Quality control in nuclear fuel fabrication on the inspection basis

International Nuclear Information System (INIS)

Fuentes S, A.

1997-01-01

Every plant productive of electric power requires the use of energetics for the transformation to electricity. In the nucleo electric plant the energetic is the uranium, in which it makes ensembles and is used as fuel in the reactor. To assure that the fuel ensembles fulfill the specifications and requirements of design stipulated in the nucleo electric plant is that under a quality control through inspections during the fabrication process. The purpose of this work is to study and verify that the lineaments of the standard 10 CFR 50 appendix B 'Quality assurement for nuclear plants' specially in the criteria 'Inspections' that is used to guarantee the quality of the ensembles. This standard is the one that rules every activity and operation inside the pilot plant and its established in the quality program in the production of nuclear fuel for the Laguna Verde plant. The quality of the assemble is verified through each one of the tests or inspections due to the importance of it in the fabrication of fuel. (Author)

Material control in nuclear fuel fabrication facilities. Part II. Accountability, instrumentation and measurement techniques in fuel fabrication facilities

International Nuclear Information System (INIS)

Borgonovi, G.M.; McCartin, T.J.; McDaniel, T.; Miller, C.L.; Nguyen, T.

1978-01-01

This report describes the measurement techniques, the instrumentation, and the procedures used in accountability and control of nuclear materials, as they apply to fuel fabrication facilities. A general discussion is given of instrumentation and measurement techniques which are presently used being considered for fuel fabrication facilities. Those aspects which are most significant from the point of view of satisfying regulatory constraints have been emphasized. Sensors and measurement devices have been discussed, together with their interfacing into a computerized system designed to permit real-time data collection and analysis. Estimates of accuracy and precision of measurement techniques have been given, and, where applicable, estimates of associated costs have been presented. A general description of material control and accounting is also included. In this section, the general principles of nuclear material accounting have been reviewed first (closure of material balance). After a discussion of the most current techniques used to calculate the limit of error on inventory difference, a number of advanced statistical techniques are reviewed. The rest of the section deals with some regulatory aspects of data collection and analysis, for accountability purposes, and with the overall effectiveness of accountability in detecting diversion attempts in fuel fabrication facilities. A specific example of application of the accountability methods to a model fuel fabrication facility is given. The effect of random and systematic errors on the total material uncertainty has been discussed, together with the effect on uncertainty of the length of the accounting period

Material control in nuclear fuel fabrication facilities. Part II. Accountability, instrumentation and measurement techniques in fuel fabrication facilities

Energy Technology Data Exchange (ETDEWEB)

Borgonovi, G.M.; McCartin, T.J.; McDaniel, T.; Miller, C.L.; Nguyen, T.

1978-01-01

This report describes the measurement techniques, the instrumentation, and the procedures used in accountability and control of nuclear materials, as they apply to fuel fabrication facilities. A general discussion is given of instrumentation and measurement techniques which are presently used being considered for fuel fabrication facilities. Those aspects which are most significant from the point of view of satisfying regulatory constraints have been emphasized. Sensors and measurement devices have been discussed, together with their interfacing into a computerized system designed to permit real-time data collection and analysis. Estimates of accuracy and precision of measurement techniques have been given, and, where applicable, estimates of associated costs have been presented. A general description of material control and accounting is also included. In this section, the general principles of nuclear material accounting have been reviewed first (closure of material balance). After a discussion of the most current techniques used to calculate the limit of error on inventory difference, a number of advanced statistical techniques are reviewed. The rest of the section deals with some regulatory aspects of data collection and analysis, for accountability purposes, and with the overall effectiveness of accountability in detecting diversion attempts in fuel fabrication facilities. A specific example of application of the accountability methods to a model fuel fabrication facility is given. The effect of random and systematic errors on the total material uncertainty has been discussed, together with the effect on uncertainty of the length of the accounting period.

The fabrication of nuclear fuel elements in Mexico

International Nuclear Information System (INIS)

Guerrero Morillo, H.L.

1977-01-01

The situation of nuclear electricity generation in Mexico in 1976 is described: two nuclear reactors were under construction but no definite programme on the type and start-up dates for the next power plants existed. However, the existence of a general plan on nuclear power plants is mentioned, which, according to the latest estimates, will provide 10,000MW installed by 1990. The national intention, as laid down in an appropriate Law, is to supply domestic nuclear fuel to the power reactors operating in the country, starting with the first reloading of the two BWRs at the first national station in Laguna Verde, required at the end of 1981 and 1982, respectively. Before this can be achieved and to provide the relatively small amounts of fuel elements for the two reactors, Mexico must adopt a strategy of fuel elements fabrication. The two main options are analysed: (1) to delay local fabrication until a national nuclear programme has been defined, meanwhile purchasing abroad the necessary initial cores and refuelling; (2) to start local fabrication of fuel elements as soon as possible in order to provide the first refuelling of the first unit of Laguna Verde, confronting the economic risks of such a decision with the advantages of immediate action. Both options are analysed in detail, comparing them especially from the economic point of view. Current information from potential licensors for design and manufacture are used in the analysis. (author)

Tuning up and fabrication of U3Si2 nuclear material

International Nuclear Information System (INIS)

Pasqualini, Enrique E.; Echenique, Patricia N.; Rossi, Gustavo S.; Canil, Eduardo E.; Esteban, Adolfo; Lopez, Marisol; Adelfang, Pablo

2000-01-01

This work describes the tuning up and fabrication of uranium-silicide powder for its utilization as nuclear fuel in material testing reactors taking in account NUREG-1313 recommendations, the experience of several suppliers and the one acquired in this work.Several alloy compositions were melted with natural uranium at temperatures of about 1800 degree C for adjusting composition and ingot homogeneity. Alumina, magnesia and zirconia-5% stabilized yttria crucibles were tested to evaluate the degree of contamination introduced by chemical attack of molten uranium and silicon. The fabrication procedure of 20% enriched uranium-silicide powder was established for building up the P-06 fuel element that actually is being irradiated at the RA-3 reactor facility. The selected procedures of fabrication and the critical analysis for the interpretation of several specifications are discussed. Results are shown of the obtained ingots and powder produced with the enriched uranium-silicide. (author)

The fabrication of nozzles for nuclear components by welding

International Nuclear Information System (INIS)

Moraes, M.M.; Krausser, P.; Echeverria, J.A.V.

1986-01-01

A nozzle with medium outside diameter of 1000 mm and medium thickness of 150 mm composed integrally by deposited metal by submerged-arc (wire S3NiMo1, 0.5mm) was fabricated in NUCLEP. The nondestructive, mechanical, metallographic and chemical testing carried out in a test sample made by the same procedure and welding parameters, showed results according to specifications established for primary components for nuclear power plants, and the tests presented mechanical properties and tenacity better than similar nozzle samples. This nozzle is cheapest concerning to importations, in respecting to its forged similar. (M.C.K.) [pt

Establishing QC/QA system in the fabrication of nuclear fuel assemblies

International Nuclear Information System (INIS)

Suh, K.S.; Choi, S.K.; Park, H.G.; Park, T.G.; Chung, J.S.

1980-01-01

Quality control instruction manuals and inspection methods for UO 2 powder and zircaloy materials as the material control, and for UO 2 pellets and nuclear fuel rods as the process control were established. And for the establishment of Q.A programme, the technical specifications of the purchased materials, the control regulation of the measuring and testing equipments, and traceability chart as a part of document control have also been provided and practically applied to the fuel fabrication process

The data acquisition system for the management of nuclear materials involved in the fabrication of MOX fuel at the Cogema plant in Cadarache

International Nuclear Information System (INIS)

Crousilles, M.; Beche, M.; Dalverny, G.

2001-01-01

This article presents the follow-up system of all the nuclear materials that are involved in the industrial process of MOX fuel fabrication. This system, called Concerto, allows the management of MOX fabrication but also of any nuclear material transfer and of the stockpile of nuclear materials with taking into account their own specificity such as the risk of criticality. Operators that intervene on the different steps of the fabrication process, supply Concerto with information so Concerto can be considered as a near real-time system providing and recording the localization, the composition, the weight, the container,... of any batch of nuclear materials. Concerto complies with the requirements of quality assurance but also of nuclear safety by forbidding any transfer whenever the maximal authorized quantity would be exceeded. (A.C.)

Flexible manufacturing systems and their relevance in nuclear fuel fabrication in India

International Nuclear Information System (INIS)

Ramakumar, M.S.

1989-01-01

Fabrication of nuclear reactor fuel bundle involves several materials and a number of complicated technologies and the process of manufacture has to conform to stringent standards. The Indian Nuclear Programme relies heavily on indigeneous capability of manufacture of nuclear fuels as well as automation of the related facilities. Automation of the existing nuclear facilities is a challenge in view of the characteristic plant environments and process demands as well as the various mechanical and metallurgical steps involved. This paper discusses their requirements and the measures initiated for achieving a high order of automation in Indian nuclear facilities. As a first step, specific automation steps are being incorporated in the existing plants. Such interface automation will enhance productivity and avoid the need for building new totally automated palnts. Flexible manufacturing system as applied here, has a different connotation vis-a-vis conventional manufacturing industry. Robotic devices, even for stacking jobs, have not been used on a large scale the world over. (author). 6 figs

Applications of ultrasonic phased array technique during fabrication of nuclear tubing and other components for the Indian nuclear power program

International Nuclear Information System (INIS)

Kapoor, K.

2015-01-01

Ultrasonic phased array technique has been applied in fabrication of nuclear fuel and structural at NFC. The integrity of the nuclear fuel and structural components is most crucial as they are exposed to severe environment during operation leading to rapid degradation of its properties during its lifecycle. Nuclear Fuel Complex has mandate for the fabrication of the nuclear fuel and core structurals for Indian PHWRs/BWR, sub-assemblies for the PFBR and steam generator tubing for PFBR and PHWRs which are the most critical materials for the Indian Nuclear Power program. NDE during fabrication of these materials is thus most crucial as it provides the confidence to the designer for safe operation during its lifetime. Many of these techniques have to be developed in-house to meet unique requirements of high sensitivity, resolution and shape of the components. Some of the advancements in the NDE during the fabrication include use of ultrasonic phased array which is detailed in this paper

Environmental aspects based on operation performance of nuclear fuel fabrication facilities

International Nuclear Information System (INIS)

2001-07-01

This publication was prepared within the framework of the IAEA Project entitled Development and Upgrading of Guidelines, Databases and Tools for Integrating Comparative Assessment into Energy System Analysis and Policy Making, which included the collection, review and input of data into a database on health and environmental impacts related to operation of nuclear fuel cycle facilities. The objectives of the report included assembling environmental data on operational performance of nuclear fabrication facilities in each country; compiling and arranging the data in a database, which will be easily available to experts and the public; and presenting data that may be of value for future environmental assessment of nuclear fabrication facilities

Radiological surveillance in the nuclear fuel fabrication in Mexico

International Nuclear Information System (INIS)

Garcia A, J.; Reynoso V, R.; Delgado A, G.

1996-01-01

The objective of this report is to present the obtained results related to the application of the radiological safety programme established at the Nuclear Fuel Fabrication Pilot Plant (NFFPF) in Mexico, such as: surveillance methods, radiological protection criteria and regulations, radiation control and records and the application of ALARA recommendation. During the starting period from April 1994 to April 1995, at the NFFPF were made two nuclear fuel bundles a Dummy and other to be burned up in a BWR the mainly process activities are: UO 2 powder receiving, powder pressing for the pellets formation, pellets grinding, cleaning and drying, loading into a rod, Quality Control testing, nuclear fuel bundles assembly. The NFFPF is divided into an unsealed source area (pellets manufacturing plant) and into a sealed source area (rods fabrication plant). The control followed have helped to detect failures and to improve the safety programme and operation. (authors). 1 ref., 3 figs

Material control in nuclear fuel fabrication facilities. Part I. Fuel descriptions and fabrication processes, P.O. 1236909 Final report

International Nuclear Information System (INIS)

Borgonovi, G.M.; McCartin, T.J.; Miller, C.L.

1978-12-01

The report presents information on foreign nuclear fuel fabrication facilities. Fuel descriptions and fuel fabrication information for three basic reactor types are presented: The information presented for LWRs assumes that Pu--U Mixed Oxide Fuel (MOX) will be used as fuel

Review of training methods employed in nuclear fuel fabrication plants

International Nuclear Information System (INIS)

Box, W.D.; Browder, F.N.

1975-01-01

A search of the literature through the Nuclear Safety Information Center revealed that 86 percent of the incidents that have occurred in fuel fabrication plants can be traced directly or indirectly to insufficient operator training. In view of these findings, a review was made of the training programs now employed by the nuclear fuel fabrication industry. Most companies give the new employee approximately 20 hours of orientation courses, followed by 60 to 80 hours of on-the-job training. It was concluded that these training programs should be expanded in both scope and depth. A proposed program is outlined to offer guidance in improving the basic methods currently in use

Fabrication of fuel elements interplay between typical SNR Mark Ia specifications and the fuel element fabrication

International Nuclear Information System (INIS)

Biermann, W.K.; Heuvel, H.J.; Pilate, S.; Vanderborck, Y.; Pelckmans, E.; Vanhellemont, G.; Roepenack, H.; Stoll, W.

1987-01-01

The core and fuel were designed for the SNR-300 first core by Interatom GmbH and Belgonucleaire. The fuel was fabricated by Alkem/RBU and Belgonucleaire. Based on the preparation of drawings and specifications and on the results of the prerun fabrication, an extensive interplay took place between design requirements, specifications, and fabrication processes at both fuel plants. During start-up of pellet and pin fabrication, this solved such technical questions as /sup 239/Pu equivalent linear weight, pellet density, stoichiometry of the pellets, and impurity content. Close cooperation of designers and manufacturers has allowed manufacture of 205 fuel assemblies without major problems

The Role of Friction Stir Welding in Nuclear Fuel Plate Fabrication

International Nuclear Information System (INIS)

Burkes, D.; Medvedev, P.; Chapple, M.; Amritkar, A.; Wells, P.; Charit, I

2009-01-01

The friction bonding process combines desirable attributes of both friction stir welding and friction stir processing. The development of the process is spurred on by the need to fabricate thin, high density, reduced enrichment fuel plates for nuclear research reactors. The work seeks to convert research and test reactors currently operating on highly enriched uranium fuel to operate on low enriched uranium fuel without significant loss in reactor performance, safety characteristics, or significant increase in cost. In doing so, the threat of global nuclear material proliferation will be reduced. Feasibility studies performed on the process show that this is a viable option for mass production of plate-type nuclear fuel. Adapting the friction stir weld process for nuclear fuel fabrication has resulted in the development of several unique ideas and observations. Preliminary results of this adaptation and process model development are discussed

Radiation doses and cause-specific mortality among workers at a nuclear materials fabrication plant

International Nuclear Information System (INIS)

Checkoway, H.; Pearce, N.; Crawford-Brown, D.J.; Cragle, D.L.

1988-01-01

A historical cohort mortality study was conducted among 6781 white male employees from a nuclear weapons materials fabrication plant for the years 1947-1979. Exposures of greatest concern are alpha and gamma radiation emanating primarily from insoluble uranium compounds. Among monitored workers, the mean cumulative alpha radiation dose to the lung was 8.21 rem, and the mean cumulative external whole body penetrating dose from gamma radiation was 0.96 rem. Relative to US white males, the cohort experienced mortality deficits from all causes combined, cardiovascular diseases, and from most site-specific cancers. Mortality excesses of lung and brain and central nervous system cancers were seen from comparisons with national and state rates. Dose-response trends were detected for lung cancer mortality with respect to cumulative alpha and gamma radiation, with the most pronounced trend occurring for gamma radiation among workers who received greater than or equal to 5 rem of alpha radiation. These trends diminished in magnitude when a 10-year latency assumption was applied. Under a zero-year latency assumption, the rate ratio for lung cancer mortality associated with joint exposure of greater than or equal to 5 versus less than 1 rem of both types of radiation is 4.60 (95% confidence limits (CL) 0.91, 23.35), while the corresponding result, assuming a 10-year latency, is 3.05 (95% CL 0.37, 24.83). While these rate ratios, which are based on three and one death, respectively, lack statistical precision, the observed dose-response trends indicate potential carcinogenic effects to the lung of relatively low-dose radiation. There are no dose-response trends for mortality from brain and central nervous system cancers

Review of training methods employed in nuclear fuel fabrication plants

International Nuclear Information System (INIS)

Box, W.D.; Browder, F.N.

A search of the literature through the Nuclear Safety Information Center revealed that approximately 86 percent of the incidents that have occurred in fuel fabrication plants can be traced directly or indirectly to insufficient operator training. In view of these findings, a review was made of the training programs now employed by the nuclear fuel fabrication industry. Most companies give the new employee approximately 20 h of orientation courses, followed by 60 to 80 h of on-the-job training. It was concluded that these training programs should be expanded in both scope and depth. A proposed program is outlined to offer guidance in improving the basic methods currently in use. (U.S.)

Regulations concerning the fabricating business of nuclear fuel materials

International Nuclear Information System (INIS)

1977-01-01

As regards an application for permission of an fabricating business of nuclear fuel materials, it should describe the site of the fabricating facilities and the structure and equipments of buildings (fire-resistant, aseismatic, waterproof, ventilating and air-tight structures), etc. The business plan to be attached to the foregoing application should contain 1) scheduled date when the fabricating business starts, 2) scheduled amounts of products classified by the kinds in each business year within 5 years since the business starts, 3) the amount and the procurement plan of funds necessary for the operation, etc. For the permission of change of a fabricating business, an application must be filed. One who wants to obtain the permission of design and construction of fabricating facilities must file an application. One who wants to undergo inspection of the construction of fabricating facilities must file an application in which various items must be written. After such inspection has been done and it is regarded as passable, a certificate of passing inspection will be given. (Rikitake, Y.)

Thermal insulation system design and fabrication specification (nuclear) for the Clinch River Breeder Reactor plant

International Nuclear Information System (INIS)

1978-01-01

This specification defines the design, analysis, fabrication, testing, shipping, and quality requirements of the Insulation System for the Clinch River Breeder Reactor Plant (CRBRP), near Oak Ridge, Tennessee. The Insulation System includes all supports, convection barriers, jacketing, insulation, penetrations, fasteners, or other insulation support material or devices required to insulate the piping and equipment cryogenic and other special applications excluded. Site storage, handling and installation of the Insulation System are under the cognizance of the Purchaser

Design of a quality assurance system in the nuclear fuel fabrication

International Nuclear Information System (INIS)

Garcia Rojas Palacios, L.

1992-01-01

A)For the first time a project on nuclear fuel fabrication is going to be lead in this country. For this reason the work is oriented to establish a quality assurance system for the different stages of fuel fabrication. C) The work of this thesis was developed first by means of an analysis of quality philosophies of Deming, Ishikawa, Juran and Crosby from which several important points were stracted to be used in the designed quality system. Metrology and normalization are so important for quality control that a study of them is made considering definitions, unit systems and type of errors (for Metrology) as well as standards for quality systems, qualification, destructive and non destructive tests, shipment, packing for nuclear power plants. With the standards as a basis, the working strategy for the system was reached, as well as the design of control cards and the design of documents for inspection control, personnel and its documentation and finally the diagrams for each one of the fabrication stages

Quality control in nuclear fuel fabrication on the inspection basis; Control de calidad para fabricacion de combustible nuclear en base a inspecciones

Energy Technology Data Exchange (ETDEWEB)

Fuentes S, A. [Instituto Tecnologico de Toluca, Toluca (Mexico)

1997-12-31

Every plant productive of electric power requires the use of energetics for the transformation to electricity. In the nucleo electric plant the energetic is the uranium, in which it makes ensembles and is used as fuel in the reactor. To assure that the fuel ensembles fulfill the specifications and requirements of design stipulated in the nucleo electric plant is that under a quality control through inspections during the fabrication process. The purpose of this work is to study and verify that the lineaments of the standard 10 CFR 50 appendix B `Quality assurement for nuclear plants` specially in the criteria `Inspections` that is used to guarantee the quality of the ensembles. This standard is the one that rules every activity and operation inside the pilot plant and its established in the quality program in the production of nuclear fuel for the Laguna Verde plant. The quality of the assemble is verified through each one of the tests or inspections due to the importance of it in the fabrication of fuel. (Author)

Quality control in nuclear fuel fabrication on the inspection basis; Control de calidad para fabricacion de combustible nuclear en base a inspecciones

Energy Technology Data Exchange (ETDEWEB)

Fuentes S, A [Instituto Tecnologico de Toluca, Toluca (Mexico)

1998-12-31

Every plant productive of electric power requires the use of energetics for the transformation to electricity. In the nucleo electric plant the energetic is the uranium, in which it makes ensembles and is used as fuel in the reactor. To assure that the fuel ensembles fulfill the specifications and requirements of design stipulated in the nucleo electric plant is that under a quality control through inspections during the fabrication process. The purpose of this work is to study and verify that the lineaments of the standard 10 CFR 50 appendix B `Quality assurement for nuclear plants` specially in the criteria `Inspections` that is used to guarantee the quality of the ensembles. This standard is the one that rules every activity and operation inside the pilot plant and its established in the quality program in the production of nuclear fuel for the Laguna Verde plant. The quality of the assemble is verified through each one of the tests or inspections due to the importance of it in the fabrication of fuel. (Author)

The fabrication of nuclear fuel elements in Mexico

International Nuclear Information System (INIS)

Guerrero Morillo, H.L.

1977-01-01

The situation of the nucleoelectrical generation in Mexico by 1976 is described: two nuclear reactors under construction but no defined program on the type and start-up dates for the next power plants. However the existence of a general plan on nuclear power plants is mentioned, which, according to the last estimates reaches to 10,000 MW installed by 1990. The national intension, definitely expressed in the Law, is to supply domestic nuclear fuel to the power reactors operating in the country, starting with the first reload for the two BWR's at the first national station in Laguna Verde, which will be required at the end of 1981 and of 1982, respectively. Before such circumstances and the relatively short amounts of fuel elements that should be produced for those two unique reactors, Mexico already has to adopt a strategy to follow in respect to fuel elements fabrication. The two main options are analyzed: 1. To delay the local fabrication until a National Nuclear Program may be defined, meanwhile purchasing abroad the necessary reloads and initial cores; and 2. To start as soon as possible the local fuel elements fabrication in order to supply fuel for the first reload of the first unit of Laguna Verde, confronting the economical risks of such posture with the advantages of an immediate action. Both options are analyzed in detail comparing them specially under the economic point of view, standing out immediately the big effect of some factors which are economically imponderable, as experience and independance that would be gained with the second option. Emphasis is made on the advantages and risks of any case. According to the first option and once a National Program is defined, the work would be heavy but of simple strategy. On the contrary, the second option requires the adoption of a more complicated strategy, as either the project of the factory as its initial operation should be made under transient conditions, in view of the expected future expansion still

Criticality Calculations for a Typical Nuclear Fuel Fabrication Plant with Low Enriched Uranium

International Nuclear Information System (INIS)

Elsayed, Hade; Nagy, Mohamed; Agamy, Said; Shaat, Mohmaed

2013-01-01

The operations with the fissile materials such as U 235 introduce the risk of a criticality accident that may be lethal to nearby personnel and can lead the facility to shutdown. Therefore, the prevention of a nuclear criticality accident should play a major role in the design of a nuclear facility. The objectives of criticality safety are to prevent a self-sustained nuclear chain reaction and to minimize the consequences. Sixty criticality accidents were occurred in the world. These are accidents divided into two categories, 22 accidents occurred in process facilities and 38 accidents occurred during critical experiments or operations with research reactor. About 21 criticality accidents including Japan Nuclear Fuel Conversion Co. (JCO) accident took place with fuel solution or slurry and only one accident occurred with metal fuel. In this study the nuclear criticality calculations have been performed for a typical nuclear fuel fabrication plant producing nuclear fuel elements for nuclear research reactors with low enriched uranium up to 20%. The calculations were performed for both normal and abnormal operation conditions. The effective multiplication factor (k eff ) during the nuclear fuel fabrication process (Uranium hexafluoride - Ammonium Diuranate conversion process) was determined. Several accident scenarios were postulated and the criticalities of these accidents were evaluated. The computer code MCNP-4B which based on Monte Carlo method was used to calculate neutron multiplication factor. The criticality calculations Monte Carlo method was used to calculate neutron multiplication factor. The criticality calculations were performed for the cases of, change of moderator to fuel ratio, solution density and concentration of the solute in order to prevent or mitigate criticality accidents during the nuclear fuel fabrication process. The calculation results are analyzed and discussed

Regulations concerning the fabricating business of nuclear fuel materials

International Nuclear Information System (INIS)

1979-01-01

The regulations are entirely revised under the law for the regulations of nuclear materials, nuclear fuel materials and reactors and provisions concerning the fabricating business in the order for execution of the law. Basic concepts and terms are defined, such as: exposure dose; accumulative dose; controlled area; inspected surrounding area; employee and radioactive waste. The application for permission of the fabricating business shall include: location of processing facilities; structure of building structure and equipment of chemical processing facilities; molding facilities; structure and equipment of covering and assembling facilities, storage facilities of nuclear fuel materials and disposal facilities of radioactive waste, etc. Records shall be made and kept for particular periods in each works and place of enterprise on inspection of processing facilities, control of dose, operation, maintenance, accident of processing facilities and weather. Specified measures shall be taken in controlled area and inspected surrounding area to restrict entrance. Measures shall be made not to exceed permissible exposure dose for employees defined by the Director General of Science and Technology Agency. Inspection and check up of processing facilities shall be carried on by employees more than once a day. Operation of processing facilities, transportation in the works and enterprise, storage, disposal, safety securing, report and measures in dangerous situations, etc. are in detail prescribed. (Okada, K.)

Material engineering to fabricate rare earth erbium thin films for exploring nuclear energy sources

Science.gov (United States)

Banerjee, A.; Abhilash, S. R.; Umapathy, G. R.; Kabiraj, D.; Ojha, S.; Mandal, S.

2018-04-01

High vacuum evaporation and cold-rolling techniques to fabricate thin films of the rare earth lanthanide-erbium have been discussed in this communication. Cold rolling has been used for the first time to successfully fabricate films of enriched and highly expensive erbium metal with areal density in the range of 0.5-1.0 mg/cm2. The fabricated films were used as target materials in an advanced nuclear physics experiment. The experiment was designed to investigate isomeric states in the heavy nuclei mass region for exploring physics related to nuclear energy sources. The films fabricated using different techniques varied in thickness as well as purity. Methods to fabricate films with thickness of the order of 0.9 mg/cm2 were different than those of 0.4 mg/cm2 areal density. All the thin films were characterized using multiple advanced techniques to accurately ascertain levels of contamination as well as to determine their exact surface density. Detailed fabrication methods as well as characterization techniques have been discussed.

Introduction to Exxon nuclear fuel fabrication plant

International Nuclear Information System (INIS)

Schneider, R.A.

1985-01-01

The Exxon Nuclear low-enriched uranium fuel fabrication plant in Richland, Washington produces fuel assemblies for both pressurized water and boiling water reactors. The Richland plant was the first US bulk-handling facility selected by the IAEA for inspection under the US-IAEA Safeguards Agreement. The plant was under IAEA inspection from March 1981 through October 1983. This text provides a written description of the plant layout, operation and process. The text also includes a one ton-a-day model (or reference) plant which was adapted from the Exxon Nuclear plant. The Model Plant provides a generic example of a low-enriched uranium (LEU) bulk-handling facility. The Model Plant is used to illustrate in a more quantitative way some of the key safeguards requirements for a bulk-handling facility

Control of nuclear material hold-up: The key factors for design and operation of MOX fuel fabrication plants in Europe

International Nuclear Information System (INIS)

Beaman, M.; Beckers, J.; Boella, M.

2001-01-01

Full text: Some protagonists of the nuclear industry suggest that MOX fuel fabrication plants are awash with nuclear materials which cannot be adequately safeguarded and that materials 'stuck in the plant' could conceal clandestine diversion of plutonium. In Europe the real situation is quite different: nuclear operators have gone to considerable efforts to deploy effective systems for safety, security, quality and nuclear materials control and accountancy which provide detailed information. The safeguards authorities use this information as part of the safeguards measures enabling them to give safeguards assurances for MOX fuel fabrication plants. This paper focuses on the issue of hold-up: definition of the hold-up and of the so-called 'hidden inventory'; measures implemented by the plant operators, from design to day to day operations, for minimising hold-up and 'hidden inventory'; plant operators' actions to manage the hold-up during production activities but also at PIT/PIV time; monitoring and management of the 'hidden inventory'; measures implemented by the safeguards authorities and inspectorate for verification and control of both hold-up and 'hidden inventory'. The examples of the different plant specific experiences related in this paper reveal the extensive experience gained in european MOX fuel fabrication plants by the plant operators and the safeguards authorities for the minimising and the control of both hold-up and 'hidden inventory'. MOX fuel has been fabricated in Europe, with an actual combined capacity of 2501. HM/year subject, without any discrimination, to EURATOM Safeguards, for more than 30 years and the total output is, to date, some 1000 t.HM. (author)

Development of Nuclear Fuel Remote Fabrication Technology

International Nuclear Information System (INIS)

Lee, Jung Won; Yang, M. S.; Kim, S. S. and others

2005-04-01

The aim of this study is to develop the essential technology of dry refabrication using spent fuel materials in a laboratory scale on the basis of proliferation resistance policy. The emphasis is placed on the assessment and the development of the essential technology of dry refabrication using spent fuel materials. In this study, the remote fuel fabrication technology to make a dry refabricated fuel with an enhanced quality was established. And the instrumented fuel pellets and mini-elements were manufactured for the irradiation testing in HANARO. The design and development technology of the remote fabrication equipment and the remote operating and maintenance technology of the equipment in hot cell were also achieved. These achievements will be used in and applied to the future back-end fuel cycle and GEN-IV fuel cycle and be a milestone for Korea to be an advanced nuclear country in the world

FABRICATION OF TUBE TYPE FUEL ELEMENT FOR NUCLEAR REACTORS

Science.gov (United States)

Loeb, E.; Nicklas, J.H.

1959-02-01

A method of fabricating a nuclear reactor fuel element is given. It consists essentially of fixing two tubes in concentric relationship with respect to one another to provide an annulus therebetween, filling the annulus with a fissionablematerial-containing powder, compacting the powder material within the annulus and closing the ends thereof. The powder material is further compacted by swaging the inner surface of the inner tube to increase its diameter while maintaining the original size of the outer tube. This process results in reduced fabrication costs of powdered fissionable material type fuel elements and a substantial reduction in the peak core temperatures while materially enhancing the heat removal characteristics.

