis is used to determine the fuel burnup and type from the isotopic data

State-of-the-art report on accident analysis and risk analysis of reprocessing plants in European countries

International Nuclear Information System (INIS)

Nomura, Yasushi

1985-12-01

This report summarizes informations obtained from America, England, France and FRG concerning methodology, computer code, fundamental data and calculational model on accident/risk analyses of spent fuel reprocessing plants. As a result, the followings are revealed. (1) The system analysis codes developed for reactor plants can be used for reprocessing plants with some code modification. (2) Calculational models and programs have been developed for accidental phenomenological analyses in FRG, but with insufficient data to prove them. (3) The release tree analysis codes developed in FRG are available to estimate radioactivity release amount/probability via off-gas/exhaustair lines in the case of accidents. (4) The computer codes developed in America for reactor-plant environmental transport/safety analyses of released radioactivity can be applied to reprocessing facilities. (author)

Analytical chemistry needs for nuclear safeguards in nuclear fuel reprocessing

International Nuclear Information System (INIS)

Hakkila, E.A.

1977-01-01

A fuel reprocessing plant designed to process 1500 tons of light water reactor fuel per year will recover 15 tons of Pu during that time, or approximately 40 to 50 kg of Pu per day. Conventional nuclear safeguards accountability has relied on batch accounting at the head and tail ends of the reprocessing plant with semi-annual plant cleanout to determine in-process holdup. An alternative proposed safeguards system relies on dynamic material accounting whereby in-line NDA and conventional analytical techniques provide indications on a daily basis of SNM transfers into the system and information of Pu holdup within the system. Some of the analytical requirements and problems for dynamic materials accounting in a nuclear fuel reprocessing plant are described. Some suggestions for further development will be proposed

Consolidated Reprocessing Progam. Quarterly progress report ending February 28, 1979

International Nuclear Information System (INIS)

1979-03-01

This publication continues the quarterly series presenting results of work performed under the Consolidated Reprocessing Program at General Atomic Company. Results of work on this program prior to June 1974 were included in a quarterly series on the HTGR Base Program. The work reported includes the development of unit processes and equipment for reprocessing of High-Temperature Gas-Cooled Reactor (HTGR) fuel, the design and development of an integrated pilot line to demonstrate the head end of HTGR reprocessing using unirradiated fuel materials, and design work in support of Hot Engineering Tests (HET). Work is also described on trade-off studies concerning the required design of facilities and equipment for the large-scale recycle of HTGR fuels in order to guide the development activities for HTGR fuel recycle

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

Molten salt reactors and possible scenarios for future nuclear power deployment

International Nuclear Information System (INIS)

Merle-Lucotte, E.; Mathieu, L.; Heuer, D.; Loiseaux, J. M.; Billebaud, A.; Brissot, R.; David, S.; Garzenne, C.; Laulan, O.; Le Brun, C.; Lecarpentier, D.; Liatard, E.; Meplan, O.; Michel-Sendis, F.; Nuttin, A.; Perdu, F.

2004-01-01

An important fraction of the nature energy demand may be satisfied by nuclear power. In this context, the possibilities of worldwide nuclear deployment are studied. We are convinced that the Molten Salt Reactors may play a central role in this deployment. The Molten Salt Reactor needs to be coupled to a reprocessing unit in order to extract the Fission Products which poison the core. The efficiency of this reprocessing has a crucial influence on reactor behavior especially for the breeding ratio. The Molten Salt Breeder Reactor project was based on an intensive reprocessing for high breeding purposes. A new concept of Thorium Molten Salt Reactor is presented here. Including this new concept in the worldwide nuclear deployment, to satisfy these power needs, we consider three typical scenarios, based on three reactor types: Pressurized Water Reactor, Fast Neutron Reactor and Thorium Molten Salt Reactor. The aim of this paper is to demonstrate, in a first hand that a Thorium Molten Salt Reactor can be realistic, with correct temperature coefficients and at least iso-breeder with slow reprocessing and new geometry; on the other hand that such Molten Salt Reactors enable a successful nuclear deployment, while minimizing fuel and waste management problems. (authors)

THE ECONOMICS OF REPROCESSING vs. DIRECT DISPOSAL OF SPENT NUCLEAR FUEL

International Nuclear Information System (INIS)

Bunn, Matthew; Fetter, Steve; Holdren, John P.; Zwaan, Bob van der

2003-01-01

This report assesses the economics of reprocessing versus direct disposal of spent nuclear fuel. The breakeven uranium price at which reprocessing spent nuclear fuel from existing light-water reactors (LWRs) and recycling the resulting plutonium and uranium in LWRs would become economic is assessed, using central estimates of the costs of different elements of the nuclear fuel cycle (and other fuel cycle input parameters), for a wide range of range of potential reprocessing prices. Sensitivity analysis is performed, showing that the conclusions reached are robust across a wide range of input parameters. The contribution of direct disposal or reprocessing and recycling to electricity cost is also assessed. The choice of particular central estimates and ranges for the input parameters of the fuel cycle model is justified through a review of the relevant literature. The impact of different fuel cycle approaches on the volume needed for geologic repositories is briefly discussed, as are the issues surrounding the possibility of performing separations and transmutation on spent nuclear fuel to reduce the need for additional repositories. A similar analysis is then performed of the breakeven uranium price at which deploying fast neutron breeder reactors would become competitive compared with a once-through fuel cycle in LWRs, for a range of possible differences in capital cost between LWRs and fast neutron reactors. Sensitivity analysis is again provided, as are an analysis of the contribution to electricity cost, and a justification of the choices of central estimates and ranges for the input parameters. The equations used in the economic model are derived and explained in an appendix. Another appendix assesses the quantities of uranium likely to be recoverable worldwide in the future at a range of different possible future prices

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.

Survey of economics of spent nuclear fuel reprocessing

International Nuclear Information System (INIS)

Valvoda, Z.

1976-01-01

Literature data are surveyed on the economic problems of reprocessing spent fuel from light-water reactors in the period 1970 to 1975 and on the capacity of some reprocessing plants, such as NFS, Windscale, Marcoule, etc. The sharp increase in capital and production costs is analyzed and the future trend is estimated. The question is discussed of the use of plutonium and the cost thereof. The economic advantageousness previously considered to be the primary factor is no longer decisive due to new circumstances. The main objective today is to safeguard uninterrupted operation of nuclear power plants and the separation of radioactive wastes from the fuel cycle and the safe disposal thereof. (Oy)

The economic and environmental benefits of reprocessing and recycle

International Nuclear Information System (INIS)

Masters, R.

1977-01-01

Recently published reports and papers on the economic, environmental and resource conservation benefits of reprocessing and the recycle of plutonium in thermal reactors provide an important background to the essentially political discussions in the United States on non-proliferation. Some of the main arguments and conclusions are presented. (author)

Study on reprocessing plant during transition period from LWR to FBR

International Nuclear Information System (INIS)

Shimada, Takashi; Matsui, Minefumi; Nishimura, Masashi; Ishida, Yasuhiro; Mori, Yukihide; Kuroda, Kazuhiko

2011-01-01

We have proposed a concept of a reprocessing plant suitable for the transition period from the light water reactors (LWRs) to the fast breeder reactors (FBRs) by making comparison of two plant concepts: (1) Independent Plant which processes LWR fuel and FBR fuel in separately constructed lines and (2) Modularized Plant which processes LWR fuel and FBR fuel in a same line. We made construction plans based on the reference power generation plan, and evaluated the Pu supply capability using the power generation plan as an indicator of plant operation flexibility. In general, a margin of processing capacity increases the Pu supply capability. The margin of the Modularized Plant necessary to obtain equivalent Pu supply capability is smaller than that of the Independent Plant. Also the margin of the Independent Plant results in decrease in the plant utilization factor. But the margin of the Modularized Plant results in little decrease in the plant utilization factor, because the Modularized Plant can address the types of reprocessing fuel to adjust to Pu demand and processing capacity. Therefore, the Modularized Plant has a greater potential for the reprocessing plants during transition period. (author)

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.

Status of French reactors

Energy Technology Data Exchange (ETDEWEB)

Ballagny, A. [Commissariat a l`Energie Atomique, Saclay (France)

1997-08-01

The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm{sup 3}. The OSIRIS reactor has already been converted to LEU. It will use U{sub 3}Si{sub 2} as soon as its present stock of UO{sub 2} fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (except if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU.

Nitrogen oxide closed system in the future reprocessing process

International Nuclear Information System (INIS)

Tanaka, S.; Takaoku, Y.; Sumida, Y.; Moriya, T.; Araya, S.

2006-01-01

Full text: Full text: The aqueous reprocessing process for the future type reactor like as Fast Breeder Reactor(FBR) is being developed in many institutes, while the existence of sodium nitrate as the secondary waste is considered as problematic due to an enormous quantity of sodium nitrate generated and the difficulty in its handling for disposal. As a means for solving the problem, a complete recycle of nitric acid and salt free system is considered, but it become a constraint in the process constitution. We have devised the alternative system, which shall approve the generation of sodium nitrate, and make the choice of a wide reprocessing process. Under this system nitric acid within sodium nitrate shall be reduced and made into harmless gas, while at the same time, the remaining sodium compound shall be re-used in a suitable form. In order to prevent the accumulation of radioactivity by re-use, we propose to use a part of remaining sodium compound as substitution of the fresh sodium within the glass material used for the vitrified solid waste. As a result of using this system, the waste originating from sodium nitrate can be reduced to 'zero'. We have studied a typical application case for the future reprocessing process, and got a good result at an economical point of view

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

The molten salt reactors (MSR) pyro chemistry and fuel cycle for innovative nuclear systems; Congres sur les reacteurs a sels fondus (RSF) pyrochimie et cycles des combustibles nucleaires du futur

Energy Technology Data Exchange (ETDEWEB)

Brossard, Ph. [GEDEON, Groupement de Recherche CEA CNRS EDF FRAMATOME (France); Garzenne, C.; Mouney, H. [and others

2002-07-01

In the frame of the studies on next generation nuclear systems, and especially for the molten salt reactors and for the integrated fuel cycle (as IFR), the fuel cycle constraints must be taken into account in the preliminary studies of the system to improve the cycle and reactor optimisation. Among the purposes for next generation nuclear systems, sustainability and waste (radio-toxicity and mass) management are important goals. These goals imply reprocessing and recycling strategies. The objectives of this workshop are to present and to share the different strategies and scenarios, the needs based on these scenarios, the experimental facilities available today or in the future and their capabilities, the needs for demonstration. It aims at: identifying the needs for fuel cycle based on solid fuel or liquid fuel, and especially, the on-line reprocessing or clean up for the molten salt reactors; assessing the state-of-the-art on the pyro-chemistry applied to solid fuel and to present the research activities; assessing the state-of-the-art on liquid fuels (or others), and to present the research activities; expressing the R and D programs for pyro-chemistry, molten salt, and also to propose innovative processes; and proposing some joint activities in the frame of GEDEON and PRACTIS programs. This document brings together the transparencies of 18 contributions dealing with: scenario studies with AMSTER concept (Scenarios, MSR, breeders (Th) and burners); fuel cycle for innovative systems; current reprocessing of spent nuclear fuel (SNF) in molten salts (review of pyro-chemistry processes (non nuclear and nuclear)); high temperature NMR spectroscopies in molten salts; reductive extraction of An from molten fluorides (salt - liquid metal extraction); electrochemistry characterisation; characterisation with physical methods - extraction coefficient and kinetics; electrolytic extraction; dissolution-precipitation of plutonium in the eutectic LiCl-KCl (dissolution and

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

Idaho Chemical Processing Plant spent fuel and waste management technology development program plan: 1994 Update

International Nuclear Information System (INIS)

1994-09-01

The Department of Energy has received spent nuclear fuel (SNF) at the Idaho Chemical Processing Plant (ICPP) for interim storage since 1951 and reprocessing since 1953. Until April 1992, the major activity of the ICPP was the reprocessing of SNF to recover fissile uranium and the management of the resulting high-level wastes (HLW). In 1992, DOE chose to discontinue reprocessing SNF for uranium recovery and shifted its focus toward the continued safe management and disposition of SNF and radioactive wastes accumulated through reprocessing activities. Currently, 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), 3,800 cubic meters of calcine waste, and 289 metric tons heavy metal of SNF are in inventory at the ICPP. Disposal of SNF and high-level waste (HLW) is planned for a repository. Preparation of SNF, HLW, and other radioactive wastes for disposal may include mechanical, physical, and/or chemical processes. This plan outlines the program strategy of the ICPP spent Fuel and Waste Management Technology Development Program (SF ampersand WMTDP) to develop and demonstrate the technology required to ensure that SNF and radioactive waste will be properly stored and prepared for final disposal in accordance with regulatory drivers. This Plan presents a brief summary of each of the major elements of the SF ampersand WMTDP; identifies key program assumptions and their bases; and outlines the key activities and decisions that must be completed to identify, develop, demonstrate, and implement a process(es) that will properly prepare the SNF and radioactive wastes stored at the ICPP for safe and efficient interim storage and final disposal

Remote systems and remote maintenance of a reprocessing plant in Japan

International Nuclear Information System (INIS)

Funaya, T.

1977-01-01

The design concept and overall maintenance philosophy applied in the Power Reactor and Nuclear Fuel Development Corporation Reprocessing Plant at Tokai-mura, Japan, are briefly introduced. Details on remote systems and remote maintenance in mechanical processing areas are described

EdF speaks about economic advantages of fuel reprocessing as compared with interim storage

International Nuclear Information System (INIS)

Anon.

1997-01-01

The French company Electricite de France (EdF) will prefer nuclear fuel reprocessing and plutonium recycling to spent fuel storage also in the years after 2000. This option is economically advantageous if the proportional cost of reprocessing does not exceed 1900 FRF/kg heavy metal. Economic analysis shows that this is feasible. EdF will soon have to reprocess annually about 1000 Mt spent fuel to supply enough plutonium for MOX fuel fabrication to feed as many as 28 PWR units and the Superphenix reactor. Spent fuel reprocessing is seen as promising as long as the efficiency of the MOX fuel approaches that of natural uranium based fuel. The French national industrial, political and legal context of EdF operations is also considered. (P.A.)

Possible use of dual purpose dry storage casks for transportation and future storage of spent nuclear fuel from IRT-Sofia

International Nuclear Information System (INIS)

Manev, L.; Baltiyski, M.

2003-01-01

Objectives: The main objective of the present paper is related to one of the priority goals stipulated in Bulgarian Governmental Decision No.332 from May 17, 1999 - removal of SNF from IRT-Sofia site and its exporting for reprocessing and/or for temporary storage at Kozloduy NPP site. The variant of using dual purpose dry storage casks for transportation and future temporary storage of SNF from IRT-Sofia aims to find out a reasonable alternative of the existing till now variant for temporary SNF storage under water in the existing Kozloduy NPP Spent Fuel Storage Facility until its export for reprocessing. Results: Based on the given data for the condition of 73 Spent Nuclear Fuel Assemblies (SNFA) stored in the storage pool and technical data as well as data for available equipment and IRT-Sofia layout the following framework are specified: draft technical features of dual purpose dry storage casks and their overall dimensions; the suitability of the available equipment for safety and reliable performance of transportation and handling operations of assemblies from storage pool to dual purpose dry storage casks; the necessity of new equipment for performance of the above mentioned operations; Assemblies' transportation and handling operations are described; requirements to and conditions for future safety and reliable storage of SNFA loaded casks are determined. When selecting the technical solutions for safety assurance during performance of site handling operations of IRT-Sofia and for description of the exemplary casks the Effective Bulgarian Regulations are considered. The experience of other countries in performance of transfer and transportation of SNFA from such types of research reactors is taken into account. Also, Kozloduy NPP experience in SNF handling operations is taken into account. Conclusions: The Decision of Council of Minister for refurbishment of research reactor into a low power one and its future utilization for experimental and training

Air conditioning facilities in a fuel reprocessing plant

International Nuclear Information System (INIS)

Kawasaki, Michitaka; Oka, Tsutomu

1987-01-01

Reprocessing plants are the facilities for separating the plutonium produced by nuclear reaction and unconsumed remaining uranium from fission products in the spent fuel taken out of nuclear reactors and recovering them. The fuel reprocessing procedure is outlined. In order to ensure safety in handling radioactive substances, triple confinement using vessels, concrete cells and buildings is carried out in addition to the prevention of criticality and radiation shielding, and stainless steel linings and drip trays are installed as occasion demands. The ventilation system in a reprocessing plant is roughly divided into three systems, that is, tower and tank ventilation system to deal with offgas, cell ventilation system for the cells in which main towers and tanks are installed, and building ventilation system. Air pressure becomes higher from tower and tank system to building system. In a reprocessing plant, the areas in a building are classified according to dose rate. The building ventilation system deals with green and amber areas, and the cell ventilation system deals with red area. These three ventilation systems are explained. Radiation monitors are installed to monitor the radiation dose rate and air contamination in working places. The maintenance and checkup of ventilation systems are important. (Kako, I.)