Fabrication and Testing of CERMET Fuel Materials for Nuclear Thermal Propulsion

Science.gov (United States)

Hickman, Robert; Broadway, Jeramie; Mireles, Omar

2012-01-01

A first generation Nuclear Cryogenic Propulsion Stage (NCPS) based on Nuclear Thermal Propulsion (NTP) is currently being developed for Advanced Space Exploration Systems. The overall goal of the project is to address critical NTP technology challenges and programmatic issues to establish confidence in the affordability and viability of NTP systems. The current technology roadmap for NTP identifies the development of a robust fuel form as a critical near term need. The lack of a qualified nuclear fuel is a significant technical risk that will require a considerable fraction of program resources to mitigate. Due to these risks and the cost for qualification, the development and selection of a primary fuel must begin prior to Authority to Proceed (ATP) for a specific mission. The fuel development is a progressive approach to incrementally reduce risk, converge the fuel materials, and mature the design and fabrication process of the fuel element. A key objective of the current project is to advance the maturity of CERMET fuels. The work includes fuel processing development and characterization, fuel specimen hot hydrogen screening, and prototypic fuel element testing. Early fuel materials development is critical to help validate requirements and fuel performance. The purpose of this paper is to provide an overview and status of the work at Marshall Space Flight Center (MSFC).

Fabrication and Characterization of Surrogate Glasses Aimed to Validate Nuclear Forensic Techniques

Science.gov (United States)

2017-12-01

the glass formed during a nuclear event, trinitite [14]. The SiO2 composition is generally greater than 50% for trinitite and can vary appreciably...CHARACTERIZATION OF SURROGATE GLASSES AIMED TO VALIDATE NUCLEAR FORENSIC TECHNIQUES by Ken G. Foos December 2017 Thesis Advisor: Claudia...December 2017 3. REPORT TYPE AND DATES COVERED Master’s thesis 4. TITLE AND SUBTITLE FABRICATION AND CHARACTERIZATION OF SURROGATE GLASSES AIMED TO

Fabrication of the fuel elements cladding for utilization in the fluidized bed nuclear reactor

International Nuclear Information System (INIS)

Schaeffer, L.; Sefidvash, F.

1986-01-01

A method for the fabrication of cladding of the spherical fuel elements for the utilization in the fluidized bed nuclear reactor is presented. Some prelimminary experiments were performed to adopt a method which adapt itself to mass production with the desired high quality. Still methods for cladding fabrication are under study. (Author) [pt

Fabrication of High Temperature Cermet Materials for Nuclear Thermal Propulsion

Science.gov (United States)

Hickman, Robert; Panda, Binayak; Shah, Sandeep

2005-01-01

Processing techniques are being developed to fabricate refractory metal and ceramic cermet materials for Nuclear Thermal Propulsion (NTP). Significant advances have been made in the area of high-temperature cermet fuel processing since RoverNERVA. Cermet materials offer several advantages such as retention of fission products and fuels, thermal shock resistance, hydrogen compatibility, high conductivity, and high strength. Recent NASA h d e d research has demonstrated the net shape fabrication of W-Re-HfC and other refractory metal and ceramic components that are similar to UN/W-Re cermet fuels. This effort is focused on basic research and characterization to identify the most promising compositions and processing techniques. A particular emphasis is being placed on low cost processes to fabricate near net shape parts of practical size. Several processing methods including Vacuum Plasma Spray (VPS) and conventional PM processes are being evaluated to fabricate material property samples and components. Surrogate W-Re/ZrN cermet fuel materials are being used to develop processing techniques for both coated and uncoated ceramic particles. After process optimization, depleted uranium-based cermets will be fabricated and tested to evaluate mechanical, thermal, and hot H2 erosion properties. This paper provides details on the current results of the project.

Facile Fabrication of Animal-Specific Positioning Molds For Multi-modality Molecular Imaging

International Nuclear Information System (INIS)

Park, Jeong Chan; Oh, Ji Eun; Woo, Seung Tae

2008-01-01

Recently multi-modal imaging system has become widely adopted in molecular imaging. We tried to fabricate animal-specific positioning molds for PET/MR fusion imaging using easily available molding clay and rapid foam. The animal-specific positioning molds provide immobilization and reproducible positioning of small animal. Herein, we have compared fiber-based molding clay with rapid foam in fabricating the molds of experimental animal. The round bottomed-acrylic frame, which fitted into microPET gantry, was prepared at first. The experimental mice was anesthetized and placed on the mold for positioning. Rapid foam and fiber-based clay were used to fabricate the mold. In case of both rapid foam and the clay, the experimental animal needs to be pushed down smoothly into the mold for positioning. However, after the mouse was removed, the fabricated clay needed to be dried completely at 60 .deg. C in oven overnight for hardening. Four sealed pipe tips containing [ 18 F]FDG solution were used as fiduciary markers. After injection of [ 18 F]FDG via tail vein, microPET scanning was performed. Successively, MRI scanning was followed in the same animal. Animal-specific positioning molds were fabricated using rapid foam and fiber-based molding clay for multimodality imaging. Functional and anatomical images were obtained with microPET and MRI, respectively. The fused PET/MR images were obtained using freely available AMIDE program. Animal-specific molds were successfully prepared using easily available rapid foam, molding clay and disposable pipet tips. Thanks to animal-specific molds, fusion images of PET and MR were co-registered with negligible misalignment

Improved fabrication of HgI2 nuclear radiation detectors by machine-cleaving

International Nuclear Information System (INIS)

Levi, A.; Burger, A.; Schieber, M.; Vandenberg, L.; Yellon, W.B.; Alkire, R.W.

1982-01-01

The perfection of machine-cleaved sections from HgI 2 bulk crystals was examined. The perfection of the machine-cleaved sections as established by gamma diffraction rocking curves was found to be much better than the perfection of hand-cleaved sections or as grown thin platelets, reaching a perfection similar to that of the wire-sawn sections of HgI 2 . A correlation between the perfection and the thickness of the machine-cleaved section was also found, i.e., the thicker the cleaved-section the more perfect it is. The reproducibility of the fabrication was significantly improved by using machine cleaving in the process of fabrication. Large single crystals of HgI 2 weighing 20 to 200 g, can be grown from the vapor phase using the TOM Technique. In order to fabricate nuclear radiation detectors from these single crystals, thin sections of about 0.4 to 0.8 mm thickness have to be prepared. Up till now, the state-of-the-art of fabricating HgI 2 nuclear radiation detectors involved two methods to get thin sections from the large single crystals: (1) hand-cleaving using a razor-blade and (2) solution wire sawing. The chemical wire sawing method involves a loss of about 50% of the crystal volume and is usually followed by a chemical polishing process which involves a significant loss of volume of the original volume. This procedure is complicated and wasteful. The traditional fabrication method, i.e., hand-cleaving followed by rapid nonselective chemical etching, is simpler and less wasteful

Aspects for selection of materials and fabrication processes for nuclear component manufacturing

International Nuclear Information System (INIS)

Pernstich, K.

1980-01-01

For components of the Nuclear steam supply System of Light Water Reactors an extremely high safety standard is required. These requirements only can be met by adequate selection of materials and fabrication processes and their proper application in combination with strict quality assurance and control measurements. A general overview of the basic aspects to be considered in this connection is presented together with an indication of the present state of art for the main materials and fabrication processes. (author) [pt

Design, fabrication and transportation of Si rotating device

International Nuclear Information System (INIS)

Kimura, Nobuaki; Imaizumi, Tomomi; Takemoto, Noriyuki; Tanimoto, Masataka; Saito, Takashi; Hori, Naohiko; Tsuchiya, Kunihiko; Romanova, Nataliya; Gizatulin, Shamil; Martyushov, Alexandr; Nakipov, Darkhan; Chakrov, Petr; Tanaka, Futoshi; Nakajima, Takeshi

2012-06-01

Si semiconductor production by Neutron Transmutation Doping (NTD) method using the Japan Materials Testing Reactor (JMTR) has been investigated in Neutron Irradiation and Testing Reactor Center, Japan Atomic Energy Agency (JAEA) in order to expand industry use. As a part of investigations, irradiation test of silicon ingot for development of NTD-Si with high quality was planned using WWR-K in Institute of Nuclear Physics (INP), National Nuclear Center of Republic of Kazakhstan (NNC-RK) based on one of specific topics of cooperation (STC), Irradiation Technology for NTD-Si (STC No.II-4), on the implementing arrangement between NNC-RK and the JAEA for 'Nuclear Technology on Testing/Research Reactors' in cooperation in research and development in nuclear energy and technology. As for the irradiation test, Si rotating device was fabricated in JAEA, and the fabricated device was transported with irradiation specimens from JAEA to INP-NNC-RK. This report described the design, the fabrication, the performance test of the Si rotating device and transportation procedures. (author)

CDMS Detector Fabrication Improvements and Low Energy Nuclear Recoil Measurements in Germanium

Energy Technology Data Exchange (ETDEWEB)

Jastram, Andrew [Texas A & M Univ., College Station, TX (United States)

2015-12-01

As the CDMS (Cryogenic Dark Matter Search) experiment is scaled up to tackle new dark matter parameter spaces (lower masses and cross-sections), detector production efficiency and repeatability becomes ever more important. A dedicated facility has been commissioned for SuperCDMS detector fabrication at Texas A&M University (TAMU). The fabrication process has been carefully tuned using this facility and its equipment. Production of successfully tested detectors has been demonstrated. Significant improvements in detector performance have been made using new fabrication methods, equipment, and tuning of process parameters. This work has demonstrated the capability for production of next generation CDMS SNOLAB detectors. Additionally, as the dark matter parameter space is probed further, careful calibrations of detector response to nuclear recoil interactions must be performed in order to extract useful information (in relation to dark matter particle characterzations) from experimental results. A neutron beam of tunable energy is used in conjunction with a commercial radiation detector to characterize ionization energy losses in germanium during nuclear recoil events. Data indicates agreement with values predicted by the Lindhard equation, providing a best-t k-value of 0.146.

Application of gas shielded arc welding and submerged arc welding for fabrication of nuclear reactor vessels

International Nuclear Information System (INIS)

Gehani, M.L.; Rodrigues, W.D.

1976-01-01

The remarkable progress made in the development of knowhow and expertise in the manufacture of equipment for nuclear power plants in India is outlined. Some of the specific advances made in the application of higher efficiency weld processes for fabrication of nuclear reactor vessels and the higher level of quality attained are discussed in detail. Modifications and developments in submerged arc, gas tungsten arc and gas metal arc processes for welding of Calandria which have been a highly challenging and rewarding experience are discussed. Future scope for making the gas metal arc process more economical by using various gas-mixes like Agron + Oxygen, Argon + Carbon Dioxide, Argon + Nitrogen (for Copper Alloys) etc., in various proportions are outlined. Quality and dimensional control exercised in these jobs of high precision are highlighted. (K.B.)

Basic tendencies of restructured UO2 nuclear fuels fabrication industry for water-moderated reactors

International Nuclear Information System (INIS)

Makhova, V.A.; Bokshitskij, V.I.; Blinova, I.V.

2002-01-01

Processes of reformation and consolidation of firms and frontier nuclear fuels fabrication industry associated with processes of globalization and deregulation of electric power market are analyzed. Current state of nuclear fuel market and basic factors influenced on the market are presented. The role of nuclear fuel in increasing competition of NPP and fundamental directions of innovation action on the creation of perspective kinds of fuel were considered [ru

Fabrication and characterization of joined silicon carbide cylindrical components for nuclear applications

Science.gov (United States)

Khalifa, H. E.; Deck, C. P.; Gutierrez, O.; Jacobsen, G. M.; Back, C. A.

2015-02-01

The use of silicon carbide (SiC) composites as structural materials in nuclear applications necessitates the development of a viable joining method. One critical application for nuclear-grade joining is the sealing of fuel within a cylindrical cladding. This paper demonstrates cylindrical joint feasibility using a low activation nuclear-grade joint material comprised entirely of β-SiC. While many papers have considered joining material, this paper takes into consideration the joint geometry and component form factor, as well as the material performance. Work focused specifically on characterizing the strength and permeability performance of joints between cylindrical SiC-SiC composites and monolithic SiC endplugs. The effects of environment and neutron irradiation were not evaluated in this study. Joint test specimens of different geometries were evaluated in their as-fabricated state, as well as after being subjected to thermal cycling and partial mechanical loading. A butted scarf geometry supplied the best combination of high strength and low permeability. A leak rate performance of 2 × 10-9 mbar l s-1 was maintained after thermal cycling and partial mechanical loading and sustained applied force of 3.4 kN, or an apparent strength of 77 MPa. This work shows that a cylindrical SiC-SiC composite tube sealed with a butted scarf endplug provides out-of-pile strength and permeability performance that meets light water reactor design requirements.

The Hanau atomic energy laws. Nuclear fuel fabrication and the administrative law system

International Nuclear Information System (INIS)

Becker-Neetz, G.; Uebersohn, G.

1989-01-01

The review concentrates on administrative law aspects in the discussion of problems relating to the licences and preliminary notices of approval issued for the Hanau nuclear industry. The authors deal with the licences granted in 1974 (according to sec. 9 Atomic Energy Act), with the extended licensing requirements of sec. 7 Atomic Energy Act as amended by the 3rd amendment (concerning fabrication and handling of nuclear fuels), and the criminal court proceedings examining the conduct of the Alkem management and senior officers of the Hessian Ministry of Economics. Specific aspects investigated in the review include continuation of existing operations in accordance with transitory provisions, replacement of existing by new installations, and preliminary notice of approval. The preliminary notices of approval given up to the date of December 31, 1977 are said to have been illegal and extinct at that date, but the court's decision to abstain from punishment is accepted. The authors outline some possibilities of giving more concrete shape to the judicial control by administrative courts. (RST) [de

Material control in nuclear fuel fabrication facilities. Part II. Accountability, instrumntation, and measurement techniques in fuel fabrication facilities, P.O.1236909. Final report

International Nuclear Information System (INIS)

Borgonovi, G.M.; McCartin, T.J.; McDaniel, T.; Miller, C.L.; Nguyen, T.

1978-12-01

This report describes the measurement techniques, the instrumentation, and the procedures used in accountability and control of nuclear materials, as they apply to fuel fabrication facilities. Some of the material included has appeared elswhere and it has been summarized. An extensive bibliography is included. A spcific example of application of the accountability methods to a model fuel fabrication facility which is based on the Westinghouse Anderson design

Material control in nuclear fuel fabrication facilities. Part II. Accountability, instrumntation, and measurement techniques in fuel fabrication facilities, P. O. 1236909. Final report

Energy Technology Data Exchange (ETDEWEB)

Borgonovi, G.M.; McCartin, T.J.; McDaniel, T.; Miller, C.L.; Nguyen, T.

1978-12-01

This report describes the measurement techniques, the instrumentation, and the procedures used in accountability and control of nuclear materials, as they apply to fuel fabrication facilities. Some of the material included has appeared elswhere and it has been summarized. An extensive bibliography is included. A spcific example of application of the accountability methods to a model fuel fabrication facility which is based on the Westinghouse Anderson design.

Novel fabrication of silicon carbide based ceramics for nuclear applications

Science.gov (United States)

Singh, Abhishek Kumar

Advances in nuclear reactor technology and the use of gas-cooled fast reactors require the development of new materials that can operate at the higher temperatures expected in these systems. These materials include refractory alloys based on Nb, Zr, Ta, Mo, W, and Re; ceramics and composites such as SiC--SiCf; carbon--carbon composites; and advanced coatings. Besides the ability to handle higher expected temperatures, effective heat transfer between reactor components is necessary for improved efficiency. Improving thermal conductivity of the fuel can lower the center-line temperature and, thereby, enhance power production capabilities and reduce the risk of premature fuel pellet failure. Crystalline silicon carbide has superior characteristics as a structural material from the viewpoint of its thermal and mechanical properties, thermal shock resistance, chemical stability, and low radioactivation. Therefore, there have been many efforts to develop SiC based composites in various forms for use in advanced energy systems. In recent years, with the development of high yield preceramic precursors, the polymer infiltration and pyrolysis (PIP) method has aroused interest for the fabrication of ceramic based materials, for various applications ranging from disc brakes to nuclear reactor fuels. The pyrolysis of preceramic polymers allow new types of ceramic materials to be processed at relatively low temperatures. The raw materials are element-organic polymers whose composition and architecture can be tailored and varied. The primary focus of this study is to use a pyrolysis based process to fabricate a host of novel silicon carbide-metal carbide or oxide composites, and to synthesize new materials based on mixed-metal silicocarbides that cannot be processed using conventional techniques. Allylhydridopolycarbosilane (AHPCS), which is an organometal polymer, was used as the precursor for silicon carbide. Inert gas pyrolysis of AHPCS produces near-stoichiometric amorphous

Application of vacuum technology during nuclear fuel fabrication, inspection and characterization

International Nuclear Information System (INIS)

Majumdar, S.

2003-01-01

Full text: Vacuum technology plays very important role during various stages of fabrication, inspection and characterization of U, Pu based nuclear fuels. Controlled vacuum is needed for melting and casting of U, Pu based alloys, picture framing of the fuel meat for plate type fuel fabrication, carbothermic reduction for synthesis of (U-Pu) mixed carbide powder, dewaxing of green ceramic fuel pellets, degassing of sintered pellets and encapsulation of fuel pellets inside clad tube. Application of vacuum technology is also important during inspection and characterization of fuel materials and fuel pins by way of XRF and XRD analysis, Mass spectrometer Helium leak detection etc. A novel method of low temperature sintering of UO 2 developed at BARC using controlled vacuum as sintering atmosphere has undergone successful irradiation testing in Cirus. The paper will describe various fuel fabrication flow sheets highlighting the stages where vacuum applications are needed

Characterization of aerosols from industrial fabrication of mixed-oxide nuclear reactor fuels

International Nuclear Information System (INIS)

Hoover, M.D.; Newton, G.J.

1997-01-01

Recycling plutonium into mixed-oxide (MOX) fuel for nuclear reactors is being given serious consideration as a safe and environmentally sound method of managing plutonium from weapons programs. Planning for the proper design and safe operation of the MOX fuel fabrication facilities can take advantage of studies done in the 1970s, when recycling of plutonium from nuclear fuel was under serious consideration. At that time, it was recognized that the recycle of plutonium and uranium in irradiated fuel could provide a significant energy source and that the use of 239 Pu in light water reactor fuel would reduce the requirements for enriched 235 U as a reactor fuel. It was also recognized that the fabrication of uranium and plutonium reactor fuels would not be risk-free. Despite engineered safety precautions such as the handling of uranium and plutonium in glove-box enclosures, accidental releases of radioactive aerosols from normal containment might occur. Workers might then be exposed to the released materials by inhalation

Control and balance of nuclear matters used for core fabrication of Super Phenix

International Nuclear Information System (INIS)

Beche, M.; Guillet, H.; Heyraud, H.; Levrard, J.; Pajot, J.

1987-05-01

The fabrication of the core of the fast breeder reactor set up at Creys Malville ended in March 1984. It started in 1978 and it required, for the fabrication of the 410 assemblies, the utilization of 7438 kg of plutonium. To satisfy national and international regulations, DPFER/SFER has used a methodology to follow and to control the movements of the nuclear materials. These controls are achieved by physical methods, chemical methods and empiric methods. Euratom has conducted a succession of inspections during the 5.5 years of that campaign. The inventory difference, in the fabrication of that core, represents about 0.1% of the total mass of the plutonium handled [fr

Specification for a total quality assurance programme for nuclear power plants

International Nuclear Information System (INIS)

1980-01-01

This British Standard specifies principles for the establishment and implementation of quality assurance programmes during all phases of design, procurement, fabrication, construction, commissioning, operation, maintenance and decommissioning of structures, systems and components of nuclear power plants. These principles apply to activities affecting the quality of items, such as designing, purchasing, fabricating, handling, shipping, storing, cleaning, erecting, installing, testing, commissioning, operating, inspecting, maintaining, repairing, refuelling, modifying and, eventually decommissioning. (author)

Fabrication and characterization of CeO{sub 2} pellets for simulation of nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

García-Ostos, C.; Rodríguez-Ortiz, J.A. [Department of Mechanical and Materials Engineering, School of Engineering, University of Seville, Seville (Spain); Arévalo, C., E-mail: carevalo@us.es [Department of Mechanical and Materials Engineering, School of Engineering, University of Seville, Seville (Spain); Cobos, J. [CIEMAT, Avenida Complutense, 40, Madrid (Spain); Gotor, F.J. [Materials Science Institute of Seville (CSIC-US), Av. Américo Vespucio, 49, 41092 Seville (Spain); Torres, Y. [Department of Mechanical and Materials Engineering, School of Engineering, University of Seville, Seville (Spain)

2016-03-15

Highlights: • CeO{sub 2} is presented as a surrogate material for UO{sub 2} to study nuclear fuel. • Powder-metallurgy methods are applied to fabricate CeO{sub 2} pellets with controlled porosity. • An optimization of the fabrication parameters is established. • Microstructural and tribo-mechanical characterizations are performed. • Properties are compared to those of the nuclear fuel. - Abstract: Cerium Oxide, CeO{sub 2}, has been shown as a surrogate material to understand irradiated Mixed Oxide (MOX) based matrix fuel for nuclear power plants due to its similar structure, chemical and mechanical properties. In this work, CeO{sub 2} pellets with controlled porosity have been developed through conventional powder-metallurgy process. Influence of the main processing parameters (binder content, compaction pressure, sintering temperature and sintering time) on porosity and volumetric contraction values has been studied. Microstructure and physical properties of sintered compacts have also been characterized through several techniques. Mechanical properties such as dynamic Young's modulus, hardness and fracture toughness have been determined and connected to powder-metallurgy parameters. Simulation of nuclear fuel after reactor utilization with radial gradient porosity is proposed.

Standard Specification for Nuclear Facility Transient Worker Records

CERN Document Server

American Society for Testing and Materials. Philadelphia

1995-01-01

1.1 This specification covers the required content and provides retention requirements for records needed for in-processing of nuclear facility transient workers. 1.2 This specification applies to records to be used for in-processing only. 1.3 This specification is not intended to cover specific skills records (such as equipment operating licenses, ASME inspection qualifications, or welding certifications). 1.4 This specification does not reduce any regulatory requirement for records retention at a licensed nuclear facility. Note 1—Nuclear facilities operated by the U.S. Department of Energy (DOE) are not licensed by the U.S. Nuclear Regulatory Commission (NRC), nor are other nuclear facilities that may come under the control of the U.S. Department of Defense (DOD) or individual agreement states. The references in this specification to licensee, the U.S. NRC Regulatory Guides, and Title 10 of the U.S. Code of Federal Regulations are to imply appropriate alternative nomenclature with respect to DOE, DOD...

Design, fabrication and erection of steel structures important to safety of nuclear facilities

International Nuclear Information System (INIS)

2001-10-01

Civil engineering structures in nuclear installations form an important feature having implications to safety performance of these installations. The objective and minimum requirements for the design of civil engineering buildings/structures to be fulfilled to provide adequate assurance for safety of nuclear installations in India (such as pressurised heavy water reactor and related systems) are specified in the Safety Standard for Civil Engineering Structures Important to Safety of Nuclear Facilities. This standard is written by AERB to specify guidelines for implementation of the above civil engineering safety standard in the design, fabrication and erection of steel structures important to safety

Application of plasma deposition technology for nuclear fuel fabrication

International Nuclear Information System (INIS)

Jung, I. H.; Moon, J. S.; Park, H. S.; Song, K. C.; Lee, C. Y.; Kang, K. H.; Ryu, H. J.; Kim, H. S.; Yang, M. S.

2001-01-01

Yttria-stabilized-zirconia (m.p. 2670.deg. C), was deposited by induction plasma spraying system with a view to develop a new nuclear fuel fabrication technology. To fabricate the dense pellets, the spraying condition was optimized through the process parameters such as, chamber pressure, plasma plate power, powder spraying distance, sheath gas composition, probe position particle size and its morphology. The results with a 5mm thick deposit on rectangular planar graphite substrates showed 97.11% theoretical density, when the sheath gas flow rate was Ar/H 2 120/20 L/min, probe position 8cm, particle size-75 μm and spraying distance 22cm. The microstructure of YSZ deposit by ICP was lamellae and columnar perpendicular to the spraying direction. In the bottom part near the substrate, small equiaxed grains bounded in a layer. In the middle part, relatively regular size of columnar grains with excellent bonding each other were distinctive

Statistical methods to assess and control processes and products during nuclear fuel fabrication

International Nuclear Information System (INIS)

Weidinger, H.

1999-01-01

Very good statistical tools and techniques are available today to access the quality and the reliability of fabrication process as the original sources for a good and reliable quality of the fabricated processes. Quality control charts of different types play a key role and the high capability of modern electronic data acquisition technologies proved, at least potentially, a high efficiency in the more or less online application of these methods. These techniques focus mainly on stability and the reliability of the fabrication process. In addition, relatively simple statistical tolls are available to access the capability of fabrication process, assuming they are stable, to fulfill the product specifications. All these techniques can only result in as good a product as the product design is able to describe the product requirements necessary for good performance. Therefore it is essential that product design is strictly and closely performance oriented. However, performance orientation is only successful through an open and effective cooperation with the customer who uses or applies those products. During the last one to two decades in the west, a multi-vendor strategy has been developed by the utility, sometimes leading to three different fuel vendors for one reactor core. This development resulted in better economic conditions for the user but did not necessarily increase an open attitude with the vendor toward the using utility. The responsibility of the utility increased considerably to ensure an adequate quality of the fuel they received. As a matter of fact, sometimes the utilities had to pay a high price because of unexpected performance problems. Thus the utilities are now learning that they need to increase their knowledge and experience in the area of nuclear fuel quality management and technology. This process started some time ago in the west. However, it now also reaches the utilities in the eastern countries. (author)

Fabrication characteristics of zircaloy tubes for nuclear reactors

International Nuclear Information System (INIS)

Haydt, H.M.

1980-11-01

The production sequence for zircaloy cladding tubes to be used in nuclear reactors is described, with emphasis on the texture after reduction and on the variation in the hydrides orientation. The qualities requested for the cladding tubes are presented and reference is made to the quality control applied in the process. The destructive tests as well as the final inspection to which those tubes are subjected are related. A Fabrication Quality Project is requested from the manufacturers by reason of what Quality Control Plans are submitted to be clients. At last an evaluation of the quality to be obtained and of the control performed is mentioned. (Author) [pt

Nuclear fuel control in fuel fabrication plants

International Nuclear Information System (INIS)

Seki, Yoshitatsu

1976-01-01

The basic control problems of measuring uranium and of the environment inside and outside nuclear fuel fabrication plants are reviewed, excluding criticality prevention in case of submergence. The occurrence of loss scraps in fabrication and scrap-recycling, the measuring error, the uranium going cut of the system, the confirmation of the presence of lost uranium and the requirement of the measurement control for safeguard make the measurement control very complicated. The establishment of MBA (material balance area) and ICA (item control area) can make clearer the control of inventories, the control of loss scraps and the control of measuring points. Besides the above basic points, the following points are to be taken into account: 1) the method of confirmation of inventories, 2) the introduction of reliable NDT instruments for the rapid check system for enrichment and amount of uranium, 3) the introduction of real time system, and 4) the clarification of MUF analysis and its application to the reliability check of measurement control system. The environment control includes the controls of the uranium concentration in factory atmosphere, the surface contamination, the space dose rate, the uranium concentration in air and water discharged from factories, and the uranium in liquid wastes. The future problems are the practical restudy of measurement control under NPT, the definite plan of burglary protection and the realization of the disposal of solid wastes. (Iwakiri, K.)

Fabrication of Cerium Oxide and Uranium Oxide Microspheres for Space Nuclear Power Applications

Energy Technology Data Exchange (ETDEWEB)

Jeffrey A. Katalenich; Michael R. Hartman; Robert C. O' Brien

2013-02-01

Cerium oxide and uranium oxide microspheres are being produced via an internal gelation sol-gel method to investigate alternative fabrication routes for space nuclear fuels. Depleted uranium and non-radioactive cerium are being utilized as surrogates for plutonium-238 (Pu-238) used in radioisotope thermoelectric generators and for enriched uranium required by nuclear thermal rockets. While current methods used to produce Pu-238 fuels at Los Alamos National Laboratory (LANL) involve the generation of fine powders that pose a respiratory hazard and have a propensity to contaminate glove boxes, the sol-gel route allows for the generation of oxide microsphere fuels through an aqueous route. The sol-gel method does not generate fine powders and may require fewer processing steps than the LANL method with less operator handling. High-quality cerium dioxide microspheres have been fabricated in the desired size range and equipment is being prepared to establish a uranium dioxide microsphere production capability.

Development of a facility for fabricating nuclear waste canisters from radioactively contaminated steel

International Nuclear Information System (INIS)

Logan, J.A.; Larsen, M.M.

1986-01-01

This paper describes design of a facility and processes capable of using radioactively contaminated waste steel as the principal raw material for fabricating stainless steel canisters to be used for disposal of nuclear high-level waste. By such action, expenditure (i.e., permanent loss to society) of thousands of tons of uncontaminated chromium and nickel to fabricate such canisters can be avoided. Moreover, the cost and risks involved in disposing of large accumulations of radioactively contaminated steel as low-level radioactive waste (LLRW), that would otherwise be necessary, can also be avoided. The canister fabrication processes (involving centrifugal casting) described herein have been tested and proven for this application. The performance characteristics of stainless steel canisters so fabricated have been tested and agreed to by the organizations that have been involved in this development work (Battelle Memorial Institute, DuPont, EGandG and the Savannah River Laboratory) as equivalent to the performance characteristics of canisters fabricated of uncontaminated wrought stainless steel. It is estimated that the production cost for fabricating canisters by the methods described will not differ greatly from the production cost using uncontaminated wrought steel, and the other costs avoided by not having to dispose of the contaminated steel as LLRW could cause this method to produce the lowest ultimate overall costs

Measurement of specific heat and specific absorption rate by nuclear magnetic resonance

Energy Technology Data Exchange (ETDEWEB)

Gultekin, David H., E-mail: david.gultekin@aya.yale.edu [Department of Electrical Engineering, Yale University, New Haven, CT 06520 (United States); Department of Medical Physics, Memorial Sloan-Kettering Cancer Center, New York, NY 10065 (United States); Department of Radiology, Memorial Sloan-Kettering Cancer Center, New York, NY 10065 (United States); Institute of Imaging Science, Vanderbilt University, Nashville, TN 37232 (United States); Gore, John C. [Department of Biomedical Engineering, Vanderbilt University, Nashville, TN 37232 (United States); Department of Radiology and Radiological Sciences, Vanderbilt University, Nashville, TN 37232 (United States); Department of Molecular Physiology and Biophysics, Vanderbilt University, Nashville, TN 37232 (United States); Department of Physics and Astronomy, Vanderbilt University, Nashville, TN 37232 (United States); Institute of Imaging Science, Vanderbilt University, Nashville, TN 37232 (United States)

2010-05-20

We evaluate a nuclear magnetic resonance (NMR) method of calorimetry for the measurement of specific heat (c{sub p}) and specific absorption rate (SAR) in liquids. The feasibility of NMR calorimetry is demonstrated by experimental measurements of water, ethylene glycol and glycerol using any of three different NMR parameters (chemical shift, spin-spin relaxation rate and equilibrium nuclear magnetization). The method involves heating the sample using a continuous wave laser beam and measuring the temporal variation of the spatially averaged NMR parameter by non-invasive means. The temporal variation of the spatially averaged NMR parameter as a function of thermal power yields the ratio of the heat capacity to the respective nuclear thermal coefficient, from which the specific heat can be determined for the substance. The specific absorption rate is obtained by subjecting the liquid to heating by two types of radiation, radiofrequency (RF) and near-infrared (NIR), and by measuring the change in the nuclear spin phase shift by a gradient echo imaging sequence. These studies suggest NMR may be a useful tool for measurements of the thermal properties of liquids.

Quality control in nuclear fuel fabrication

International Nuclear Information System (INIS)

Abdelhalim, A.S.; Elsayed, A.A.; Shaaban, H.I.

1988-01-01

The department of metallurgy, NRC Inchass is embarking on a programme of on a laboratory scale, fuel pins containing uranium dioxide pellets are going to be produced. The department is making use of the expertise and equipment at present available and is going to utilize the new fuel pin fabrication unit which would be shortly in operation. The fabrication and testing of uranium dioxide pellets then gradually adapt them and develop, a national know how in this field. This would also involve building up of indigenous experience through proper training of qualified personnel. That are applied to ensure quality of U o 2 pellets, the techniques implemented, the equipment used and the specifications of the equipment presently available. The following parameters are subject to quality control tests: density. O/U ration, hydrogen content, microstructure, each property will be discussed, measurements related to U o 2 powders, including flow ability, bulk density, O/U ratio, bet surface area and water content will be critically discussed. Relevant tests to ensure Q C of pellets are reviewed. These include surface integrity, density, dimensions, microstructure.4 fig., 1 tab

PHWR fuel fabrication with imported uranium - procedures and processes

International Nuclear Information System (INIS)

Rao, R.V.R.L.V.; Rameswara Rao, A.; Hemantha Rao, G.V.S.; Jayaraj, R.N.

2010-01-01

Following the 123 agreement and subsequent agreements with IAEA & NSG, Government of India has entered into bilateral agreements with different countries for nuclear trade. Department of Atomic Energy (DAE), Government of India, has entered into contract with few countries for supply of uranium material for use in the safeguarded PHWRs. Nuclear Fuel Complex (NFC), an industrial unit of DAE, established in the early seventies, is engaged in the production of Nuclear Fuel and Zircaloy items required for Nuclear Power Reactors operating in the country. NFC has placed one of its fuel fabrication facilities (NFC, Block-A, INE-) under safeguards. DAE has opted to procure uranium material in the form of ore concentrate and fuel pellets. Uranium ore concentrate was procured as per the ASTM specifications. Since no international standards are available for PHWR fuel pellets, Specifications have to be finalized based on the present fabrication and operating experience. The process steps have to be modified and fine tuned for handling the imported uranium material especially for ore concentrate. Different transportation methods are to be employed for transportation of uranium material to the facility. Cost of the uranium material imported and the recoveries at various stages of fuel fabrication have impact on the fuel pricing and in turn the unit energy costs. Similarly the operating procedures have to be modified for safeguards inspections by IAEA. NFC has successfully manufactured and supplied fuel bundles for the three 220 MWe safeguarded PHWRs. The paper describes various issues encountered while manufacturing fuel bundles with different types of nuclear material. (author)

Direct characterization of cotton fabrics treated with di-epoxide by nuclear magnetic resonance.

Science.gov (United States)

Xiao, Min; Chéry, Joronia; Keresztes, Ivan; Zax, David B; Frey, Margaret W

2017-10-15

A non-acid-based, di-functional epoxide, neopentyl glycol diglycidyl ether (NPGDGE), was used to modify cotton fabrics. Direct characterization of the modified cotton was conducted by Nuclear Magnetic Resonance (NMR) without grinding the fabric into a fine powder. NaOH and MgBr 2 were compared in catalyzing the reaction between the epoxide groups of NPGDGE and the hydroxyl groups of cellulose. Possible reaction routes were discussed. Scanning electron microscopy (SEM) images showed that while the MgBr 2 -catalyzed reaction resulted in self-polymerization of NPGDGE, the NaOH-catalyzed reaction did not. Fourier transform infrared spectroscopy (FTIR) showed that at high NaOH concentration cellulose restructures from allomorph I to II. NMR studies verified the incorporation of NPGDGE into cotton fabrics with a clear NMR signal, and confirmed that at higher NaOH concentration the efficiency of grafting of NPGDGE was increased. This demonstrates that use of solid state NMR directly on woven fabric samples can simultaneously characterize chemical modification and crystalline polymorph of cotton. No loss of tensile strength was observed for cotton fabrics modified with NPGDGE. Copyright © 2017 Elsevier Ltd. All rights reserved.

Evaluation of methods for seismic analysis of nuclear fuel reprocessing and fabrication facilities

International Nuclear Information System (INIS)

Arthur, D.F.; Dong, R.G.; Murray, R.C.; Nelson, T.A.; Smith, P.D.; Wight, L.H.

1978-01-01

Methods of seismic analysis for critical structures and equipment in nuclear fuel reprocessing plants (NFRPs) and mixed oxide fuel fabrication plants (MOFFPs) are evaluated. The purpose of this series of reports is to provide the NRC with a technical basis for assessing seismic analysis methods and for writing regulatory guides in which methods ensuring the safe design of nuclear fuel cycle facilities are recommended. The present report evaluates methods of analyzing buried pipes and wells, sloshing effects in large pools, earth dams, multiply supported equipment, pile foundations, and soil-structure interactions

Present state and problems of uranium fuel fabrication businesses

International Nuclear Information System (INIS)

Yuki, Akio

1981-01-01

The businesses of uranium fuel fabrication converting uranium hexafluoride to uranium dioxide powder and forming fuel assemblies are the field of most advanced industrialization among nuclear fuel cycle industries in Japan. At present, five plants of four companies engage in this business, and their yearly sales exceeded 20 billion yen. All companies are planning the augmentation of installation capacity to meet the growth of nuclear power generation. The companies of uranium fuel fabrication make the nuclear fuel of the specifications specified by reactor manufacturers as the subcontractors. In addition to initially loaded fuel, the fuel for replacement is required, therefore the demand of uranium fuel is relatively stable. As for the safety of enriched uranium flowing through the farbicating processes, the prevention of inhaling uranium powder by workers and the precaution against criticality are necessary. Also the safeguard measures are imposed so as not to convert enriched uranium to other purposes than peacefull ones. The strict quality control and many times of inspections are carried out to insure the soundness of nuclear fuel. The growth of the business of uranium fuel fabrication and the regulation of the businesses by laws are described. As the problems for the future, the reduction of fabrication cost, the promotion of research and development and others are pointed out. (Kako, I.)

Literature on fabrication of tungsten for application in pyrochemical processing of spent nuclear fuels

International Nuclear Information System (INIS)

Edstrom, C.M.; Phillips, A.G.; Johnson, L.D.; Corle, R.R.

1980-01-01

The pyrochemical processing of nuclear fuels requires crucibles, stirrers, and transfer tubing that will withstand the temperature and the chemical attack from molten salts and metals used in the process. This report summarizes the literature that pertains to fabrication (joining, chemical vapor deposition, plasma spraying, forming, and spinning) is the main theme. This report also summarizes a sampling of literature on molbdenum and the work previously performed at Argonne National Laboratory on other container materials used for pyrochemical processing of spent nuclear fuels

Maintenance of nuclear chemical and fuel fabrication plants [Invited talk no. IT-3

International Nuclear Information System (INIS)

Prasad, A.M.

1981-01-01

Though the objective of the maintenance practices followed in nuclear facilities is to optimise production as in other conventional production plants, the radioactivity associated with nuclear materials is a major constraint in all maintenance jobs on equipment of the nuclear facility. Often non-routine maintenance have to be adopted. Maintenance aspect has to be taken into consideration at the design stage of the nuclear facility. The maintenance concept adopted in a nuclear facility depends on the type of plant and varies from full indirect remote maintenance to direct contact maintenance. This is illustrated by discussing maintenance practices followed in a fuel reprocessing plant, a high level radioactive waste management facility, a fuel fabrication plant, and a heavy water plant. Exposure of maintenance staff to radiation has to be kept within limits governed by safety regulations. Along with planning and scheduling of maintenance, training of manpower with mock-up facilities assumes importance and the maintenance jobs must be carried out under strict supervision. (M.G.B.)

Process for the fabrication of nuclear fuel oxide pellets

International Nuclear Information System (INIS)

Francois, Bernard; Paradis, Yves.

1977-01-01

Process for the fabrication of nuclear fuel oxide pellets of the type for which particles charged with an organic binder -selected from the group that includes polyvinyl alcohol, carboxymethyl cellulose, polyvinyl compounds and methyl cellulose- are prepared from a powder of such an oxide, for instance uranium dioxide. These particles are then compressed into pellets which are then sintered. Under this process the binder charged particles are prepared by stirring the powder with a gas, spraying on to the stirred powder a solution or a suspension in a liquid of this organic binder in order to obtain these particles and then drying the particles so obtained with this gas [fr

Method of fabricating self-powered nuclear radiation detector assemblies

International Nuclear Information System (INIS)

Playfoot, K.; Bauer, R.F.; Sekella, Y.M.

1982-01-01

In a method of fabricating a self-powered nuclear radiation detector assembly an emitter electrode wire and signal cable center wire are connected and disposed within the collector electrode tubular sheath with compressible insulating means disposed between the wires and the tubular sheath. The above assembly is reduced in diameter while elongating the tubular sheath and the emitter wire and signal cable wire. The emitter wire is reduced to a predetermined desired diameter, and is trimmed to a predetermined length. An end cap is hermetically sealed to the tubular sheath at the extending end of the emitter with insulating means between the emitter end and the end cap. (author)

Cold-crucible fabrication of nuclear glasses

International Nuclear Information System (INIS)

Boen, R.

2010-01-01

Vitrification has stood the nuclear industry in good stead, for many years now, as a safe long-term conditioning technology for high-level waste. Major advances are nonetheless still being made, with the development of the cold-crucible technology, affording as it does new possibilities, in terms of volume reduction, and of extending the range of waste products amenable to incorporation. Indeed, by allowing higher melting temperatures to be achieved (1200 - 1400 C degrees), this process opens the way to a considerable increase in glass production capacities, and the fabrication of novel matrices, involving higher incorporation rates than current glasses. In the cold-crucible technology, materials put into the crucible are heated directly through induction. The crucible made of metal is cooled by water circulation. Where the glass comes into contact with the cold wall, a thin layer of solidified glass forms, with a thickness of 5-10 mm preventing the metal forming the crucible from coming into contact with the molten glass. A full scale pilot of the cold crucible was constructed at the La Hague vitrification workshop

Nondestructive assay of special nuclear material for uranium fuel-fabrication facilities

International Nuclear Information System (INIS)

Smith, H.A. Jr.; Schillebeeckx, P.

1997-01-01

A high-quality materials accounting system and effective international inspections in uranium fuel-fabrication facilities depend heavily upon accurate nondestructive assay measurements of the facility's nuclear materials. While item accounting can monitor a large portion of the facility inventory (fuel rods, assemblies, storage items), the contents of all such items and mass values for all bulk materials must be based on quantitative measurements. Weight measurements, combined with destructive analysis of process samples, can provide highly accurate quantitative information on well-characterized and uniform product materials. However, to cover the full range of process materials and to provide timely accountancy data on hard-to-measure items and rapid verification of previous measurements, radiation-based nondestructive assay (NDA) techniques play an important role. NDA for uranium fuel fabrication facilities relies on passive gamma spectroscopy for enrichment and U isotope mass values of medium-to-low-density samples and holdup deposits; it relies on active neutron techniques for U-235 mass values of high-density and heterogeneous samples. This paper will describe the basic radiation-based nondestructive assay techniques used to perform these measurements. The authors will also discuss the NDA measurement applications for international inspections of European fuel-fabrication facilities

Process for the fabrication of a nuclear fuel

International Nuclear Information System (INIS)

Hirose, Yasuo.

1970-01-01

Herein disclosed is a process for fabricating a nuclear fuel incorporating either uranium or plutonium. A pellet-like substrate consisting of a packed powder ceramic fuel such as uranium or plutonium is prepared with the horizontal surface of the body provided with a masking. Next, after impregnating the substrate voids with a solution consisting of a fissile material or mixture of fissile material and poison, the solvent is removed by a chemical deposition process which causes the impregnated material to migrate through capillary action toward the vicinity of the fuel body surface. Sintering and pyrolysis of the deposited material and masking are subsequently carried out to yield a fuel body having adjacent to its surface an intensely concentrated layer of either fissile material or a mixture of fissile material and poison. (Owens, K.J.)

25 years of NDE in fabrication of zirconium alloy mill products and nuclear fuel in the Nuclear Fuel Complex

International Nuclear Information System (INIS)

Mistry, R.K.; Laxminarayana, B.; Srivastava, R.K.

1996-01-01

Failure of nuclear fuel is highly undesirable from both economic and operational aspects. Hence all the components require rigorous QC and inspection checks. NDT plays a major role in assuring the quality of the products both at final and intermediate stages. This paper gives an overall review of NDT methods employed in achieving the integrity of nuclear products. The NDE procedures followed in NFC are visual inspection, radiography, penetrant testing, eddy current testing, ultrasonic testing and helium leak testing. NFC's quality assurance programme is organised to achieve the desired objectives by carrying out in process and final inspection at all critical steps of fabrication. (author)

Hybrid pellets: an improved concept for fabrication of nuclear fuel

International Nuclear Information System (INIS)

Matthews, R.B.; Hart, P.E.

1979-09-01

The feasibility of fabricating fuel pellets using gel-derived microspheres as press feed was evaluated. By using gel-derived microspheres as press feed, the potential exists for eliminating dusty operations like milling, slugging, and granulation, from the pelleting process. The free-flowing character of the spheres should also result in limited dust generation during powder transport and pressing operations. The results of this study clearly demonstrate that fuel pellets can be successfully fabricated on a laboratory scale using UO 2 gel microspheres as press feed. Under moderate sintering conditions, 1,500 0 C for 4 h in Ar-4% H 2 , UO 2 pellets with densities up to 96% TD were fabricated. A range of pellet microstructures and densities were achieved depending on sphere forming and calcining conditions. Based on these results, a set of necessary sphere properties are suggested: O/U less than 2.20, crystallite size less than 500 A, specific surface area greater than 8 m 2 /g, and sphere size 200 and 400 μm. Preliminary attempts to fabricate ThO 2 and ThO 2 -UO 2 pellets using microspheres were unsuccessful because the requisite sphere properties were not achieved. Areas requiring additional development include: demonstration of the process on scaled-up equipment suitable for use in a remote fuel fabrication facility and evaluation of the irradiation performance of pellet fuels from gel-spheres

Pulsed TIG welding in the fabrication of nuclear components and structures

International Nuclear Information System (INIS)

Lucas, W.; Males, B.O.

1979-01-01

TIG welding is an important welding technique in nuclear plant fabrication for the welding of critical components and structures where a high level of weld integrity is demanded. Whilst the process is ideally suited to precision welding, since the arc is a small intense heat source, it has proved to be somewhat intolerant to production variations in 'difficult' applications, such as tube to tube plate welding and orbital tube welding with tube in the fixed position. Whilst the problems directly associated with this intolerance (of the welding process) are less frequently observed when used manually, difficulties are experienced in fully mechanised welding operations particularly when welding to a relatively rigid approved procedure. Pulsing of the welding current was developed as a technique to achieve greater control of the behaviour of the weld pool. Instead of moving the weld pool in a continuous motion around the joint, welding was conducted intermittently in the form of overlapping spots. This technique, which offers significant advantages over continuous current welding has been exploited in nuclear fabrication for welding those components which demand a high level of weld quality. In this paper, the essential features of this technique are described and, in indicating its advantages, examples have been drawn from recent experiences on the welding of two types of joint for the Advanced Gas Cooled Reactor, a tube sheet and a butt joint in the G Position. (author)

Robotic fabrication and inspection for power plants

International Nuclear Information System (INIS)

Date, Ranjit

2002-01-01

The usage of Robotic Automation is now an integral part of the modern manufacturing systems. Applications in nuclear power plants is no exception. As a matter of fact, as a result of the hazards of radiations for the human workers makes automation of the on-site working highly desirable. This presentation will focus on the broad benefits by use of automation in Power plants. Various processes and technologies for robotic applications in fabrication, maintenance and inspection will be highlighted. The specific technology features for use in nuclear environments will be highlighted

INMACS - An approach to on-line nuclear materials accounting and control in a fuel fabrication environment

International Nuclear Information System (INIS)

Yan, G.; L'Archeveque, J.V.R.; Paul, R.N.

1977-08-01

Taking advantage of modern system technologies, the concept of an Integrated Nuclear Materials Accounting and Control System (INMACS) was formulated as an alternative solution to manual inventory procedures. The selected approach offers prospects for tackling the more general fissile materials inventory problem while satisfying the immediate requirements of the Fuel Fabrication Pilot Line at CRNL. A PDP-11/40 minicomputer system was purchased, and a Data Base Management System (DBMS) was designed and implemented to provide a uniform file handling capability. The specific requirements of the Pilot Line were met by a package of application programs. About 16 man-years have been spent on the project. INMACS has been installed in the field and its usefulness as an on-line inventory system will be demonstrated in the Pilot Line. (author)

Does international nuclear trade law have a specificity

International Nuclear Information System (INIS)

David, J.L.

1988-01-01

This study on the specificity of international nuclear trade law covers public international and private international aspects. As regards the first, international organisations and agreements (bilateral and multilateral) are reviewed. In the context of the second, the international organisations with a scientific, legal or commercial vocation are briefly listed. Commercial contracts are then studied in greater detail from the viewpoint of contractual nuclear liability and that outside the contracts. In addition, special aspects are examined, relating to the flexibility of supply contracts, swap agreements in the nuclear field, and other more particular clauses such as the ''Consensus'' framework for export credits. The authors' conclusion is that while there is no specificity properly speaking in international nuclear trade law, it nevertheless has original features (NEA) [fr

Ultrahigh Specific Impulse Nuclear Thermal Propulsion

Energy Technology Data Exchange (ETDEWEB)

Anne Charmeau; Brandon Cunningham; Samim Anghaie

2009-02-09

Research on nuclear thermal propulsion systems (NTP) have been in forefront of the space nuclear power and propulsion due to their design simplicity and their promise for providing very high thrust at reasonably high specific impulse. During NERVA-ROVER program in late 1950's till early 1970's, the United States developed and ground tested about 18 NTP systems without ever deploying them into space. The NERVA-ROVER program included development and testing of NTP systems with very high thrust (~250,000 lbf) and relatively high specific impulse (~850 s). High thrust to weight ratio in NTP systems is an indicator of high acceleration that could be achieved with these systems. The specific impulse in the lowest mass propellant, hydrogen, is a function of square root of absolute temperature in the NTP thrust chamber. Therefor optimizing design performance of NTP systems would require achieving the highest possible hydrogen temperature at reasonably high thrust to weight ratio. High hydrogen exit temperature produces high specific impulse that is a diret measure of propellant usage efficiency.

Development of Automatic Quality Check Software in Mailbox Declaration For Nuclear Fuel Fabrication Plants

International Nuclear Information System (INIS)

Kim, Minsu; Shim, Hye Won; Jo, Seong Yeon; Lee, Kwang Yeol; Ban, Myoung Jin

2014-01-01

Short Notice Random Inspection (SNRI) is a new IAEA safeguards inspection regime for bulk handing facility, which utilities random inspection through a mailbox system. Its main objective is to verify 100% of the flow components of the safeguarded nuclear material at such a facility. To achieve the SNRI objective, it is required to provide daily mailbox declaration, by a facility's operator, to the IAEA with regard to information, such as the receipt and shipment of nuclear materials. Mailbox declarations are then later compared with accounting records so as to examine the accuracy and consistency of the facility operator's declaration at the time of the SNRI. The IAEA has emphasized the importance of accurate mailbox declarations and recommended that the ROK initiate its own independent quality control system in order to improve and maintain its mailbox declarations as a part of the SSAC activities. In an effort to improve the transparency of operational activities at fuel fabrication plants and to satisfy IAEA recommendation, an automatic quality check software application has been developed to improve mailbox declarations at fabrication plants in Korea. The ROK and the IAEA have recognized the importance of providing good quality mailbox declaration for an effective and efficient SNRI at fuel fabrication plants in Korea. The SRA developed an automatic quality check software program in order to provide an independent QC system of mailbox declaration, as well as to improve the quality of mailbox declaration. Once the automatic QC system is implemented, it will improve the quality of an operator's mailbox declaration by examining data before sending it to the IAEA. The QC system will be applied to fuel fabrication plants in the first half of 2014

Nuclear power plants - Quality assurance

International Nuclear Information System (INIS)

1980-01-01

This International Standard defines principles for the establishment and implementation of quality assurance programmes during all phases of design, procurement, fabrication, construction, commissioning, operation, maintenance and decommissioning of structures, systems and components of nuclear power plants. These principles apply to activities affecting the quality of items, such as designing, purchasing, fabricating, handling, shipping, storing, cleaning, erecting, installing, testing, commissioning, operating, inspecting, maintaining, repairing, refuelling and modifying and eventually decommissioning. The manner in which the principles described in this document will be implemented in different organizations involved in a specific nuclear power project will depend on regulatory and contractual requirements, the form of management applied to a nuclear power project, and the nature and scope of the work to be performed by different organizations

Mortality and cancer incidence experience of employees in a nuclear fuels fabrication plant

International Nuclear Information System (INIS)

Hadjimichael, O.C.; Ostfeld, A.M.; D'Atri, D.A.; Brubaker, R.E.

1983-01-01

The mortality and cancer incidence experience of 4,106 employees in a nuclear fuels fabrication plant was evaluated in this retrospective cohort study. Standardized mortality (SMR) and incidence ratios were calculated for groups of employees holding different jobs in the company associated with various types of industrial exposures and with low levels of radiation. Connecticut population mortality rates and Connecticut Tumor Registry incidence rates, specific for age-sex, calendar year and cause of death or cancer site, were used for the calculation of expected rates. Results showed the SMR for all male employees to be significantly lower than expected for all causes and what would be expected for all cancer deaths. More deaths were observed than expected from diseases of the central and peripheral nervous system and from obstructive pulmonary disease. The overall cancer incidence experience of the male employees was significantly lower than expected among the industrial employees. There was no risk associated with any particular job exposure group. Log linear models analysis showed no significant effect from industrial and radiation exposures or from their combined influence

Fabrication development for high-level nuclear waste containers for the tuff repository

International Nuclear Information System (INIS)

Domian, H.A.; Holbrook, R.L.; LaCount, D.F.; Babcock and Wilcox Co., Alliance, OH

1990-09-01

This final report completes Phase 1 of an engineering study of potential manufacturing processes for the fabrication of containers for the long-term storage of nuclear waste. An extensive literature and industry review was conducted to identify and characterize various processes. A technical specification was prepared using the American Society of Mechanical Engineers Boiler ampersand Pressure Vessel Code (ASME BPVC) to develop the requirements. A complex weighting and evaluation system was devised as a preliminary method to assess the processes. The system takes into account the likelihood and severity of each possible failure mechanism in service and the effects of various processes on the microstructural features. It is concluded that an integral, seamless lower unit of the container made by back extrusion has potential performance advantages but is also very high in cost. A welded construction offers lower cost and may be adequate for the application. Recommendations are made for the processes to be further evaluated in the next phase when mock-up trials will be conducted to address key concerns with various processes and materials before selecting a primary manufacturing process. 43 refs., 26 figs., 34 tabs

Design, fabrication and installation of irradiation facilities -Advanced nuclear material development-

International Nuclear Information System (INIS)

Kim, Yong Seong; Lee, Jeong Yeong; Lee, Seong Ho; Ji, Dae Yeong; Kim, Seok Hoon; An, Seong Ho; Kim, Dong Hoon; Seok, Ho Cheon; Kim, Joon Yeon; Yang, Seong Hong

1994-07-01

The objective of this study is to design and construct the steady state fuel test loop and non-instrumented capsules to be installed in KMRR. The principle contents of this project are to design, fabricate the steady-state fuel test loop and non-instrumented capsule to be installed in KMRR for nuclear technology development. This project will be completed in 1996, so preparation of design criteria for fuel test loop have been performed in 1993 as the first year of the first phase in implementing this project. Also design and pressure drop test of non-instrumented capsule have been performed in 1993

Engineering and Fabrication Considerations for Cost-Effective Space Reactor Shield Development

International Nuclear Information System (INIS)

Berg, Thomas A.; Disney, Richard K.

2004-01-01

Investment in developing nuclear power for space missions cannot be made on the basis of a single mission. Current efforts in the design and fabrication of the reactor module, including the reactor shield, must be cost-effective and take into account scalability and fabricability for planned and future missions. Engineering considerations for the shield need to accommodate passive thermal management, varying radiation levels and effects, and structural/mechanical issues. Considering these challenges, design principles and cost drivers specific to the engineering and fabrication of the reactor shield are presented that contribute to lower recurring mission costs

Scanning tunnelling microscope fabrication of phosphorus array in silicon for a nuclear spin quantum computer

International Nuclear Information System (INIS)

O'Brien, J.L.; Schofield, S.R.; Simmons, M.Y.; Clark, R.G.; Dzurak, A.S.; Prawer, S.; Adrienko, I.; Cimino, A.

2000-01-01

Full text: In the vigorous worldwide effort to experimentally build a quantum computer, recent intense interest has focussed on solid state approaches for their promise of scalability. Particular attention has been given to silicon-based proposals that can readily be integrated into conventional computing technology. For example the Kane design uses the well isolated nuclear spin of phosphorous donor nuclei (I=1/2) as the qubits embedded in isotopically pure 28 Si (I=0). We demonstrate the ability to fabricate a precise array of P atoms on a clean Si surface with atomic-scale resolution compatible with the fabrication of the Kane quantum computer

Fabrication and closure development of nuclear waste containers for storage at the Yucca Mountain, Nevada repository

International Nuclear Information System (INIS)

Russell, E.W.; Nelson, T.A.; Domian, H.A.; LaCount, D.F.; Robitz, E.S.; Stein, K.O.

1989-04-01

US Congress and the President have determined that the Yucca Mountain site in Nevada is to be characterized to determine its suitability for construction of the first US high-level nuclear waste repository. Work in connection with this site is carried out within the Yucca Mountain Project (YMP). Lawrence Livermore National Laboratory (LLNL) has the responsibility for designing, developing, and projecting the performance of the waste package for the permanent storage of high-level nuclear waste. Babcock ampersand Wilcox (B ampersand W) is involved with the YMP as a subcontractor to LLNL. B ampersand W's role is to recommend and demonstrate a method for fabricating the metallic waste container and a method for performing the final closure of the container after it has been filled with waste. Various fabrication and closure methods are under consideration for the production of containers. This paper presents progress to date in identifying and evaluating the candidate manufacturing processes. 2 refs., 1 fig., 7 tabs

Thoria-based nuclear fuels thermophysical and thermodynamic properties, fabrication, reprocessing, and waste management

CERN Document Server

Bharadwaj, S R

2013-01-01

This book presents the state of the art on thermophysical and thermochemical properties, fabrication methodologies, irradiation behaviours, fuel reprocessing procedures, and aspects of waste management for oxide fuels in general and for thoria-based fuels in particular. The book covers all the essential features involved in the development of and working with nuclear technology. With the help of key databases, many of which were created by the authors, information is presented in the form of tables, figures, schematic diagrams and flow sheets, and photographs. This information will be useful for scientists and engineers working in the nuclear field, particularly for design and simulation, and for establishing the technology. One special feature is the inclusion of the latest information on thoria-based fuels, especially on the use of thorium in power generation, as it has less proliferation potential for nuclear weapons. Given its natural abundance, thorium offers a future alternative to uranium fuels in nuc...

EDF specifications on nuclear grade resins

International Nuclear Information System (INIS)

Mascarenhas, Darren; Gressier, Frederic; Taunier, Stephane; Le-Calvar, Marc; Ranchoux, Gilles; Marteau, Herve; Labed, Veronique

2012-09-01

Ion exchange resins are widely used across EDF, especially within the nuclear division for the purification of water. Important applications include primary circuit, secondary circuit and effluent treatment, which require high quality nuclear grade resins to retain the dissolved species, some of which may be radioactive. There is a need for more and more efficient purification in order to decrease worker dose during maintenance but also to decrease volumes of radioactive resin waste. Resin performance is subject to several forms of degradation, including physical, chemical, thermal and radioactive, therefore appropriate resin properties have to be selected to reduce such effects. Work has been done with research institutes, manufacturers and on EDF sites to select these properties, create specifications and to continuously improve on these specifications. An interesting example of research regarding resin performance is the resin degradation under irradiation. Resins used in the CVCS circuit of EDF nuclear power plants are subject to irradiation over their lifetime. A study was carried out on the effects of total integrated doses of 0.1, 1 and 10 MGy on typically used EDF mixed bed resins in a 'mini-CVCS' apparatus to simultaneously test actual primary circuit fluid. The tests confirmed that the resins still perform efficiently after a typical CVCS radiation dose. Certain resins also need additional specifications in order to maintain the integrity of the particular circuits they are used in. Recently, EDF has updated its requirements on these high purity nuclear grade resins, produced generic doctrines for all products and materials used on site which include resins of all grades, and as a result have also updated a guide on recommended resin usage for the French fleet of reactors. An overview of the evolutions will be presented. (authors)

Fabrication and closure development of nuclear waste disposal containers for the Yucca Mountain Project: Status report

International Nuclear Information System (INIS)

Domian, H.A.; Robitz, E.S.; Conrardy, C.C.; LaCount, D.F.; McAninch, M.D.; Fish, R.L.; Russell, E.W.

1991-09-01

In GFY 89, a project was underway to determine and demonstrate a suitable method for fabricating thin-walled monolithic waste containers for service within the potential repository at Yucca Mountain. A concurrent project was underway to determine and demonstrate a suitable closure process for these containers after they have been filled with high-level nuclear waste. Phase 1 for both the fabrication and closure projects was a screening phase in which candidate processes were selected for further laboratory testing in Phase 2. This report describes the final results of the Phase 1 efforts. It also describes the preliminary results of Phase 2 efforts

Fabrication and optical characterization of cadmium sulfide needles using nuclear track membrane

International Nuclear Information System (INIS)

Peng, L.Q.; Wang, S.C.; Ju, X.; Xiao, H.; Chen, H.; He, Y.J.