Phenix: A Reprocessing and Multiple Recycling Experiment Unique in the World

International Nuclear Information System (INIS)

Guidez, J.

2013-01-01

Fast reactor fuel cycle principle: The Phenix fresh fuel is MOX with about 20% of Plutonium, and 80% of uranium. It can be natural or even depleted uranium. After reprocessing of this burned fuel, you obtain more plutonium(surgenerator). If you use this plutonium with depleted uranium, you can produce new MOX for the reactor. And again… at the end you operate the plant, burning only natural and depleted uranium

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

The regulations concerning the reprocessing business of spent fuels

International Nuclear Information System (INIS)

1978-01-01

In compliance with ''The law for the regulations of nuclear source material, nuclear fuel material and reactors'' these regulations prescribe concerning reprocessing facilities: The procedures to apply for the approval of the design and method of construction and the approval of the change thereof; as well as the procedure to apply for the inspection of the facilities, and details of the inspection (in sections 2-6). After that, the regulations require the enterpriser of reprocessing business to keep necessary records and take necessary measures for safety concerning the facilities, operation of reprocessing equipments, and transportation, storage on disposal of used fuel, materials separated therefrom or materials contaminated by either of them (in sections 8-16). Further, the regulations prescribe the procedure to apply for the approval of the safety rule required to the enterpriser of reprocessing business by above mentioned law and specifies items which should be included into the rule (section 17). Moreover, the regulations require the enterpriser to submit reports of each use of the internationally controllled material and specifies the items which should be included into these reports (section 19). (Matsushima, A.)

ERDA activities related to reprocessing and plutonium recycle

International Nuclear Information System (INIS)

Spurgeon, D.R.

1977-01-01

ERDA has redirected its program in support of the LWR fuel cycle from one emphasizing the commercialization of existing fuel cycle technology to a broader based assessment of alternative fuel cycle concepts with the emphasis on safeguardability and avoidance of proliferation risks. As part of this program, ERDA will evaluate a number of possible technical and institutional options to reduce proliferation risks. ERDA will continue its current program of LWR fuel reprocessing R and D with added emphasis on improved safeguards capability as well as the applicability of conventional reprocessing technology to large multinational plants. These activities and supporting design studies will provide the basis for a decision regarding the design of an optimized system for the management of spent LWR fuel. Such a system would provide a model for the development of future domestic and foreign facilities and programs. A recently completed ERDA study of the benefits of LWR reprocessing and recycle would also be expected to be factored into such a decision. The study concluded that based on currently available data, recycle of uranium and plutonium in LWR's is attractive from the standpoint of economics and resource utilization relative to the discarding of spent fuel. The LWR reprocessing/recycle picture today is clouded by several unresolved policy issues. These include the need for adequate spent fuel storage capacity for both domestic and foreign reactors; the possibility of foreign reprocessing of U.S. produced fuel; the possibility of the disposal of foreign fuel in the U.S.; the possible need to dispose of wastes generated by multinational reprocessing plants; and finally, determination of the optimum balance between recycling recovered plutonium and saving it for the breeder

Why reprocess

International Nuclear Information System (INIS)

Greenwood, T.

1977-01-01

Prospective costs of reprocessing, waste management, and mixed oxide fuel fabrication have risen so much that the costs of U/P recycle and of spent fuel storage are nearly equal. This paper reviews the current state of the reprocessing industry, with a list of facilities all over the world, and examines the incentives and disincentives other than short-term economics that will affect the decision of states to acquire their own reprocessing facilities. Finally, it examines the possibility of avoiding a widespread commercial reprocessing industry

Economic evaluation of reprocessing - Indicative German position

International Nuclear Information System (INIS)

1979-05-01

This paper, which also appears as an Appendix to the final Working Group 4 report, forms part of the overall economic assessment of reprocessing. The indicative national position and illustrative ''phase diagram'' for the Federal Republic of Germany is presented. Advantages to be gained from the recycling of plutonium in thermal reactors are identified and it is concluded that many of these are not explicitly taken into account in the ''phase diagram''. Under the conditions pertaining in the Federal Republic of Germany thermal recycle will be economic under present day uranium prices. On the other hand the fast breeder reactor might become commercially economic around the year 2000

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

Role of the consolidated fuel reprocessing program in the United States Breeder Reactor Program

International Nuclear Information System (INIS)

Ballard, W.W.; Burch, W.D.

1980-01-01

While present US policy precludes the commercial reprocessing of LWR fuels and the recycle of plutonium, the policy does encompass the need to continue a program to develop the technology for reprocessing breeder fuels. Some questions have again risen this year as to the pace of the entire breeder program, including recycle, and the answers are evolving. This paper and the other companion papers which describe several aspects of the Consolidated Fuel Reprocessing Program take a longer-range perspective on the total program. Whether the program is implemented in the general time frame described is dependent on future government actions dedicated to carrying out a systematic program that would permit breeders to be commercialized early in the next century

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

Fuel fabrication and reprocessing at UKAEA Dounreay

International Nuclear Information System (INIS)

Anderson, B.

1994-01-01

The Dounreay fuel plants, which are the most flexible anywhere in the world, will continue to carry out work for foreign commercial customers. A number of German companies are important customers of UKAEA and examples of the wide variety of the work currently being carried out for them in the Dounreay plants is given (reprocessing and fabrication of fuel elements from and for research reactors). (orig./HP) [de

Nitrate conversion and supercritical fluid extraction of UO{sub 2}-CeO{sub 2} solid solution prepared by an electrolytic reduction-coprecipitation method

Energy Technology Data Exchange (ETDEWEB)

Zhu, L.Y. [Tsinghua Univ., Beijing (China). Inst. of Nuclear and New Energy Technology; China Institute of Atomic Energy, Beijing (China); Duan, W.H.; Wen, M.F.; Xu, J.M.; Zhu, Y.J. [Tsinghua Univ., Beijing (China). Inst. of Nuclear and New Energy Technology

2014-04-01

A low-waste technology for the reprocessing of spent nuclear fuel (SNF) has been developed recently, which involves the conversion of actinide and lanthanide oxides with liquid N{sub 2}O{sub 4} into their nitrates followed by supercritical fluid extraction of the nitrates. The possibility of the reprocessing of SNF from high-temperature gas-cooled reactors (HTGRs) with nitrate conversion and supercritical fluid extraction is a current area of research in China. Here, a UO{sub 2}-CeO{sub 2} solid solution was prepared as a surrogate for a UO{sub 2}-PuO{sub 2} solid solution, and the recovery of U and Ce from the UO{sub 2}-CeO{sub 2} solid solution with liquid N{sub 2}O{sub 4} and supercritical CO{sub 2} containing tri-n-butyl phosphate (TBP) was investigated. The UO{sub 2}-CeO{sub 2} solid solution prepared by electrolytic reduction-coprecipitation method had square plate microstructures. The solid solution after heat treatment was completely converted into nitrates with liquid N{sub 2}O{sub 4}. The XRD pattern of the nitrates was similar to that of UO{sub 2}(NO{sub 3}){sub 2} . 3H{sub 2}O. After 120 min of online extraction at 25 MPa and 50 , 99.98% of the U and 98.74% of the Ce were recovered from the nitrates with supercritical CO{sub 2} containing TBP. The results suggest a promising potential technology for the reprocessing of SNF from HTGRs. (orig.)

Linear accelerator driven (LADR) and regenerative reactors (LARR) for nuclear non-proliferation

International Nuclear Information System (INIS)

Steinberg, M.; Takahashi, H.; Powell, J.R.; Kouts, H.J.C.

1977-09-01

Linear accelerator breeders (LAB) could be used to produce fissile fuel in two modes, either with fuel reprocessing or without fuel reprocessing. With fuel reprocessing, the fissile material would be separated from the target and refabricated into a fuel element for use in a burner power reactor. Without reprocessing, the fissile material would be produced in-situ, either in a fresh fuel element or in a depleted or burned element after use in a power reactor. In the latter mode the fissile material would be increased in concentration for reuse in a power reactor. This system is called a Linear Accelerator Regenerative Reactor (LARR). The LAB can also be conceived of operating in a power production mode in which the spallation neutrons would be used to drive a subcritical assembly to produce power. This is called a Linear Accelerator Driven Reactor (LADR). A discussion is given of the principles and some of the technical problems of both types of accelerator breeders

Studies and research concerning BNFP: converting reprocessing plant's fuel receiving and storage area to an away-from-reactor (AFR) storage facility. Final report

International Nuclear Information System (INIS)

Cottrell, J.E.; Shallo, F.A.; Musselwhite, E.L.; Wiedemann, G.F.; Young, M.

1979-09-01

Converting a reprocessing plant's fuel receiving and storage station into an Away-From-Reactor storage facility is evaluated in this report. An engineering analysis is developed which includes (1) equipment modifications to the facility including the physical protection system, (2) planning schedules for licensing-related activities, and (3) cost estimates for implementing such a facility conversion. Storage capacities are evaluated using the presently available pools of the existing Barnwell Nuclear Fuel Plant-Fuel Receiving and Storage Station (BNFP-FRSS) as a model

Fuel reprocessing of the fast molten salt reactor: actinides et lanthanides extraction

International Nuclear Information System (INIS)

Jaskierowicz, S.

2012-01-01

The fuel reprocessing of the molten salt reactor (Gen IV concept) is a multi-steps process in which actinides and lanthanides extraction is performed by a reductive extraction technique. The development of an analytic model has showed that the contact between the liquid fuel LiF-ThF 4 and a metallic phase constituted of Bi-Li provide firstly a selective and quantitative extraction of actinides and secondly a quantitative extraction of lanthanides. The control of this process implies the knowledge of saline phase properties. Studies of the physico-chemical properties of fluoride salts lead to develop a technique based on potentiometric measurements to evaluate the fluoro-acidity of the salts. An acidity scale was established in order to classify the different fluoride salts considered. Another electrochemical method was also developed in order to determine the solvation properties of solutes in fluoride F- environment (and particularly ThF 4 by F-) in reductive extraction technique, a metallic phase is also involved. A method to prepare this phase was developed by electro-reduction of lithium on a bismuth liquid cathode in LiCl-LiF melt. This technique allows to accurately control the molar fraction of lithium introduced into the liquid bismuth, which is a main parameter to obtain an efficient extraction. (author)

Transmutation Strategy Using Thorium-Reprocessed Fuel ADS for Future Reactors in Vietnam

Directory of Open Access Journals (Sweden)

Thanh Mai Vu

2013-01-01

Full Text Available Nuclear power is believed to be a key to the energy security for a developing country like Vietnam where the power demanding increases rapidly every year. Nevertheless, spent nuclear fuel from nuclear power plants is the source of radiotoxic and proliferation risk. A conceptual design of ADS utilizing thorium fuel as a based fuel and reprocessed fuel as a seed for nuclear waste transmutation and energy production is proposed as one of the clean, safe, and economical solutions for the problem. In the design, 96 seed assemblies and 84 blanket assemblies were inserted into the core to make a heterogeneous subcritical core configuration. Introducing thorium fuel into the core offers an effective way to transmute plutonium and minor actinide (MA and gain energy from this process. Transmutation rate as a function of burnup is estimated using MCNPX 2.7.0 code. Results show that by using the seed-blanket designed ADS, at 40 GWd/t burnup, 192 kg of plutonium and 156 kg of MA can be eliminated. Equivalently, 1  ADS can be able to transmute the transuranic (TRU waste from 2  LWRs. 14 units of ADS would be required to eliminate TRUs from the future reactors to be constructed in Vietnam.

Optimal measurement uncertainties for materials accounting in a fast breeder reactor spent-fuel reprocessing plant

International Nuclear Information System (INIS)

Dayem, H.A.; Kern, E.A.; Markin, J.T.

1982-01-01

Optimization techniques are used to calculate measurement uncertainties for materials accountability instruments in a fast breeder reactor spent-fuel reprocessing plant. Optimal measurement uncertainties are calculated so that performance goals for detecting materials loss are achieved while minimizing the total instrument development cost. Improved materials accounting in the chemical separations process (111 kg Pu/day) to meet 8-kg plutonium abrupt (1 day) and 40-kg plutonium protracted (6 months) loss-detection goals requires: process tank volume and concentration measurements having precisions less than or equal to 1%; accountability and plutonium sample tank volume measurements having precisions less than or equal to 0.3%, short-term correlated errors less than or equal to 0.04%, and long-term correlated errors less than or equal to 0.04%; and accountability and plutonium sample tank concentration measurements having precisions less than or equal to 0.4%, short-term correlated errors less than or equal to 0.1%, and long-term correlated errors less than or equal to 0.05%

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

Integrated international safeguards concepts for fuel reprocessing

International Nuclear Information System (INIS)

Hakkila, E.A.; Gutmacher, R.G.; Markin, J.T.; Shipley, J.P.; Whitty, W.J.; Camp, A.L.; Cameron, C.P.; Bleck, M.E.; Ellwein, L.B.

1981-12-01

This report is the fourth in a series of efforts by the Los Alamos National Laboratory and Sandia National Laboratories, Albuquerque, to identify problems and propose solutions for international safeguarding of light-water reactor spent-fuel reprocessing plants. Problem areas for international safeguards were identified in a previous Problem Statement (LA-7551-MS/SAND79-0108). Accounting concepts that could be verified internationally were presented in a subsequent study (LA-8042). Concepts for containment/surveillance were presented, conceptual designs were developed, and the effectiveness of these designs was evaluated in a companion study (SAND80-0160). The report discusses the coordination of nuclear materials accounting and containment/surveillance concepts in an effort to define an effective integrated safeguards system. The Allied-General Nuclear Services fuels reprocessing plant at Barnwell, South Carolina, was used as the reference facility

Handling of spent nuclear fuel and final storage of vitrified high level reprocessing waste

International Nuclear Information System (INIS)

1978-01-01

The report gives a general summary of the Swedish KBS-project on management and disposal of vitrified reprocessed waste. Its final aim is to demostrate that the means of processing and managing power reactor waste in an absolutely safe way, as stipulated in the Swedish so called Conditions Act, already exist. Chapters on Storage facility for spent fuel, Intermidiate storage of reprocessed waste, Geology, Final repository, Transportation, Protection, and Siting. (L.E.)

Simulation tools and new developments of the molten salt fast reactor

International Nuclear Information System (INIS)

Heuer, D.; Merle-Lucotte, E.; Allibert, M.; Doligez, X.; Ghetta, V.

2010-01-01

In the MSFR (Molten Salt Fast Reactor), the liquid fuel processing is part of the reactor where a small side stream of the molten salt is processed for fission product removal and then returned to the reactor. Because of this design characteristic, the MSFR can thus operate with a widely varying fuel composition. Our reactor's studies of the MSFR concept rely on numerical simulations making use of the MCNP neutron transport code coupled with a code for Bateman's equations computing the population of any nucleus inside any part of the reactor at any moment. The classical Bateman's equations have been modified by adding 2 terms representing the reprocessing capacities and an online addition. We have thus coupled neutronic and reprocessing simulation codes in a numerical tool used to calculate the extraction efficiencies of fission products, their location in the whole system and radioprotection issues. The very preliminary results show the potential of the neutronic-reprocessing coupling we have developed. We also show that these studies are limited by the uncertainties on the design and the knowledge of the chemical reprocessing processes. (A.C.)

Management of regenerant effluent waste at reprocessing plant, Tarapur- a new approach

Energy Technology Data Exchange (ETDEWEB)

Chandra, Munish; Bajpai, D D; Mudaiya, Avinash; Varadarajan, N [Power Reactor Fuel Reprocessing Plant, Tarapur (India)

1994-06-01

Power Reactor Fuel Reprocessing (PREFRE) Plant, Tarapur has been processing zircaloy clad spent fuel arising from PHWR namely RAPS and MAPS. The plant has been provided with a water pool to receive and store the irradiated fuel assemblies from the reactor site for an interim period before they are taken up for chop-leach and further reprocessing by PUREX process. This paper highlights the important and innovative modifications like introduction of a cation exchanger for water polishing and using nitric acid as regenerant. The regenerant effluent (nitric acid) is recycled to the main process cells where it is mixed and further treated along with process waste stream. This is a step towards minimising effluent generation. The paper describes the advantages of modified system like operational simplification, manpower, man-rem saving and minimising release of activity to environment. 3 figs., 4 tabs.