1999-01-01

Cadmium sulfide needles with a diameter of 0.2 μm have been fabricated in nuclear track polyethylene-terephthalate (PET) membrane by electrochemically depositing from organic solvent dimethylsulfoxide (DMSO) containing CdCl 2 and elemental sulfur at the temperature 110 deg. C. The characterization of the sample of CdS needles was studied by scanning electron microscope, X-ray diffraction, absorption and photoluminescence spectra. The optical experiments show that in the sample of CdS needles there is an absorption peak that could be assigned to the interface states of the CdS needles

Fabrication and optical characterization of cadmium sulfide needles using nuclear track membrane

Energy Technology Data Exchange (ETDEWEB)

Peng, L.Q.; Wang, S.C.; Ju, X.; Xiao, H.; Chen, H.; He, Y.J

1999-06-01

Cadmium sulfide needles with a diameter of 0.2 {mu}m have been fabricated in nuclear track polyethylene-terephthalate (PET) membrane by electrochemically depositing from organic solvent dimethylsulfoxide (DMSO) containing CdCl{sub 2} and elemental sulfur at the temperature 110 deg. C. The characterization of the sample of CdS needles was studied by scanning electron microscope, X-ray diffraction, absorption and photoluminescence spectra. The optical experiments show that in the sample of CdS needles there is an absorption peak that could be assigned to the interface states of the CdS needles.

Characterisation and fabrication of zirconia and thoria based ceramics for nuclear applications

International Nuclear Information System (INIS)

Barrier, D.C.

2005-11-01

The reduction of the long term radiotoxicity of nuclear waste during disposal is the aim of the research called ''Partitioning and Transmutation of Minor actinides (MAs)'', which also requires the development of inert ceramic support materials. Moreover, after separation, if the transmutation is not available, the actinides can be conditioned into stable dedicated solid matrices (Partitioning and Conditioning strategy). Yttrium-stabilized zirconia and thoria are discussed in the international nuclear community as candidates for the fixation of long-lived actinides as target material for transmutation and as stable materials for long-term final disposal. The aims of the following work are twofold: determine the impact of the addition of actinides, simulated by cerium on the properties of the matrices and study the possibility of synthesising homogeneous ceramics using simple fabrication routes. Within this framework, (ZrY)O 2-x -CeO 2 and ThO 2 -CeO 2 powders with variable ceria contents (from 0 to 100 %) were synthesised by a co-precipitation method of nitrate solution. The influence of ceria concentration on the powder' properties, such as thermal behaviour and the evolution of material crystallisation during annealing, was investigated in detail by thermogravimetry (TG) coupled with differential scanning calorimetry (DSC) and X-ray diffraction (XRD). Both systems crystallise at high temperature in a stable solid solution, fcc, fluorite type structure and follow the Vegard's law for the complete range of ceria. For both systems a critical concentration of 20 mol% has been established. For ceria concentrations lower than 20%, the properties of the system are mainly controlled by the matrix. Pellets with different ceria concentrations were compacted from these powders by using different technological cycles. In order to obtain materials with reliable properties, the technological parameters of each chosen fabrication route, have been optimised. By employing mild wet

Characterisation and fabrication of zirconia and thoria based ceramics for nuclear applications

Energy Technology Data Exchange (ETDEWEB)

Barrier, D C

2005-11-01

The reduction of the long term radiotoxicity of nuclear waste during disposal is the aim of the research called ''Partitioning and Transmutation of Minor actinides (MAs)'', which also requires the development of inert ceramic support materials. Moreover, after separation, if the transmutation is not available, the actinides can be conditioned into stable dedicated solid matrices (Partitioning and Conditioning strategy). Yttrium-stabilized zirconia and thoria are discussed in the international nuclear community as candidates for the fixation of long-lived actinides as target material for transmutation and as stable materials for long-term final disposal. The aims of the following work are twofold: determine the impact of the addition of actinides, simulated by cerium on the properties of the matrices and study the possibility of synthesising homogeneous ceramics using simple fabrication routes. Within this framework, (ZrY)O{sub 2-x}-CeO{sub 2} and ThO{sub 2}-CeO{sub 2} powders with variable ceria contents (from 0 to 100 %) were synthesised by a co-precipitation method of nitrate solution. The influence of ceria concentration on the powder' properties, such as thermal behaviour and the evolution of material crystallisation during annealing, was investigated in detail by thermogravimetry (TG) coupled with differential scanning calorimetry (DSC) and X-ray diffraction (XRD). Both systems crystallise at high temperature in a stable solid solution, fcc, fluorite type structure and follow the Vegard's law for the complete range of ceria. For both systems a critical concentration of 20 mol% has been established. For ceria concentrations lower than 20%, the properties of the system are mainly controlled by the matrix. Pellets with different ceria concentrations were compacted from these powders by using different technological cycles. In order to obtain materials with reliable properties, the technological parameters of each chosen fabrication route, have been optimised. By

Characterisation and fabrication of zirconia and thoria based ceramics for nuclear applications

Energy Technology Data Exchange (ETDEWEB)

Barrier, D.C.

2005-11-01

The reduction of the long term radiotoxicity of nuclear waste during disposal is the aim of the research called ''Partitioning and Transmutation of Minor actinides (MAs)'', which also requires the development of inert ceramic support materials. Moreover, after separation, if the transmutation is not available, the actinides can be conditioned into stable dedicated solid matrices (Partitioning and Conditioning strategy). Yttrium-stabilized zirconia and thoria are discussed in the international nuclear community as candidates for the fixation of long-lived actinides as target material for transmutation and as stable materials for long-term final disposal. The aims of the following work are twofold: determine the impact of the addition of actinides, simulated by cerium on the properties of the matrices and study the possibility of synthesising homogeneous ceramics using simple fabrication routes. Within this framework, (ZrY)O{sub 2-x}-CeO{sub 2} and ThO{sub 2}-CeO{sub 2} powders with variable ceria contents (from 0 to 100 %) were synthesised by a co-precipitation method of nitrate solution. The influence of ceria concentration on the powder' properties, such as thermal behaviour and the evolution of material crystallisation during annealing, was investigated in detail by thermogravimetry (TG) coupled with differential scanning calorimetry (DSC) and X-ray diffraction (XRD). Both systems crystallise at high temperature in a stable solid solution, fcc, fluorite type structure and follow the Vegard's law for the complete range of ceria. For both systems a critical concentration of 20 mol% has been established. For ceria concentrations lower than 20%, the properties of the system are mainly controlled by the matrix. Pellets with different ceria concentrations were compacted from these powders by using different technological cycles. In order to obtain materials with reliable properties, the technological parameters of each chosen fabrication

Probabilistic safety analysis for nuclear fuel cycle facilities, an exemplary application for a fuel fabrication plant

International Nuclear Information System (INIS)

Gmal, B.; Gaenssmantel, G.; Mayer, G.; Moser, E.F.

2013-01-01

In order to assess the risk of complex technical systems, the application of the Probabilistic Safety Assessment (PSA) in addition to the Deterministic Safety Analysis becomes of increasing interest. Besides nuclear installations this applies to e. g. chemical plants. A PSA is capable of expanding the basis for the risk assessment and of complementing the conventional deterministic analysis, by which means the existing safety standards of that facility can be improved if necessary. In the available paper, the differences between a PSA for a nuclear power plant and a nuclear fuel cycle facility (NFCF) are discussed in shortness and a basic concept for a PSA for a nuclear fuel cycle facility is described. Furthermore, an exemplary PSA for a partial process in a fuel assembly fabrication facility is described. The underlying data are partially taken from an older German facility, other parts are generic. Moreover, a selected set of reported events corresponding to this partial process is taken as auxiliary data. The investigation of this partial process from the fuel fabrication as an example application shows that PSA methods are in principle applicable to nuclear fuel cycle facilities. Here, the focus is on preventing an initiating event, so that the system analysis is directed to the modeling of fault trees for initiating events. The quantitative results of this exemplary study are given as point values for the average occurrence frequencies. They include large uncertainties because of the limited documentation and data basis available, and thus have only methodological character. While quantitative results are given, further detailed information on process components and process flow is strongly required for robust conclusions with respect to the real process. (authors)

Fabrication and closure development of corrosion resistant containers for Nevada's Yucca Mountain high-level nuclear waste repository

International Nuclear Information System (INIS)

Russell, E.W.; Nelson, T.A.; Domian, H.A.; LaCount, D.F.; Robitz, E.S.; Stein, K.O.

1989-11-01

US Congress and the President have determined that the Yucca Mountain site in Nevada is to be characterized to determine its suitability for construction of the first US high-level nuclear waste repository. Work in connection with this site is carried out within the Yucca Mountain Project (YMP). Lawrence Livermore National Laboratory (LLNL) has the responsibility for designing, developing, and projecting the performance of the waste package for the permanent storage of high-level nuclear waste. Babcock ampersand Wilcox (B ampersand W) is involved with the YMP as a subcontractor to LLNL. B ampersand W's role is to recommend and demonstrate a method for fabricating the metallic waste container and a method for performing the final closure of the container after it has been filled with waste. Various fabrication and closure methods are under consideration for the production of containers. This paper presents progress to date in identifying and evaluating the candidate manufacturing processes. 2 refs., 2 figs., 4 tabs

The second answers and questions on the licence of the fabrication project for the nuclear fuel of research reactors

International Nuclear Information System (INIS)

Park, Hee Dae; Kim, C. K.; Kim, K. H.

2002-07-01

KINS has examined the application for licensing of research reactor fuel fabrication for seven months, from May to Dec. 2000. The most hot issues during examination, in order to understand whether the design and facilities are fitted to the regulation criteria or not, were the availability of basic ground, design criteria on safety, availability and methodology of design, seismic criteria, availability of nuclear fuel fabrication, safety related criticality, safety related the process, availability of nuclear waste management, validity of organization and procedure for radioactivity management, and the validity of both selection and analysis about predicted accident. Moreover, another issues such as the radioactivity inspection plan for waste treatment, effect on both radioactive material and accidant, method of decrease of damage on environment, and environmental inspection plan of radioactivity, were severely examined

A proposal of ITER vacuum vessel fabrication specification and results of the full-scale partial mock-up test

International Nuclear Information System (INIS)

Nakahira, M.; Takeda, N.; Kakudate, S.; Onozuka, M.

2008-01-01

The structure and fabrication methods of the ITER vacuum vessel (VV) have been investigated and defined by the ITER International Team (IT). However, some of the current technical specifications are difficult to be achieved from the manufacturing point of view. To solve such an issue, this paper proposes an alternative specification of the VV to the IT's design. A series of the fabrication and assembly procedures for the mock-up are presented, together with candidates of welding configurations. Finally, the paper summarizes the results of mock-up fabrication, such as non-destructive examination of weld lines, obtained welding deformation and issues revealed from the fabrication experience. Based on the results, it is suggested that several issues such as clarification of conditions of repair welding, demonstration of welding distortion control and detectability/localization of internal defects should be solved before manufacturing the ITER VV

Physical Properties of Niobium and Specifications for Fabrication of Superconducting Cavities

International Nuclear Information System (INIS)

Antoine, C.; Foley, M.; Dhanaraj, N.

2011-01-01

It is important to distinguish among the properties of niobium, the ones that are related to the cavity's SRF performances, the formability of the material, and the mechanical behavior of the formed cavity. In general, the properties that dictate each of the above mentioned characteristics have a detrimental effect on one another and in order to preserve the superconducting properties without subduing the mechanical behavior, a balance has to be established. Depending on the applications, some parameters become less important and an understanding of the physical origin of the requirements might help in this optimization. SRF applications require high purity niobium (high RRR), but pure niobium is very soft from fabrication viewpoint. Moreover conventional fabrication techniques tend to override the effects of any metallurgical process meant to strengthen it. As those treatments dramatically affect the forming of the material they should be avoided. These unfavorable mechanical properties have to be accounted for in the design of the cavities rather than in the material specification. The aim of this paper is to review the significance of the important mechanical properties used to characterize niobium and to present the optimal range of values. Most of the following information deals with the specification of sheets for cell forming unless otherwise noted.

Physical Properties of Niobium and Specifications for Fabrication of Superconducting Cavities

Energy Technology Data Exchange (ETDEWEB)

Antoine, C.; Foley, M.; Dhanaraj, N.; /Fermilab

2011-07-01

It is important to distinguish among the properties of niobium, the ones that are related to the cavity's SRF performances, the formability of the material, and the mechanical behavior of the formed cavity. In general, the properties that dictate each of the above mentioned characteristics have a detrimental effect on one another and in order to preserve the superconducting properties without subduing the mechanical behavior, a balance has to be established. Depending on the applications, some parameters become less important and an understanding of the physical origin of the requirements might help in this optimization. SRF applications require high purity niobium (high RRR), but pure niobium is very soft from fabrication viewpoint. Moreover conventional fabrication techniques tend to override the effects of any metallurgical process meant to strengthen it. As those treatments dramatically affect the forming of the material they should be avoided. These unfavorable mechanical properties have to be accounted for in the design of the cavities rather than in the material specification. The aim of this paper is to review the significance of the important mechanical properties used to characterize niobium and to present the optimal range of values. Most of the following information deals with the specification of sheets for cell forming unless otherwise noted.

CVD diamond for nuclear detection applications

International Nuclear Information System (INIS)

Bergonzo, P.; Brambilla, A.; Tromson, D.; Mer, C.; Guizard, B.; Marshall, R.D.; Foulon, F.

2002-01-01

Chemically vapour deposited (CVD) diamond is a remarkable material for the fabrication of radiation detectors. In fact, there exist several applications where other standard semiconductor detectors do not fulfil the specific requirements imposed by corrosive, hot and/or high radiation dose environments. The improvement of the electronic properties of CVD diamond has been under intensive investigations and led to the development of a few applications that are addressing specific industrial needs. Here, we report on CVD diamond-based detector developments and we describe how this material, even though of a polycrystalline nature, is readily of great interest for applications in the nuclear industry as well as for physics experiments. Improvements in the material synthesis as well as on device fabrication especially concern the synthesis of films that do not exhibit space charge build up effects which are often encountered in CVD diamond materials and that are highly detrimental for detection devices. On a pre-industrial basis, CVD diamond detectors have been fabricated for nuclear industry applications in hostile environments. Such devices can operate in harsh environments and overcome limitations encountered with the standard semiconductor materials. Of these, this paper presents devices for the monitoring of the alpha activity in corrosive nuclear waste solutions, such as those encountered in nuclear fuel assembly reprocessing facilities, as well as diamond-based thermal neutron detectors exhibiting a high neutron to gamma selectivity. All these demonstrate the effectiveness of a demanding industrial need that relies on the remarkable resilience of CVD diamond

Recycling of nuclear fuel swarf at the fabrication of UO sub(2)-pellets and its influence on the irradiation behavior

International Nuclear Information System (INIS)

Dias, M.S.; Lameiras, F.S.; Santos, A.M.M. dos

1991-01-01

From the fabrication of UO sub(2) pellets for light water reactor fuel rods, nuclear fuel scraps results in form of UO sub(2) grinding swarf and UO sub(2) sinter scraps oxidized to U sub(3)O sub(8) powder. Detailed investigations on five types of UO sub(2) pellets fabricated with different portions of this scrap kinds added to the UO sub(2) press powder showed that there is only a small influence of such scrap additions on the irradiation behavior, especially for the fission gas release. This allows to recycle the fabrication scrap in a simple and economic way. (author)

A high-temperature, short-duration method of fabricating surrogate fuel microkernels for carbide-based TRISO nuclear fuels

International Nuclear Information System (INIS)

Vasudevamurthy, G.; Radecka, A.; Massey, C.

2015-01-01

High-temperature gas-cooled reactor technology is a frontrunner among generation IV nuclear reactor designs. Among the advanced nuclear fuel forms proposed for these reactors, dispersion-type fuel consisting of microencapsulated uranium di-oxide kernels, popularly known as tri-structural isotropic (TRISO) fuel, has emerged as the fuel form of choice. Generation IV gas-cooled fast reactors offer the benefit of recycling nuclear waste with increased burn-ups in addition to producing the required power and hydrogen. Uranium carbide has shown great potential to replace uranium di-oxide for use in these fast spectrum reactors. Uranium carbide microkernels for fast reactor TRISO fuel have traditionally been fabricated by long-duration carbothermic reduction and sintering of precursor uranium dioxide microkernels produced using sol-gel techniques. These long-duration conversion processes are often plagued by issues such as final product purity and process parameters that are detrimental to minor actinide retention. In this context a relatively simple, high-temperature but relatively quick-rotating electrode arc melting method to fabricate microkernels directly from a feedstock electrode was investigated. The process was demonstrated using surrogate tungsten carbide on account of its easy availability, accessibility and the similarity of its melting point relative to uranium carbide and uranium di-oxide.

A high-temperature, short-duration method of fabricating surrogate fuel microkernels for carbide-based TRISO nuclear fuels

Energy Technology Data Exchange (ETDEWEB)

Vasudevamurthy, G.; Radecka, A.; Massey, C. [Virginia Commonwealth Univ., Richmond, VA (United States). High Temperature Materials Lab.

2015-07-01

High-temperature gas-cooled reactor technology is a frontrunner among generation IV nuclear reactor designs. Among the advanced nuclear fuel forms proposed for these reactors, dispersion-type fuel consisting of microencapsulated uranium di-oxide kernels, popularly known as tri-structural isotropic (TRISO) fuel, has emerged as the fuel form of choice. Generation IV gas-cooled fast reactors offer the benefit of recycling nuclear waste with increased burn-ups in addition to producing the required power and hydrogen. Uranium carbide has shown great potential to replace uranium di-oxide for use in these fast spectrum reactors. Uranium carbide microkernels for fast reactor TRISO fuel have traditionally been fabricated by long-duration carbothermic reduction and sintering of precursor uranium dioxide microkernels produced using sol-gel techniques. These long-duration conversion processes are often plagued by issues such as final product purity and process parameters that are detrimental to minor actinide retention. In this context a relatively simple, high-temperature but relatively quick-rotating electrode arc melting method to fabricate microkernels directly from a feedstock electrode was investigated. The process was demonstrated using surrogate tungsten carbide on account of its easy availability, accessibility and the similarity of its melting point relative to uranium carbide and uranium di-oxide.

A proposal of ITER vacuum vessel fabrication specification and results of the full-scale partial mock-up test

Energy Technology Data Exchange (ETDEWEB)

Nakahira, Masataka; Takeda, Nobukazu; Onozuka, Masanori [Japan Atomic Energy Agency (Japan); Kakudate, Satoshi [Mitsubishi Heavy Industries, Ltd. (Japan)

2007-07-01

The structure and fabrication methods of the ITER vacuum vessel have been investigated and defined by the ITER international team. However, some of the current specifications are very difficult to be achieved from the manufacturing point of view and will lead to cost increase. In the mock-up fabrication, it is planned to conduct the following items: 1. Feasibility of the Japanese proposed VV structure and fabrication methods and the applicability to the ITER are to be confirmed; 2. Assembly procedure and inspection procedure are to be confirmed; 3. Manufacturing tolerances are to be assessed; 4. Manufacturing schedule is to be assessed. This report summarizes the Japanese proposed specification of the VV mock-up describing differences between the ITER supplied design. General scope of the mock-up fabrication and the detailed dimensions are also shown. In the VV fabrication, several types of weld joint configuration will be used. This report shows the joint configurations proposed by Japan to be used for the inner shell connection, the rib-to-shell connection and outer shell connection, and the housing-to-shell connection, respectively. Non-destructive testing considered to be applied to each joint configuration is also presented. A series of the fabrication and assembly procedures for the mock-up are presented in this report, together with candidates of welding configurations. Finally, the report summarizes the results of mock-up fabrication, including results of nondestructive examination of weld lines, obtained welding deformation and issues revealed from the fabrication experience. (orig.)

Interfacing robotics with plutonium fuel fabrication

International Nuclear Information System (INIS)

Bowen, W.W.; Moore, F.W.

1986-01-01

Interfacing robotic systems with nuclear fuel fabrication processes resulted in a number of interfacing challenges. The system not only interfaces with the fuel process, but must also interface with nuclear containment, radiation control boundaries, criticality control restrictions, and numerous other safety systems required in a fuel fabrication plant. The robotic system must be designed to allow operator interface during maintenance and recovery from an upset as well as normal operations

A proposal of ITER vacuum vessel fabrication specification and results of the full-scale partial mock-up test

Energy Technology Data Exchange (ETDEWEB)

Nakahira, M. [Japan Atomic Energy Agency, Mukouyama 801-1, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan)], E-mail: nakahira.masataka@jaea.go.jp; Takeda, N.; Kakudate, S. [Japan Atomic Energy Agency, Mukouyama 801-1, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan); Onozuka, M. [Mitsubishi Nuclear Energy Systems, Inc., 1700K Street NW, Suite 440, Washington, DC 20006 (United States)

2008-12-15

The structure and fabrication methods of the ITER vacuum vessel (VV) have been investigated and defined by the ITER International Team (IT). However, some of the current technical specifications are difficult to be achieved from the manufacturing point of view. To solve such an issue, this paper proposes an alternative specification of the VV to the IT's design. A series of the fabrication and assembly procedures for the mock-up are presented, together with candidates of welding configurations. Finally, the paper summarizes the results of mock-up fabrication, such as non-destructive examination of weld lines, obtained welding deformation and issues revealed from the fabrication experience. Based on the results, it is suggested that several issues such as clarification of conditions of repair welding, demonstration of welding distortion control and detectability/localization of internal defects should be solved before manufacturing the ITER VV.

Spent nuclear fuel project product specification

International Nuclear Information System (INIS)

PAJUNEN, A.L.

1999-01-01

This document establishes the limits and controls for the significant parameters that could potentially affect the safety and/or quality of the Spent Nuclear Fuel (SNF) packaged for processing, transport, and storage. The product specifications in this document cover the SNF packaged in Multi-Canister Overpacks to be transported throughout the SNF Project

Siemens technology transfer and cooperation in the nuclear fuel area

International Nuclear Information System (INIS)

Holley, H.-P.; Fuchs, J. H.; Rothenbuecher, R. A.

1997-01-01

Siemens is a full-range supplier in the area of nuclear power generation with broad experience and activities in the field of nuclear fuel. Siemens has developed advanced fuel technology for all types fuel assemblies used throughout the world and has significant experience worldwide in technology transfer in the field of nuclear fuel. Technology transfer and cooperation has ranged between the provision of mechanical design advice for a specific fuel design and the erection of complete fabrication plants for commercial operation in 3 countries. In the following the wide range of Siemens' technology transfer activities for both fuel design and fuel fabrication technologies are shown

Support of the radioactive waste treatment nuclear fuel fabrication facility

International Nuclear Information System (INIS)

Park, H.H.; Han, K.W.; Lee, B.J.; Shim, G.S.; Chung, M.S.

1982-01-01

Technical service of radioactive waste treatment in Daeduck Engineering Center includes; 1) Treatment of radioactive wastes from the nuclear fuel fabrication facility and from laboratories. 2) Establishing a process for intermediate treatment necessary till the time when RWTF is in completion. 3) Technical evaluation of unit processes and equipments concerning RWTF. About 35 drums (8 m 3 ) of solid wastes were treated and stored while more than 130 m 3 of liquid wastes were disposed or stored. A process with evaporators of 10 1/hr in capacity, a four-stage solvent washer, storage tanks and disposal system was designed and some of the equipments were manufactured. Concerning RWTF, its process was reviewed technically and emphasis were made on stability of the bituminization process against explosion, function of PAAC pump, decontamination, and finally on problems to be solved in the comming years. (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.)

An evaluation of UO2-CNT composites made by SPS as an accident tolerant nuclear fuel pellet and the feasibility of SPS as an economical fabrication process for the nuclear fuel cycle

Science.gov (United States)

Cartas, Andrew R.

The innovative and advanced purpose of this study is to understand and establish proper sintering procedures for Spark Plasma Sintering process in order to fabricate high density, high thermal conductivity UO2 -CNT pellets. Mixing quality and chemical reactions have been investigated by field emission scanning electron microscopy (FESEM), wavelength dispersive spectroscopy (WDS), and X-ray diffraction (XRD). The effect of various types of CNTs on the mixing and sintering quality of UO2-CNT pellets with SPS processing have been examined. The Archimedes Immersion Method, laser flash method, and FE-SEM will be used to investigate the density, thermal conductivity, grain size, pinning effects, and CNT dispersion of fabricated UO2-CNT pellets. Pre-fabricated CNT's were added to UO 2 powder and dispersed via sonication and/or ball milling and then made into composite nuclear pellets. An investigation of the economic impact of SPS on the nuclear fuel cycle for producing pure and composite UO2 fuels was conducted.

Nuclear electronic equipment for control and monitoring panel. Specifications and methods for testing radiation detectors

International Nuclear Information System (INIS)

Roguin, Andre.

1976-02-01

This document will be of interest to users and makers of neutron and gamma radiations detectors in the field of nuclear reactor control and protection. Information is given which will enable users to optimize their choice and methods of using equipment, and makers to optimize their methods of fabrication. It should also serve as a model from which official specifications, technical instructions and test methods for these detectors, could be established. A detailed description is given of the appropriate parameters, terminology and notations. General specifications, operating conditions and test methods are indicated. The following detectors are studied: in-core detectors: fission ionization chambers, self powered neutron detectors (S.P.N.D.); out-core detectors: boron ionization chambers (for monitoring), boron trifluoride proportional counter tubes, boron lined proportional counter tubes, helium-3 proportional counter tubes. The devices described in the document are intended for industrial radiation monitoring applications and not for calibration standards (dosimetry) or for health physics measurement purposes. They are characterized by their fidelity, fast response, reliability and long lifetimes [fr

Romanian regulatory requirements on nuclear field specific education needs

International Nuclear Information System (INIS)

Biro, L.; Velicu, O.

2004-01-01

This work is intended as a general presentation of the educational system and research field, with reference to nuclear sciences, and the legal system, with reference to requirements established by the regulatory body for the professional qualification and periodic training of personnel involved in different activities in the nuclear field. Thus, part 2 and 3 of the work present only public information regarding the education in nuclear sciences and nuclear research in Romania; in part 4 the CNCAN requirements for the personnel training, specific to nuclear activities are slightly detailed; part 5 consists of few words about the public information activities in Romania; and part 6 tries to draw a conclusion. (authors)

Nuclear fuel fabrication - developing indigenous capability

International Nuclear Information System (INIS)

Gupta, U.C.; Jayaraj, R.N.; Meena, R.; Sastry, V.S.; Radhakrishna, C.; Rao, S.M.; Sinha, K.K.

1997-01-01

Nuclear Fuel Complex (NFC), established in early 70's for production of fuel for PHWRs and BWRs in India, has made several improvements in different areas of fuel manufacturing. Starting with wire-wrap type of fuel bundles, NFC had switched over to split spacer type fuel bundle production in mid 80's. On the upstream side slurry extraction was introduced to prepare the pure uranyl nitrate solution directly from the MDU cake. Applying a thin layer of graphite to the inside of the tube was another modification. The Complex has developed cost effective and innovative techniques for these processes, especially for resistance welding of appendages on the fuel elements which has been a unique feature of the Indian PHWR fuel assemblies. Initially, the fuel fabrication plants were set-up with imported process equipment for most of the pelletisation and assembly operations. Gradually with design and development of indigenous equipment both for production and quality control, NFC has demonstrated total self reliance in fuel production by getting these special purpose machines manufactured indigenously. With the expertise gained in different areas of process development and equipment manufacturing, today NFC is in a position to offer know-how and process equipment at very attractive prices. The paper discusses some of the new processes that are developed/introduced in this field and describes different features of a few PLC based automatic equipment developed. Salient features of innovative techniques being adopted in the area Of UO 2 powder production are also briefly indicated. (author)

Induction plasma deposition technology for nuclear fuel fabrication

International Nuclear Information System (INIS)

Jung, I. H.; Bae, K. K.; Lee, J. W.; Kim, T. K.; Yang, M. S.

1998-01-01

A study on induction plasma deposition with ceramic materials, yttria-stabilized-zirconia ZrO 2 -Y 2 O 3 (m.p. 2640 degree C), was conducted with a view of developing a new method for nuclear fuel fabrication. Before making dense pellets of more than 96%T.D., the spraying condition was optimized through the process parameters, such as chamber pressure, plasma plate power, powder spraying distance, sheath gas composition, probe position, particle size and powders of different morphology. The results with a 5mm thick deposit on rectangular planar graphite substrates showed a 97.11% theoretical density when the sheath gas flow rate was Ar/H 2 120/20 l/min, probe position 8cm, particle size -75 μm and spraying distance 22cm by AMDRY146 powder. The degree of influence of the main effects on density were powder morphology, particle size, sheath gas composition, plate power and spraying distance, in that order. Among the two parameter interactions, the sheath gas composition and chamber pressure affects density greatly. By using the multi-pellets mold of wheel type, the pellet density did not exceed 94%T.D., owing to the spraying angle

Experience in developing countries in monitoring procurement and fabrication

International Nuclear Information System (INIS)

Csik, B.J.

1977-01-01

Owner's responsibility in monitoring procurement and fabrication. Monitoring ectivity, tasks, knowledge and personnel requirements, scope and organization. Contractual arrangements, commitments, responsibilities, rights and obligations. Domestic and foreign supplies. Staff and consultants. Experience in developing countries. Problem areas: availability of qualified staff, organization, methodology standards, codes, specifications, availability and flow of information, language, technical knowledge, access to suppliers' facilities, delays, nuclear safety related components, modifications and additionals. (orig.) [de

Site-Specific Atmospheric Dispersion Characteristics of Korean Nuclear Power Plant Sites

International Nuclear Information System (INIS)

Han, M. H.; Kim, E. H.; Suh, K. S.; Hwang, W. T.; Choi, Y. G.

2001-01-01

Site-specific atmospheric dispersion characteristics have been analyzed. The northwest and the southwest wind prevail on nuclear sites of Korea. The annual isobaric surface averaged for twenty years around Korean peninsula shows that west wind prevails. The prevailing west wind is profitable in the viewpoint of radiation protection because three of four nuclear sites are located in the east side. Large scale field tracer experiments over nuclear sites have been conducted for the purpose of analyzing the atmospheric dispersion characteristics and validating a real-time atmospheric dispersion and dose assessment system FADAS. To analyze the site-specific atmospheric dispersion characteristics is essential for making effective countermeasures against a nuclear emergency

Stainless steel fabrications: past and present

International Nuclear Information System (INIS)

Daniels, R.

1986-01-01

The paper deals with stainless steel fabrications of Fairey Engineering Company for the nuclear industry. The manufacture of stainless steel containers for Magnox and Advanced Gas Cooled Reactors, flexible fabrication facility, and welding development, are all briefly described. (U.K.)

Technical specification of the NRPB Nuclear Emulsion Dosemeter

International Nuclear Information System (INIS)

Bartlett, D.T.; Bird, T.V.

1978-08-01

This document is a formal specification of the NRPB Nuclear Emulsion Dosemeter. The dosemeter specified in this report replaces the NRPB Fast Neutron Personal Dosemeter specified in NRPB-R50. (author)

Nuclear utility education and training becoming too plant specific?