Reprocessing of spent nuclear fuel; Prerada isluzenog nuklearnog goriva

Energy Technology Data Exchange (ETDEWEB)

Gal, I [Institute of Nuclear Sciences Boris Kidric, Laboratorija za visoku aktivnost, Vinca, Beograd (Serbia and Montenegro)

1963-12-15

This report covers: chemical-technology investigation of modified purex process for reprocessing of spent fuel; implementation of the procedure for obtaining plutonium peroxide and oxalate; research in the field of uranium, plutonium, and fission products separation by inorganic ion exchangers and extraction by organic solutions; study of the fission products in the heavy water RA reactor.

Radioactive Semivolatiles in Nuclear Fuel Reprocessing

Energy Technology Data Exchange (ETDEWEB)

Jubin, R. T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Strachan, D. M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Ilas, G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Spencer, B. B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Soelberg, N. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2014-09-01

In nuclear fuel reprocessing, various radioactive elements enter the gas phase from the unit operations found in the reprocessing facility. In previous reports, the pathways and required removal were discussed for four radionuclides known to be volatile, 14C, 3H, 129I, and 85Kr. Other, less volatile isotopes can also report to the off-gas streams in a reprocessing facility. These were reported to be isotopes of Cs, Cd, Ru, Sb, Tc, and Te. In this report, an effort is made to determine which, if any, of 24 semivolatile radionuclides could be released from a reprocessing plant and, if so, what would be the likely quantities released. As part of this study of semivolatile elements, the amount of each generated during fission is included as part of the assessment for the need to control their emission. Also included in this study is the assessment of the cooling time (time out of reactor) before the fuel is processed. This aspect is important for the short-lived isotopes shown in the list, especially for cooling times approaching 10 y. The approach taken in this study was to determine if semivolatile radionuclides need to be included in a list of gas-phase radionuclides that might need to be removed to meet Environmental Protection Agency (EPA) and Nuclear Regulatory Commission (NRC) regulations. A list of possible elements was developed through a literature search and through knowledge and literature on the chemical processes in typical aqueous processing of nuclear fuels. A long list of possible radionuclides present in irradiated fuel was generated and then trimmed by considering isotope half-life and calculating the dose from each to a maximum exposed individual with the US EPA airborne radiological dispersion and risk assessment code CAP88 (Rosnick 1992) to yield a short list of elements that actually need to be considered for control because they require high decontamination factors to meet a reasonable fraction of the regulated release. Each of these elements is

Back-end of the research reactor fuel cycle

International Nuclear Information System (INIS)

Gruber, Gehard J.

1996-01-01

This paper outlines the status of topics and issues related to: (1) Research Reactor Spent Nuclear Fuel Return to the U.S., including policy, shipments and ports of entry, management sites, fees, storage technologies, contracts, actual shipment, and legal process, (2) UKAEA: MTR Spent Nuclear Fuel Reprocessing, (3) COGEMA: MTR Spent Nuclear Fuel Reprocessing, and (4) Intermediate Storage + Direct Disposal for Research Reactors. (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

Simulation tools and new developments of the molten salt fast reactor

International Nuclear Information System (INIS)

Merle-Lucotte, E.; Doligez, X.; Heuer, D.; Allibert, M.; Ghetta, V.

2010-01-01

Starting from the Molten Salt Breeder Reactor project of Oak-Ridge, we have performed parametric studies in terms of safety coefficients, reprocessing requirements and breeding capabilities. In the frame of this major re-evaluation of the molten salt reactor (MSR), we have developed a new concept called Molten Salt Fast Reactor or MSFR, based on the Thorium fuel cycle and a fast neutron spectrum. This concept has been selected for further studies by the MSR steering committee of the Generation IV International Forum in 2009. Our reactor's studies of the MSFR concept rely on numerical simulations making use of the MCNP neutron transport code coupled with a code for materials evolution which resolves the Bateman's equations giving the population of each nucleus inside each part of the reactor at each moment. Because of MSR's fundamental characteristics compared to classical solid-fuelled reactors, the classical Bateman equations have to be modified by adding two terms representing the reprocessing capacities and the fertile or fissile alimentation. We have thus coupled neutronic and reprocessing simulation codes in a numerical tool used to calculate the extraction efficiencies of fission products, their location in the whole system (reactor and reprocessing unit) and radioprotection issues. (authors)

Molten salt reactors

International Nuclear Information System (INIS)

Bouchter, J.C.; Dufour, P.; Guidez, J.; Simon, N.; Renault, C.

2014-01-01

Molten salt reactors are one of the 6 concepts retained for the 4. generation of nuclear reactors. The principle of this reactor is very innovative: the nuclear fuel is dissolved in the coolant which allows the online reprocessing of the fuel and the online recovery of the fission products. A small prototype: the Molten Salt Reactor Experiment (MSRE - 8 MWt) was operating a few years in the sixties in the USA. The passage towards a fast reactor by the suppression of the graphite moderator leads to the concept of Molten Salt Fast Reactor (MSFR) which is presently studied through different European projects such as MOST, ALISIA and EVOL. Worldwide the main topics of research are: the adequate materials resisting to the high level of corrosiveness of the molten salts, fuel salt reprocessing, the 3-side coupling between neutron transport, thermohydraulics and thermo-chemistry, the management of the changing chemical composition of the salt, the enrichment of lithium with Li 7 in the case of the use of lithium fluoride salt and the use of MSFR using U 233 fuel (thorium cycle). The last part of the article presents a preliminary safety analysis of the MSFR. (A.C.)

United Reprocessors' arrangements for storage and reprocessing of irradiated fuel from power reactors

International Nuclear Information System (INIS)

Anon.

1974-01-01

United Reprocessors are of the opinion thate the European demand for reprocessing can be covered despite the numerous technical, financial and commercial problems yet to be solved, on the condition that new safety requirements and conditions do not lead to extensive delays. Considering the exceptionally high investments to be transacted, a financial support from the customers will be necessary. It is expected that in the second half of the 80's, reprocessing plants will be built with the technical support of United Reprocessors in other countries. (orig./LH) [de

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

Proposal of a fast gas-cooled reactor using transuranics

International Nuclear Information System (INIS)

Macedo, Anderson Altair Pinheiro de

2016-01-01

In the last two decades, nations that have invested in research and energy generation through nuclear source have devoted part of their efforts in developing new technologies for nuclear reactors. Part of this investment focuses on new material testing, particularly regarding new fuels. In a world view that breaths sustainability, the reprocess and reuse of spent fuel from conventional reactors comes alive in nuclear technology, presenting itself as a real alternative of energy source for the latest generation of reactors. Different concepts of fourth generation reactors have been proposed and must meet some basic requirements, such as: extended burnup, improvement of passive safety, better radioactive waste management, possibility to use reprocessed fuel and proliferation resistance. In this context, the GFR (Gas-cooled Fast Reactor) is one of the future promises, presenting satisfactory neutronic results on the use of type of fuel (U, Pu) C. In the present work, the fuel of a traditional GFR reactor that uses (U, Pu)C was sub was replaced by a transuranic reprocessed fuel (TRU), obtained by non-proliferation reprocessing technology. The UO 2 fuel initially enriched by 3.1% was burned in a standard PWR, with full burn of 33,000 MWd/T. Afterward it was left in a pool for 5 years and finally reprocessed by UREX + method. Two fuels were studied and evaluated, one diluted with depleted uranium (U, TRU)C, and the other diluted in thorium (Th, TRU)C. Assessments were done in steady state and as well as during burning and were compared with results obtained using the standard fuel, (U, Pu) C. The outcome shows that the use of TRU as a fuel, in GFR type reactors, is a real possibility. The research was done using the SCALE 6.0 code modules. (author)

Method of reprocessing spent nuclear fuels

International Nuclear Information System (INIS)

Kamiyama, Hiroaki; Inoue, Tadashi; Miyashiro, Hajime.

1987-01-01

Purpose: To facilitate the storage management for the wastes resulting from reprocessing by chemically separating transuranium elements such as actionoid elements together with uranium and plutonium. Method: Spent fuels from a nuclear reactor are separated into two groups, that is, a mixture of uranium, plutonium and transuranium elements and cesium, strontium and other nuclear fission products. Virgin uranium is mixed to adjust the mixture of uranium, plutonium and transuranium elements in the first group, which is used as the fuels for the nuclear reactor. After separating to recover useful metals such as cesium and strontium are separated from short half-decay nuclear fission products of the second group, other nuclear fission products are stored and managed. This enables to shorten the storage period and safety storage and management for the wastes. (Takahashi, M.)

Spent nuclear fuel reprocessing and international law. Germany's obligations under international law in matters of spent fuel reprocessing and the relevant contracts concluded with France and the United Kingdom

International Nuclear Information System (INIS)

Heintschel v Heinegg, W.

1999-01-01

The review presented is an excerpt from an expert opinion written by the author in December last year, in response to changes in nuclear energy policy announced by the new German government. The reprocessing of spent nuclear fuels from German power reactors in the reprocessing facilities of France (La Hague) and the UK (Sellafield) is not only based on contracts concluded by the German electric utilities and the French COGEMA or the British BNFL, but has been agreed as well by an exchange of diplomatic notes between the French Ministry of Foreign Affairs and the German ambassador in Paris, the German Foreign Ministry and the French ambassador as well as the British ambassador in Bonn. The article therefore first examines from the angle of international law the legal obligations binding the states involved, and Germany in particular, in matters of spent fuel reprocessing contracts. The next question arising in this context and discussed by the article is that of whether and how much indemnification can be demanded by the reprocessing companies, or their governments, resp., if Germany should discontinue spent fuel reprocessing and thus might be made liable for breach of the bilateral agreements. (orig/CB) [de

Environmental assessment for Breeder Reprocessing Engineering Test (BRET): Revision 1

International Nuclear Information System (INIS)

Lerch, R.E.

1989-03-01

This Environmental Assessment (EA) is for the proposed installation and operation of an integrated breeder fuel reprocessing test system in the shielded cells of the Fuels and Materials Examination Facility (FMEF) at Hanford and the associated modifications to the FMEF to accommodate BRET. These modifications would begin in FY-1986 subject to Congressional authorization. Hot operations would be scheduled to start in the early 1990's. The system, called the Breeder Reprocessing Engineering Test (BRET), is being designed to provide a test capability for developing the demonstrating fuel reprocessing, remote maintenance, and safeguards technologies for breeder reactor fuels. This EA describes (1) the action being proposed, (2) the existing environment which would be affected, (3) the potential environmental impacts from normal operations and severe accidents from the proposed action, (4) potential conflicts with federal, state, regional, and/or local plans for the area, and (5) environmental implications of alternatives considered to the proposed action. 41 refs., 10 figs., 31 tabs

Remotex and servomanipulator needs in nuclear fuel reprocessing plants

International Nuclear Information System (INIS)

Garin, J.

1981-01-01

Work on the conceptual design of a pilot-scale plant for reprocessing breeder reactor fuels is being performed at Oak Ridge National Laboratory. The plant design will meet all current federal regulations for repocessing plants and will serve as prototype for future production plants. A unique future of the concept is the incorporation of totally remote operation and maintenance of the process equipment within a large barn-like hot cell. This approach, caled Remotex, utilizes servomanipulators coupled with television viewing to extend man's capabilities into the hostile cell environment. The Remotex concept provides significant improvements for fuel reprocessing plants and other nuclear facilities in the areas of safeguarding nuclear materials, reducing radiation exposure, improving plant availability, recovering from unplanned events, and plant decommissioning

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

Prospect of spent fuel reprocessing and back-end cycling in China in 1990's

International Nuclear Information System (INIS)

Ke Youzhi; Wang Rengtao

1987-01-01

According to the CHinese Program of nuclear energy in 1990's, the amount of spent fuel by the year 2000 is estimated in this paper. Reprocessing is considered as an important link in the back-end fuel cycle. A pilot plant is scheduled for hot start up in 1996. The main goal of the study is LWR spent fuel reprocessing. We will use the experience gained from reprocessing of production reactor fuel and last research results. The advanced foreign technigue and experience will be introduced. The study emphasizes on the test of technology, equipments, instrumentation and automation, development of remote maintenance and decontamination. China will start to demonstrate the way for fuel cycle. (author)

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.

Fast reactors - Dounreay and the future

International Nuclear Information System (INIS)

Jordan, G.

1988-01-01

In 1960 at Dounreay, the Dounreay Fast Reactor (DFR) supplied the world's first fast reactor grid electricity, and went on to a highly successful career as a test facility, as fuel designs advanced. In the 1960s, the Prototype Fast Reactor (PFR) was designed and built, beginning operation in 1974. The PFR was built to provide a sound technical and experienced base to support the UK's future Fast Reactor development and design. The in-vessel fuel handling facilities have demonstrated the flexibility of the pool design and a considerable body of in-core fuel handling experience is available. A key issue for further Fast Reactor application is the performance of fuel and, because PFR was designed to take full-scale fuel assemblies, the fuel performance experience is directly relevant to commercial designs. The original PFR design irradiation target of 60000 MWd/t U (equivalent to 7.5 % burn-up) has already been exceeded by a factor of more than two and a 15.9 % burn-up sub-assembly has been discharged and reprocessed without difficulty. Soon a 20 % sub-assembly will follow. Also the PFR reprocessing plant has demonstrated the safety and efficiency of this essential adjunct to Fast Reactor operation. The safety and the environmental protection features of both the PFR and its fuel reprocessing plant have been demonstrated over the last 14 years. 2 refs., 3 figs

Water storage of liquid-metal fast-breeder-reactor fuel

International Nuclear Information System (INIS)

Meacham, S.A.

1982-01-01

The purpose of this paper is to present a general overview of a concept proposed for receiving and storing liquid metal fast breeder reactor (LMFBR) spent fuel. This work was done as part of the Consolidated Fuel Reprocessing Program (CFRP) at the Oak Ridge National Laboratory (ORNL). The CFRP has as its major objective the development of technology for reprocessing advanced nuclear reactor fuels. The program plans that research and development will be carried through to a sufficient scale, using irradiated spent fuel under plant operating conditions, to establish a basis for confident projection of reprocessing capability to support a breeder industry

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

International Nuclear Information System (INIS)

Ozawa, T.

1974-01-01

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

Plutonium, proliferation, and the price of reprocessing

International Nuclear Information System (INIS)

Gilinsky, V.

1978-01-01

France and Britain disagree with the US on whether deferring fuel reprocessing that provides plutonium for export can help contain proliferation. The US has veto power over reprocessing of US-supplied fuels for non-EURATOM countries, but exceptions will be made for movement within the EURATOM community. Political issues will be influenced by the magnitude of the financial investments, however, and commercial considerations have until recently dominated and complicated international safeguards. The author notes that US policy was reversed by the gradual acknowledgment that the same international inspection of plutonium stockpiles would not work as it had for low-enriched fuel and that economic interests must have a lower priority to avoiding proliferation. He cites the combination of sudden policy shifts, failure to prove that present reactors are best, and long-term distrust of US economic motives as failing to persuade either the French or British, who feel the best safeguard is provided by their high-security reprocessing facilities. Still to be resolved are the conditions under which plutonium must be returned to its owners, a problem that must determine safe international transport and storage and international management. Technical fixes, such as the CIVEX process, cannot contribute to the solution for several decades, while reprocessing is no longer considered a first step in waste disposal and would be more expensive and complicated than present waste disposal procedures. The author concedes merit in President Carter's requirement of separating ''the legitimate and necessary use of uranium'' and nuclear fuels that are also explosives

Management of radioactive waste from reprocessing including disposal aspects

International Nuclear Information System (INIS)

Malherbe, J.