International Nuclear Information System (INIS)

Wicks, F.

1986-01-01

As the Supervisor of a university nuclear reactor and operations curriculum, the author has also been offering education and training programs for nuclear utility technical support and operations personnel. Similar results have been reported by other universities offering similar programs. These programs also provide very important benefits to university nuclear engineering departments in terms of much needed revenues during this time of declining student enrollment and also by the information flow from the nuclear utility participants to the university personnel, which can yield both improved courses and identify research opportunities. University programs serve an important complementary function to plant-specific programs and should be continued and supported

Fabrication of nuclear fuel by powder injection moulding: Study of the binders systems and the de-binding of feedstock containing actinide powder

International Nuclear Information System (INIS)

Bricout, J.

2012-01-01

Powder Injection Moulding (PIM) is identified as an innovative process for the nuclear fuel fabrication. Technological breakthrough compared to the current process of powder metallurgy, the impact of actinide powder's specificities on the different steps of PIM is performed. Alumina powders simulating actinide powder have been implemented with a reference binders system. Thermal and rheological studies show the injectability and the de-binding of feedstocks with adequate solid loading (≥50 %vol), thanks to the de-agglomeration during the mixing step, which allow to obtain net shape fuel pellet. Specific surface area of powders, acting as a key role in behaviour's feedstocks, has been integrated in analysis models of viscosity prediction according to the shear rate. Also conducted studies on uranium oxide powder show that the selected binders systems, which have a compatible rheological behaviour with PIM process, impact the de-agglomeration of powder and final microstructure of the fuel pellet, consistent with the results obtained on alumina powders. Independent behaviour of binders and uranium oxide powder, showing no adverse chemical reaction against the PIM process, show a residual mass of carbon of about 150 ppm after sintering. Binders system using polystyrene, resistant to radiolysis phenomena and loadable more than 50 %(vol) of actinide powder, shows the promising potential of PIM process for the fuel fabrication. (author) [fr

Nuclear power reactor technology

International Nuclear Information System (INIS)

1978-09-01

Risoe National Laboratory was established more than twenty years ago with research and development of nuclear reactor technology as its main objective. The Laboratory has by now accumulated many years of experience in a number of areas vital to nuclear reactor technology. The work and experience of, and services offered by the Laboratory within the following fields are described: Health physics site supervision; Treatment of low and medium level radioactive waste; Core performance evaluation; Transient analysis; Accident analysis; Fuel management; Fuel element design, fabrication and performance evaluation; Non-destructive testing of nuclear fuel; Theoretical and experimental structural analysis; Reliability analysis; Site evaluation. Environmental risk and hazard calculation; Review and analysis of safety documentation. Risoe has already given much assistance to the authorities, utilities and industries in such fields, carrying out work on both light and heavy water reactors. The Laboratory now offers its services to others as a consultant, in education and training of staff, in planning, in qualitative and quantitative analysis, and for the development and specification of fabrication techniques. (author)

Technical specifications review of nuclear power plants: a risk-informed evaluation

International Nuclear Information System (INIS)

Saldanha, Pedro Luiz da Cruz; Sousa, Anna Leticia; Frutuoso e Melo, Paulo Fernando Ferreira; Duarte, Juliana Pacheco

2012-01-01

The use of risk information by a regulatory body as part of an integrated decision making process addresses the way in which risk information is being used as part of an integrated process in making decisions about safety issues at nuclear plants – commonly referred to as risk-informed decision making. The risk-informed approach aims to integrate in a systematic manner quantitative and qualitative, deterministic and probabilistic safety considerations to obtain a balanced decision. Probabilistic Safety Assessment (PSA) is a methodology that can be applied to provide a structured analysis process to evaluate the frequency and consequences of accidents scenarios in nuclear power plants. Technical Specifications (TS) are specifications regarding the characteristics of nuclear power plants (variables, systems or components) of overriding importance to nuclear safety and radiation protection, which is an integral part of plant operation authorization. Limiting Conditions of Operation (LCO) are the minimum levels of performance or capacity or operating system components required for the safe operation of nuclear plants, as defined in technical specifications. The Maintenance Rule (MR) is a requirement established by the U. S. Nuclear Regulatory Commission (NRC) to check the effectiveness of maintenance carried out in nuclear plants, and plant configuration control. The control of plant configuration is necessary to verify the impact of the maintenance of a safety device out of service on plant safety. The Electric Power Research Institute (EPRI) has assessed the role of probabilistic safety analysis in the regulation of nuclear power plants with the following objectives: a) to provide utilities with an approach for developing and implementing nuclear power station risk-managed technical specification programs; and b) to complement and supplement existing successful configuration risk management applications such as MR. This paper focuses on the evaluation of EPRI

Preparation of radiological effluent technical specifications for nuclear power plants. a guidance manual for users of standard technical specifications

International Nuclear Information System (INIS)

Boegli, J.S.; Bellamy, R.R.; Britz, W.L.; Waterfield, R.L.

1978-10-01

The purpose of this manual is to describe methods found acceptable to the staff of the U.S. Nuclear Regulatory Commission (NRC) for the calculation of certain key values required in the preparation of proposed radiological effluent Technical Specifications using the Standard Technical Specifications for light-water-cooled nuclear power plants. This manual also provides guidance to applicants for operating licenses for nuclear power plants in the preparation of proposed radiological effluent Technical Specifications or in preparing requests for changes to existing radiological effluent Technical Specifications for operating licenses. The manual additionally describes current staff positions on the methodology for estimating radiation exposure due to the release of radioactive materials in effluents and on the administrative control of radioactive waste treatment systems

Nuclear materials accountancy in an industrial MOX fuel fabrication plant safeguards versus commercial aspects

International Nuclear Information System (INIS)

Canck, H. de; Ingels, R.; Lefevre, R.

1991-01-01

In a modern MOX Fuel Fabrication Plant, with a large throughput of nuclear materials, computerized real-time accountancy systems are applied. Following regulations and prescriptions imposed by the Inspectorates EURATOM-IAEA, the State and also by internal plant safety rules, the accountancy is kept in plutonium element, uranium element and 235 U for enriched uranium. In practice, Safeguards Authorities are concerned with quantities of the element (U tot , Pu tot ) and to some extent with its fissile content. Custom Authorities are for historical reasons, interested in fissile quantities (U fiss , Pu fiss ) whereas owners wish to recover the energetic value of their material (Pu equivalent). Balancing the accountancy simultaneously in all these related but not proportional units is a new problem in a MOX-plant where pool accountancy is applied. This paper indicates possible ways to solve the balancing problem created by these different units used for expressing nuclear material quantities

Radiological Effluent Technical Specifications (RETS) implementation, Kewaunee Nuclear Power Plant

International Nuclear Information System (INIS)

Serrano, W.; Akers, D.W.

1985-06-01

A review of the Radiological Effluent Technical Specifications (RETS) for the Kewaunee Nuclear Power Plant was performed. The principal review guidelines used were NUREG-0133, ''Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants,'' and Draft 7'' of NUREG-0472, Revision 3, ''Radiological Effluent Technical Specifications for Pressurized Water Reactors.'' Draft submittals were discussed with the Licensee by the NRC staff until all items requiring changes to the Technical Specifications were resolved. The Licensee then submitted final proposed RETS to the NRC which were evaluated and found to be in compliance with the NRC review guidelines. The proposed Offsite Dose Calculation Manual and the Radiological Environmental Monitoring Manual were reviewed and generally found to be in compliance with the NRC review guidelines

Interpretation of bioassay data from nuclear fuel fabrication workers

International Nuclear Information System (INIS)

Melo, D.; Xavier, M.

2005-01-01

Full text: In nuclear fuel fabrication facilities, workers are exposed to different compounds of enriched uranium. Although in this kind of facility the main route of intake is inhalation, ingestion may occur in some situations. The interpretation of the bioassay data is very complex, since it is necessary taking into account all the different parameters, which is a big challenge. Due to the high cost of the individual monitoring programme for internal dose assessment in the routine monitoring programmes, usually only one type of measurement is assigned. In complex situations like the one described in this paper, where several parameters can compromise the accuracy of the bioassay interpretation it is need to have a combination of techniques to evaluate the internal dose. According to ICRP 78 (1997), the general order of preference in terms of accuracy of interpretation is: body activity measurement, excreta analysis and personal air sampling. Results of monitoring of working environment may provide information that assists in interpretation on particle size, chemical form and solubility, time of intake. A group of seventeen workers from controlled area of the fuel fabrication facility was selected to evaluate the internal dose using all different available techniques during a certain period. The workers were monitored for determination of uranium content in the daily urinary and faecal excretion (collected over a period of 3 consecutive days), chest counting and personal air sampling. The results have shown that at least two types of sensitivity techniques must be used, since there are some sources of uncertainties on the bioassay interpretation, like mixture of uranium compounds intake and different routes of intake. The combination of urine and faeces analysis has shown to be the more appropriate methodology for assessing internal dose in this situation. (author)

Guidelines for preparing specifications for nuclear power plants (NCIG-04): Final report

International Nuclear Information System (INIS)

1988-04-01

This document provides guidance for preparing technical requirements used in procurement and installation specifications. It is a compilation of recommend practices for writing specifications to preserve the best guidance coming out of recent years experience from construction of nuclear plants. It is intended to: Establish good practices for the content of specifications used for nuclear power plants; Be applicable to a wide range of specifications used for initial construction of plants and modifications to existing plants, including equipment replacement; and Provide guidance to specification preparers and reviewers

Specific schedule conditions for the formation of personnel of A or B category working in nuclear facilities. Option nuclear reactor

CERN Document Server

Int. At. Energy Agency, Wien

2002-01-01

This document describes the specific dispositions relative to the nuclear reactor domain, for the formation to the conventional and radiation risks prevention of personnel of A or B category working in nuclear facilities. The application domain, the applicable documents, the liability, the specificity of the nuclear reactor and of the retraining, the Passerelle formation, are presented. (A.L.B.)

Development and fabrication of patient-specific knee implant using additive manufacturing techniques

Science.gov (United States)

Zammit, Robert; Rochman, Arif

2017-10-01

Total knee replacement is the most effective treatment to relief pain and restore normal function in a diseased knee joint. The aim of this research was to develop a patient-specific knee implant which can be fabricated using additive manufacturing techniques and has reduced wear rates using a highly wear resistant materials. The proposed design was chosen based on implant requirements, such as reduction in wear rates as well as strong fixation. The patient-specific knee implant improves on conventional knee implants by modifying the articulating surfaces and bone-implant interfaces. Moreover, tribological tests of different polymeric wear couples were carried out to determine the optimal materials to use for the articulating surfaces. Finite element analysis was utilized to evaluate the stresses sustained by the proposed design. Finally, the patient-specific knee implant was successfully built using additive manufacturing techniques.

Powder fabrication of U-Mo alloys for nuclear dispersion fuels

Energy Technology Data Exchange (ETDEWEB)

Durazzo, Michelangelo; Rocha, Claudio Jose da; Mestnik Filho, Jose; Leal Neto, Ricardo Mendes, E-mail: mdurazzo@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

For the last 30 years high uranium density dispersion fuels have been developed in order to accomplish the low enrichment goals of the Reduced Enrichment for Research and Test Reactors (RERTR) Program. Gamma U-Mo alloys, particularly with 7 to 10 wt% Mo, as a fuel phase dispersed in aluminum matrix, have shown good results concerning its performance under irradiation tests. That's why this fissile phase is considered to be used in the nuclear fuel of the Brazilian Multipurpose Research Reactor (RMB), currently being designed. Powder production from these ductile alloys has been attained by atomization, mechanical (machining, grinding, cryogenic milling) and chemical (hydriding-de hydriding) methods. This work is a part of the efforts presently under way at IPEN to investigate the feasibility of these methods. Results on alloy fabrication by induction melting and gamma-stabilization of U-10Mo alloys are presented. Some results on powder production and characterization are also discussed. (author)

Powder fabrication of U-Mo alloys for nuclear dispersion fuels

International Nuclear Information System (INIS)

Durazzo, Michelangelo; Rocha, Claudio Jose da; Mestnik Filho, Jose; Leal Neto, Ricardo Mendes

2011-01-01

For the last 30 years high uranium density dispersion fuels have been developed in order to accomplish the low enrichment goals of the Reduced Enrichment for Research and Test Reactors (RERTR) Program. Gamma U-Mo alloys, particularly with 7 to 10 wt% Mo, as a fuel phase dispersed in aluminum matrix, have shown good results concerning its performance under irradiation tests. That's why this fissile phase is considered to be used in the nuclear fuel of the Brazilian Multipurpose Research Reactor (RMB), currently being designed. Powder production from these ductile alloys has been attained by atomization, mechanical (machining, grinding, cryogenic milling) and chemical (hydriding-de hydriding) methods. This work is a part of the efforts presently under way at IPEN to investigate the feasibility of these methods. Results on alloy fabrication by induction melting and gamma-stabilization of U-10Mo alloys are presented. Some results on powder production and characterization are also discussed. (author)

Financial aspects of the nuclear fuel cycle

International Nuclear Information System (INIS)

Lurf, G.

1975-01-01

A nuclear power plant has a forward supply of several years as a consequence of the long processing time of the uranium from mining to delivery of fabricated fuel elements and of the long insertion time in the reactor. This leads to a considerable capital requirement although the specific fuel costs for nuclear fuel are considerably lower then for a conventional power plant and present only 15% of the total generating costs. (orig./RW) [de

An Overview of Current and Past W-UO[2] CERMET Fuel Fabrication Technology

International Nuclear Information System (INIS)

Douglas E. Burkes; Daniel M. Wachs; James E. Werner; Steven D. Howe

2007-01-01

Studies dating back to the late 1940s performed by a number of different organizations and laboratories have established the major advantages of Nuclear Thermal Propulsion (NTP) systems, particularly for manned missions. A number of NTP projects have been initiated since this time; none have had any sustained fuel development work that appreciably contributed to fuel fabrication or performance data from this era. As interest in these missions returns and previous space nuclear power researchers begin to retire, fuel fabrication technologies must be revisited, so that established technologies can be transferred to young researchers seamlessly and updated, more advanced processes can be employed to develop successful NTP fuels. CERMET fuels, specifically W-UO2, are of particular interest to the next generation NTP plans since these fuels have shown significant advantages over other fuel types, such as relatively high burnup, no significant failures under severe transient conditions, capability of accommodating a large fission product inventory during irradiation and compatibility with flowing hot hydrogen. Examples of previous fabrication routes involved with CERMET fuels include hot isostatic pressing (HIPing) and press and sinter, whereas newer technologies, such as spark plasma sintering, combustion synthesis and microsphere fabrication might be well suited to produce high quality, effective fuel elements. These advanced technologies may address common issues with CERMET fuels, such as grain growth, ductile to brittle transition temperature and UO2 stoichiometry, more effectively than the commonly accepted 'traditional' fabrication routes. Bonding of fuel elements, especially if the fabrication process demands production of smaller element segments, must be investigated. Advanced brazing techniques and compounds are now available that could produce a higher quality bond segment with increased ease in joining. This paper will briefly address the history of CERMET

New polymorphous computing fabric

International Nuclear Information System (INIS)

Wolinski, Christophe; Gokhale, Maya; McCabe, Kevin P.

2002-01-01

This paper introduces a new polymorphous computing Fabric well suited to DSP and Image Processing and describes its implementation on a Configurable System on a Chip (CSOC). The architecture is highly parameterized and enables customization of the synthesized Fabric to achieve high performance for a specific class of application. For this reason it can be considered to be a generic model for hardware accelerator synthesis from a high level specification. Another important innovation is the Fabric uses a global memory concept, which gives the host processor random access to all the variables and instructions on the Fabric. The Fabric supports different computing models including MIMD, SPMD and systolic flow and permits dynamic reconfiguration. We present a specific implementation of a bank of FIR filters on a Fabric composed of 52 cells on the Altera Excalibur ARM running at 33 MHz. The theoretical performance of this Fabric is 1.8 GMACh. For the FIR application we obtain 1.6 GMAC/s real performance. Some automatic tools have been developed like the tool to provide a host access utility and assembler.

Selection of engineering materials and fabrication of liquid metal fast breeder reactors

International Nuclear Information System (INIS)

Patriarca, P.

1975-01-01

Information is presented graphically and pictorially concerning the need for nuclear power; basic nuclear concepts including BWR, PWR, HTGR, and LMFBR; the fissioning process; nuclear reactor fuel; fabrication of reactor vessels for LMFBR's; fabrication of intermediate heat exchangers for LMFBR's; piping fabrication for LMFBR's; transition welds; steam generators for LMFBR demonstration plants worldwide; stress corrosion cracking of steam generator materials and weldments; post--test examination of the Alco/BLH sodium-heated steam generator; alternate steam generator designs; and alternate structural materials. (DCC)

Construction, fabrication, and installation

International Nuclear Information System (INIS)

1992-05-01

This standard specifies the construction, fabrication, and installation requirements that apply to concrete containment structures of a containment system designated as class containment components, parts and appurtenances for nuclear power plants

Development of fabrication technology for ceramic nuclear fuel

International Nuclear Information System (INIS)

Lee, Young Woo; Sohn, D. S.; Na, S. H.

2003-05-01

The purpose of the study is to develop the fabrication technology of MOX fuel. The researches carried out during the last stage(1997. 4.∼2003. 3.) mainly consisted of ; study of MOX pellet fabrication technology for application and development of characterization technology for the aim of confirming the development of powder treatment technology and sintering technology and of the optimization of the above technologies and fabrication of Pu-MOX pellet specimens through an international joint collaboration between KAERI and PSI based on the fundamental technologies developed in KAERI. Based on the studies carried out and the results obtained during the last stage, more extensive studies for the process technologies of the unit processes were performed, in this year, for the purpose of development of indigenous overall MOX pellet fabrication process technology, relating process parameters among the unit processes and integrating these unit process technologies. Furthermore, for the preparation of transfer of relevant technologies to the industries, a feasibility study was performed on the commercialization of the technology developed in KAERI with the relevant industry in close collaboration

Standard Specification for Nuclear Grade Zirconium Oxide Pellets

CERN Document Server

American Society for Testing and Materials. Philadelphia

2009-01-01

1.1 This specification applies to pellets of stabilized zirconium oxide used in nuclear reactors. 1.2 The values stated in SI units are to be regarded as the standard. The values given in parentheses are for information only.

Electrochemical/Pyrometallurgical Waste Stream Processing and Waste Form Fabrication

Energy Technology Data Exchange (ETDEWEB)

Steven Frank; Hwan Seo Park; Yung Zun Cho; William Ebert; Brian Riley

2015-07-01

This report summarizes treatment and waste form options being evaluated for waste streams resulting from the electrochemical/pyrometallurgical (pyro ) processing of used oxide nuclear fuel. The technologies that are described are South Korean (Republic of Korea – ROK) and United States of America (US) ‘centric’ in the approach to treating pyroprocessing wastes and are based on the decade long collaborations between US and ROK researchers. Some of the general and advanced technologies described in this report will be demonstrated during the Integrated Recycle Test (IRT) to be conducted as a part of the Joint Fuel Cycle Study (JFCS) collaboration between US Department of Energy (DOE) and ROK national laboratories. The JFCS means to specifically address and evaluated the technological, economic, and safe guard issues associated with the treatment of used nuclear fuel by pyroprocessing. The IRT will involve the processing of commercial, used oxide fuel to recover uranium and transuranics. The recovered transuranics will then be fabricated into metallic fuel and irradiated to transmutate, or burn the transuranic elements to shorter lived radionuclides. In addition, the various process streams will be evaluated and tested for fission product removal, electrolytic salt recycle, minimization of actinide loss to waste streams and waste form fabrication and characterization. This report specifically addresses the production and testing of those waste forms to demonstrate their compatibility with treatment options and suitability for disposal.

South Korea's nuclear fuel industry

International Nuclear Information System (INIS)

Clark, R.G.

1990-01-01

March 1990 marked a major milestone for South Korea's nuclear power program, as the country became self-sufficient in nuclear fuel fabrication. The reconversion line (UF 6 to UO 2 ) came into full operation at the Korea Nuclear Fuel Company's fabrication plant, as the last step in South Korea's program, initiated in the mid-1970s, to localize fuel fabrication. Thus, South Korea now has the capability to produce both CANDU and pressurized water reactor (PWR) fuel assemblies. This article covers the nuclear fuel industry in South Korea-how it is structures, its current capabilities, and its outlook for the future

Fabrication of HTTR first loading fuel

International Nuclear Information System (INIS)

Kato, S.; Yoshimuta, S.; Hasumi, T.; Sato, K.; Sawa, K.; Suzuki, S.; Mogi, H.; Shiozawa, S.; Tanaka, T.

2001-01-01

This paper summarizes the fabrication of the first loading fuel for HTTR, High Temperature engineering Test Reactor constructed by JAERI, Japan Atomic Energy Research Institute. The fuel fabrication started at the HTR fuel facility of NFI, Nuclear Fuel Industries, Ltd., June 1995. 4,770 fuel rods were fabricated through the fuel kernel, coated fuel particle and fuel compaction process, then 150 fuel elements were assembled in the reactor building December 1997. Fabrication technology for the fuel was established through a lot of R and D activities and fabrication experience of irradiation examination samples spread over about 30 years. Most of all, very high quality and production efficiency of fuel were achieved by the development of the fuel kernel process using the vibration dropping technology, the continuous 4-layer coating process and the automatic compaction process. As for the inspection technology, the development of the automatic measurement equipment for coated layer thickness of a coated fuel particle and uranium content of a fuel compact contributed to the higher reliability and rationalization of the inspection process. The data processing system for the fabrication and quality control, which was originally developed by NFI, made possible not only quick feedback of statistical quality data to the fabrication processes, but also automatic document preparation, such as inspection certificates and accountability control reports. The quality of the first loading fuel fully satisfied the design specifications for the fuel. In particular, average bare uranium fraction and SiC defective fraction of fuel compacts were 2x10 -6 and 8x10 -5 , respectively. According to the preceding irradiation examinations being performed at JMTR, Japan Materials Testing Reactor of JAERI, the specimen sampled from the first loading fuel shows good irradiation performance. (author)

Influence of Fabric Parameters on Thermal Comfort Performance of Double Layer Knitted Interlock Fabrics

Directory of Open Access Journals (Sweden)

Afzal Ali

2017-03-01

Full Text Available The aim of this study was to analyse the effects of various fabric parameters on the thermal resistance, thermal conductivity, thermal transmittance, thermal absorptivity and thermal insulation of polyester/cotton double layer knitted interlock fabrics. It was found that by increasing fibre content with higher specific heat increases the thermal insulation while decreases the thermal transmittance and absorptivity of the fabric. It was concluded that double layer knitted fabrics developed with higher specific heat fibres, coarser yarn linear densities, higher knitting loop length and fabric thickness could be adequately used for winter clothing purposes.

The ORSEC arrangement and the 'nuclear' intervention specific plan

International Nuclear Information System (INIS)

Guenon, C.

2010-01-01

In order to take the specific character of a nuclear emergency situation into account, France has developed planning tools within the so-called Crisis National Organisation (ONC, organisation nationale de crise). This organisation involves public bodies, agencies and companies. Thus, intervention specific plans (PPI, plans particuliers d'intervention) are included in the ORSEC general arrangement. The assessment of geographical and chronological consequences of a nuclear accident has lead to the definition of two main categories of measures, depending on the fact they are immediately or progressively applied. They involve the intervention of specialised means. This report also indicates how new measures have been introduced in the ORSEC arrangement to manage the post-accident phase. The author also outlines that crisis communication must also be prepared and tested

Metallic Reactor Fuel Fabrication for SFR

Energy Technology Data Exchange (ETDEWEB)

Song, Hoon; Kim, Jong-Hwan; Ko, Young-Mo; Woo, Yoon-Myung; Kim, Ki-Hwan; Lee, Chan-Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

The metal fuel for an SFR has such advantages such as simple fabrication procedures, good neutron economy, high thermal conductivity, excellent compatibility with a Na coolant, and inherent passive safety 1. U-Zr metal fuel for SFR is now being developed by KAERI as a national R and D program of Korea. The fabrication technology of metal fuel for SFR has been under development in Korea as a national nuclear R and D program since 2007. The fabrication process for SFR fuel is composed of (1) fuel slug casting, (2) loading and fabrication of the fuel rods, and (3) fabrication of the final fuel assemblies. Fuel slug casting is the dominant source of fuel losses and recycled streams in this fabrication process. Fabrication on the rod type metallic fuel was carried out for the purpose of establishing a practical fabrication method. Rod-type fuel slugs were fabricated by injection casting. Metallic fuel slugs fabricated showed a general appearance was smooth.

Specific schedule conditions for the formation of personnel of A or B category working in nuclear facilities. Option nuclear reactor-borne

CERN Document Server

Int. At. Energy Agency, Wien

2002-01-01

This document describes the specific dispositions relative to the nuclear reactor-borne domain, for the formation to the conventional and radiation risks prevention of personnel of A or B category working in nuclear facilities. The application domain, the applicable documents, the liability, the specificity of the nuclear reactor-borne and of the retraining, the Passerelle formation, are presented. (A.L.B.)

Characteristics and fabrication of cermet spent nuclear fuel casks: ceramic particles embedded in steel

International Nuclear Information System (INIS)

Forsberg, C.W.; Swaney, P.M.; Tiegs, T.N.

2004-01-01

Cermets are being investigated as an advanced material of construction for casks that can be used for storage, transport, or disposal of spent nuclear fuel (SNF). Cermets, which consist of ceramic particles embedded in steel, are a method to incorporate brittle ceramics with highly desirable properties into a strong ductile metal matrix with a high thermal conductivity, thus combining the best properties of both materials. Traditional applications of cermets include tank armor, vault armor, drill bits, and nuclear test-reactor fuel. Cermets with different ceramics (DUO 2 , Al 2 O 3 , Gd 2 O 3 , etc.) are being investigated for the manufacture of SNF casks. Cermet casks offer four potential benefits: greater capacity (more SNF assemblies) for the same gross weight cask, greater capacity (more SNF assemblies) for the same external dimensions, improved resistance to assault, and superior repository performance. These benefits are achieved by varying the composition, volume fraction, and particulate size of the ceramic particles in the cermet with position in the cask body. Addition of depleted uranium dioxide (DUO 2 ) to the cermet increases shielding density, improves shielding effectiveness, and increases cask capacity for a given cask weight or size. Addition of low-density aluminium oxide (Al 2 O 3 ) to the outer top and bottom sections of the cermet cask, where the radiation levels are lower, can lower cask weight without compromising shielding. The use of Al2O3 and other oxides, in appropriate locations, can increase resistance to assault. Repository performance may be improved by compositional control of the cask body to (1) create a local geochemical environment that slows the long-term degradation of the SNF and (2) enables the use of DUO 2 for longterm criticality control. While the benefits of using cermets follow directly from their known properties, the primary challenge is to develop low-cost methods to fabricate casks with variable cermet compositions

Results from a Field Trial of the Radio Frequency Based Cylinder Accountability and Tracking System at the Global Nuclear Fuel Americas Fuel Fabrication Facility

International Nuclear Information System (INIS)

Fitzgerald, Peter; Laughter, Mark D.; Martyn, Rose; Pickett, Chris A.; Rowe, Nathan C.; Younkin, James R.; Shephard, Adam M.

2010-01-01

The Cylinder Accountability and Tracking System (CATS) is a tool designed for use by the International Atomic Energy Agency (IAEA) to improve overall inspector efficiency through real-time unattended monitoring of cylinder movements, site specific rules-based event detection, and the capability to integrate many types of monitoring technologies. The system is based on the tracking of cylinder movements using (radio frequency) RF tags, and the collection of data, such as accountability weights, that can be associated with the cylinders. This presentation will cover the installation and evaluation of the CATS at the Global Nuclear Fuels (GNF) fuel fabrication facility in Wilmington, NC. This system was installed to evaluate its safeguards applicability, operational durability under operating conditions, and overall performance. An overview of the system design and elements specific to the GNF deployment will be presented along with lessons learned from the installation process and results from the field trial.

Informal presentations by fuel fabricators and others [contributed by A. Nishiyama, Nuclear Fuel Industries, Ltd.

International Nuclear Information System (INIS)

Nishiyama, A.

1993-01-01

This paper contains a brief summary of activities in the field of research reactor fuel fabrication in Nuclear Fuel Industries Sumitomo and Furukawa Industries. Since 1956 2 million dollars were spent for development of nuclear fuels and plant facilities including complete manufacturing and testing capabilities. Now this company is the only fuel supplier for the research reactors in Japan. The fabrication process starts with the melting, alloying, and casting of U-Al. The uranium billets are prepared by foreign fabricators. The uranium content varies from 13 to 22 wt % according to the purchaser's specifications. In making fuel plates, the picture frame method is applied. In this case, the original procedure is sufficiently effective in avoiding dogboning. The plates are finished by hot and cold roll milling and inspected dimensionally, metallurgically, and mechanically, and at the same time the blister test and X-ray radiographic tests are performed. Fuel elements are assembled by rolling flat or curved plates into side plate grooves and end-fit welding. Finished elements are tested dimensionally and hydraulically. Nominal losses during operation are less than 1% of the uranium metal. Our present capacity licensed by the Japanese Government is approximately 950 fuel elements a year. About 35 employees including engineers are engaged in development and manufacturing of fuels. Owing to the small limited demand of the research reactor fuels in Japan during the past 20 years (mostly in last 10 years), we processed only about 350 kg of highly enriched uranium and supplied approximately 1000 fuel elements to JAERI, Kyoto University, and others, and we have been suffering red-ink balance of budget every year. Some of trials in development are briefly discussed. In case of UO 2 -Al metal fuel plates, the vibratory compacting method was very popular among many researchers about 10 years ago. A lot of time and money was spent to study the economic fabrication process of

Safety of nuclear fuel cycle facilities. Safety requirements

International Nuclear Information System (INIS)

2008-01-01

This publication covers the broad scope of requirements for fuel cycle facilities that, in light of the experience and present state of technology, must be satisfied to ensure safety for the lifetime of the facility. Topics of specific reference include aspects of nuclear fuel generation, storage, reprocessing and disposal. Contents: 1. Introduction; 2. The safety objective, concepts and safety principles; 3. Legal framework and regulatory supervision; 4. The management system and verification of safety; 5. Siting of the facility; 6. Design of the facility; 7. Construction of the facility; 8. Commissioning of the facility; 9. Operation of the facility; 10. Decommissioning of the facility; Appendix I: Requirements specific to uranium fuel fabrication facilities; Appendix II: Requirements specific to mixed oxide fuel fabrication facilities; Appendix III: Requirements specific to conversion facilities and enrichment facilities

Spent nuclear fuel project product specification

International Nuclear Information System (INIS)

Pajunen, A.L.