1991-01-01

Based on a hypothetical scenario including a reactor park of 20 GWe consisting of Pressurised-Water-Reactors with a resulting annual production of 600 tonnes of heavy metal of spent fuel, all aspects of management of resulting wastes are studied. Waste streams from reprocessing include gaseous and liquid effluents, and a number of solid conditioned waste types. Disposal of waste is supposed to be performed either in a near-surface engineered repository, as long as the content of alpha-emitting radionuclides is low enough, and in a deep geological granite formation. After having estimated quantities, cost and radiological consequences, the sensitivity of results to modification in reactor park size, burn-up and the introduction of mixed-oxide fuel (MOX) is evaluated

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

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)

Energies and media nr 30. Conditions for the nuclear sector. The fuel cycle and wastes. The usefulness of fuel reprocessing. Wastes

International Nuclear Information System (INIS)

2009-10-01

After some comments on recent events in the nuclear sector in different countries (energy policy and projects in the USA, Europe, China, India, Russia), this issue proposes some explanations on the nuclear fuel cycle and on nuclear wastes: involved processes and products from mining to reprocessing and recycling, usefulness of reprocessing (future opportunities of fast neutron reactors, present usefulness of reprocessing with the recycling of separated fissile materials), impact of reprocessing on the environment in La Hague (gas and liquid releases, release standard definition), and the destiny of wastes

The tritium and the controlled fusion reactors

International Nuclear Information System (INIS)

Leger, D.; Rouyer, J.L.

1986-04-01

It is shown how tritium is used how it is circulating in a fusion reactor. The great functions of tritium circuits are detailed: reprocessing of burnt gases, reprocessing of gases coming from neutral injectors, reprocessing from gaseous wastes, detritiation of cooling fluids. Current technologic developments are quoted. Then tritium confinement and containment, in normal or accidental situations, are displayed. Limitation devices of effluents and release for normal operating (noticeably the reprocessing systems of atmosphere) and safety and protection systems in case of accident are described [fr

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.

A survey of methods to immobilize tritium and carbon-14 arising from a nuclear fuel reprocessing plant

International Nuclear Information System (INIS)

Taylor, P.

1991-02-01

This report reviews the literature on methods to separate and immobilize tritium ( 3 H) and carbon-14 ( 14 C) released from U0 2 fuel in a nuclear fuel reprocessing plant. It was prepared as part of a broader review of fuel reprocessing waste management methods that might find future application in Canada. The calculated inventories of both 3 H and 14 C in used fuel are low; special measures to limit releases of these radionuclides from reprocessing plants are not currently in place, and may not be necessary in future. If required, however, several possible approaches to the concentration and immobilization of both radionuclides are available for development. Technology to control these radionuclides in reactor process streams is in general more highly developed than for reprocessing plant effluent, and some control methods may be adaptable to reprocessing applications

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

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)

International safeguards for a light-water reactor fuels reprocessing plant: containment and surveillance concepts

International Nuclear Information System (INIS)

Cameron, C.P.; Bleck, M.E.

1980-12-01

Concepts for containment/surveillance for reprocessing plants are described, conceptual designs are developed, and their effectiveness is evaluated. A technical approach to design of containment/surveillance systems is presented, and design considerations are discussed. This is the second in a series of reports. The first described the basis for the study of international safeguards for reprocessing plants. In this second report, only containment/surveillance is discussed. The third report will discuss the integration of concepts for containment/surveillance and material accountancy

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

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.

Tritium control in reprocessing plants

International Nuclear Information System (INIS)

Goumondy, J.P.; Miquel, P.

1977-01-01

There is a danger that the T which is formed in water reactors will prove detrimental to the environment over the next few years, and studies have been undertaken to develop techniques to contain and process it where possible. In order to retain T, which is present largely in the fuel and on the possible to adapt for use in the conventional design of reprocessing plant. In this process T is maintained in the form of an aqueous solution in the high-active area of the plant. Control is achieved by restricting as far as possible the ingress of non-tritiated water into this area, and by setting up a tritiated water barrier at the first U and Pu extraction stage, stripping the tritium-containing solvent at that point with ordinary water. In this way the T can be extracted in a small volume of water with a view to intermediate storage, disposal at sea additional processing to remove the T from the water. Experiments carried out so far have demonstrated the effectiveness of the T barrier and have shown what equipment would be required for the application of the process in new reprocessing plants. (orig.) [de

Concerning results of environmental monitoring around the reprocessing facilities of Power Reactor and Nuclear Fuel Development Corporation

International Nuclear Information System (INIS)

1989-01-01

The Central Evaluation Expert Group for Environmental radiation Monitoring has been engaged in examinations of plants for and results of the environmental radiation monitoring performed by Power Reactor and Nuclear Fuel Development Corporation around its reprocessing facilities. The present report outlines an examination of the results of monitoring carried out in 1987 (January to December). It is concluded that the methods used for the monitoring and its technical level are satisfactory in meeting the objectives of the monitoring plans. Expept for tritium in seawater, the level of radiations stays within the normal variation determined based on preliminary measurements of the background radiation. The procedure used for the calculation of exposure dose is also satisfactory in meeting the requirements specified in the monitoring plants. It is confirmed that the exposure dose of the residents around the facilities is well below the permissible exposure dose limite specified in law. (Nogami. K.)

Peculiarities of highly burned-up NPP SNF reprocessing and new approach to simulation of solvent extraction processes

Energy Technology Data Exchange (ETDEWEB)

Fedorov, Y.S.; Zilberman, B.Y.; Goletskiy, N.D.; Puzikov, E.A.; Ryabkov, D.V.; Rodionov, S.A.; Beznosyuk, V.I.; Petrov, Y.Y.; Saprykin, V.F.; Murzin, A.A.; Bibichev, B.A.; Aloy, A.S.; Kudinov, A.S.; Blazheva, I.V. [RPA ' V.G.Khlopin Radium Institute' , 28, 2 Murinsky av., St-Petersburg, 194 021 (Russian Federation); Kurenkov, N.V. [Institute of Industrial Nuclear Technology NRNU MEPHI, 31, Kashirskoye shosse, Moscow, 115409 (Russian Federation)

2013-07-01

Substantiation, general description and performance characteristics of a reprocessing flowsheet for WWER-1000 spent fuel with burn-up >60 GW*day/t U is given. Pu and U losses were <0.1%, separation factor > 10{sup 4}; their decontamination factor from γ-emitting fission products was 4*10{sup 4} and 3*10{sup 7}, respectively. Zr, Tc, Np removal was >98% at U and Pu losses <0.05%. A new approach to simulation of extraction equilibrium has been developed. It is based on a set of simultaneous chemical reactions characterized by apparent concentration constants. A software package was created for simulation of spent fuel component distribution in multistage countercurrent extraction processes in the presence of salting out agents. (authors)

Experience of in-cell visual inspection using CCD camera in hot cell of Reprocessing Plant

International Nuclear Information System (INIS)

Reddy, Padi Srinivas; Amudhu Ramesh Kumar, R.; Geo Mathews, M.; Ravisankar, A.

2013-01-01

This paper describes the selection, customization and operating experience of the visual inspection system for the hot cell of a Reprocessing Plant. For process equipment such as fuel chopping machine, dissolver, centrifuge, centrifugal extractors etc., viewing of operations and maintenance using manipulators is required. For this, the service of in-cell camera is essential. The ambience of the hot cell of Compact facility for Reprocessing of Advanced fuels in Lead cell (CORAL) for the reprocessing of fast reactor spent fuel has high gamma radiation and acidic vapors. Black and white Charge Coupled Device (CCD) camera has been used in CORAL incorporating in-house modifications to suit the operating ambient conditions, thereby extending the operating life of the camera. (author)

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)

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.

Reprocessing of nuclear fuels

International Nuclear Information System (INIS)

Hatfield, G.W.

1960-11-01

One of the persistent ideas concerning nuclear power is that the fuel costs are negligible. This, of course, is incorrect and, in fact, one of the major problems in the development of economic nuclear power is to get the cost of the fuel cycles down to an acceptable level. The irradiated fuel removed from the nuclear power reactors must be returned as fresh fuel into the system. Aside from the problems of handling and shipping involved in the reprocessing cycles, the two major steps are the chemical separation and the refabrication. The chemical separation covers the processing of the spent fuel to separate and recover the unburned fuel as well as the new fuel produced in the reactor. This includes the decontamination of these materials from other radioactive fission products formed in the reactor. Refabrication involves the working and sheathing of recycled fuel into the shapes and forms required by reactor design and the economics of the fabrication problem determines to a large extent the quality of the material required from the chemical treatment. At present there appear to be enough separating facilities in the United States and the United Kingdom to handle the recycling of fuel from power reactors for the next few years. However, we understand the costs of recycling fuel in these facilities will be high or low depend ing on whether or not the capital costs of the plant are included in the processing cost. Also, the present plants may not be well adapted to carry out the chemical processing of the very wide variety of power reactor fuel elements which are being considered and will continue to be considered over the years to come. (author)

HTGR fuel reprocessing technology

International Nuclear Information System (INIS)

Brooks, L.H.; Heath, C.A.; Shefcik, J.J.

1976-01-01

The following aspects of HTGR reprocessing technology are discussed: characteristics of HTGR fuels, criteria for a fuel reprocessing flowsheet; selection of a reference reprocessing flowsheet, and waste treatment

Safety aspects of LWR fuel reprocessing and mixed oxide fuel fabrication plants

International Nuclear Information System (INIS)

Fischer, M.; Leichsenring, C.H.; Herrmann, G.W.; Schueller, W.; Hagenberg, W.; Stoll, W.

1977-01-01

The paper is focused on the safety and the control of the consequences of credible accidents in LWR fuel reprocessing plants and in mixed oxide fuel fabrication plants. Each of these plants serve for many power reactor (about 50.000 Mwel) thus the contribution to the overall risk of nuclear energy is correspondingly low. Because of basic functional differences between reprocessing plants, fuel fabrication plants and nuclear power reactors, the structure and safety systems of these plants are different in many respects. The most important differences that influence safety systems are: (1) Both fuel reprocessing and fabrication plants do not have the high system pressure that is associated with power reactors. (2) A considerable amount of the radioactivity of the fuel, which is in the form of short-lived radionuclides has decayed. Therefore, fuel reprocessing plants and mixed oxide fuel fabrication plants are designed with multiple confinement barriers for control of radioactive materials, but do not require the high-pressure containment systems that are used in LWR plants. The consequences of accidents which may lead to the dispersion of radioactive materials such as chemical explosions, nuclear excursions, fires and failure of cooling systems are considered. A reasonable high reliability of the multiple confinement approach can be assured by design. In fuel reprocessing plants, forced cooling is necessary only in systems where fission products are accumulated. However, the control of radioactive materials can be maintained during normal operation and during the above mentioned accidents, if the dissolver off-gas and vessel off-gas treatment systems provide for effective removal of radioactive iodine, radioactive particulates, nitrogen oxides, tritium and krypton 85. In addition, the following incidents in the dissolver off-gas system itself must be controlled: failures of iodine filters, hydrogen explosion in O 2 - and NOsub(x)-reduction component, decomposition of

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...

Review of fast reactor activities in India

International Nuclear Information System (INIS)

Paranjpe, S.R.

1982-01-01

A review of fast reactor activities in India is introduced. One stage of construction of the Fast Breeder Test Reactor (FBTR) and design studies for 500MWe Prototype Fast Breeder Reactor (PFBR) are briefly summarized. The emphasis is on fast reactor physics, materials studies, radiochemistry, and the safety and fuel reprocessing programme

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.

Chemistry and technology of Molten Salt Reactors - history and perspectives

International Nuclear Information System (INIS)

Uhlir, Jan

2007-01-01

Molten Salt Reactors represent one of promising future nuclear reactor concept included also in the Generation IV reactors family. This reactor type is distinguished by an extraordinarily close connection between the reactor physics and chemical technology, which is given by the specific features of the chemical form of fuel, representing by molten fluoride salt and circulating through the reactor core and also by the requirements of continuous 'on-line' reprocessing of the spent fuel. The history of Molten Salt Reactors reaches the period of fifties and sixties, when the first experimental Molten Salt Reactors were constructed and tested in ORNL (US). Several molten salt techniques dedicated to fresh molten salt fuel processing and spent fuel reprocessing were studied and developed in those days. Today, after nearly thirty years of discontinuance, a renewed interest in the Molten Salt Reactor technology is observed. Current experimental R and D activities in the area of Molten Salt Reactor technology are realized by a relatively small number of research institutions mainly in the EU, Russia and USA. The main effort is directed primarily to the development of separation processes suitable for the molten salt fuel processing and reprocessing technology. The techniques under development are molten salt/liquid metal extraction processes, electrochemical separation processes from the molten salt media, fused salt volatilization techniques and gas extraction from the molten salt medium

Performance of metallic fuels in liquid-metal fast reactors

International Nuclear Information System (INIS)

Seidel, B.R.; Walters, L.C.; Kittel, J.H.

1984-01-01

Interest in metallic fuels for liquid-metal fast reactors has come full circle. Metallic fuels are once again a viable alternative for fast reactors because reactor outlet temperature of interest to industry are well within the range where metallic fuels have demonstrated high burnup and reliable performance. In addition, metallic fuel is very tolerant of off-normal events of its high thermal conductivity and fuel behavior. Futhermore, metallic fuels lend themselves to compact and simplified reprocessing and refabrication technologies, a key feature in a new concept for deployment of fast reactors called the Integral Fast Reactor (IFR). The IFR concept is a metallic-fueled pool reactor(s) coupled to an integral-remote reprocessing and fabrication facility. The purpose of this paper is to review recent metallic fuel performance, much of which was tested and proven during the twenty years of EBR-II operation

Studies and research concerning BNFP: converting reprocessing plant's fuel receiving and storage area to an away-from-reactor (AFR) storage facility. Final report

Energy Technology Data Exchange (ETDEWEB)

Cottrell, Jim E.; Shallo, Frank A.; Musselwhite, E Larry; Wiedemann, George F.; Young, Moylen

1979-09-01

Converting a reprocessing plant's fuel receiving and storage station into an Away-From-Reactor storage facility is evaluated in this report. An engineering analysis is developed which includes (1) equipment modifications to the facility including the physical protection system, (2) planning schedules for licensing-related activities, and (3) cost estimates for implementing such a facility conversion. Storage capacities are evaluated using the presently available pools of the existing Barnwell Nuclear Fuel Plant-Fuel Receiving and Storage Station (BNFP-FRSS) as a model.

Radioactive waste management in a fuel reprocessing facility in fiscal 1982

International Nuclear Information System (INIS)

1984-01-01

In the fuel reprocessing facility of the Power Reactor and Nuclear Fuel Development Corporation, radioactive gaseous and liquid waste are released not exceeding the respective permissible levels. Radioactive concentrated solutions are stored at the site. Radioactive solid waste are stored appropriately at the site. In fiscal 1982, the released quantities of radioactive gaseous and liquid waste were both below the permissible levels. The results of radioactive waste management in the fuel reprocessing facility in fiscal 1982 are given in the tables: the released quantities of radioactive gaseous and liquid waste, the produced quantities of radioactive solid waste, and the stored quantities of radioactive concentrated solutions and of radioactive solid waste as of the end of fiscal 1982. (Mori, K.)

Aqueous reprocessing - some dreams!

International Nuclear Information System (INIS)

Srinivasan, T.G.

2015-01-01

India has been pursuing a aqueous reprocessing based closed fuel cycle for both thermal and fast reactor fuels employing the PUREX process. Though the country has more than six decades of experience, the dreams or wish lists such as, a highly efficient process with textbook specifications of 99.9% recovery of U and Pu, a DF of more than 10 7 for both U and Pu from the fission products, operating with name plate capacity with high safety, low waste generation, recovery of useful fission products and minor actinides from high level waste are never ceasing and ever growing. The talk will cover safety precautions and actions to be taken in the steps listed below, to ensure a safe and successful process

Fast reactor fuel design and development

International Nuclear Information System (INIS)

Bishop, J.F.W.; Chamberlain, A.; Holmes, J.A.G.

1977-01-01

Fuel design parameters for oxide and carbide fast reactor fuels are reviewed in the context of minimising the total uranium demands for a combined thermal and fast reactor system. The major physical phenomena conditioning fast reactor fuel design, with a target of high burn-up, good breeding and reliable operation, are characterised. These include neutron induced void swelling, irradiation creep, pin failure modes, sub-assembly structural behaviour, behaviour of defect fuel, behaviour of alternative fuel forms. The salient considerations in the commercial scale fabrication and reprocessing of the fuels are reviewed, leading to the delineation of possible routes for the manufacture and reprocessing of Commercial Reactor fuel. From the desiderata and restraints arising from Surveys, Performance and Manufacture, the problems posed to the Designer are considered, and a narrow range of design alternatives is proposed. The paper concludes with a consideration of the development areas and the conceptual problems for fast reactors associated with those areas

Thorex reprocessing characterization

International Nuclear Information System (INIS)

1978-11-01

The purpose of this report is to bring together, in highly condensed form, information which would need to be considered in planning a commercial reprocessing plant for recovering 233 U-Th reactor fuel. This report does not include a discussion of process modifications which would be required for thorium-base fuels that contain plutonium (such as would be required for thorium fuels containing 235 U or 233 U denatured with 238 U). It is the intent of this paper to address only the basic Thorex process for treating 233 U-Th fuels. As will be pointed out, the degree of development of the various proposed operations varies widely, from preliminary laboratory experiments for the dissolution of Zircaloy-clad thoria to engineering scale demonstration of the recovery of moderately irradiated thorium by a solvent extraction process (Thorex)

The reprocessing-recycling of spent nuclear fuel. Actinides separation - Application to wastes management

International Nuclear Information System (INIS)

2008-01-01

After its use in the reactor, the spent fuel still contains lot of recoverable material for an energetic use (uranium, plutonium), but also fission products and minor actinides which represent the residues of nuclear reactions. The reprocessing-recycling of the spent fuel, as it is performed in France, implies the chemical separation of these materials. The development and the industrial implementation of this separation process represent a major contribution of the French science and technology. The reprocessing-recycling allows a good management of nuclear wastes and a significant saving of fissile materials. With the recent spectacular rise of uranium prices, this process will become indispensable with the development of the next generation of fast neutron reactors. This book takes stock of the present and future variants of the chemical process used for the reprocessing of spent fuels. It describes the researches in progress and presents the stakes and recent results obtained by the CEA. content: the separation of actinides, a key factor for a sustainable nuclear energy; the actinides, a discovery of the 20. century; the radionuclides in nuclear fuels; the aquo ions of actinides; some redox properties of actinides; some complexing properties of actinide cations; general considerations about treatment processes; some characteristics of nuclear fuels in relation with their reprocessing; technical goals and specific constraints of the PUREX process; front-end operations of the PUREX process; separation and purification operations of the PUREX process; elaboration of finite products in the framework of the PUREX process; management and treatment of liquid effluents; solid wastes of the PUREX process; towards a joint management of uranium and plutonium: the COEX TM process; technical options of treatment and recycling techniques; the fuels of generation IV reactors; front-end treatment processes of advanced fuels; hydrometallurgical processes for future fuel cycles

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

Fast reactors fuel Cycle: State in Europe

International Nuclear Information System (INIS)

1991-01-01

In this SFEN day we treat all aspects (economics-reactor cores, reprocessing, experience return) of the LMFBR fuel cycle in Europe and we discuss about the development of this type of reactor (EFR project) [fr

Indian program on management of high level radioactive waste - emphasis on value recovery for societal applications

International Nuclear Information System (INIS)

Kaushik, C.P.; Tomar, Neelima Singh; Kumar, Amar; Wadhwa, S.; Diwan, Jyoti

2017-01-01

Nuclear Power Programme in India is based on 'closed fuel cycle'. Closed fuel cycle involves reprocessing and recycling of Spent Nuclear Fuel (SNF) coming out of nuclear reactors. During reprocessing, uranium and plutonium, constituting bulk of the SNF are separated and subsequently recycled. The remaining small portion constitutes high level radioactive waste containing most of the fission products and minor actinides. A three-step strategy involving immobilization, interim storage followed by ultimate disposal has been adopted in India for management of High Level Waste (HLW). Borosilicate glass matrix has been identified for immobilization of HLW owing to optimal waste loading, adequate leach resistance and long term stability of the product. An interim storage facility is in operation for storage and surveillance of VWP. A comprehensive program based on screening of different materials like granite, argillite, clay with respect to sorption of different radionuclides is being pursued to identify the suitable areas of disposal of the conditioned waste products. Separation of useful radionuclides like "1"3"7Cs, "9"0Sr, "9"0Y, "1"0"6Ru etc and its utilization for societal applications is being practiced in India. (author)

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

Fuel cycle cost analysis on molten-salt reactors

International Nuclear Information System (INIS)

Shimazu, Yoichiro

1976-01-01

An evaluation is made of the fuel cycle costs for molten-salt reactors (MSR's), developed at Oak Ridge National Laboratory. Eight combinations of conditions affecting fuel cycle costs are compared, covering 233 U-Th, 235 U-Th and 239 Pu-Th fuels, with and without on-site continuous fuel reprocessing. The resulting fuel cycle costs range from 0.61 to 1.18 mill/kWh. A discussion is also given on the practicability of these fuel cycles. The calculations indicate that somewhat lower fuel cycle costs can be expected from reactor operation in converter mode on 235 U make-up with fuel reprocessed in batches every 10 years to avoid fission product precipitation, than from operation as 233 U-Th breeder with continuous reprocessing. (auth.)

Extending Spent Fuel Storage until Transport for Reprocessing or Disposal

Energy Technology Data Exchange (ETDEWEB)

Carlsen, Brett; Chiguer, Mustapha; Grahn, Per; Sampson, Michele; Wolff, Dietmar; Bevilaqua, Arturo; Wasinger, Karl; Saegusa, Toshiari; Seelev, Igor

2016-09-01

Spent fuel (SF) must be stored until an end point such as reprocessing or geologic disposal is imple-mented. Selection and implementation of an end point for SF depends upon future funding, legisla-tion, licensing and other factors that cannot be predicted with certainty. Past presumptions related to the availability of an end point have often been wrong and resulted in missed opportunities for properly informing spent fuel management policies and strategies. For example, dry cask storage systems were originally conceived to free up needed space in reactor spent fuel pools and also to provide SFS of up to 20 years until reprocessing and/or deep geological disposal became available. Hundreds of dry cask storage systems are now employed throughout the world and will be relied upon well beyond the originally envisioned design life. Given present and projected rates for the use of nuclear power coupled with projections for SF repro-cessing and disposal capacities, one concludes that SF storage will be prolonged, potentially for several decades. The US Nuclear Regulatory Commission has recently considered 300 years of storage to be appropriate for the characterization and prediction of ageing effects and ageing management issues associated with extending SF storage and subsequent transport. This paper encourages addressing the uncertainty associated with the duration of SF storage by de-sign – rather than by default. It suggests ways that this uncertainty may be considered in design, li-censing, policy, and strategy decisions and proposes a framework for safely extending spent fuel storage until SF can be transported for reprocessing or disposal – regardless of how long that may be. The paper however is not intended to either encourage or facilitate needlessly extending spent fuel storage durations. Its intent is to ensure a design and safety basis with sufficient margin to accommodate the full range of potential future scenarios. Although the focus is primarily on

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....

Operation of Nuclear Fuel Based on Reprocessed Uranium for VVER-type Reactors in Competitive Nuclear Fuel Cycles

Energy Technology Data Exchange (ETDEWEB)

Troyanov, V.; Molchanov, V.; Tuzov, A. [TVEL Corporation, 49 Kashirskoe shosse, Moscow 115409 (Russian Federation); Semchenkov, Yu.; Lizorkin, M. [RRC ' Kurchatov Institute' (Russian Federation); Vasilchenko, I.; Lushin, V. [OKB ' Gidropress' (Russian Federation)

2009-06-15

Current nuclear fuel cycle of Russian nuclear power involves reprocessed low-enriched uranium in nuclear fuel production for some NPP units with VVER-type LWR. This paper discusses design and performance characteristics of commercial nuclear fuel based on natural and reprocessed uranium. It presents the review of results of commercial operation of nuclear fuel based on reprocessed uranium on Russian NPPs-unit No.2 of Kola NPP and unit No.2 of Kalinin NPP. The results of calculation and experimental validation of safe fuel operation including necessary isotope composition conformed to regulation requirements and results of pilot fuel operation are also considered. Meeting the customer requirements the possibility of high burn-up achieving was demonstrated. In addition the paper compares the characteristics of nuclear fuel cycles with maximum length based on reprocessed and natural uranium considering relevant 5% enrichment limitation and necessity of {sup 236}U compensation. The expedience of uranium-235 enrichment increasing over 5% is discussed with the aim to implement longer fuel cycles. (authors)

Commissioning and Operational Experience in Power Reactor Fuel Reprocessing Plant

Energy Technology Data Exchange (ETDEWEB)

Pradhan, S., E-mail: spradhan@barctara.gov.in [Tarapur Based Reprocessing Plant, Bhabha Atomic Research Centre, Tarapur (India)

2014-10-15

After completing design, construction, commissioning, operation and maintenance experience of the reprocessing plants at Tarapur, Mumbai and Kalpakkam a new reprocessing plant is commissioned and put into operation at BARC, Tarapur since 2011. Subsequent to construction clearance, commissioning of the plant is taken in many steps with simultaneous review by design and safety committees. In spite of vast experience, all the staff was retrained in various aspects of process and utility operations and in operation of innovative changes incorporated in the design. Operating personnel are licensed through an elaborate procedure consisting of various check lists followed by personnel interview. Commissioning systems were divided in sub-systems. Sub-systems were commissioned independently and later integrated testing was carried out. For commissioning, extreme operating conditions were identified in consultation with designers and detailed commissioning procedures were made accordingly. Commissioning was done in different conditions to ensure safety, smooth operation and maintainability. Few modifications were carried out based on commissioning experience. Technical specifications for operation of the plant are made in consultation with designers and reviewed by safety committees. Operation of the plant was carried out after successful commissioning trials with Deep Depleted Uranium (DDU). Emergency operating procedures for each design basis accident were made. Performance of various systems, subsystems are quite satisfactory and the plant has given very good capacity factor. (author)

Commercial U.S. Vendors Focus on Reducing the Cost of Fast Reactors

International Nuclear Information System (INIS)

Fraser, Dan; Kouhestani, Amir; Parmentola, John; Prince, Robert; Reynolds, Roger

2013-01-01

A US commercial perspective: • Focus on economic benefits of fast reactors (Market Determined): • Compact design, high energy density (modularity); • High temperature output; • Improved energy conversion efficiency; • Process heat application market; • Inherent safety capabilities. • Focus on burn or “breed and burn”; • Helps avoid some non-proliferation challenges. • Focus on reactor alone, not on a re-processing plant; • Reactor is the first place for economic payback; • Happy to burn reprocessed fuel if available

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

The uranium-plutonium breeder reactor fuel cycle

International Nuclear Information System (INIS)

Salmon, A.; Allardice, R.H.

1979-01-01

All power-producing systems have an associated fuel cycle covering the history of the fuel from its source to its eventual sink. Most, if not all, of the processes of extraction, preparation, generation, reprocessing, waste treatment and transportation are involved. With thermal nuclear reactors more than one fuel cycle is possible, however it is probable that the uranium-plutonium fuel cycle will become predominant; in this cycle the fuel is mined, usually enriched, fabricated, used and then reprocessed. The useful components of the fuel, the uranium and the plutonium, are then available for further use, the waste products are treated and disposed of safely. This particular thermal reactor fuel cycle is essential if the fast breeder reactor (FBR) using plutonium as its major fuel is to be used in a power-producing system, because it provides the necessary initial plutonium to get the system started. In this paper the authors only consider the FBR using plutonium as its major fuel, at present it is the type envisaged in all, current national plans for FBR power systems. The corresponding fuel cycle, the uranium-plutonium breeder reactor fuel cycle, is basically the same as the thermal reactor fuel cycle - the fuel is used and then reprocessed to separate the useful components from the waste products, the useful uranium and plutonium are used again and the waste disposed of safely. However the details of the cycle are significantly different from those of the thermal reactor cycle. (Auth.)

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.

Papers presented at the 7th status report of the Reprocessing and Waste Treatment Project

International Nuclear Information System (INIS)

1988-10-01

The report contains all lectures on the present state of the R+D work which were delivered on the occasion of the 7th seminar on the present state of the Reprocessing and Waste Treatment Project in the Karlsruhe Nuclear Research Centre on 15/16 March 1988. The project is aimed at improving methods for nuclear waste disposal with regard to the German reprocessing plant for light-water reactor fuels currently under construction in Wackersdorf. The individual contributions were separately adapted for the INIS and EDB data banks. (RB) [de

Safety problems in fuel reprocessing plants

International Nuclear Information System (INIS)

Amaury, P.; Jouannaud, C.; Niezborala, F.

1979-01-01

The document first situates the reprocessing in the fuel cycle as a whole. It shows that a large reprocessing plant serves a significant number of reactors (50 for a plant of 1500 tonnes per annum). It then assesses the potential risks with respect to the environment as well as with respect to the operating personnel. The amounts of radioactive matter handled are very significant and their easily dispersible physical form represents very important risks. But the low potential energy likely to bring about this dispersion and the very severe and plentiful confinement arrangements are such that the radioactive risks are very small, both with respect to the environment and the operating personnel. The problems of the interventions for maintenance or repairs are mentioned. The intervention techniques in a radioactive environment are perfected, but they represent the main causes of operating personnel irradiation. The design principle applied in the new plants take this fact into account, involving a very significant effort to improve the reliability of the equipment and ensuring the provision of devices enabling the failing components to be replaced without causing irradiation of the personnel [fr

Reprocessing and clearance as ways for reducing radioactive waste from fusion reactors

Energy Technology Data Exchange (ETDEWEB)

Rocco, P.; Zucchetti, M. [Energetics Dept., Polytechnic of Turin (Italy); Zucchetti, M. [European Commission, JRC, Institute for Advanced Material, Ispra, Vatican City State, Holy See (Italy)

1998-07-01

The irradiated material chosen to investigate this concept is V-4Cr-4Ti, used as in-vessel structural material of PM-1, one of the three reactor models of SEAFP-2 assessments. The analysis performed deserves the following comments: in general, the modified concentrations of the parent isotopes have been assumed equal to the actual detection limits of the concerned elements. The nuclides investigated are those having the higher activities in the irradiated alloy (dominant nuclides). The contribution to the clearance index of other nuclides, present in the irradiated alloy and not yet examined should also be assessed. Even having reduced the concentrations of the impurities to the very low levels hypothesized, the ratios (A/L{sub c}){sub m} of some nuclides are too great. Beside K-42 and Ar-42, deriving from titanium, this happens for nuclides deriving from niobium, silver, nickel, copper, strontium. Hence the clearance index of the modified alloy is greater than the unity and clearance conditions are not achieved. Additional purification process have to be envisaged after the irradiation, with an elemental dilution of the noxious nuclides and then performing a purification process. As an example, it is found from table 2 that the sum of the ratios (A/L{sub c}){sub m} of Nb-93m, Nb-91 and Nb-94, nuclides deriving from a concentration of 2 x 10{sup -2} ppm of niobium, is 7.75. An addition of 5/10 ppm of natural niobium to the molten irradiated material and a further purification of niobium to levels near to the initial one could achieve a 100-fold reduction of the ratios (A/L{sub c}){sub m}. 10000 t of V-4Cr4Ti, representing the total amount (maintenance + decommissioning) arising from the in vessel structures of a power reactor should be reprocessed with 100 kg of niobium, in the highly conservative hypothesis that all material is irradiated in first wall conditions. The secondary waste arising from the purification processes would not exceed a few ton which could be

Future fuel cycle development for CANDU reactors

International Nuclear Information System (INIS)

Hatcher, S.R.; McDonnell, F.N.; Griffiths, J.; Boczar, P.G.

1987-01-01

The CANDU reactor has proven to be safe and economical and has demonstrated outstanding performance with natural uranium fuel. The use of on-power fuelling, coupled with excellent neutron economy, leads to a very flexible reactor system with can utilize a wide variety of fuels. The spectrum of fuel cycles ranges from natural uranium, through slightly enriched uranium, to plutonium and ultimately thorium fuels which offer many of the advantages of the fast breeder reactor system. CANDU can also burn the recycled uranium and/or the plutonium from fuel discharged from light water reactors. This synergistic relationship could obviate the need to re-enrich the reprocessed uranium and allow a simpler reprocessing scheme. Fule management strategies that will permit future fuel cycles to be used in existing CANDU reactors have been identified. Evolutionary design changes will lead to an even greater flexibility, which will guarantee the continued success of the CANDU system. (author)

Nuclear Burning Wave Modular Fast Reactor Concept

International Nuclear Information System (INIS)

Kodochigov, N.G.; Sukharev, Yu.P.