1998-01-01

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

Fabrication development for high-level nuclear waste containers for the tuff repository; Phase 1 final report

Energy Technology Data Exchange (ETDEWEB)

Domian, H.A.; Holbrook, R.L.; LaCount, D.F. [Babcock and Wilcox Co., Lynchburg, VA (USA). Nuclear Power Div.]|[Babcock and Wilcox Co., Alliance, OH (USA). Research and Development Div.

1990-09-01

This final report completes Phase 1 of an engineering study of potential manufacturing processes for the fabrication of containers for the long-term storage of nuclear waste. An extensive literature and industry review was conducted to identify and characterize various processes. A technical specification was prepared using the American Society of Mechanical Engineers Boiler & Pressure Vessel Code (ASME BPVC) to develop the requirements. A complex weighting and evaluation system was devised as a preliminary method to assess the processes. The system takes into account the likelihood and severity of each possible failure mechanism in service and the effects of various processes on the microstructural features. It is concluded that an integral, seamless lower unit of the container made by back extrusion has potential performance advantages but is also very high in cost. A welded construction offers lower cost and may be adequate for the application. Recommendations are made for the processes to be further evaluated in the next phase when mock-up trials will be conducted to address key concerns with various processes and materials before selecting a primary manufacturing process. 43 refs., 26 figs., 34 tabs.

Subsoil exploration of the estimated building site for nuclear fuel development and fabrication facility

Energy Technology Data Exchange (ETDEWEB)

Song, In Taek [Korea Atomic Energy Research Institute, Taejeon (Korea)

2000-01-01

The objective of this report, as the result of subsoil exploration, is to provide basic design data of structural plan for nuclear fuel development and fabrication facility that is builded on Duckjin 150, Yusong, Taejeon, Korea, and provide basic data for execution of work. The soft rock level of estimated building site is deep(18.0m:BH-1, 20.5m:BH-2, 25.5m:BH-3) and the hard rock level of it is very deep (33.0m:BH-1, 46.0m:BH-2, 34.5m:BH-3) , for structural design, the hard rock shall be the bottom of foundation. 9 figs., 19 tabs. (Author)

Design and fabrication of NDA standards

International Nuclear Information System (INIS)

Long, S.M.; Hsue, S.T.

1996-01-01

The Plutonium Facility, TA-55, at Los Alamos National Laboratory is currently producing NDA calibration standards used by various laboratories in the DOE complex. These NIST traceable standards have been produced to calibrate NDA instruments for accountability measurements used for resolving shipper/receiver differences, and for accountability in process residues and process waste. Standards are needed to calibrate various NDA (Non-destructive Assay) instruments such as neutron coincidence counters, gamma-ray counters, and calorimeters. These instruments measure various ranges of nuclear material being produced in the DOE nuclear community. Los Alamos National Laboratory has taken a lead role in fabrication of uranium and plutonium standards, along with other actinides such as neptunium and americium. These standards have been fabricated for several laboratories within the complex. This paper will summarize previous publications detailing the careful planning encompassing components such as precise weighing, destructive analysis, and the use of post fabrication NDA measurements to confirm that the standards meet all preliminary expectations before use in instrument calibration. The paper will also describe the specialized containers, diluents, and the various amount of nuclear materials needed to accommodate the calibration ranges of the instruments

Advanced fabrication technology

International Nuclear Information System (INIS)

Sheely, W.F.

1986-01-01

The Fuel Cycle Plant is a multipurpose nuclear facility located on the Hanford Nuclear Reservation in eastern Washington state. The facility is part of the Hanford Engineering Development Laboratory which is operated by Westinghouse Hanford Company for the Department of Energy. The Fuel Cycle Plant is currently being prepared to support the Liquid Metal Reactors Program with fuel fabrication services for the Fast Flux Test Facility and other LMR programs. This report describes the technical innovations to be utilized in the operation of this plant

Nuclear power and the nuclear fuel cycle

International Nuclear Information System (INIS)

Hardy, C.J.; Silver, J.M.

1985-09-01

The report provides data and assessments of the status and prospects of nuclear power and the nuclear fuel cycle. The report discusses the economic competitiveness of nuclear electricity generation, the extent of world uranium resources, production and requirements, uranium conversion and enrichment, fuel fabrication, spent fuel treatment and radioactive waste management. A review is given of the status of nuclear fusion research

LLNL/YMP Waste Container Fabrication and Closure Project

International Nuclear Information System (INIS)

1990-10-01

The Department of Energy's Office of Civilian Radioactive Waste Management (OCRWM) Program is studying Yucca Mountain, Nevada as a suitable site for the first US high-level nuclear waste repository. Lawrence Livermore National Laboratory (LLNL) has the responsibility for designing and developing the waste package for the permanent storage of high-level nuclear waste. This report is a summary of the technical activities for the LLNL/YMP Nuclear Waste Disposal Container Fabrication and Closure Development Project. Candidate welding closure processes were identified in the Phase 1 report. This report discusses Phase 2. Phase 2 of this effort involved laboratory studies to determine the optimum fabrication and closure processes. Because of budget limitations, LLNL narrowed the materials for evaluation in Phase 2 from the original six to four: Alloy 825, CDA 715, CDA 102 (or CDA 122) and CDA 952. Phase 2 studies focused on evaluation of candidate material in conjunction with fabrication and closure processes

Improved technical specifications and related improvements to safety in commercial Nuclear power plants

International Nuclear Information System (INIS)

Hoffman, D.R.; Demitrack, T.; Schiele, R.; Jones, J.C.

2004-01-01

Many of the commercial nuclear power plants in the United States (US) have been converting a portion of the plant operating license known as the Technical Specifications (TS) in accordance with a document published by the US Nuclear Regulatory Commission (NRC). The TS prescribe commercial nuclear power plant operating requirements. There are several types of nuclear power plants in the US, based on the technology of different vendors, and there is an NRC document that supports each of the five different vendor designs. The NRC documents are known as the Improved Standard Technical Specifications (ISTS) and are contained in a separate document (NUREG series) for each one of the designs. EXCEL Services Corporation (hereinafter EXCEL) has played a major role in the development of the ISTS and in the development, licensing, and implementation of the plant specific Improved Technical Specifications (ITS) (which is based on the ISTS) for the commercial nuclear power plants in the US that have elected to make this conversion. There are currently 103 operating commercial nuclear power plants in the US and 68 of them have successfully completed the conversion to the ITS and are now operating in accordance with their plant specific ITS. The ISTS is focused mainly on safety by ensuring the commercial nuclear reactors can safely shut down and mitigate the consequences of any postulated transient and accident. It accomplishes this function by including requirements directly associated with safety in a document structured systematically and taking into account some key human factors and technical initiatives. This paper discusses the ISTS including its format, content, and detail, the history of the ISTS, the ITS development, licensing, and implementation process, the safety improvements resulting from a plant conversion to ITS, and the importance of the ITS Project to the industry. (Author)

Improved technical specifications and related improvements to safety in commercial Nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Hoffman, D.R.; Demitrack, T.; Schiele, R.; Jones, J.C. [EXCEL Services Corporation, 11921 Rockville Pike, Suite 100, Rockville, MD 20852 (United States)]. e-mail: donaldh@excelservices.com

2004-07-01

Many of the commercial nuclear power plants in the United States (US) have been converting a portion of the plant operating license known as the Technical Specifications (TS) in accordance with a document published by the US Nuclear Regulatory Commission (NRC). The TS prescribe commercial nuclear power plant operating requirements. There are several types of nuclear power plants in the US, based on the technology of different vendors, and there is an NRC document that supports each of the five different vendor designs. The NRC documents are known as the Improved Standard Technical Specifications (ISTS) and are contained in a separate document (NUREG series) for each one of the designs. EXCEL Services Corporation (hereinafter EXCEL) has played a major role in the development of the ISTS and in the development, licensing, and implementation of the plant specific Improved Technical Specifications (ITS) (which is based on the ISTS) for the commercial nuclear power plants in the US that have elected to make this conversion. There are currently 103 operating commercial nuclear power plants in the US and 68 of them have successfully completed the conversion to the ITS and are now operating in accordance with their plant specific ITS. The ISTS is focused mainly on safety by ensuring the commercial nuclear reactors can safely shut down and mitigate the consequences of any postulated transient and accident. It accomplishes this function by including requirements directly associated with safety in a document structured systematically and taking into account some key human factors and technical initiatives. This paper discusses the ISTS including its format, content, and detail, the history of the ISTS, the ITS development, licensing, and implementation process, the safety improvements resulting from a plant conversion to ITS, and the importance of the ITS Project to the industry. (Author)

Introduction to the 'CAS' nuclear propulsion plant for ships: specific safety options

International Nuclear Information System (INIS)

Verdeau, J.J.; Baujat, J.

1978-01-01

After a brief review of the development of nuclear propulsion in FRANCE (Land Based Prototype PAT 1964 - Navy nuclear ships - Advanced Nuclear Boiler Prototype CAP 1975 and now the CAS nuclear plant), the specific safety options of CAS are presented: cold, compartmented fuel (plates); reduced flow during LOCA; permanent cooling of fuel during LOCA; pressurized, entirely passive containment; no control rod ejection and possibility of temporary storage of spent fuel on board [fr

Chemical characterization of nuclear materials: recent trends

International Nuclear Information System (INIS)

Prakash, Amrit; Nandi, C.; Patil, A.B.; Khan, K.B.

2013-01-01

Analytical chemistry plays a very important role for nuclear fuel activities be it fuel fabrication, waste management or reprocessing. Nuclear fuels are selected based on the type of reactor. The nuclear fuel has to conform to stringent chemical specifications like boron, cadmium, rare earths, hydrogen, oxygen to metal ratio, total gas, heavy metal content, chlorine and fluorine etc. Selection of technique is very important to evaluate the true specification. This is important particularly when the analyses have to perform inside leak tight enclosure. The present paper describes the details of advanced analytical techniques being developed and used in chemical characterization of nuclear materials specially fuels during their fabrication. Nuclear fuels comprise of fuels based on UO 2 , PUO 2 , ThO 2 and combination of (U+Pu)O 2 , (Th+U)O 2 , (Th+Pu)O 2 , (U+Pu)C, (U+Pu)N etc depending on the type of reactors chosen Viz. Pressurized Heavy water Reactor (PHWR), Boiling Water Reactor (BWR), Fast Breeder Test Reactor and Prototype Fast Breeder Reactor (PFBR). Chemical characterization of these fuels is very important for performance of fuel in the reactor. It provides means to ascertain that the quality of the fabricated fuel conforms to the chemical specifications for the fuel laid down by the designer. The batches of sintered/degassed pellets are subjected to comprehensive chemical quality control for trace constituents, stoichiometry and isotopic composition. Chemical Quality Control of fuel is carried out at different stages of manufacture namely feed materials, sintering, vacuum degassing and fuel element welding. Advanced analytical technique based on titrimetry, spectroscopy, thermogravimetry, XRF and XRD have largely been used for this purpose. Since they have to be handled inside special enclosures, extreme care are being taken during handling. Instruments are being developed/modified for ease of handling and maintenance. The method should be fast to reduce

Fabric based supercapacitor

International Nuclear Information System (INIS)

Yong, S; Tudor, M J; Beeby, S P; Owen, J R

2013-01-01

Flexible supercapacitors with electrodes coated on inexpensive fabrics by the dipping technique. This paper present details of the design, fabrication and characterisation of fabric supercapacitor. The sandwich structured supercapacitors can achieve specific capacitances of 11.1F/g, area capacitance 105 mF.cm −2 and maintain 95% of the initial capacitance after cycling the device for more than 15000 times

Design guide for heat transfer equipment in water-cooled nuclear reactor systems

International Nuclear Information System (INIS)

1975-07-01

Information pertaining to design methods, material selection, fabrication, quality assurance, and performance tests for heat transfer equipment in water-cooled nuclear reactor systems is given in this design guide. This information is intended to assist those concerned with the design, specification, and evaluation of heat transfer equipment for nuclear service and the systems in which this equipment is required. (U.S.)

Role of non-destructive examinations in leak testing of glove boxes for industrial scale plutonium handling at nuclear fuel fabrication facility along with case study

International Nuclear Information System (INIS)

Aher, Sachin

2015-01-01

Non Destructive Examinations has the prominent role at Nuclear Fuel Fabrication Facilities. Specifically NDE has contributed at utmost stratum in Leak Testing of Glove Boxes and qualifying them as a Class-I confinement for safe Plutonium handling at industrial scale. Advanced Fuel Fabrication Facility, BARC, Tarapur is engaged in fabrication of Plutonium based MOX (PuO 2 , DDUO 2 ) fuel with different enrichments for first core of PFBR reactor. Alpha- Leak Tight Glove Boxes along with HEPA Filters and dynamic ventilation form the promising engineering system for safe and reliable handling of plutonium bearing materials considering the radiotoxicity and risk associated with handling of plutonium. Leak Testing of Glove Boxes which involves the leak detection, leak rectification and leak quantifications is major challenging task. To accomplish this challenge, various Non Destructive Testing methods have assisted in promising way to achieve the stringent leak rate criterion for commissioning of Glove Box facilities for plutonium handling. This paper highlights the Role of various NDE techniques like Soap Solution Test, Argon Sniffer Test, Pressure Drop/Rise Test etc. in Glove Box Leak Testing along with procedure and methodology for effective rectification of leakage points. A Flow Chart consisting of Glove Box leak testing procedure starting from preliminary stage up to qualification stage along with a case study and observations are discussed in this paper. (author)

Decision no. 2011-DC-0214 of the French nuclear safety authority from May 5, 2011, ordering CIS bio international company to proceed to a complementary safety evaluation of its basic nuclear facility in the eyes of the Fukushima Daiichi nuclear power plant accident

International Nuclear Information System (INIS)

2011-01-01

As a consequence of the accident of the Fukushima Daiichi nuclear power plant (Japan), the French Prime Minister entrusted the French nuclear safety authority (ASN) with the mission to carry out a safety analysis re-evaluation of the French nuclear facilities, and in particular the nuclear power plants. A decision has been addressed by the ASN to each nuclear operator with the specifications of this safety re-evaluation analysis and the list of facilities in concern. This document is the decision addressed to CIS bio international company, operator of the radiopharmaceuticals fabrication facility (INB 29) of Saclay (France). (J.S.)

Technical Requirements for Fabrication and Installation of Removable Shield for CNRF in HANARO

Energy Technology Data Exchange (ETDEWEB)

Ryu, Jeong Soo; Cho, Yeong Garp; Lee, Jung Hee; Shin, Jin Won

2008-04-15

This report details the technical requirements for the fabrication and installation of the removable shield for the Cold Neutron Research Facility (CNRF) in HANARO reactor hall. The removable shield is classified as non-nuclear safety (NNS), seismic category II, and quality class T. The main function of the removable shield is to do the biological shielding of neutrons and gamma from the CN port and the guides. The removable shield consists of block type walls and roofs that can be necessarily assembled, disassembled and moveable. These will be installed between the reactor pool wall and the CNS guide bunker in. This report describes technical requirements for the removable shield such as quality assurance, seismic analysis requirements, configuration, concrete compositions, fabrication and installation requirements, test and inspection, shipping, delivery, etc. Appendix is the technical specification of structural design and analysis. Attachments are composed of the technical specification for the fabrication of the removable shield, shielding design drawings and procurement quality requirements. These technical requirements will be provided to a contract for the manufacturing and installation.

On the Generalized Correlation Equation of Welding Current for the Tig Welding Machine Used in Nuclear Fuel Fabrication

International Nuclear Information System (INIS)

Umar, Efrizon

1995-01-01

In nuclear fuel fabrication, welding plays a very important role to join the end cap to the tube. In order to determine the welding current in TIG welding process for various materials, weld geometries and welding rates, the correlation between the welding current and the other parameters are needed. This paper presents the correlation of those parameters mentioned above. The proposed correlation was tested and produced satisfactory results. (author). 8 refs., 2 tabs., 2 figs

Steel, specially for the fabrication of welded structure working under pressure in nuclear installations

International Nuclear Information System (INIS)

Dolbenko, E.T.; Astafiev, A.A.; Kark, G.S.

1981-01-01

The present invention is in the field of metallurgy. Steels have found an increasing number of applications in mechanical constructions, and notably in the construction of materials for the production of energy and for the fabrication of welded structures operating under pressure at temperatures as high as 450 0 C. A possible application is the pressurized vessels of nuclear facilities. The steels of interest contain carbon, silicon, manganese, nickel, molybdenum, vanadium, aluminium, nitrogen, phosphorus and iron, but are characterized by the fact that they also contain arsenic, tin and calcium. The sum of the weighted percentages of nickel and manganese and the weighted percentage of phosphorous are related as follows: (Ni + Mn) . P [fr

Nuclear data for specific problems. Part 1: Methods

International Nuclear Information System (INIS)

Leszczynski, Francisco

1999-01-01

The growing volume of basic nuclear data, methods and codes for processing these data, and the wide variety of problems where these data and codes are required, oblige to have an efficient system for managing all this information. In this work we present a new methodology for nuclear data processing, applied to neutron and photon transport calculations for specific problems. The base of the new methodology is the analysis of the requirements, following the chain: Problem-Components-Materials-Elements-Isotopes-Process-Tests-Final product (a library with processed data). This order is the inverse of the normal order followed up to date where, for performing a specific calculation, the first step is the choice of an existing data library for general purposes, without the previous steps of pre-processing data, and tests of the final library. Then, the used data are limited to the isotope content of this library, and the adaptation of material compositions and components to the data availability is necessary , performing finally the required calculations in a rather approximated form, depending on the available data. An interactive computer program for PC , is developed, for managing all the information generated by nuclear data processing, with the additional advantage of having a help tool for performing the needed analysis, before processing data calculations for specific applications. These analyses are based on the particular characteristics of each application, and the processed information of previous cases, is stored in conveniently designed data bases for an easy inspection of its contents. By means of an example of application of the new method, in this paper the methods of analysis and calculations and the tools used (computer programs, data bases and documents) are describes. (author)

Storage of Spent Nuclear Fuel. Specific Safety Guide

International Nuclear Information System (INIS)

2012-01-01

This Safety Guide provides recommendations and guidance on the storage of spent nuclear fuel. It covers all types of storage facilities and all types of spent fuel from nuclear power plants and research reactors. It takes into consideration the longer storage periods that have become necessary owing to delays in the development of disposal facilities and the decrease in reprocessing activities. It also considers developments associated with nuclear fuel, such as higher enrichment, mixed oxide fuels and higher burnup. The Safety Guide is not intended to cover the storage of spent fuel if this is part of the operation of a nuclear power plant or spent fuel reprocessing facility. Guidance is provided on all stages for spent fuel storage facilities, from planning through siting and design to operation and decommissioning, and in particular retrieval of spent fuel. Contents: 1. Introduction; 2. Protection of human health and the environment; 3. Roles and responsibilities; 4. Management system; 5. Safety case and safety assessment; 6. General safety considerations for storage of spent fuel. Appendix I: Specific safety considerations for wet or dry storage of spent fuel; Appendix II: Conditions for specific types of fuel and additional considerations; Annex: I: Short term and long term storage; Annex II: Operational and safety considerations for wet and dry spent fuel storage facilities; Annex III: Examples of sections of operating procedures for a spent fuel storage facility; Annex IV: Site conditions, processes and events for consideration in a safety assessment (external human induced phenomena); Annex V: Site conditions, processes and events for consideration in a safety assessment (external natural phenomena); Annex VI: Site conditions, processes and events for consideration in a safety assessment (external human induced phenomena); Annex VII: Postulated initiating events for consideration in a safety assessment (internal phenomena).

10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

Science.gov (United States)

2010-01-01

...; Ineligibility of Certain Applicants § 50.36a Technical specifications on effluents from nuclear power reactors..., including expected occurrences, as low as is reasonably achievable, each licensee of a nuclear power reactor... the design, construction, and operation of nuclear power reactors indicates that compliance with the...

Fabrication drawings of fuel pins for FUJI project among PSI, JNC and NRG. Revised version

International Nuclear Information System (INIS)

Ozawa, Takayuki; Nakazawa, Hiroaki; Abe, Tomoyuki; Nagayama, Masahiro

2002-02-01

Irradiation tests and post-irradiation examinations in the framework of JNC-PSI-NRG collaboration project will be performed in 2003-2005. Irradiation fuel pins will be fabricated by the middle of 2003. The fabrication procedure for irradiation fuel pins has been started in 2001. Several fabrication tests and qualification tests in JNC and PSI (Paul Scherrer Institute, Switzerland) have been performed before the fuel pin fabrication. According to the design assignment between PSI and JNC in the frame of this project, PSI should make a specification document for the fuel pellet, the sphere-pac fuel particles, the vipac fuel particles, and the fuel pin. JNC should make a fabrication drawing for irradiation pins. JNC has been performed the fuel design in cooperation with PSI and NRG (Nuclear Research and Consultancy Group, Netherlands). In this project, the pelletized fuel, the sphere-pac fuel, and the vipac fuel will be simultaneously irradiated on HFR (High Flux Reactor, Netherlands). This fabrication drawing has been made under the design assignment with PSI, and consists of the drawing of MOX pellet, thermal insulator pellet, pin components, fuel segments, and the constructed pin. The fabrication drawings were approved in October 2001, but after that, the optimization of specifications has been discussed and agreed among all partners. In this report, the revised fabrication drawings will be shown. Based on the commission of Plutonium Fuel Technology Group, Advanced Fuel Recycle Technology Division, this design work has been performed in Fuel Design and Evaluation Group, Plutonium Fuel Fabrication Division, Plutonium Fuel Center. (author)

ASME Code requirements for multi-canister overpack design and fabrication

International Nuclear Information System (INIS)

SMITH, K.E.

1998-01-01

The baseline requirements for the design and fabrication of the MCO include the application of the technical requirements of the ASME Code, Section III, Subsection NB for containment and Section III, Subsection NG for criticality control. ASME Code administrative requirements, which have not historically been applied at the Hanford site and which have not been required by the US Nuclear Regulatory Commission (NRC) for licensed spent fuel casks/canisters, were not invoked for the MCO. As a result of recommendations made from an ASME Code consultant in response to DNFSB staff concerns regarding ASME Code application, the SNF Project will be making the following modifications: issue an ASME Code Design Specification and Design Report, certified by a Registered Professional Engineer; Require the MCO fabricator to hold ASME Section III or Section VIII, Division 2 accreditation; and Use ASME Authorized Inspectors for MCO fabrication. Incorporation of these modifications will ensure that the MCO is designed and fabricated in accordance with the ASME Code. Code Stamping has not been a requirement at the Hanford site, nor for NRC licensed spent fuel casks/canisters, but will be considered if determined to be economically justified

Decision no. 2011-DC-0223 of the French nuclear safety authority from May 5, 2011, ordering the MELOX SA company to proceed to a complementary safety evaluation of some of its basic nuclear facilities in the eyes of the Fukushima Daiichi nuclear power plant accident

International Nuclear Information System (INIS)

2011-01-01

As a consequence of the accident of the Fukushima Daiichi nuclear power plant (Japan), the French Prime Minister entrusted the French nuclear safety authority (ASN) with the mission to carry out a safety analysis re-evaluation of the French nuclear facilities, and in particular the nuclear power plants. A decision has been addressed by the ASN to each nuclear operator with the specifications of this safety re-evaluation analysis and the list of facilities in concern. This document is the decision addressed to MELOX SA company, operator of the Melox MOX fuel fabrication plant of Marcoule (France). (J.S.)

Tissue specificity of the hormonal response in sex accessory tissues is associated with nuclear matrix protein patterns.

Science.gov (United States)

Getzenberg, R H; Coffey, D S

1990-09-01

The DNA of interphase nuclei have very specific three-dimensional organizations that are different in different cell types, and it is possible that this varying DNA organization is responsible for the tissue specificity of gene expression. The nuclear matrix organizes the three-dimensional structure of the DNA and is believed to be involved in the control of gene expression. This study compares the nuclear structural proteins between two sex accessory tissues in the same animal responding to the same androgen stimulation by the differential expression of major tissue-specific secretory proteins. We demonstrate here that the nuclear matrix is tissue specific in the rat ventral prostate and seminal vesicle, and undergoes characteristic alterations in its protein composition upon androgen withdrawal. Three types of nuclear matrix proteins were observed: 1) nuclear matrix proteins that are different and tissue specific in the rat ventral prostate and seminal vesicle, 2) a set of nuclear matrix proteins that either appear or disappear upon androgen withdrawal, and 3) a set of proteins that are common to both the ventral prostate and seminal vesicle and do not change with the hormonal state of the animal. Since the nuclear matrix is known to bind androgen receptors in a tissue- and steroid-specific manner, we propose that the tissue specificity of the nuclear matrix arranges the DNA in a unique conformation, which may be involved in the specific interaction of transcription factors with DNA sequences, resulting in tissue-specific patterns of secretory protein expression.

Typical IAEA operations at a fuel fabrication plant

International Nuclear Information System (INIS)

Morsy, S.

1984-01-01

The IAEA operations performed at a typical Fuel Fabrication Plant are explained. To make the analysis less general the case of Low Enriched Uranium (LEU) Fuel Fabrication Plants is considered. Many of the conclusions drawn from this analysis could be extended to other types of fabrication plants. The safeguards objectives and goals at LEU Fuel Fabrication Plants are defined followed by a brief description of the fabrication process. The basic philosophy behind nuclear material stratification and the concept of Material Balance Areas (MBA's) and Key Measurement Points (KMP's) is explained. The Agency operations and verification methods used during physical inventory verifications are illustrated

Characteristics and fabrication of cermet spent nuclear fuel casks: ceramic particles embedded in steel

Energy Technology Data Exchange (ETDEWEB)

Forsberg, C.W.; Swaney, P.M.; Tiegs, T.N. [Oak Ridge National Lab., Oak Ridge, TN (United States)

2004-07-01

Cermets are being investigated as an advanced material of construction for casks that can be used for storage, transport, or disposal of spent nuclear fuel (SNF). Cermets, which consist of ceramic particles embedded in steel, are a method to incorporate brittle ceramics with highly desirable properties into a strong ductile metal matrix with a high thermal conductivity, thus combining the best properties of both materials. Traditional applications of cermets include tank armor, vault armor, drill bits, and nuclear test-reactor fuel. Cermets with different ceramics (DUO{sub 2}, Al{sub 2}O{sub 3}, Gd{sub 2}O{sub 3}, etc.) are being investigated for the manufacture of SNF casks. Cermet casks offer four potential benefits: greater capacity (more SNF assemblies) for the same gross weight cask, greater capacity (more SNF assemblies) for the same external dimensions, improved resistance to assault, and superior repository performance. These benefits are achieved by varying the composition, volume fraction, and particulate size of the ceramic particles in the cermet with position in the cask body. Addition of depleted uranium dioxide (DUO{sub 2}) to the cermet increases shielding density, improves shielding effectiveness, and increases cask capacity for a given cask weight or size. Addition of low-density aluminium oxide (Al{sub 2}O{sub 3}) to the outer top and bottom sections of the cermet cask, where the radiation levels are lower, can lower cask weight without compromising shielding. The use of Al2O3 and other oxides, in appropriate locations, can increase resistance to assault. Repository performance may be improved by compositional control of the cask body to (1) create a local geochemical environment that slows the long-term degradation of the SNF and (2) enables the use of DUO{sub 2} for longterm criticality control. While the benefits of using cermets follow directly from their known properties, the primary challenge is to develop low-cost methods to fabricate

Chemical characterization of nuclear fuel materials

International Nuclear Information System (INIS)

Ramakumar, K.L.

2011-01-01

India is fabricating nuclear fuels for various types of reactors, for example, (U-Pu) MOX fuel of varying Pu content for boiling water reactors (BWRs), pressurized heavy water reactors (PHWRs), prototype fast breeder reactors (PFBRs), (U-Pu) carbide fuel fast breeder test reactor (FBTR), and U-based fuels for research reactors. Nuclear fuel being the heart of the reactor, its chemical and physical characterisation is an important component of this design. Both the fuel materials and finished fuel products are to be characterised for this purpose. Quality control (both chemical and physical) provides a means to ensure that the quality of the fabricated fuel conforms to the specifications for the fuel laid down by the fuel designer. Chemical specifications are worked out for the major and minor constituents which affect the fuel properties and hence its performance under conditions prevailing in an operating reactor. Each fuel batch has to be subjected to comprehensive chemical quality control for trace constituents, stoichiometry and isotopic composition. A number of advanced process and quality control steps are required to ensure the quality of the fuels. Further more, in the case of Pu-based fuels, it is necessary to extract maximum quality data by employing different evaluation techniques which would result in minimum scrap/waste generation of valuable plutonium. The task of quality control during fabrication of nuclear fuels of various types is both challenging and difficult. The underlying philosophy is total quality control of the fuel by proper mix of process and quality control steps at various stages of fuel manufacture starting from the feed materials. It is also desirable to adapt more than one analytical technique to increase the confidence and reliability of the quality data generated. This is all the most required when certified reference materials are not available. In addition, the adaptation of non-destructive techniques in the chemical quality

Analysis of general specifications for nuclear facilities environmental monitoring vehicles

International Nuclear Information System (INIS)

Xu Xiaowei

2014-01-01

At present, with the nuclear energy more increasingly extensive application, the continuous stable radiation monitoring has become the focus of the public attention. The main purpose of the environmental monitoring vehicle for the continuous monitoring of the environmental radiation dose rate and the radionuclides concentration in the medium around nuclear facilities is that the environmental radiation level and the radioactive nuclides activity in the environment medium are measured. The radioactive pollution levels, the scope contaminated and the trends of the pollution accumulation are found out. The change trends for the pollution are observed and the monitoring results are explained. The domestic demand of the environmental monitoring for the nuclear facilities is shown in this report. The changes and demands of the routine environmental monitoring and the nuclear emergency monitoring are researched. The revision opinions for EJ/T 981-1995 General specifications for nuclear facilities environmental monitoring vehicles are put forward. The purpose is to regulate domestic environmental monitoring vehicle technical criterion. The criterion makes it better able to adapt and serve the environmental monitoring for nuclear facilities. The technical guarantee is provided for the environmental monitoring of the nuclear facilities. (authors)

Plant Design Nuclear Fuel Element Production Capacity Optimization to Support Nuclear Power Plant in Indonesia

International Nuclear Information System (INIS)

Bambang Galung Susanto

2007-01-01

The optimization production capacity for designing nuclear fuel element fabrication plant in Indonesia to support the nuclear power plant has been done. From calculation and by assuming that nuclear power plant to be built in Indonesia as much as 12 NPP and having capacity each 1000 MW, the optimum capacity for nuclear fuel element fabrication plant is 710 ton UO 2 /year. The optimum capacity production selected, has considered some aspects such as fraction batch (cycle, n = 3), length of cycle (18 months), discharge burn-up value (Bd) 35,000 up 50,000 MWD/ton U, enriched uranium to be used in the NPP (3.22 % to 4.51 %), future market development for fuel element, and the trend of capacity production selected by advances country to built nuclear fuel element fabrication plant type of PWR. (author)

Technical specifications for the successful fabrication of laminated seismic isolation bearings

International Nuclear Information System (INIS)

Kulak, R.F.