2014-01-01

The necessity to provide nuclear power industry, comparable in a scope with power industry based on a traditional fuel, inspired studies of an open-cycle fast reactor aimed at: - solution of the problem of fuel provision by implementing the highest breeding characteristics of new fissile materials of raw isotopes in a fast reactor and applying accumulated fissile isotopes in the same reactor, independently on a spent fuel reprocessing rate in the external fuel cycle; - application of natural or depleted uranium for makeup fuel, which, with no spent fuel reprocessing, forms the most favorable non-proliferation conditions; - application of inherent properties of the core and reactor for safety provision. The present report, based on previously published papers, gives the theoretical backgrounds of the concept of the reactor with a nuclear burning wave, in which an enriched-fuel core (driver) is replaced by a blanket, and basic conditions for nuclear burning wave initiating and keeping are shown. (author)

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

Development of fuel cycle technology for molten-salt reactor systems

International Nuclear Information System (INIS)

Uhlir, J.

2006-01-01

Full text: Full text: The Molten-Salt Reactor (MSR) represents one of promising advanced reactor type assigned to the GEN IV reactor systems. It can be operated either as thorium breeder within the Th -133U fuel cycle or as actinide transmuter incinerating transuranium fuel. Essentially the main advantage of MSR comes out from the prerequisite, that this reactor type should be directly connected with the 'on-line' reprocessing of circulating liquid (molten-salt) fuel. This principle should allow very effective extraction of freshly constituted fissile material (233U). Besides, the on-line fuel salt clean up is necessary within a long run to keep the reactor in operation. As a matter of principle, it permits to clear away typical reactor poisons like xenon, krypton, lanthanides etc. and possibly also other products of burned plutonium and transmuted minor actinides. The fuel salt clean up technology should be linked with the fresh MSR fuel processing to continuously refill the new fuel (thorium or transuranics) into the reactor system. On the other hand, the technologies of fresh transuranium molten-salt fuel processing from the current LWR spent fuel and of the on-line reprocessing of MSR fuel represent two killing points of the whole MSR technology, which have to be successfully solved before MSR deployment in the future. There are three main pyrochemical partitioning techniques proposed for processing and/or reprocessing of MSR fuel: Fluoride volatilization processes, Molten salt / liquid metal extraction processes and Electrochemical separation processes. Two of them - Fluoride Volatility Method and Electrochemical separation process from fluoride media are under development in the Nuclear Research Institute Rez pic. R and D in the field of Fluoride Volatility Method is concentrated to the development and verification of experimental semi-pilot technology for LWR spent fuel reprocessing, which may result in a product the form and composition of which might be

Research reactors fuel cycle problems and dilemma

International Nuclear Information System (INIS)

Romano, R.

2004-01-01

During last 10 years, some problems appeared in different steps of research reactors fuel cycle. Actually the majority of these reactors have been built in the 60s and these reactors were operated during all this long period in a cycle with steps which were dedicated to this activity. Progressively and for reasons often economical, certain steps of the cycle became more and more difficult to manage due to closing of some specialised workshops in the activities of scraps recycling, irradiated fuel reprocessing, even fuel fabrication. Other steps of the cycle meet or will meet difficulties, in particular supplying of fissile raw material LEU or HEU because this material was mostly produced in enrichment units existing mainly for military reason. Rarefaction of fissile material lead to use more and more enriched uraniums said 'of technical quality', that is to say which come from mixing of varied qualities of enriched material, containing products resulting from reprocessing. Actually, problems of end of fuel cycle are increased, either consisting of intermediary storage on the site of reactor or on specialised sites, or consisting of reprocessing. This brief summary shows most difficulties which are met today by a major part of industrials of the fuel cycle in the exercise of their activities

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.

Safety aspects in fuel reprocessing and radioactive waste management

International Nuclear Information System (INIS)

Agarwal, K.

2018-01-01

Nuclear energy is used for generation of electricity and for production of a wide range of radionuclides for use in research and development, healthcare and industry. Nuclear industry uses nuclear fission as source of energy so a large amount of energy is available from very small amount of fuel. As India has adopted c losed fuel cycle , spent nuclear fuel from nuclear reactor is considered as a material of resource and reprocessed to recovery valuable fuel elements. Main incentive of reprocessing is to use the uranium resources effectively by recovering/recycling Pu and U present in the spent fuel. This finally leads to a very small percentage of residual material present in spent nuclear fuel requiring their management as radioactive waste. Another special feature of the Indian Atomic Energy Program is the attention paid from the very beginning to the safe management of radioactive waste

The integration of fast reactor to the fuel cycle in Slovakia

International Nuclear Information System (INIS)

Zajac, R.; Darilek, P.; Necas, V.

2009-01-01

A very topical problem of nuclear power is the fuel cycle back-end. One of the options is a LWR spent fuel reprocessing and a fissile nuclides re-use in the fast reactor. A large amount of spent fuel has been stored in the power plant intermediate storage during the operation of WWER-440 reactors in Slovakia. This paper is based on an analysis of Pu and minor actinides content in actual WWER-440 spent fuel stored in Slovakia. The next part presents the possibilities of reprocessing and Pu re-use in fast reactor under Slovak conditions. The fuel cycle consisting of the WWER-440 reactor, PUREX reprocessing plant and a sodium fast reactor was designed. The last section compares two parts of this fuel cycle: one is UOX cycle in WWER-440 reactor and the other is cycle in the fast reactor - SUPER PHENIX loaded with MOX fuel (Pu + Minor Actinides). The starting point is a single recycling of Pu from WWER-440 in the fission products. The next step is multi recycling of Pu in the fission products to obtain equilibrium cycle. This article is dealing with the solution of power production and fuel cycle indicators. All kinds of calculations were performed by computer code HELIOS 1.10. (Authors)

The integration of fast reactor to the fuel cycle in Slovakia

International Nuclear Information System (INIS)

Zajac, R.; Darilek, P.; Necas, V.

2009-01-01

A very topical problem of nuclear power is the fuel cycle back-end. One of the options is a LWR spent fuel reprocessing and a fissile nuclides re-use in the fast reactor. A large amount of spent fuel has been stored in the power plant intermediate storage during the operation of VVER-440 reactors in Slovakia. This paper is based on an analysis of Pu and minor actinides content in actual VVER-440 spent fuel stored in Slovakia. The next part presents the possibilities of reprocessing and Pu re-use in fast reactor under Slovak conditions. The fuel cycle consisting of the VVER-440 reactor, PUREX reprocessing plant and a sodium fast reactor was designed. The last section compares two parts of this fuel cycle: one is UOX cycle in VVER-440 reactor and the other is cycle in the fast reactor - SUPER PHENIX loaded with MOX fuel (Pu + Minor Actinides). The starting point is a single recycling of Pu from VVER-440 in the FR. The next step is multirecycling of Pu in the FR to obtain equilibrium cycle. This article is dealing with the solution of power production and fuel cycle indicators. All kinds of calculations were performed by computer code HELIOS 1.10. (authors)

A review of the UK fast reactor programme

International Nuclear Information System (INIS)

Smith, R.D.

1982-01-01

A review of the United Kingdom Fast Reactor Programme is introduced. Operational experience with the Prototype Fast Reactor (PFR) is briefly summarized. The design concept of the Commercial Demonstration Fast Reactor (CDFR) is given in some detail. The emphasis is on materials development, chemical engineering/sodium technology, fuel reprocessing and fuel cycle, engineering component development and reactor safety

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

Estimation of gamma dose rate from hulls and shield design for the hull transport cask of Fuel Reprocessing Plant (FRP)

International Nuclear Information System (INIS)

Chandrasekaran, S.; Rajagopal, V.; Jose, M.T.; Venkatraman, B.

2012-01-01

In Fuel Reprocessing Plant (FRP), un-dissolved clad of fuel pins known as hulls are the major sources of high level solid waste. Safe handling, transport and disposal require the estimation of radioactivity as a consequent of gamma dose rate from hulls in fast reactor fuel reprocessing plant in comparison with thermal reactor fuel. Due to long irradiation time and low cooling of spent fuel, the evolution of activation products 51 Cr, 58 Co, 54 Mn and 59 Fe present as impurities in the fuel clad are the major sources of gamma radiation. Gamma dose rate from hull container with hulls from Fuel Sub Assembly (FSA) and Radial Sub Assembly (RSA) of Fuel Reprocessing Plant (FRP) was estimated in order to design the hull transport cask. Shielding computations were done using point kernel code, IGSHIELD. This paper describes the details of source terms, estimation of dose rate and shielding design of hull transport cask in detail. (author)

Supply of enriched uranium for research reactors

Energy Technology Data Exchange (ETDEWEB)

Mueller, H. [NUKEM GmbH, Alzenau (Germany)

1997-08-01

Since the RERTR-meeting In Newport/USA in 1990 the author delivered a series of papers in connection with the fuel cycle for research reactors dealing with its front-end. In these papers the author underlined the need for unified specifications for enriched uranium metal suitable for the production of fuel elements and made proposals with regard to the re-use of in Europe reprocessed highly enriched uranium. With regard to the fuel cycle of research reactors the research reactor community was since 1989 more concentrating on the problems of its back-end since the USA stopped the acceptance of spent research reactor fuel on December 31, 1988. Now, since it is apparent that these back-end problem have been solved by AEA`s ability to reprocess and the preparedness of the USA to again accept physically spent research reactor fuel the author is focusing with this paper again on the front-end of the fuel cycle on the question whether there is at all a safe supply of low and high enriched uranium for research reactors in the future.

Supply of enriched uranium for research reactors

International Nuclear Information System (INIS)

Mueller, H.

1997-01-01

Since the RERTR-meeting In Newport/USA in 1990 the author delivered a series of papers in connection with the fuel cycle for research reactors dealing with its front-end. In these papers the author underlined the need for unified specifications for enriched uranium metal suitable for the production of fuel elements and made proposals with regard to the re-use of in Europe reprocessed highly enriched uranium. With regard to the fuel cycle of research reactors the research reactor community was since 1989 more concentrating on the problems of its back-end since the USA stopped the acceptance of spent research reactor fuel on December 31, 1988. Now, since it is apparent that these back-end problem have been solved by AEA's ability to reprocess and the preparedness of the USA to again accept physically spent research reactor fuel the author is focusing with this paper again on the front-end of the fuel cycle on the question whether there is at all a safe supply of low and high enriched uranium for research reactors in the future

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

About redistribution of Ionic streams in imitation experiments on plasma separation

International Nuclear Information System (INIS)

Yuferov, V.B.; Svichkar', A.S.; Sharij, S.V.; Tkacheva, T.I.; Il'icheva, V.O.; Katrechko, V.V.; Shapoval, A.I.; Khizhnyak, S.N.

2013-01-01

The third stage of spent nuclear fuel (SNF) reprocessing by plasma methods with use of imitation environments is considered. The trajectories of motion of selective ions in the one-particle approaching are calculated. The scheme of SNF separator is presented

Preliminary Study for Inventories of Minor Actinides in Thorium Molten Salt Reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Choong Wie; Kim, Hee Reyoung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

2015-05-15

It has different characteristic with the conventional reactors which use a solid fuel. It can continually supply the fuel by online refueling and reprocessing of minor actinides so that those can be separated and eliminated from the reactor. The MSR maintains steady state except initial stage and the reactor becomes stable. In this research, considering online refueling, bubbling and reprocessing, the basic concept for evaluation of the inventory of minor actinide in the molten salt reactor is driven using the Bateman equation. The simulation results, where REM and MCNP code from CNRS (Centre National de la Recherche Scientifique) applied to the concept equation are analyzed. The analysis of the basic concept was carried out for evaluation of the inventory of the minor actinides in MSR. It was thought that the inventories of the minor actinides should be evaluated by solving the modified Bateman equation due to the MSR characteristic of online refueling, chemical reprocessing and bubbling.

Preliminary Study for Inventories of Minor Actinides in Thorium Molten Salt Reactor

International Nuclear Information System (INIS)

Lee, Choong Wie; Kim, Hee Reyoung

2015-01-01

It has different characteristic with the conventional reactors which use a solid fuel. It can continually supply the fuel by online refueling and reprocessing of minor actinides so that those can be separated and eliminated from the reactor. The MSR maintains steady state except initial stage and the reactor becomes stable. In this research, considering online refueling, bubbling and reprocessing, the basic concept for evaluation of the inventory of minor actinide in the molten salt reactor is driven using the Bateman equation. The simulation results, where REM and MCNP code from CNRS (Centre National de la Recherche Scientifique) applied to the concept equation are analyzed. The analysis of the basic concept was carried out for evaluation of the inventory of the minor actinides in MSR. It was thought that the inventories of the minor actinides should be evaluated by solving the modified Bateman equation due to the MSR characteristic of online refueling, chemical reprocessing and bubbling

Noble gas atmospheric monitoring for international safeguards at reprocessing plants

International Nuclear Information System (INIS)

Nakhleh, C.W.; Poths, J.; Stanbro, W.D.; Perry, R.T. Jr.; Wilson, W.B.; Fearey, B.L.

1997-01-01

The use of environmental sampling is a major component of the improvements of International Atomic Energy Agency safeguards being carried out under Program 93+2. Nonradioactive noble gas isotopic measurements in the effluent stream of large reprocessing facilities may provide useful confirmatory information on the burnup and reactor type of the spent fuel undergoing reprocessing. The authors have taken and analyzed stack samples at an operating facility. The data show clear fission signals. The authors are currently applying a maximum-likelihood estimation procedure to determine the fuel burnup from these data. They anticipate that the general features involved in the table noble gas problem--selection of appropriate signals, measurement of those signals under realistic conditions, and inverse calculation of parameters of interest from the environmental data--will be present in all environmental sampling problems. These methods should therefore be widely applicable

Advanced Research Reactor Fuel Development

Energy Technology Data Exchange (ETDEWEB)

Kim, C. K.; Park, H. D.; Kim, K. H. (and others)

2006-04-15

RERTR program for non-proliferation has propelled to develop high-density U-Mo dispersion fuels, reprocessable and available as nuclear fuel for high performance research reactors in the world. As the centrifugal atomization technology, invented in KAERI, is optimum to fabricate high-density U-Mo fuel powders, it has a great possibility to be applied in commercialization if the atomized fuel shows an acceptable in-reactor performance in irradiation test for qualification. In addition, if rod-type U-Mo dispersion fuel is developed for qualification, it is a great possibility to export the HANARO technology and the U-Mo dispersion fuel to the research reactors supplied in foreign countries in future. In this project, reprocessable rod-type U-Mo test fuel was fabricated, and irradiated in HANARO. New U-Mo fuel to suppress the interaction between U-Mo and Al matrix was designed and evaluated for in-reactor irradiation test. The fabrication process of new U-Mo fuel developed, and the irradiation test fuel was fabricated. In-reactor irradiation data for practical use of U-Mo fuel was collected and evaluated. Application plan of atomized U-Mo powder to the commercialization of U-Mo fuel was investigated.

Summary of estimated doses and risks resulting from routine radionuclide releases from fast breeder reactor fuel cycle facilities

International Nuclear Information System (INIS)

Miller, C.W.; Meyer, H.R.

1985-01-01

A project is underway at Oak Ridge National Laboratory to assess the human health and environment effects associated with operation of Liquid Metal Fast Breeder Reactor fuel cycle. In this first phase of the work, emphasis was focused on routine radionuclide releases from reactor and reprocessing facilities. For this study, sites for fifty 1-GW(e) capacity reactors and three reprocessing plants were selected to develop scenarios representative of US power requirements. For both the reactor and reprocessing facility siting schemes selected, relatively small impacts were calculated for locality-specific populations residing within 100 km. Also, the results of these analyses are being used in the identification of research priorities. 13 refs., 2 figs., 3 tabs

Nickel based alloys for molten salt applications in pyrochemical reprocessing applications

International Nuclear Information System (INIS)

Ningshen, S.; Ravi Shankar, A.; Rao, Ch. Jagadeeswara; Mallika, C.; Kamachi Mudali, U.