1992-01-01

High damping laminated elastomeric bearings are becoming one of the preferred devices for isolating large buildings and structures. IN the United States, the current reference design for the Advanced Liquid Metal Reactor uses laminated bearings for seismic isolation. These bearing are constructed from alternating layers of rubber and steel plates. They are typically designed for shear strains between 50 to 100 percent and expected to sustain two to three times these levels for beyond design basis loading considerations. The technical specifications used to procure these bearings are an important factor in assuring that the bearings that are installed under nuclear structures meet the performance requirements of the design. The key aspects of the current version of the Technical Specifications are discussed in this paper

Porous nuclear fuel element for high-temperature gas-cooled nuclear reactors

Science.gov (United States)

Youchison, Dennis L [Albuquerque, NM; Williams, Brian E [Pacoima, CA; Benander, Robert E [Pacoima, CA

2011-03-01

Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

Fabrication Flaw Density and Distribution in Weld Repairs

International Nuclear Information System (INIS)

Doctor, Steven R.

2009-01-01

The Pacific Northwest National Laboratory (PNNL) is developing a generalized flaw distribution for the population of nuclear reactor pressure vessels and for piping welds in the U. S. operating reactors. The purpose of the generalized flaw distribution is to predict component-specific flaw densities. The estimates of fabrication flaws are intended for use in fracture mechanics structural integrity assessments. Structural integrity assessments, such as estimating the frequency of loss-of-coolant accidents, are performed by computer codes that require, as input, accurate estimates of flaw densities. Welds from four different cancelled reactor pressure vessels and a collection of archived pipes have been studied to develop empirical estimates of fabrication flaw densities. This paper describes the fabrication flaw distribution and characterization in the repair weld metal of vessels and piping. This work indicates that large flaws occur in these repairs which are complex in composition and sometimes include cracks on the ends of the repair cavities. Parametric analysis using an exponential fit is performed on the data. Construction records where available were reviewed. It is difficult to make conclusions due to the limited number of construction records reviewed. However, the records reviewed to date show a significant change in repair frequency over the years when the components in this study were fabricated. A description of repair flaw morphology is provided with a discussion of fracture mechanics significance.

Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Spanish Edition)

International Nuclear Information System (INIS)

2012-01-01

This publication is a revision of Safety Requirements No. NS-R-1, Safety of Nuclear Power Plants: Design. It establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in the design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Russian Edition)

International Nuclear Information System (INIS)

2012-01-01

This publication is a revision of Safety Requirements No. NS-R-1, Safety of Nuclear Power Plants: Design. It establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in the design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

Fuel Fabrication Capability Research and Development Plan

Energy Technology Data Exchange (ETDEWEB)

Senor, David J.; Burkes, Douglas

2013-06-28

The purpose of this document is to provide a comprehensive review of the mission of the Fuel Fabrication Capability (FFC) within the Global Threat Reduction Initiative (GTRI) Convert Program, along with research and development (R&D) needs that have been identified as necessary to ensuring mission success. The design and fabrication of successful nuclear fuels must be closely linked endeavors.

Maintenance and Fabrication of Nuclear Electronic Equipment

Energy Technology Data Exchange (ETDEWEB)

Chung, Chong Eun; Koo, In Soo; Hong, Seok Boong (and others)

2006-12-15

Based on the development of Hanaro RMS software, the RMS software system for another radiation facility(RI and IMEF) at KAERI will be developed next year. Development and its application of a wire-mesh sensor could contribute to the safe operation of nuclear power plants and could be used to develop other precision sensors for nuclear applications. Development of an ERMS could be used not only in nuclear facility but also in other radiation application institution such as acceleration facility etc.

Surveillance robot for nuclear power plants

International Nuclear Information System (INIS)

White, J.R.; Harvey, H.W.; Satterlee, P.E. Jr.

1985-01-01

A demonstration project to determine the feasibility and cost-effectiveness of mobile surveillance robots in nuclear power plants is being conducted by the Remote Technology Corporation (REMOTEC) for the US Nuclear Regulatory Commission. Phase I of the project was completed in March 1984 and included a survey of currently used robotic equipment and the development of a robotics application methodology. Three Tennessee Valley Authority plants were analyzed to identify specific plant areas with a high potential for surveillance robotics. Based on these results, a number of robotic system applications were prepared and evaluated for cost-effectiveness. The system with the highest potential, a mobile surveillance robot, was selected for fabrication and in-plant demonstration testing in a phase II effort. The design, fabrication, and assembly of SURBOT has been completed and cold testing is in process. It will be installed at the Browns Ferry Nuclear Plant early in 1986 for demonstration testing. Current projections are that SURBOT can be used in approx.40 rooms within the auxiliary building and will have annual savings of over 100 person-rem exposure, 1000 sets of C-zone clothing, and 1000 person-hours of labor

Nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Nakano, H [Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan)

1976-10-01

It is expected that nuclear power generation will reach 49 million kW in 1985 and 129 million kW in 1995, and the nuclear fuel having to be supplied and processed will increase in proportion to these values. The technical problems concerning nuclear fuel are presented on the basis of the balance between the benefit for human beings and the burden on the human beings. Recently, especially the downstream of nuclear fuel attracts public attention. Enriched uranium as the raw material for light water reactor fuel is almost monopolized by the U.S., and the technical information has not been published for fear of the diversion to nuclear weapons. In this paper, the present situations of uranium enrichment, fuel fabrication, transportation, reprocessing and waste disposal and the future problems are described according to the path of nuclear fuel cycle. The demand and supply of enriched uranium in Japan will be balanced up to about 1988, but afterwards, the supply must rely upon the early establishment of the domestic technology by centrifugal separation method. No problem remains in the fabrication of light water reactor fuel, but for the fabrication of mixed oxide fuel, the mechanization of the production facility and labor saving are necessary. The solution of the capital risk for the construction of the second reprocessing plant is the main problem. Japan must develop waste disposal techniques with all-out efforts.

Readiness Review of BWXT for Fabrication of AGR 5/6/7 Compacts

Energy Technology Data Exchange (ETDEWEB)

Marshall, Douglas William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sharp, Michelle Tracy [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2017-02-01

In support of preparations for fabricating compacts for the Advanced Gas Reactor (AGR) fuel qualification irradiation experiments (AGR-5/6/7), Idaho National Laboratory (INL) conducted a readiness review of the BWX Technology (BWXT) procedures, processes, and equipment associated with compact fabrication activities at the BWXT Nuclear Operations Group (BWXT-NOG) facility outside Lynchburg, VirginiaVA. The readiness review used quality assurance requirements taken from the American Society of Mechanical Engineers (ASME) Nuclear Quality Assurance Standard (NQA-1-2008/1a-2009) as a basis to assess readiness to start compact fabrication.

OPO fabric decontamination

International Nuclear Information System (INIS)

Severa, J.; Bar, J.; Grujbar, V.

1978-01-01

Samples of five polypropylene-based man-made fabrics were studied with regard to the degree of contamination and possibilities of decontamination in order to assess their suitability as material for protective clothing in the nuclear industry. The contamination degree of the fabrics in an aqueous solution of a fission product mixture was found to be low. Soaking in a mixture of the Sapon detergent and sodium hexametaphosphate at a concentration of both materials of 1 g/l with subsequent washing in a solution of the Zenit detergent at a concentration of 3 g/l was suggested as the most suitable decontamination procedure. It reduces the initial contamination by almost 99%. (Z.M.)

The quality challenge for fuel fabrication

International Nuclear Information System (INIS)

Lannegrace, J.-P.

1990-01-01

Fuel fabrication is a key segment of the nuclear fuel cycle, since safe and economic operation of reactors is highly dependent on the quality of the fuel. Achieving and controlling quality is, therefore, of paramount importance to fuel fabricators dominating nearly every aspect of the business. The quality policy, concepts and assurance system at three French plants are outlined. The need for integrated inspection, process optimization and good employee motivation is stressed. (author)

Safety of Nuclear Power Plants: Design. Specific Safety Requirements

International Nuclear Information System (INIS)

2012-01-01

On the basis of the principles included in the Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, this Safety Requirements publication establishes requirements applicable to the design of nuclear power plants. It covers the design phase and provides input for the safe operation of the power plant. It elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

Remote fabrication of nuclear fuel: a secure automated fabrication overview

International Nuclear Information System (INIS)

Nyman, D.H.; Benson, E.M.; Yatabe, J.M.; Nagamoto, T.T.

1981-01-01

An automated line for the fabrication of breeder reactor fuel pins is being developed. The line will be installed in the Fuels and Materials Examination Facility (FMEF) presently under construction at the Hanford site near Richland, Washington. The application of automation and remote operations to fuel processing technology is needed to meet program requirements of reduced personnel exposure, enhanced safeguards, improved product quality, and increased productivity. Commercially available robots are being integrated into operations such as handling of radioactive material within a process operation. These and other automated equipment and chemistry analyses systems under development are described

Foil fabrication for the ROMANO event. Revision 1

International Nuclear Information System (INIS)

Romo, J.G. Jr.; Weed, J.W.; Griggs, G.E.; Brown, T.G.; Tassano, P.L.

1984-01-01

The Vacuum Processes Lab (VPL), of LLNL's M.E. Dept. - Material Fabrication Division (MFD), conducted various vacuum related support activities for the ROMANO nuclear physics experiment. This report focuses on the foil fabrication activities carried out between July and November 1983 for the ROMANO event. Other vacuum related activities for ROMANO, such as outgassing tests of materials, are covered in separate documentation. VPL was asked to provide 270 coated Parylene foils for the ROMANO event. However, due to the developmental nature of some of the procedures, approximately 400 coated foils were processed. In addition, VPL interacted with MFD's Plastics Shop to help supply Parylene substrates to other organizations (i.e., LBL and commercial vendors) which had also been asked to provide coated foils for ROMANO. The purposes of this report are (A) to document the processes developed and the techniques used to produce the foils, and (B) to suggest future directions. The report is divided into four sections describing: (1) nuclear target foil fabrication, (2) Parylene substrate preparation and production, (3) calibration foil fabrication, and (4) foil and substrate inspections

Evaluation of Watts Bar Nuclear Plant Unit 1 Technical Specifications

International Nuclear Information System (INIS)

Baxter, D.E.; Bruske, S.J.

1985-08-01

This document was prepared for the Nuclear Regulatory Commission (NRC) to assist them in determining whether the Watts Bar Nuclear Plant Unit 1 Technical Specifications (T/S), which govern plant systems configurations and operations, are in conformance with the assumption of the Final Safety Analysis Report (FSAR) as amended, and the requirements of the Safety Evaluation Report (SER) as supplemented. A comparative audit of the FSAR as amended, and the SER as supplemented was performed with the Watts Bar T/S. Several discrepancies were identified and subsequently resolved through discussions with the cognizant NRC reviewer, NRC staff reviewers and/or utility representatives. The Watts Bar Nuclear Plant Unit 1 T/S, to the extent reviewed, are in conformance with the FSAR and SER

Evaluation of Shoreham Nuclear Power Station, Unit 1 technical specifications

International Nuclear Information System (INIS)

Baxter, D.E.; Bruske, S.J.

1985-08-01

This document was prepared for the Nuclear Regulatory Commission (NRC) to assist them in determining whether the Shoreham Nuclear Power Station Unit 1 Technical Specifications (T/S), which govern plant systems configurations and operations, are in conformance with the assumptions of the Final Safety Analysis Report (FSAR) as amended, and the requirements of the Safety Evaluation Report (SER) as supplemented. A comparative audit of the FSAR as amended, and the SER as supplemented was performed with the Shoreham T/S. Several discrepancies were identified and subsequently resolved through discussions with the cognizant NRC reviewer, NRC staff reviewers and/or utility representatives. The Shoreham Nuclear Power Station Unit 1 T/S, to the extent reviewed, are in conformance with the FSAR and SER

Standard specification for nuclear-grade silver-indium-cadmium alloy

CERN Document Server

American Society for Testing and Materials. Philadelphia

2003-01-01

1.1 This specification covers silver-indium-cadmium alloy for use as a control material in light-water nuclear reactors. 1.2 The scope of this specification excludes the use of this material in applications where material strength of this alloy is a prime requisite. Also, this material must be protected from the primary water by a corrosion and wear resistant cladding. 1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.

Development of PHWR fuel fabrication in Korea

International Nuclear Information System (INIS)

Suh, K.S.; Yang, M.S.; Kim, D.H.; Rim, C.S.

1988-01-01

Korea Advanced Energy Research Institute (KAERI) started a research project to develop the PHWR (CANDU) nuclear fuel fabrication technology in 1981. Based on the results of the intensive developmental work, several prototype fuel bundles were fabricated and tested in the Hot Test Loop at KAERI continuously in 1983 and 1984. After that, irradiation test and post-irradiation examination were carried out for two KAERI-made fuel bundles at Chalk River Nuclear Laboratories in Canada in 1984. Since the results of in-pile and out-of-pile tests with prototype fuel bundles proved to be satisfactory, 48 additional fuel bundles were loaded in Wolsung reactor (CANDU) in 1984 and 1985, and all of them were discharged without a defect after excellent performance in the power reactor. In 1985, the Korean government decided that KAERI supplies all the fuel necessary for the Wolsung reactor. For the mass production of nuclear fuel bundle, several process equipment, facilities and automation methods have been improved making use of experience accumulated during research. A quality assurance program was also established, and quality inspection technology was reviewed and improved to fit the mass production. This paper deals with the development experience so far obtained with the design and fabrication of the Korean PHWR fuel

Radiation exposure and cause specific mortality among nuclear workers in Belgium (1969-1994)

International Nuclear Information System (INIS)

Engels, H.; Swaen, G. M. H.; Slangen, J.; Van Amersvoort, L.; Holmstock, L.; Van Mieghem, E.; Van Regenmortel, I.; Wambersie, A.

2005-01-01

Cause specific mortality was studied in nuclear workers from five nuclear facilities in Belgium and compared to the general population. For the 1969-1994 period, mortality in male nuclear workers is significantly lower for all causes of death and for all cancer deaths. The same conclusions are reached if one assumes a latency period of 20 y between the first irradiation and cancer induction. In female workers, mortality due to all causes and all cancer deaths is not different from that of the general population. Analysis of cause specific mortality was performed for male and female workers for three endpoints: specific cancer sites, cardiovascular and respiratory diseases. No significant increase in mortality was observed. In male workers, the influence of cumulative dose was also investigated using four dose levels: No significant correlation was found. Smoking habits may be a confounding factor in smoking related health conditions. (authors)

Nuclear power and the nuclear fuel cycle

International Nuclear Information System (INIS)

Scurr, I.F.; Silver, J.M.

1990-01-01

Australian Nuclear Science and Technology Organization maintains an ongoing assessment of the world's nuclear technology developments, as a core activity of its Strategic Plan. This publication reviews the current status of the nuclear power and the nuclear fuel cycle in Australia and around the world. Main issues discussed include: performances and economics of various types of nuclear reactors, uranium resources and requirements, fuel fabrication and technology, radioactive waste management. A brief account of the large international effort to demonstrate the feasibility of fusion power is also given. 11 tabs., ills

Hydrothermal synthesis for fabrication and reprocessing of MOX nuclear fuel

International Nuclear Information System (INIS)

Ohta, Suguru; Yamamura, Tomoo; Shirasaki, Kenji; Satoh, Isamu; Shikama, Tatsuo

2011-01-01

To improve the nuclear proliferation resistance and to minimize use of chemicals, a new reprocessing and fabrication process of 'mixed oxide' (MOX) fuel was proposed and studied by using simulated spent fuel solutions. The process is consisting of the two steps, i.e. the removal of fission product (FP) from dissolved spent fuel by using carbonate solutions (Step-1), and hydrothermal synthesis of uranium dioxides (Step-2). In Step-1, rare earth (the precipitation ratio: 90%) and alkaline earth (10-50% for Sr) as FP were removed based on their low solubility of hydroxides and carbonate salts, with uranium kept dissolved for the certain carbonate solutions of weak base (Type 2) or mixtures of relatively strong base and weak base (Type 3). In Step-2, the features of uranium dioxides UO 2+x particles, i.e. stoichiometry (x=0.05-0.2), size (0.2-3 μm) and shape (cubic, spherical, rectangular parallelpiped, etc.), were controlled, and the cesium was removed down to 40 ppm by an addition of organic additives. The decontamination factors (DF) for cesium exceeds 10 5 , whereas the total DF of all the simulated FP were as low as the order of 10 which requires future studies for removal of alkaline earth, Re and Tc etc. (author)

OSMOSE experiment: high minor actinides contents pellets and pins fabrication

Energy Technology Data Exchange (ETDEWEB)

Jankowiak, A.; Leorier, C.; Desmouliere, F.; Donnet, L. [Commissariat a l' Energie Atomique, CEA/DEN/VRH/DTEC/SDTC/LEMA, 30207 Bagnols-sur-Ceze cedex (France); Antony, M. [Commissariat a l' Energie Atomique, CEA/DEN/CAD/DER/SPEX/LPE, 13108 St Paul Lez Durance cedex (France); Bernard, D. [Commissariat a l' Energie Atomique, CEA/DEN/ CAD/DER /SPRC/LEPh, 13108 St Paul Lez Durance cedex (France)

2008-07-01

The OSMOSE program aims to provide accurate experimental data on integral neutron cross-sections of isotopes (i.e.: Th{sup 232}, U{sup 233}, U{sup 234}, U{sup 235}, U{sup 236}, U{sup 238}, Np{sup 237}, Pu{sup 238}, Pu{sup 239}, Pu{sup 240}, Pu{sup 241}, Pu{sup 242}, Am{sup 241}, Am{sup 243}, Cm{sup 244} and Cm{sup 245}). The study of these nuclides is performed on a large range of neutron spectra corresponding to specific experimental conditions (thermal, epithermal, moderated/fast, and fast spectra). This program will be used to provide guidance to all nuclear data programs in the world. This program has led to an optimized fabrication process for OSMOSE pellets and pins which were fabricated by the LEMA (Actinide based Materials Study Laboratory) in the ATALANTE facility both in glove box and shielded cell. The fabrication process made possible to obtain the required material characteristics including a high density, a good distribution of the isotopes in the uranium oxide matrices. A particular attention was paid to reduce chemical pollution of the samples. The program has been successfully achieved in July 2007 with the fabrication of the last two Cm doped samples. (authors)

Safety in manufacturing of nuclear fuel

International Nuclear Information System (INIS)

Daste, Bernard

1980-01-01

Production of low enriched uranium fuel raises specific safety problems resulting from the very nature of the manufacturing process as from the industrial size generally given to the new facilities for this kind of production. The author exposes the experience so far acquired by F.B.F.C. (Societe franco-belge de fabrication du combustible) which is making important investments in order to meet the fuel needs of the French nuclear programme. After a short description of the fuel and the principal stages of its production, he analyses the potential nuclear hazards of the F.B.F.C. facilities operation and the adequate safety measures taken [fr

Geomorphologic specificities of selected sites for nuclear power plants in Czechoslovakia

International Nuclear Information System (INIS)

Kalvoda, J.; Demek, J.

1991-01-01

The contribution of geomorphology to the complex evaluation of properties of sites for the construction and operation of nuclear facilities is demonstrated. The unique manifestation of the present geodynamics at the Jaslovske Bohunice nuclear power plant locality and the spatial correlations of annals of the specific morphotectonic development of georeliefs of that nuclear power plant with the location of the epicentral earthquake zones are shown. The results of the geomorphological survey in the surroundings of the Temelin nuclear power plant construction site are described and a drawing is reproduced showing how the georelief of this locality divides into areas with different categories of occurrence of morpho-structural formations. For the Tetov locality, where the construction of a nuclear power plant is planned, the changes in the course of the Labe (Elbe) river which occurred in the Pleistocene are of importance in the assessment of the intensity of geodynamic processes. The geomorphological and geotectonic complexity of the planned Blahutovice nuclear power plant construction site is demonstrated. A drawing shows the morphotectonic situation in the surroundings of that construction site. (Z.S.). 4 figs

Development of uranium reduction system for incineration residue generated at LWR nuclear fuel fabrication plants in Japan

International Nuclear Information System (INIS)

Sampei, T.; Sato, T.; Suzuki, N.; Kai, H.; Hirata, Y.

1993-01-01

The major portion of combustible solid wastes generated at LWR nuclear fuel fabrication plants in Japan is incinerated and stored in a warehouse. The uranium content in the incineration residue is higher compared with other categories of wastes, although only a small amount of incineration residue is generated. Hence, in the future uranium should be removed from incineration residues before they are reduced to a level appropriate for the final disposal. A system for processing the incineration residue for uranium removal has been developed and tested based on the information obtained through laboratory experiments and engineering scale tests

Evaluation of Perry Nuclear Power Plant Unit 1 technical specifications

International Nuclear Information System (INIS)

Baxter, D.E.; Bruske, S.J.

1985-11-01

This document was prepared for the Nuclear Regulatory Commission (NRC) to assist them in determining whether the Perry Nuclear Power Plant Unit 1 Technical Specifications (T/S), which govern plant systems configurations and operations, are in conformance with the requirements of the Final Safety Analysis Report (FSAR) as amended, and the requirements of the Safety Evaluation Report (SER) as supplemented. A comparative audit of the FSAR as amended, and the SER as supplemented was performed with the Perry T/S. Several discrepancies were identified and subsequently resolved through telephone conversations with the staff reviewer and the utility representative. Pending completion of the resolutions noted in Parts 3 and 4 of this report, the Perry Nuclear Power Plant Unit 1 T/S, to the extent reviewed, are in conformance with the FSAR and SER

Role of analytical chemistry in the development of nuclear fuels

International Nuclear Information System (INIS)

Ramakumar, K.L.

2012-01-01

Analytical chemistry is indispensable and plays a pivotal role in the entire gamut of nuclear fuel cycle activities starting from ore refining, conversion, nuclear fuel fabrication, reactor operation, nuclear fuel reprocessing to waste management. As the fuel is the most critical component of the reactor where the fissions take place to produce power, extreme care should be taken to qualify the fuel. For example, in nuclear fuel fabrication, depending upon the reactor system, selection of nuclear fuel has to be made. The fuel for thermal reactors is normally uranium oxide either natural or slightly enriched. For research reactors it can be uranium metal or alloy. The fuel for FBR can be metal, alloy, oxide, carbide or nitride. India is planning an advanced heavy water reactor for utilization of vast resources of thorium in the country. Also research is going on to identify suitable metallic/alloy fuels for our future fast reactors and possible use in fast breeder test reactor. Other advanced fuel materials are also being investigated for thermal reactors for realizing increased performance levels. For example, advanced fuels made from UO 2 doped with Cr 2 O 3 and Al 2 O 3 are being suggested in LWR applications. These have shown to facilitate pellet densification during sintering and enlarge the pellet grain size. The chemistry of these materials has to be understood during the preparation to the stringent specification. A number of analytical parameters need to be determined as a part of chemical quality control of nuclear materials. Myriad of analytical techniques starting from the classical to sophisticated instrumentation techniques are available for this purpose. Insatiable urge of the analytical chemist enables to devise and adopt new superior methodologies in terms of reduction in the time of analysis, improvement in the measurement precision and accuracy, simplicity of the technique itself etc. Chemical quality control provides a means to ensure that the

MICRO-SEISMOMETERS VIA ADVANCED MESO-SCALE FABRICATION

Energy Technology Data Exchange (ETDEWEB)

Garcia, Caesar A; Onaran, Guclu; Avenson, Brad; Hall, Neal

2014-11-07

The Department of Energy (DOE) and the National Nuclear Security Administration (NNSA) seek revolutionary sensing innovations for the monitoring of nuclear detonations. Performance specifications are to be consistent with those obtainable by only an elite few products available today, but with orders of magnitude reduction in size, weight, power, and cost. The proposed commercial innovation calls upon several technologies including the combination of meso-scale fabrication and assembly, photonics-based displacement / motion detection methods, and the use of digital control electronics . Early Phase II development has demonstrated verified and repeatable sub 2ng noise floor from 3Hz to 100Hz, compact integration of 3-axis prototypes, and robust deployment exercises. Ongoing developments are focusing on low frequency challenges, low power consumption, ultra-miniature size, and low cross axis sensitivity. We are also addressing the rigorous set of specifications required for repeatable and reliable long-term explosion monitoring, including thermal stability, reduced recovery time from mass re-centering and large mechanical shocks, sensitivity stability, and transportability. Successful implementation will result in small, hand-held demonstration units with the ability to address national security needs of the DOE/NNSA. Additional applications envisioned include military/defense, scientific instrumentation, oil and gas exploration, inertial navigation, and civil infrastructure monitoring.

Development of IAEA safeguards at low enrichment uranium fuel fabrication plants

International Nuclear Information System (INIS)

Badawy, I.

1988-01-01

In this report the nuclear material at low enrichment uranium fuel fabrication plants under IAEA safeguards is studied. The current verification practices of the nuclear material and future improvements are also considered. The problems met during the implementation of the the verification measures of the nuclear material - particularly for the fuel assemblies are discussed. The additional verification activities as proposed for future improvements are also discussed including the physical inventory verification and the verification of receipts and shipments. It is concluded that the future development of the present IAEA verification practices at low enrichment uranium fuel fabrication plants would necessitate the application of quantitative measures of the nuclear material and the implementation of advanced measurement techniques and instruments. 2 fig., 4 tab

Risks, costs and benefits analysis for exhumation of buried radioactive materials at a nuclear fuel fabrication facility

International Nuclear Information System (INIS)

Kirk, J.S.; Moore, R.A.; Huston, T.E.

1996-01-01

A Risks, Costs and Benefits analysis provides a tool for selecting a cost-effective remedial action alternative. This analysis can help avoid transferring risks to other populations and can objectively measure the benefits of a specific remedial action project. This paper describes the methods and results of a Risks, Costs and Benefits analysis performed at a nuclear fuel fabrication facility. The analysis examined exhuming and transporting radioactive waste to an offsite disposal facility. Risks evaluated for the remedial action project were divided into two categories: risks posed to the worker and risks posed to public health. Risks to workers included exposure to radioactive contaminants during excavation and packaging of waste materials and the use of heavy machinery. Potential public health risks included exposure to radioactive materials during transport from the exhumation site to the disposal facility. Methods included use of site-specific and published data, and existing computer models. Occupational risks were quantified using data from similar onsite remedial action projects. Computer modeling was used to evaluate public health risks from transporting radioactive materials; the consequences or probability of traffic accidents; and radiation exposure to potential inhabitants occupying the site considering various land use scenarios. A costs analysis was based on data obtained from similar onsite remedial action projects. Scenarios used to identify benefits resulting from the remedial action project included (1) an evaluation of reduction in risks to human health; (2) cost reductions associated with the unrestricted release of the property; and (3) benefits identified by evaluating regulatory mandates applicable to decommissioning. This paper will provide an overview of the methods used and a discussion of the results of a Risks, Costs and Benefits analysis for a site-specific remedial action scenario

Review of qualifications for fuel assembly fabrication

International Nuclear Information System (INIS)

Slabu, Dan; Zemek, Martin; Hellwig, Christian

2013-01-01

The required quality of nuclear fuel in industrial production can only be assured by applying processes in fabrication and inspection, which are well mastered and have been proven by an appropriate qualification. The present contribution shows the understanding and experiences of Axpo with respect to qualifications in the frame of nuclear fuel manufacturing and reflects some related expectations of the operator. (orig.)

Electromagnetic and nuclear radiation detector using micromechanical sensors

Science.gov (United States)

Thundat, Thomas G.; Warmack, Robert J.; Wachter, Eric A.

2000-01-01

Electromagnetic and nuclear radiation is detected by micromechanical sensors that can be coated with various interactive materials. As the micromechanical sensors absorb radiation, the sensors bend and/or undergo a shift in resonance characteristics. The bending and resonance changes are detected with high sensitivity by any of several detection methods including optical, capacitive, and piezoresistive methods. Wide bands of the electromagnetic spectrum can be imaged with picoJoule sensitivity, and specific absorptive coatings can be used for selective sensitivity in specific wavelength bands. Microcantilevers coated with optical cross-linking polymers are useful as integrating optical radiation dosimeters. Nuclear radiation dosimetry is possible by fabricating cantilevers from materials that are sensitive to various nuclear particles or radiation. Upon exposure to radiation, the cantilever bends due to stress and its resonance frequency shifts due to changes in elastic properties, based on cantilever shape and properties of the coating.

The sources of the specificity of nuclear law and environmental law

International Nuclear Information System (INIS)

Rainaud, J.M.; Cristini, R.

1983-01-01

This paper analyses the sources of the specificity of nuclear law and its relationship with environmental law as well as with ordinary law. The characteristics of nuclear law are summarized thus: recent discovery of the atom's uses and mandatory protection against its effects; internationalization of its use, leading to a limitation of national authorities competence. Several international treaties are cited (Antarctic Treaty, NPT, London Dumping Convention etc.) showing the link between radiation protection and the environment. (NEA) [fr

Fabrication of high specificity hollow mesoporous silica nanoparticles assisted by Eudragit for targeted drug delivery.

Science.gov (United States)

She, Xiaodong; Chen, Lijue; Velleman, Leonora; Li, Chengpeng; Zhu, Haijin; He, Canzhong; Wang, Tao; Shigdar, Sarah; Duan, Wei; Kong, Lingxue

2015-05-01

Hollow mesoporous silica nanoparticles (HMSNs) are one of the most promising carriers for effective drug delivery due to their large surface area, high volume for drug loading and excellent biocompatibility. However, the non-ionic surfactant templated HMSNs often have a broad size distribution and a defective mesoporous structure because of the difficulties involved in controlling the formation and organization of micelles for the growth of silica framework. In this paper, a novel "Eudragit assisted" strategy has been developed to fabricate HMSNs by utilising the Eudragit nanoparticles as cores and to assist in the self-assembly of micelle organisation. Highly dispersed mesoporous silica spheres with intact hollow interiors and through pores on the shell were fabricated. The HMSNs have a high surface area (670 m(2)/g), small diameter (120 nm) and uniform pore size (2.5 nm) that facilitated the effective encapsulation of 5-fluorouracil within HMSNs, achieving a high loading capacity of 194.5 mg(5-FU)/g(HMSNs). The HMSNs were non-cytotoxic to colorectal cancer cells SW480 and can be bioconjugated with Epidermal Growth Factor (EGF) for efficient and specific cell internalization. The high specificity and excellent targeting performance of EGF grafted HMSNs have demonstrated that they can become potential intracellular drug delivery vehicles for colorectal cancers via EGF-EGFR interaction. Copyright © 2014 Elsevier Inc. All rights reserved.

Artificial vision in nuclear fuel fabrication

International Nuclear Information System (INIS)

Dorado, P.