2016-01-01

Pyrochemical reprocessing route is one of the best option for reprocessing of spent metallic nuclear fuel from future fast breeder in many countries, especially in the US (Integral fast reactor, IFR), Russia (Research Institute of Atomic Reactors, RIAR), Japan, Korea and India. This technology with intrinsic nuclear proliferation resistance is regarded as one of the most promising nuclear fuel cycle technologies of the next-generation. However, the selection of materials of construction for pyrochemical reprocessing plants is challenging because of the extreme environments, i.e., high radiation, corrosive molten salt (LiCl-KCl, LiCl-KCl-CsCl, KCl-NaCl-MgCl 2 , etc.), reactive molten metals, and high temperature. Efforts have been made to develop compatible materials for various unit operations like salt preparation, electrorefining, cathode processing and alloy casting in pyrochemical reprocessing. Nickel and its alloy are the candidate materials for salt purification exposed to molten LiCl-KCl under Cl 2 bubbling, in air or ultra high purity argon environment. In the present study, the corrosion behavior of candidate materials like Inconel 600, Inconel 625, Inconel 690 exposed to molten LiCl-KCl eutectic salt environment at 500 to 600 °C have been carried out. The surface morphology of the exposed samples and scales were examined by SEM/EDX and XRD. The weight loss results indicated that Inconel 600 and Inconel 690 offer better corrosion resistance compared to Inconel 625 in air and chlorine environment. Higher corrosion of Inconel 625 is attributed to development of Mo rich salt layers. However, Ni base alloys exhibited a decreasing trend of weight loss with increasing time of exposure and weight gain was observed under UHP Ar environment. The mechanism of corrosion of Ni base alloys appeared to be due to formation of Cr rich and Ni rich layers of Cr 2 O 3 , NiO and spinel oxides at the surface and subsequent spallation. Based on the present studies, Inconel 690

Reprocessing of irradiated fuel: pros and cons

International Nuclear Information System (INIS)

Lebedev, O.G.; Novikov, V.M.

1991-01-01

The acceptable-safety nuclear reactors (APWR, LMFBR, MSBR, MSCR) can be provided by the enrichment industry and by plutonium reserves. But steady accumulation of spent fuel will inevitably make to return to the problems of fuel recycle. PUREX-processing increases a danger of radionuclides spreading due to the presence of large buffer tanks. Using of compact fluoride - volatility process will sharply reduce a nuclide leakage likewise permit to reprocess a fuel with a burnup as high as possible. Success of a powerful robots development give an opportunity to design a fluoride-volatility plant twice cheaper than PUREX. (author)

Basis for criteria for exemption of decommissioning waste: reprocessing of dust from recycling of steel

International Nuclear Information System (INIS)

Elert, M.; Wiborgh, M.

1992-06-01

This study is a part of a larger study with the purpose to provide the authority concerned, the Swedish Radiation Protection Institute (SSI), with technical background material needed for future decisions concerning exemption levels for recycling and disposal of material originating from decommissioned nuclear power reactors. The dismantling of nuclear power reactors will give rise to large amounts of steel scrap with a very low activity concentration. It is of interest to exempt this material from regulatory control in order to make recycling possible. During the melting of steel scrap dust will be formed which is collected in the off-gas cleaning system of the furnace. Radionuclides may be enriched in this dust, and thereby obtain a higher activity concentration than the melted scrap. Presently, there is a strong interest to reprocess these dusts with the objectives to recover valuable metals and to reduce the amounts of waste harmful to the environment. During the reprocessing of dusts collected from the melting process a further redistribution and reconcentration of radionuclides may occur. In this report the treatment of dust from steel melting is described, the potential reconcentration of radionuclides is analyzed, and the potential radiological consequences are estimated. The study has focussed on the reprocessing of dust with the plasma method used at ScanDust in Sweden, and with the Waelz process used in, for example Spain and Germany. Various factors as economical, political and future developments of dust treatment and steel processes will determine the amounts of dusts produced and also to what degree dust will be reprocessed in the future. (au)

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)

Plant for retention of 14C in reprocessing plants for LWR fuel elements

International Nuclear Information System (INIS)

Braun, H.; Gutowski, H.; Bonka, H.; Gruendler, D.

1983-01-01

The 14 C produced from nuclear power plants is actually totally emitted from nuclear power plants and reprocessing plants. Using the radiation protection principles proposed in ICRP 26, 14 C should be retained at heavy water moderated reactors and reprocessing plants due to a cost-benefit analysis. In the frame of a research work to cost-benefit analysis, which was sponsored by the Federal Minister of the Interior, an industrial plant for 14 C retention at reprocessing plants for LWR fuel elements has been planned according to the double alkali process. The double alkali process has been chosen because of the sufficient operation experience in the conventional chemical technique. In order to verify some operational parameters and to gain experiences, a cold test plant was constructed. The experiment results showed that the double alkali process is a technically suitable method with high operation security. Solidifying CaCO 3 with cement gives a product fit for final disposal

Nuclear reactor fuel sub-assemblies

International Nuclear Information System (INIS)

Ford, J.; Bishop, J.F.W.

1981-01-01

An improved fuel sub-assembly for liquid metal cooled fast breeder nuclear reactors is described which facilitates dismantling operations for reprocessing purposes. The method of dismantling is described. (U.K.)

Electrochemical reprocessing of nuclear fuels

International Nuclear Information System (INIS)

Brambilla, G.; Sartorelli, A.

1980-01-01

A method is described for the reprocessing of irradiated nuclear fuel which is particularly suitable for use with fuel from fast reactors and has the advantage of being a dry process in which there is no danger of radiation damage to a solvent medium as in a wet process. It comprises the steps of dissolving the fuel in a salt melt under such conditions that uranium and plutonium therein are converted to sulphate form. The plutonium sulphate may then be thermally decomposed to PuO 2 and removed. The salt melt is then subjected to electrolysis conditions to achieve cathodic deposition of UO 2 (and possibly PuO 2 ). The salt melt can then be recycled or conditioned for final disposal. (author)

The law for the regulations of nuclear source materials, nuclear fuel materials and reactors

International Nuclear Information System (INIS)

1977-01-01

Concerning refining, fabrication and reprocessing operations of such materials as well as the installation and operation of reactors, necessary regulations are carried out. Namely, in case of establishing the business of refining, fabricating and reprocessing nuclear materials as well as installing nuclear reactors, applications for the permission of the Prime Minister and the Minister of International Trade and Industry should be filed. Change of such operations should be permitted after filing applications. These permissions are retractable. As regards the reactors installed aboard foreign ships, it must be reported to enter Japanese waters and the permission by the Prime Minister must be obtained. In case of nuclear fuel fabricators, a chief technician of nuclear fuel materials (qualified) must be appointed per each fabricator. In case of installing nuclear reactors, the design and methods of construction should be permitted by the Prime Minister. The standard for such permission is specified, and a chief engineer for operating reactors (qualified) must be appointed. Successors inherit the positions of ones who have operated nuclear material refining, fabrication and reprocessing businesses or operated nuclear reactors. (Rikitake, Y.)

Molten salt reactors: reactor cores

International Nuclear Information System (INIS)

1983-01-01

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

The estimation of heavy metal concentration in FBR reprocessing solvent streams by density measurement

International Nuclear Information System (INIS)

Brown, M.L.; Savage, D.J.

1986-04-01

The application of density measurement to heavy metal monitoring in the solvent phase is described, including practical experience gained during three fast reactor fuel reprocessing campaigns. An experimental algorithm relating heavy metal concentration and sample density was generated from laboratory-measured density data, for uranyl nitrate dissolved in nitric acid loaded tri-butyl phosphate in odourless kerosene. Differences in odourless kerosene batch densities are mathematically interpolated, and the algorithm can be used to estimate heavy metal concentrations from the density to within +1.5 g/l. An Anton Paar calculating digital densimeter with remote cell operation was used for all density measurements, but the algorithm will give similar accuracy with any density measuring device capable of a precision of better than 0.0005 g/cm 3 . For plant control purposes, the algorithm was simplified using a density referencing system, whereby the density of solvent not yet loaded with heavy metal is subtracted from the sample density. This simplified algorithm compares very favourably with empirical algorithms, derived from numerical analysis of density data and chemically measured uranium and plutonium data obtained during fuel reprocessing campaigns, particularly when differences in the acidity of the solvent are considered before and after loading with heavy metal. This simplified algorithm had been successfully used for plant control of heavy metal loaded solvent during four fast reactor fuel reprocessing campaigns. (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

Breeder reactors

International Nuclear Information System (INIS)

Gollion, H.

1977-01-01

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

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

Radioactive wastes management in fiscal year 1983 in the fuel reprocessing plant

International Nuclear Information System (INIS)

1985-01-01

In the nuclear fuel reprocessing plant of Power Reactor and Nuclear Fuel Development Corporation, the releases of radioactive gaseous and liquid wastes are so managed not to exceed the respective objective release levels. Of the radioactive liquid wastes, the high level concentrated wastes are stored in tanks and the low level wastes are stored in tanks or asphalt solidified. For radioactive solid wastes, high level solid wastes are stored in casks, low level solid wastes and asphalt solids in drums etc. The releases of radioactive gaseous and liquid wastes in the fiscal year 1983 were below the objective release levels. The radioactive wastes management in the fuel reprocessing plant in fiscal year 1983 is given in tables, the released quantities, the stored quantities, etc. (Mori, K.)

Flexible fuel cycle initiative for the transition period from current reactors to next generation reactors

International Nuclear Information System (INIS)

Yamashita, Junichi; Fukasawa, Tetsuo; Hoshino, Kuniyoshi; Kawamura, Fumio; Shiina, Kouji; Sasahira, Akira

2005-01-01

A sustainable electricity supply by fast breeder reactors (FBRs) is essential to ensure energy security and prevent global warming. Transition from light water reactors (LWRs) to FBRs and establishment of an FBR cycle are indispensable, which requires plutonium (Pu) for the introduction of FBRs. The authors propose advanced system called 'Flexible Fuel Cycle Initiative (FFCI)' which can respond flexibly the future expected technical and social uncertainties, can hold no surplus Pu, and can achieve an economical FBR cycle. In the new concept of FFCI, 2nd LWR reprocessing which would succeed Rokkasho Reprocessing Plant is a simple facility to carry out only uranium (U) removal and residual 'recycle material' is stored or utilized. According to FBRs introduction status, recycle material is immediately treated in an FBR reprocessing to fabricate FBR fuel or temporarily stored for the utilization in FBRs at necessary timing. FFCI has high flexibility by having several options for future uncertainties by the introduction of recycle material as a buffer material between LWR and FBR cycles. (author)

Analysis of triso packing fraction and fissile material to DB-MHR using LWR reprocessed fuel

International Nuclear Information System (INIS)

Silva, Clarysson A.M. da; Pereira, Claubia; Costa, Antonella L.; Veloso, Maria Auxiliadora F.; Gual, Maritza R.

2013-01-01

Gas-cooled and graphite-moderated reactor is being considered the next generation of nuclear power plants because of its characteristic to operate with reprocessed fuel. The typical fuel element consists of a hexagonal block with coolant and fuel channels. The fuel pin is manufactured into compacted ceramic-coated particles (TRISO) which are used to achieve both a high burnup and a high degree of passive safety. This work uses the MCNPX 2.6.0 to simulate the active core of Deep Burn Modular Helium Reactor (DB-MHR) employing PWR (Pressurized Water Reactor) reprocessed fuel. However, before a complete study of DB-MHR fuel cycle and recharge, it is necessary to evaluate the neutronic parameters to some values of TRISO Packing Fractions (PF) and Fissile Material (FM). Each PF and FM combination would generate the best behaviour of neutronic parameters. Therefore, this study configures several PF and FM combinations considering the heterogeneity of TRISO layers and lattice. The results present the best combination of PF and FM values according with the more appropriated behaviour of the neutronic parameters during the burnup. In this way, the optimized combination can be used to future works of MHR fuel cycle and recharge. (author)

Outline of human machine interface at Rokkasho reprocessing plant

International Nuclear Information System (INIS)

Niioka, T.; Miyazaki, T.; Fujita, D.; Esashika, A.; Yoshida, Y.; Nakamura, W.; Tochigi, T.; Yoshimoto, A.; Yokoi, M.

2006-01-01

The Japan Nuclear Fuel Limited (JNFL) has been performing the active tests since the end of March, 2006, for its Rokkasho Reprocessing Plant using the spent fuels retrieved from the Light Water Reactors. At the early stage of the tests relatively low burn-up fuels have been used, and the burn-up will be increased at later stages until the start of commercial operation planned next year. The plant is operated from the main control room in the Control Building, where two types of operator consoles are located for plant monitoring and operation. The Operator Interface Station (OIS) driven by computer systems is chiefly used for instrumentation and control for production activities during normal operation. In addition to this, safety panels composed of hardware circuits are installed for nuclear safety functions such as criticality safety management, explosion protection, and confinement of radioactive materials. This paper outlines the Human Machine Interface features applied to the Rokkasho Reprocessing Plant. (authors)

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

Current status and directions for fast reactor reprocessing

International Nuclear Information System (INIS)

Burch, W.D.

1983-01-01

The development of fast breeder reactors (FBRs) for commercial electric power production has been under way in several countries for more than 20 years. In the United States as elsewhere, early work was centered on small reactors to prove the feasibility of concepts and later was followed by larger reactors to test engineering features and to develop fuel technology. In the early 1970s, with the perceived crisis in electrical generation expected late in this century, major efforts were mounted to plan and carry out comprehensive development programs to ensure the capability to develop and begin using this new form of nuclear power by the end of this century. This comprehensive effort included the first serious efforts directed toward the supporting fuel cycle activities. However, because of the effects of the oil price rise and resulting conservation, a slowdown of industrial growth, and cut-backs in energy needs, there has been a decline in program activities. Unlike the fuel cycle for light-water reactors (LWRs), where supply and the back-end recycle and/or waste disposal activities can largely be uncoupled, recovery and recycle of fissile materials in spent fuel must be accomplished in one or two years in a practical breeder system. 3 references

Gas reactor international coope--ative program. Interim report: assessment of gas-cooled reactor economics

Energy Technology Data Exchange (ETDEWEB)

1979-08-01

A computer analysis of domestic economic incentive is presented. Included are the sample computer data set for ten combinations of reprocessing and reactor assumptions; basic data set and computer output; higher uranium availability computer output; 50 percent higher GCR fabrication cost computer output; 50 percent higher GCR reprocessing cost computer output; year 1990 and year 2000 GCR introduction scenario computer outputs; 75 percent perceived capacity factor for PBR computer output; and capital cost of GCRs 1.2 times that of LWRs.

Chemistry of materials relevant to aqueous reprocessing and waste management

International Nuclear Information System (INIS)

Srinivasan, T.G.

2012-01-01

Nuclear energy option will be an inevitable one with the fossil fuels depleting fast and present coal and oil based thermal power generation resulting in unwanted green house gas emission. The utilisation of the fissile resources will be more effective with closed fuel cycle option wherein the spent reactor fuel is reprocessed and the unused uranium and plutonium formed during the reactor operation is recovered and re-used. Of the aqueous and non-aqueous routes available to reprocess the spent nuclear fuels, aqueous reprocessing method of recovering the valuable uranium and plutonium by the PUREX process is in vogue for the past six decades. The process involves chopping the fuel into small lengths, leaching uranium and plutonium with concentrated nitric acid under reflux, conditioning the dissolver solution with respect to acidity and valency of U and Pu, solvent extraction with 30%TBP/n-DD to selectively extract U(VI) and Pu(IV) leaving most of the fission products into the raffinate, partitioning plutonium from uranium and reconversion of U and Pu into oxide forms after further purification. Many reagents are used to achieve near quantitative recovery of both uranium and plutonium (>99.9%) and with high decontamination factors (>10 7 ) from highly radioactive fission products. Nevertheless, the chemistry of several reagents used and the chemical processes that take place during the entire course of reprocessing and waste management operations are yet to be fully understood and gives a lot of scope for further improvements. Some examples where research requires concerted efforts are, 1) development of new extractants conforming to CHON principle, with acceptable physical properties, high stability, selectivity and resistance to third phase formation, 2) new partitioning reagents and processes which offer good efficiency and kinetics for uranium/plutonium reduction, 3) understanding the chemistry of troublesome fission products such as Tc, Ru and Zr, 4

Worldwide reprocessing supply and demand

International Nuclear Information System (INIS)

Pinto, S.

1987-01-01

The aim of this paper is to broadly examine the current situation in the LWR fuel reprocessing services market on a worldwide basis through 2010. The main factors influencing this market (nuclear programs, fuel discharges, reprocessing capacities, buyer philosophies, etc.) are identified in the paper and the most important are highlighted and discussed in more detail. Emphasis has been placed on the situation with respect to reprocessing in those countries having a significant influence on the reprocessing market

Simplified probabilistic risk assessment in fuel reprocessing

International Nuclear Information System (INIS)

Solbrig, C.W.

1993-01-01

An evaluation was made to determine if a backup mass tracking computer would significantly reduce the probability of criticality in the fuel reprocessing of the Integral Fast Reactor. Often tradeoff studies, such as this, must be made that would greatly benefit from a Probably Risk Assessment (PRA). The major benefits of a complete PRA can often be accrued with a Simplified Probabilistic Risk Assessment (SPRA). An SPRA was performed by selecting a representative fuel reprocessing operation (moving a piece of fuel) for analysis. It showed that the benefit of adding parallel computers was small compared to the benefit which could be obtained by adding parallelism to two computer input steps and two of the weighing operations. The probability of an incorrect material moves with the basic process is estimated to be 4 out of 100 moves. The actual values of the probability numbers are considered accurate to within an order of magnitude. The most useful result of developing the fault trees accrue from the ability to determine where significant improvements in the process can be made. By including the above mentioned parallelism, the error move rate can be reduced to 1 out of 1000

Farewell to a reactor

International Nuclear Information System (INIS)

Skanborg, P.

1976-01-01

Denmark's second reactor, DR 2, whose first criticality took place the night of 18/19 December 1958 was shut down for the last time on 31 October 1975. It was a light-water moderrated and cooled reactor of swimming-pool type with a thermal power of 5 MW, using 90% enriched uranium. The operation is described. The reactor and auxiliary equipment are now being put 'in store' - all fuel elements sent for reprocessing, the reactor tank and cooling circuits emptied, and a lead shielding placed over the tank opening. The rest of the equipment will remain in place. (B.P.)

Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)

Science.gov (United States)

Clement, J. D.; Rust, J. H.

1977-01-01

Design power plant studies were carried out for two applications of the plasma core reactor: (1) As a breeder reactor, (2) As a reactor able to transmute actinides effectively. In addition to the above applications the reactor produced electrical power with a high efficiency. A reactor subsystem was designed for each of the two applications. For the breeder reactor, neutronics calculations were carried out for a U-233 plasma core with a molten salt breeding blanket. A reactor was designed with a low critical mass (less than a few hundred kilograms U-233) and a breeding ratio of 1.01. The plasma core actinide transmutation reactor was designed to transmute the nuclear waste from conventional LWR's. The spent fuel is reprocessed during which 100% of Np, Am, Cm, and higher actinides are separated from the other components. These actinides are then manufactured as oxides into zirconium clad fuel rods and charged as fuel assemblies in the reflector region of the plasma core actinide transmutation reactor. In the equilibrium cycle, about 7% of the actinides are directly fissioned away, while about 31% are removed by reprocessing.

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

Removal of actinides from high-level wastes generated in the reprocessing of commercial fuels

International Nuclear Information System (INIS)

Bond, W.D.; Leuze, R.E.

1975-09-01

Progress is reported on a technical feasibility study of removing the very long-lived actinides (uranium, neptunium, plutonium, americium, and curium) from high-level wastes generated in the commercial reprocessing of spent nuclear fuels. The study was directed primarily at wastes from the reprocessing of light water reactor (LWR) fuels and specifically to developing satisfactory methods for reducing the actinide content of these wastes to values that would make 1000-year-decayed waste comparable in radiological toxicity to natural uranium ore deposits. Although studies are not complete, results thus far indicate the most promising concept for actinide removal includes both improved recovery of actinides in conventional fuel reprocessing and secondary processing of the high-level wastes. Secondary processing will be necessary for the removal of americium and curium and perhaps some residual plutonium. Laboratory-scale studies of separations methods that appear most promising are reported and conceptual flowsheets are discussed. (U.S.)

Reprocessing of nonoptimally exposed holograms

International Nuclear Information System (INIS)

Phipps, G.S.; Robertson, C.E.; Tamashiro, F.M.

1980-01-01

Two reprocessing techniques have been investigated that are capable of correcting the effects of nonoptimum optical density of photographic amplitude holograms recorded on Agfa-Gevaert type 10E75 plates. In some cases a reprocessed hologram will exhibit a diffraction efficiency even higher than that obtainable from a hologram exposed and processed to the optimum density. The SNR of the reprocessed holograms is much higher than that of the same holograms belached with cupric bromide. In some cases the SNR approaches the optimum value for a properly exposed amplitude hologram. Subjective image quality and resolution of reprocessed hologram reconstructins appear to be no different than for normal single-development holograms. Repeated reprocessing is feasible and in some cases desirable as a means of increasing diffraction efficiency

Natural uranium utilization without enrichment and reprocessing

Energy Technology Data Exchange (ETDEWEB)

Sekimoto, H.; Toshinsky, V.; Ryu, K. [Tokyo Inst. of Tech. (Japan). Research Lab. for Nuclear Reactors

2001-07-01

Two types of fast reactor are investigated to utilize the natural uranium without enrichment and reprocessing in an equilibrium state. The first trial is SFPR. Its fuel-shuffling pattern is optimized. An obtained result gives its peak fuel burnup of 22,5%, power peaking factor of 1.5 and peak excess reactivity of 2,15%. The second trial is CANDLE burnup scheme, where distribution shapes of neutron flux and nuclide densities are constant but move in axial direction with a constant velocity. A feasible solution gives the speed of burning region of 4,1 cm/year, k{sub eff} of 1,02 and average spent fuel burnup of 41%. (author)

Pebble bed reactor fiscal year 1980: review summary report

International Nuclear Information System (INIS)

1980-07-01

Information on high-temperature reactor development is presented concerning reactor operating experience; core performance assessment; core control and shutdown; reflector and core support; maintenance and availability; safety aspects of PBR and prismatic comparison; PCRV dimensions; and fuel reprocessing cost estimate

Build-up and decay of fuel actinides in the fuel cycle of nuclear reactors

International Nuclear Information System (INIS)

Tasaka, Kanji; Kikuchi, Yasuyuki; Shindo, Ryuichi; Yoshida, Hiroyuki; Yasukawa, Shigeru

1976-05-01

For boiling water reactors, pressurized light-water reactors, pressure-tube-type heavy water reactors, high-temperature gas-cooled reactors, and sodium-cooled fast breeder reactors, uranium fueled and mixed-oxide fueled, each of 1000 MWe, the following have been studied: (1) quantities of plutonium and other fuel actinides built up in the reactor, (2) cooling behaviors of activities of plutonium and other fuel actinides in the spent fuels, and (3) activities of plutonium and other fuel actinides in the high-level reprocessing wastes as a function of storage time. The neutron cross section and decay data of respective actinide nuclides are presented, with their evaluations. For effective utilization of the uranium resources and easy reprocessing and high-level waste management, a thermal reactor must be fueled with uranium; the plutonium produced in a thermal reactor should be used in a fast reactor; and the plutonium produced in the blanket of a fast reactor is more appropriate for a fast reactor than that from a thermal reactor. (auth.)

Cleaning and extraction apparatus in a nuclear fuel reprocessing

International Nuclear Information System (INIS)

Nakamura, Yoshiaki.

1983-01-01

Purpose : To eliminate the requirement for the decomposition and cleaning of a centrifugal extractor upon re-processing of FBR type reactor fuels, by preventing solid fission products from depositing on a rotary body of the centrifugal extractor. Constitution : A cleaning and extraction apparatus comprising a combination of a centrifugal cleaner and a centrifugal extractor is used for shortening the contact time between the process liquid and the extraction solvent in FBR type reactor fuel re-processing, and variable parameters are adjusted so that the following equation can be satisfied for avoiding the deposition of solids onto the rotary body of the centrifugal extractor: lsub(e). (rsub(le) 2 + rsub(2r) 2 ) . Nsub(e) . Qsub(c)/ lsub(c) (rsub(lc) 2 + rsub(2c) 2 ) . Nsub(c) . Qsub(e) < 0.8 where Qsub(c) : flow rate to be processed in a centrifugal cleaner, lsub(c) : length of the rotary body, rsub(2c) : radius of a rotary body, rsub(le) : distance from the center to the liquid-extracting hole of the rotary body center to the liquid-extraction hole, Nsub(c) : number of revolution of the rotary body, Qsub(e) : amount of flowrate to be treated in the centrifugal extractor, lsub(e) : length of the rotary body, rsub(2e) : radius for the rotary body, rsub(le) : distance from the center of the rotary body to the liquid discharging aperture and Nsub(e) : number of rotation of the rotary body. (Ikeda, J.)

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.

Research on solvent extraction process for reprocessing of Th-U fuel from HTGR

International Nuclear Information System (INIS)

Bao Borong; Wang Gaodong; Qian Jun

1992-05-01

The unique properties of spent fuel from HTGR (high temperature gas cooled reactor) have been analysed. The single solvent extraction process using 30% TBP for separation and purification of Th-U fuel has been studied. In addition, the solvent extraction process for second uranium purification is also investigated to meet different needs of reprocessing and reproduction of Th-U spent fuel from HTGR

Report of third regular inspection of Tokai reprocessing facilities, Power Reactor and Nuclear Fuel Development Corp

International Nuclear Information System (INIS)

Anon.

1987-01-01

The reprocessing facilities passed the inspection before use on December 25, 1980, and started the full operation. Since then, this is the third regular inspection. It was begun on April 1, 1986, and finished on August 18, 1986, with the inspection of the rate of recovery of products. The reprocessing facilities which became the object of inspection were the facilities for accepting and storing spent fuel, the reprocessing facilities proper (the facilities of shearing, dissolution, separation, refining, denitration and recovery of acid and solvent), the facilities for storing products, measurement and control system, radioactive waste facilities, radiation control facilities and attached facilities (power, water, steam and testing). The main works carried out during the period of this regular inspection were the repair of an enriched uranium dissolution tank by welding, the renewal of a piping for a low activity waste liquid storing tank, and the removal of a washing tank. The total exposure dose in the first half of fiscal year 1986 was about 30.81 man-rem. (Kako, I.)

Reprocessing in Europe

Energy Technology Data Exchange (ETDEWEB)

Rossney, G [United Reprocessors G.m.b.H., Karlsruhe (F.R. Germany)

1976-04-01

The status of reprocessing activities within the member organizations of United Reprocessors is reviewed. The U.K. government has approved overseas deals by BNFL which will help to pay for their planned plant of 1000 te U p.a. at Windscale. In Germany KEWA has selected a site at Aschenburg as a fuel cycle centre where they plan to build a utility financed reprocessing plant of 1500 te U p.a. France has formed a new fuel cycle corporation, Cogema, which hopes to participate in the large volume of Japanese business negotiated by BNFL. United Reprocessors have agreed to pool their technology which may be available to organisations wishing to construct reprocessing plants in their own countries.

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.

Consolidated fuel reprocessing programme: Analysis of various options for the breeder fuel cycle in the USA

International Nuclear Information System (INIS)

Stradley, J.G.; Burch, W.D.; Yook, H.R.

1986-01-01

The United States Department of Energy (DOE) has established a programme to develop innovative liquid metal reactor (LMR) designs to assist in developing future U.S. reactor strategy. The paper describes studies in progress to examine various fuel cycle strategies that relate to the reactor strategy. Three potential fuel cycle options that focus on supporting an initial 1300 MW(e) reactor station have been defined: (1) Completion and utilization of the Breeder Reprocessing Engineering Test/Secure Automated Fabrication (BRET/SAF) in the Fuels and Materials Examination Facility (FMEF) at Hanford, Washington; (2) a co-located fuel cycle facility; and (3) delayed closure of the fuel cycle for five to ten years. The BRET, designed as a development facility, has sufficient capacity to service the needs of an initial module at an LMR station. It appears feasible to increase this capacity and to utilize SAF in the FMEF to accommodate the projected output (up to 35 MtHM/year) from the 1300 MW(e) liquid-metal concepts under study. Plans developed within the United States Consolidated Management Office for an initial reactor project have envisioned that cost savings could be realized by delaying the closure of the fuel cycle as long as supplies of plutonium could be obtained relatively inexpensively. This might prove to be only five to ten years, but even that period might be long enough for the fuel cycle costs to be spread over more than one reactor rather than loaded on the initial project. This concept is being explored as is the question of the future coupling of a light water reactor reprocessing industry for plutonium supply to breeder recycle

Economics and resources analysis of the potential use of reprocessing options by the current Spanish nuclear reactor park

Energy Technology Data Exchange (ETDEWEB)

Alvarez-Velarde, F.; Merino Rodriguez, I.; Gonzalez-Romero, E.

2014-07-01

Reprocessing of irradiated nuclear fuel serves multiple purposes, from Pu separation and recovery for MOX fuel fabrication to reduction of high level waste volume, and is nowadays being implemented in several countries like France, Japan, Russia or United Kingdom. This work is aimed at exploring the possibility (in resources and economic terms) of implementing reprocessing for MOX fabrication in Spain. (Author)

Report on the Savannah River Site aluminum-based spent nuclear fuel alternatives cost study

International Nuclear Information System (INIS)

1998-12-01

Initial estimates of costs for the interim management and disposal of aluminum-based spent nuclear fuel (SNF) were developed during preparation of the Environmental Impact Statement (EIS) on the Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel. The Task Team evaluated multiple alternatives, assessing programmatic, technical, and schedule risks, and generated life-cycle cost projections for each alternative. The eight technology alternatives evaluated were: direct co-disposal; melt and dilute; reprocessing; press and dilute; glass material oxidation dissolution system (GMODS); electrometallurgical treatment; dissolve and vitrify; and plasma arc. In followup to the Business Plan that was developed to look at SNF dry storage, WSRC prepared an addendum to the cost study. This addendum estimated the costs for the modification and use of an existing (105L) reactor facility versus a greenfield approach for new facilities (for the Direct Co-Disposal and Melt and Dilute alternatives). WSRC assessed the impacts of a delay in reprocessing due to the potential reservation of H-Canyon for other missions (i.e., down blending HEU for commercial use or the conversion of plutonium to either MOX fuel or an immobilized repository disposal form). This report presents the relevant results from these WSRC cost studies, consistent with the most recent project policy, technology implementation, canyon utilization, and inventory assumptions. As this is a summary report, detailed information on the technical alternatives or the cost assumptions raised in each of the above-mentioned cost studies is not provided. A comparison table that briefly describes the bases used for the WSRC analyses is included as Appendix A

Mechanism of 232U production in MTR fuel evolution of activity in reprocessed uranium

International Nuclear Information System (INIS)

Harbonnier, G.; Lelievre, B.; Fanjas, Y.; Naccache, S.J.P.

1993-01-01

The use of reprocessed uranium for research reactor fuel fabrication implies to keep operators safe from the hard gamma rays emitted by 232 U daughter products. CERCA has carried out, with the help of French CEA and COGEMA, a detailed study to determine the evolution of the radiation dose rate associated with the use of this material. (author)

Importance of nuclear fuel reprocessing

International Nuclear Information System (INIS)

Allday, C.

1977-01-01

The following topics are discussed: world energy requirements; energy conservation and the economics of recycle environmental considerations and the timescale of reprocessing; and problems associated with reprocessing. The conclusion is reached that reprocessing is essential to the conservation of the world's energy resources and is an environmentally, and probably an economically, more acceptable option to the ''throw away'' alternative

Current status on advanced aqueous reprocessing process (next) in FaCT project

International Nuclear Information System (INIS)

Washiya, Tadahiro; Myochin, Munetaka; Koyama, Tomozo

2009-01-01

Japan Atomic Energy Agency (JAEA) launched the Fast Reactor Cycle Technology Development (FaCT) project in cooperation with the Japanese electric utilities in 2006. An integration of the advanced aqueous reprocessing concept and the simplified pelletizing fuel fabrication was selected as the most promising fuel cycle system. In order to accomplish the integration, R and D tasks were launched as FaCT Project in 2006 by Japanese joint team. The New Extraction System for TRU Recovery (NEXT) system is an advanced aqueous reprocessing concept which was based on the well established aqueous reprocessing for LWR spent fuel and newly applied processes such as uranium crystallization and extraction chromatography for MAs recovery. Main task of the NEXT process is to develop the TRU recovery process and equipments with high reliability, criticality safety, high durability and remote maintainability. In the FaCT project, all innovative technologies are planned to be developed within the next decade focusing on the future commercialization of FBR cycle systems. The judgment of the adoption of each innovative technology will be made by 2010 based on the results of R and Ds. The development of each technology is to be completed by around 2015. By the same time, it is scheduled to present the conceptual design of commercial and demonstrative fast reactor cycle facilities. The six items (Disassembling and shearing, Fuel dissolution, Uranium Crystallization, Single cycle co-extraction of U, Pu and Np, MA recovery by extraction chromatography and Waste treatment) have been identified as the issues to be developed corresponding to each process step. Current R and D status and prospects of this system until around 2015 is reported. (author)

Canisters and nonfuel components at commercial nuclear reactors

International Nuclear Information System (INIS)