2007-01-01

The development of artificial vision techniques opens a door to the optimization of industrial processes which the nuclear industry cannot miss out on. Backing these techniques represents a revolution in security and reliability in the manufacturing of a highly technological products as in nuclear fuel. Enusa Industrias Avanzadas S. A. has successfully developed and implemented the first automatic inspection equipment for pellets by artificial vision in the European nuclear industry which is nowadays qualified and is already developing the second generation of this machine. There are many possible applications for the techniques of artificial vision in the fuel manufacturing processes. Among the practices developed by Enusa Industrias Avanzadas are, besides the pellets inspection, the rod sealing drills detection and positioning in the BWR products and the sealing drills inspection in the PWR fuel. The use of artificial vision in the arduous and precise processes of full inspection will allow the absence of human error, the increase of control in the mentioned procedures, the reduction of doses received by the personnel, a higher reliability of the whole of the operations and an improvement in manufacturing costs. (Author)

Characterization of germ cell-specific expression of the orphan nuclear receptor, germ cell nuclear factor.

Science.gov (United States)

Katz, D; Niederberger, C; Slaughter, G R; Cooney, A J

1997-10-01

Nuclear receptors, such as those for androgens, estrogens, and progesterones, control many reproductive processes. Proteins with structures similar to these receptors, but for which ligands have not yet been identified, have been termed orphan nuclear receptors. One of these orphans, germ cell nuclear factor (GCNF), has been shown to be germ cell specific in the adult and, therefore, may also participate in the regulation of reproductive functions. In this paper, we examine more closely the expression patterns of GCNF in germ cells to begin to define spatio-temporal domains of its activity. In situ hybridization showed that GCNF messenger RNA (mRNA) is lacking in the testis of hypogonadal mutant mice, which lack developed spermatids, but is present in the wild-type testis. Thus, GCNF is, indeed, germ cell specific in the adult male. Quantitation of the specific in situ hybridization signal in wild-type testis reveals that GCNF mRNA is most abundant in stage VII round spermatids. Similarly, Northern analysis and specific in situ hybridization show that GCNF expression first occurs in testis of 20-day-old mice, when round spermatids first emerge. Therefore, in the male, GCNF expression occurs postmeiotically and may participate in the morphological changes of the maturing spermatids. In contrast, female expression of GCNF is shown in growing oocytes that have not completed the first meiotic division. Thus, GCNF in the female is expressed before the completion of meiosis. Finally, the nature of the two different mRNAs that hybridize to the GCNF complementary DNA was studied. Although both messages contain the DNA binding domain, only the larger message is recognized by a probe from the extreme 3' untranslated region. In situ hybridization with these differential probes demonstrates that both messages are present in growing oocytes. In addition, the coding region and portions of the 3' untranslated region of the GCNF complementary DNA are conserved in the rat.

Fabrication of thermo-responsive cotton fabrics using poly(vinyl caprolactam-co-hydroxyethyl acrylamide) copolymer.

Science.gov (United States)

Xiao, Min; González, Edurne; Monterroza, Alexis Martell; Frey, Margaret

2017-10-15

A thermo-responsive polymer with hydrophilic to hydrophobic transition behavior, poly(vinyl caprolactam-co-hydroxyethyl acrylamide) P(VCL-co-HEAA), was prepared by copolymerization of vinyl caprolactam and N-hydroxyethyl acrylamide via free radical solution polymerization. The resulting copolymer was characterized by Fourier transform infrared spectroscopy (FTIR), 1 H nuclear magnetic resonance (NMR), gel permeation chromatography (GPC), differential scanning calorimetry (DSC), and thermogravimetric analysis (TGA). The lower critical solution temperature (LCST) of P(VCL-co-HEAA) was determined at 34.5°C. This thermo-responsive polymer was then grafted onto cotton fabrics using 1,2,3,4-butanetetracarboxylic acid (BTCA) as crosslinker and sodium hypophosphite (SHP) as catalyst. FTIR and energy dispersive X-ray spectroscopy (EDS) studies confirmed the successful grafting reaction. The modified cotton fabric exhibited thermo-responsive behavior as evidenced by water vapor permeability measurement confirming decreased permeability at elevated temperature. This is the first demonstration that a PVCL based copolymer is grafted to cotton fabrics. This study provides a new thermo-responsive polymer for fabrication of smart cotton fabrics with thermally switchable hydrophilicity. Copyright © 2017 Elsevier Ltd. All rights reserved.

Essential Specification Elements for Heat Exchanger Replacement

Energy Technology Data Exchange (ETDEWEB)

Bower, L.

2015-07-01

Performance upgrade and equipment degradation are the primary impetuses for a nuclear power plant to engage in the large capital cost project of heat exchanger replacement. Along with attention to these issues, consideration of heat exchanger Codes and Standards, material improvements, thermal redesign, and configuration are essential for developing User’s Design Specifications for successful replacement projects. The User’s Design Specification is the central document in procuring ASME heat exchangers. Properly stated objectives for the heat exchanger replacement are essential for obtaining the materials, configurations and thermal designs best suited for the nuclear power plant. Additionally, the code of construction required and the applied manufacturing standard (TEMA or HEI) affects how the heat exchanger may be designed or configured to meet the replacement goals. Knowledge of how Codes and Standards affect design and configuration details will aid in writing the User’s Design Specification. Joseph Oat Corporation has designed and fabricated many replacement heat exchangers for the nuclear power industry. These heat exchangers have been constructed per ASME Section III to various Code-Years or ASME Section VIII-1 to the current Code-Year also in accordance with TEMA and HEI. These heat exchangers have been a range of like-for-like replacement to complete thermal, material and configuration redesigns. Several examples of these heat exchangers with their Code, Standard and specification implications are presented. (Author.

The difficulties of writing procurement specifications for robots in nuclear applications

International Nuclear Information System (INIS)

Moore, F.W.; Bowen, W.W.

1986-01-01

The commercial robots available today were developed to primarily support the automotive or electronics industries. The adaptation of these robots and the current robotic technology to handle and manufacture nuclear materials has had its problems. The operational space and maintenance constraints have special consideration. The robotic systems of today tend to not have the payload capability for nuclear applications or, if the payload is sufficient, the system is very large and has several operating and maintenance accessibility requirements. The process of specifying, purchasing, and modifying a robotic system is an expensive and time-consuming process. The procurement specification is critical to obtaining competitive quotations on robots for nuclear applications resulting in the most economical robotic system

Modelling the Effect of Weave Structure and Fabric Thread Density on Mechanical and Comfort Properties of Woven Fabrics

Directory of Open Access Journals (Sweden)

Maqsood Muhammad

2016-09-01

Full Text Available The paper investigates the effects of weave structure and fabric thread density on the comfort and mechanical properties of various test fabrics woven from polyester/cotton yarns. Three different weave structures, that is, 1/1 plain, 2/1 twill and 3/1 twill, and three different fabric densities were taken as input variables whereas air permeability, overall moisture management capacity, tensile strength and tear strength of fabrics were taken as response variables and a comparison is made of the effect of weave structure and fabric density on the response variables. The results of fabric samples were analysed in Minitab statistical software. The coefficients of determinations (R-sq values of the regression equations show a good predictive ability of the developed statistical models. The findings of the study may be helpful in deciding appropriate manufacturing specifications of woven fabrics to attain specific comfort and mechanical properties.

Inspection of nuclear reactor welding by acoustic emission

International Nuclear Information System (INIS)

Prine, D.W.

1977-01-01

The objective of the work described is to evaluate in-process acoustic emission weld monitoring with the goal of upgrading the inspection techniques for assuring better weld quality in nuclear reactor piping and pressure vessels. To accomplish this overall objective, the following specific goals have been set within a three year program: (1) prove the feasibility of in-process AE under shop conditions and validate the AE findings by currently acceptable NDE techniques; (2) develop and build a prototype monitor for use in Nuclear Fabrication shops; and (3) provide data for NRC/ASME acceptance of the inspection practice

Porous nuclear fuel element with internal skeleton for high-temperature gas-cooled nuclear reactors

Science.gov (United States)

Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

2013-09-03

Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

Iran's nuclear programme

International Nuclear Information System (INIS)

Boureston, J.; Marvin, B.

2004-01-01

Iran's nuclear program is discussed, activity of enterprises connected with the nuclear industry of the country is evaluated. IAEA initiated inspection of some industrial sites with the aim of data acquisition about nuclear developments of Iran. Uranium ore mining and reducing to small size, uranium conversion, uranium enrichment, fabrication of nuclear fuel, production of plutonium: plant of heavy water production, spent fuel reprocessing are discussed [ru

A performance study of radio-opaque personal protective fabrics for the reduction of transmittance of gamma-rays and neutrons

International Nuclear Information System (INIS)

Corcoran, E.C.; William Forest; Robert Horton; Kelly, D.G.; Kristine Mattson; Curtis McDonald; Nielsen, K.S.; Kristin Topping; Weir, R.D.; Andre Yonkeu

2012-01-01

Commercial radio-opaque combat (CRC) fabrics, for incorporation into personal protective equipment used by first responders and armed forces, are marketed as having the ability to provide a level of protection against specific types of radiation. For a CRC material, a standard combat uniform and a multi-layered chemical, biological, radiological, nuclear (CBRN) protective material, the present work examines chemical composition and radiation protection against gamma-rays and neutron fluxes. Significant reduction in gamma-ray transmittance occurs only for the CRC fabric (46-514 keV) with gamma-ray attenuation coefficients of 3.10 to 2 g -1 . Reduction in neutron transmittance, for all three fabrics, could not be assessed with certainty as the measured transmittance was obscured by large statistical uncertainties. (author)

Nuclear power

International Nuclear Information System (INIS)

Abd Khalik Wood

2005-01-01

This chapter discussed the following topics related to the nuclear power: nuclear reactions, nuclear reactors and its components - reactor fuel, fuel assembly, moderator, control system, coolants. The topics titled nuclear fuel cycle following subtopics are covered: , mining and milling, tailings, enrichment, fuel fabrication, reactor operations, radioactive waste and fuel reprocessing. Special topic on types of nuclear reactor highlighted the reactors for research, training, production, material testing and quite detail on reactors for electricity generation. Other related topics are also discussed: sustainability of nuclear power, renewable nuclear fuel, human capital, environmental friendly, emission free, impacts on global warming and air pollution, conservation and preservation, and future prospect of nuclear power

Nuclear energy and the environment

International Nuclear Information System (INIS)

El-Hinnawi, E.E.

1980-01-01

Chapters are presented concerning the environmental impact of mining and milling of radioactive ores, upgrading processes, and fabrication of nuclear fuels; environmental impacts of nuclear power plants; non-radiological environmental implications of nuclear energy; radioactive releases from nuclear power plant accidents; environmental impact of reprocessing; nuclear waste disposal; fuel cycle; and the future of nuclear energy

CVD diamond for nuclear detection applications

CERN Document Server

Bergonzo, P; Tromson, D; Mer, C; Guizard, B; Marshall, R D; Foulon, F

2002-01-01

Chemically vapour deposited (CVD) diamond is a remarkable material for the fabrication of radiation detectors. In fact, there exist several applications where other standard semiconductor detectors do not fulfil the specific requirements imposed by corrosive, hot and/or high radiation dose environments. The improvement of the electronic properties of CVD diamond has been under intensive investigations and led to the development of a few applications that are addressing specific industrial needs. Here, we report on CVD diamond-based detector developments and we describe how this material, even though of a polycrystalline nature, is readily of great interest for applications in the nuclear industry as well as for physics experiments. Improvements in the material synthesis as well as on device fabrication especially concern the synthesis of films that do not exhibit space charge build up effects which are often encountered in CVD diamond materials and that are highly detrimental for detection devices. On a pre-i...

Y-12 product improvements expected to reduce metal production costs and decrease fabrication losses

International Nuclear Information System (INIS)

Parker, Elaine; Hassler, Morris

2004-01-01

Full text: The Y-12 National Security Complex supplies uranium metal and uranium oxide feed material that is then fabricated into fuel for research reactors around the world. Over the past two to three years, Y-12 has learned a great deal about its Low Enriched Uranium (LEU) product. The LEU is produced by taking U.S. surplus Highly Enriched Uranium (HEU) and blending it with depleted or natural uranium. The surplus HEU comes from dismantled U.S. weapons parts that have been declared as surplus. Those research reactors that use LEU from Y-12 are making important contributions to international nuclear non-proliferation by using LEU rather than HEU, and by helping to disposition former weapons material. We clearly understand that our customers want to keep fuel costs as low as possible. We at Y-12 are making every effort to improve efficiencies in producing the uranium through standardizing the chemical specifications as well as the product mass and dimensional qualities. This paper will discuss the new standard specification that we have proposed to existing LEU metal customers and fuel fabricators. It will also cover Y-12's progress on a new mold-design that will result in a more uniform, higher quality product that is less expensive to produce. This new product is expected to decrease overall fabrication losses by 5-10%, depending on the fabricator's process. The paper will include planned activities and the schedule associated with implementation of the new specification and product form. (author)

Mechanism for G2 phase-specific nuclear export of the kinetochore protein CENP-F.

Science.gov (United States)

Loftus, Kyle M; Cui, Heying; Coutavas, Elias; King, David S; Ceravolo, Amanda; Pereiras, Dylan; Solmaz, Sozanne R

2017-08-03

Centromere protein F (CENP-F) is a component of the kinetochore and a regulator of cell cycle progression. CENP-F recruits the dynein transport machinery and orchestrates several cell cycle-specific transport events, including transport of the nucleus, mitochondria and chromosomes. A key regulatory step for several of these functions is likely the G2 phase-specific export of CENP-F from the nucleus to the cytosol, where the cytoplasmic dynein transport machinery resides; however, the molecular mechanism of this process is elusive. Here, we have identified 3 phosphorylation sites within the bipartite classical nuclear localization signal (cNLS) of CENP-F. These sites are specific for cyclin-dependent kinase 1 (Cdk1), which is active in G2 phase. Phosphomimetic mutations of these residues strongly diminish the interaction of the CENP-F cNLS with its nuclear transport receptor karyopherin α. These mutations also diminish nuclear localization of the CENP-F cNLS in cells. Notably, the cNLS is phosphorylated in the -1 position, which is important to orient the adjacent major motif for binding into its pocket on karyopherin α. We propose that localization of CENP-F is regulated by a cNLS, and a nuclear export pathway, resulting in nuclear localization during most of interphase. In G2 phase, the cNLS is weakened by phosphorylation through Cdk1, likely resulting in nuclear export of CENP-F via the still active nuclear export pathway. Once CENP-F resides in the cytosol, it can engage in pathways that are important for cell cycle progression, kinetochore assembly and the faithful segregation of chromosomes into daughter cells.

Design specifications for ASME B and PV Code Section III nuclear class 1 piping

International Nuclear Information System (INIS)

Richardson, J.A.

1978-01-01

ASME B and PV Code Section III code regulations for nuclear piping requires that a comprehensive Design Specification be developed for ensuring that the design and installation of the piping meets all code requirements. The intent of this paper is to describe the code requirements, discuss the implementation of these requirements in a typical Class 1 piping design specification, and to report on recent piping failures in operating light water nuclear power plants in the US. (author)

Highlights of 50 years of nuclear fuels developments

International Nuclear Information System (INIS)

Simnad, M.T.

1989-01-01

The development of nuclear fuels since the discovery of nuclear fission is briefly surveyed in this paper. The fabrication of the uranium fuel for the first nuclear pile, CP-1, is described. The research and development studies and fabrication of the different types of nuclear fuels for the variety of research and power reactors are reviewed. The important factors involved to achieve low fuel cycle costs and reliable performance in the fuel elements are discussed in the historical context

Boosting nuclear fuels

International Nuclear Information System (INIS)

Demarthon, F.; Donnars, O.; Dupuy-Maury, F.

2002-01-01

This dossier gives a broad overview of the present day status of the nuclear fuel cycle in France: 1 - the revival of nuclear power as a solution to the global warming and to the increase of worldwide energy needs; 2 - the security of uranium supplies thanks to the reuse of weapon grade highly enriched uranium; 3 - the fabrication of nuclear fuels from the mining extraction to the enrichment processes, the fabrication of fuel pellets and the assembly of fuel rods; 4 - the new composition of present day fuels (UO x and chromium-doped pellets); 5 - the consumption of plutonium stocks and the Corail and Apa fuel assemblies for the reduction of plutonium stocks and the preservation of uranium resources. (J.S.)

Design, fabrication and installation of irradiation facilities

Energy Technology Data Exchange (ETDEWEB)

Sim, Bong Shick; Kim, Y. S.; Lee, C. Y. and others

1999-03-01

The principal contents of this project are to design, fabricate and install the steady-state fuel test loop in HANARO for nuclear technology development. Procurement and fabrication of main equipment, licensing and technical review for fuel test loop have been performed during 2 years(1997, 1998) for this project. Following contents are described in the report. - Procurement and fabrication of the equipment, piping for OPS - IPS manufacture - License - Technical review and evaluation of the FTL facility. As besides, as these irradiation facilities will be installed in HANARO, review of safety concern, discussion with KINS for licensing and review ofHANARO interface have been performed respectively. (author)

Quality assurance monitoring during nuclear fuel production in JSC 'TVEL'

International Nuclear Information System (INIS)

Filimonov, G.; Tchirkov, V.

2000-01-01

The paper describes Quality Assurance (QA) monitoring during fabrication of nuclear fuel in Russian Federation. Joint Stock Company 'TVEL', natural state monopoly of the type of holding that fabricates and supplies nuclear fuel for the NPPs of Russia, CIS and Europe, incorporates the major enterprises of the nuclear fuel cycle including JSC 'Mashinostroitelny zavod', Electrostal (fabrication of fuel pellets, rods and assemblies for different types of reactors), JSC 'Novosibirsky zavod khimconcentratov', Novosibirsk (fabrication of fuel rods and assemblies for WWER-440 and WWER-1000), JSC 'Tchepetsky mechanitchesky zavod', Tchepetsk (fabrication of Zr tubing). Monitoring of QA is an important element of Quality Management System (QMS) developed and implemented at the above-mentioned enterprises of the JSC 'TVEL' and it is performed on three levels including external and internal audits and author's supervision. Paper also describes short- and long-term policies of the JSC 'TVEL' in nuclear fuel quality field. (author)

Fabrication drawings of fuel pins for FUJI project among PSI, JNC and NRG. Revised version 2

International Nuclear Information System (INIS)

Ozawa, Takayuki; Nakazawa, Hiroaki; Abe, Tomoyuki; Nagayama, Masahiro

2002-10-01

Irradiation tests and post-irradiation examinations in the framework of JNC-PSI-NRG collaboration project will be performed in 2003-2005. Irradiation fuel pins will be fabricated by the middle of 2003. The fabrication procedure for irradiation fuel pins has been started in 2001. Several fabrication tests and qualification tests in JNC and PSI (Paul Scherrer Institut, Switzerland) have been performed before the fuel pin fabrication. According to the design assignment between PSI and JNC in the frame of this project, PSI should make specification documents for the fuel pellet, the sphere-pac fuel particles, the vipac fuel fragments, and the fuel segment fabrication. JNC should make the fabrication drawings for irradiation pins. JNC has been performed the fuel design in cooperation with PSI and NRG (Nuclear Research and Consultancy Group, Holland). In this project, the pelletized fuel, the sphere-pac fuel, and the vipac fuel will be simultaneously irradiated on HFR (High Flux Reactor, Holland). The fabrication drawings have been made under the design assignment with PSI, and consist of the drawings of MOX pellet, thermal insulator pellet, pin components, fuel segments, and the constructed pin. The fabrication drawings were approved in October 2001, but after that, the optimization of specifications was discussed and agreed among all partners. According to this agreement, the fabrication drawings were revised in January 2002. After the earlier revision, the shape of particle retainer to be made by PSI was modified from its drawing beforehand delivered. In this report, the fabrication drawings revised again will be shown, and the fabrication procedure (welding Qualification Tests) will be modified in accordance with the result of discussion on the 3rd technical meeting held in September 2002. These design works have been performed in Fuel Design and Evaluation Group, Plutonium Fuel Fabrication Division, Plutonium Fuel Center under the commission of Plutonium Fuel

New fabrication techniques for the nuclear fuels of tomorrow

International Nuclear Information System (INIS)

Babelot, J.F.; Bokelund, H.; Gerontopoulos, P.; Gueugnon, J.F.; Richter, K.

1995-01-01

The shift of the emphasis of the work at the Institute for Transuranium Elements (ITU) from the development of fuels based on uranium and plutonium to safety aspects concerning the use of plutonium and other of actinides, necessitates the production of targets containing appreciable amounts of minor actinides for irradiation experiments. The handling of minor actinides requires additional protective measures, combined with improved fuel fabrication techniques. The boundary conditions for a suitable process are flexibility, adaptability to remote control, and minimization of dust formation. A method based on the sol-gel fabrication technique meets these criteria, and was selected for the present developments at ITU. (author)

Inspection of NFT-type cask fabrication

International Nuclear Information System (INIS)

Takani, M.; Umegaki, O.

1998-01-01

NFT-type cask has been developed to transport the high burn-up spent fuel from Japanese nuclear power stations to the reprocessing plant of Japan Nuclear Fuel Limited which is under construction in Rokkasho-mura, Aomori prefecture. NFT placed orders of 53 casks to 5 fabricators in Japan and overseas, and these casks have been fabricated since 1994. There are two types of NFT-type casks for PWR spent fuel and four types of NFT-type cask for BWR spent fuel. These are designed in consideration of the number of spent fuels accommodated into each type of casks and the handling conditions at domestic nuclear power stations. According to Japanese notification, it is required to be confirmed by competent authority that casks are manufactured in accordance with approved designs. Furthermore, additional tests are performed such as through-gauge test for basket and pressure test on the shielding material space to ensure the performance of cask by NFT other than items inspected by the competent authority. In order to enhance maintainability of casks, replacement parts such as bolts and valves are shared as much as possible. (authors)

Nuclear power in Pakistan

International Nuclear Information System (INIS)

Siddiqui, Z.H.; Qureshi, I.H.

2005-01-01

Pakistan started its nuclear power program by installing a 137 M We Canadian Deuterium Reactor (Candu) at Karachi in 1971 which became operational in 1972. The post-contract technical support for the Karachi Nuclear Power Plant (KANUPP) was withdrawn by Canada in 196 as a consequence of Indian nuclear device test in 1974. In spite of various difficulties PAEC resolved to continue to operate KANUPP and started a process for the indigenous fabrication of spare parts and nuclear fuel. The first fuel bundle fabricated in Pakistan was loaded in the core in 1980. Since then KANUPP has been operating on the indigenously fabricated fuel. The plant computer systems and the most critical instrumentation and Control system were also replaced with up-to date technology. In 2002 KANUPP completed its original design life of 30 year. A program for the life extension of the plant had already been started. The second nuclear power plant of 300 M We pressurized water reactor purchased from China was installed in Chashma in 1997, which started commercial operations in 2001. Another unit of 300 M We will be installed at Chashma in near future. These nuclear power plants have been operating under IAEA safeguards agreements. PAEC through the long-term performance of the two power plants has demonstrated its competence to safely and successfully operate and maintain nuclear power plants. Pakistan foresees an increasingly important and significant share of nuclear power in the energy sector. The Government has recently allocated a share of 8000 MWe for nuclear energy in the total energy scenario of Pakistan by the year 2025. (author)

Prototypical spent nuclear nuclear fuel rod consolidation equipment, Phase 2: Final design report: Volume 2, Appendices: Part 1

International Nuclear Information System (INIS)

Ciez, A.P.

1987-01-01

The purpose of this specification is to establish functional and design requirements for the Prototypical Spent Nuclear Fuel Rod Consolidation System. The Department of Energy-Idaho Operations Office (DOE-ID) is responsible for the implementation of the Prototypic Dry Rod Consolidation Demonstration Project. This program is to develop and demonstrate a fully qualified, licensable, cost-effective, dry spent fuel rod consolidation system by July 1989. The work is divided into four phases as follows: Phase I--Preliminary Design, Phase II--Final Design Option, Phase III--Fabrication and System Checkout Option, and Phase IV--Installation and Hot Demonstration Option. This specification is part of the Phase II effort. The objectives of this specification are to provide functional and design requirements for the Prototypical Spent Nuclear Fuel Rod Consolidation equipment; establish specific tool and subsystem requirements such that the integrated and overall system requirements are satisfied; and establish positioning, envelope and size interface control requirements for each tool or subsystem such that the individual components will interface properly with the overall system design

Highlights of 50 years of nuclear fuel development

International Nuclear Information System (INIS)

Simnad, M.T.

1989-01-01

The development of nuclear fuels since the discovery of nuclear fission is briefly surveyed in this paper. The fabrication of the uranium fuel for the first nuclear pile, CP-1, is described. The research and development studies and fabrication of the different types of nuclear fuels for the variety of research and power reactors are reviewed. The important factors involved to achieve low fuel-cycle costs and reliable performance in the fuel elements are discussed in the historical context. 10 refs

The market for nuclear equipment - engineering and fabrication

International Nuclear Information System (INIS)

Tait, D.R.

1977-01-01

The role of electronic equipment in a CANDU power station is explained. Costs of installations and outages are outlined. The nuclear market for electronic equipment utilizes many components common to non-nuclear applications. (E.C.B.)

The nuclear age

International Nuclear Information System (INIS)

Leclercq, J.

1986-01-01

This book is divided into 7 chapters bearing on: (1) Nuclear power: an historical perspective (2) Diversity of reactor designs (3) Safety and the environment (4) Architecture and heavy construction (5) Fabrication and installation (6) Nuclear fuel (7) Working for the electric power industry

Fabrication of PWR fuel assembly and CANDU fuel bundle

International Nuclear Information System (INIS)

Lee, G.S.; Suh, K.S.; Chang, H.I.; Chung, S.H.

1980-01-01

For the project of localization of nuclear fuel fabrication, the R and D to establish the fabrication technology of CANDU fuel bundle as well as PWR fuel assembly was carried out. The suitable boss height and the prober Beryllium coating thickness to get good brazing condition of appendage were studied in the fabrication process of CANDU fuel rod. Basic Studies on CANLUB coating method also were performed. Problems in each fabrication process step and process flow between steps were reviewed and modified. The welding conditions for top and bottom nozzles, guide tube, seal and thimble screw pin were established in the fabrication processes of PWR fuel assembly. Additionally, some researches for a part of PWR grid brazing problems are also carried out

DUPIC nuclear fuel manufacturing and process technology development

International Nuclear Information System (INIS)

Yang, Myung Seung; Park, J. J.; Lee, J. W.

2000-05-01

In this study, DUPIC fuel fabrication technology and the active fuel laboratory were developed for the study of spent nuclear fuel. A new nuclear fuel using highly radioactive nuclear materials can be studied at the active fuel laboratory. Detailed DUPIC fuel fabrication process flow was developed considering the manufacturing flow, quality control process and material accountability. The equipment layout of about twenty DUPIC equipment at IMEF M6 hot cell was established for the minimization of the contamination during DUPIC processes. The characteristics of the SIMFUEL powder and pellets was studied in terms of milling conditions. The characteristics of DUPIC powder and pellet was studied by using 1 kg of spent PWR fuel at PIEF nr.9405 hot cell. The results were used as reference process conditions for following DUPIC fuel fabrication at IMEF M6. Based on the reference fabrication process conditions, the main DUPIC pellet fabrication campaign has been started at IMEF M6 using 2 kg of spent PWR fuel since 2000 January. As of March 2000, about thirty DUPIC pellets were successfully fabricated

DUPIC nuclear fuel manufacturing and process technology development

Energy Technology Data Exchange (ETDEWEB)

Yang, Myung Seung; Park, J. J.; Lee, J. W. [and others

2000-05-01

In this study, DUPIC fuel fabrication technology and the active fuel laboratory were developed for the study of spent nuclear fuel. A new nuclear fuel using highly radioactive nuclear materials can be studied at the active fuel laboratory. Detailed DUPIC fuel fabrication process flow was developed considering the manufacturing flow, quality control process and material accountability. The equipment layout of about twenty DUPIC equipment at IMEF M6 hot cell was established for the minimization of the contamination during DUPIC processes. The characteristics of the SIMFUEL powder and pellets was studied in terms of milling conditions. The characteristics of DUPIC powder and pellet was studied by using 1 kg of spent PWR fuel at PIEF nr.9405 hot cell. The results were used as reference process conditions for following DUPIC fuel fabrication at IMEF M6. Based on the reference fabrication process conditions, the main DUPIC pellet fabrication campaign has been started at IMEF M6 using 2 kg of spent PWR fuel since 2000 January. As of March 2000, about thirty DUPIC pellets were successfully fabricated.

Nuclear Safety

Energy Technology Data Exchange (ETDEWEB)

Silver, E G [ed.

1989-01-01

This document is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated.

SpinlineTM, Benefits of a nuclear specific safety-critical digital I/C platform - 15102

International Nuclear Information System (INIS)

Duthou, A.; Mouly, P.; Jegou, H.

2015-01-01

Spinline TM is Rolls-Royce modular and digital solution dedicated to developing and/or upgrading safety I/C used in nuclear reactors. From the start, Spinline TM was specifically designed for Nuclear applications. Therefore, its architecture and components satisfy, from design, the most stringent safety standards required by the local Safety authorities, while they can be adapted to various types of reactors. This is a significant advantage over suppliers who tried to adapt industrial systems to the Nuclear constraints and faced unexpected delays and costs to meet Safety authorities requirements. Spinline TM was specifically designed to implement any Class 1E and category A IEC-61226 safety I/C functions. It is qualified according to European and French nuclear standard and more recently by the US NRC, notably thanks to its Fail-safe features, deterministic behavior and Physical and Functional Separation. In 2011 EDF chose Spinline TM as its safety I/C systems technology for the modernization of 20 units of its 1300 MW PWR fleet

Nuclear waste package fabricated from concrete

International Nuclear Information System (INIS)

Pfeiffer, P.A.; Kennedy, J.M.

1987-03-01

After the United States enacted the Nuclear Waste Policy Act in 1983, the Department of Energy must design, site, build and operate permanent geologic repositories for high-level nuclear waste. The Department of Energy has recently selected three sites, one being the Hanford Site in the state of Washington. At this particular site, the repository will be located in basalt at a depth of approximately 3000 feet deep. The main concern of this site, is contamination of the groundwater by release of radionuclides from the waste package. The waste package basically has three components: the containment barrier (metal or concrete container, in this study concrete will be considered), the waste form, and other materials (such as packing material, emplacement hole liners, etc.). The containment barriers are the primary waste container structural materials and are intended to provide containment of the nuclear waste up to a thousand years after emplacement. After the containment barriers are breached by groundwater, the packing material (expanding sodium bentonite clay) is expected to provide the primary control of release of radionuclide into the immediate repository environment. The loading conditions on the concrete container (from emplacement to approximately 1000 years), will be twofold; (1) internal heat of the high-level waste which could be up to 400 0 C; (2) external hydrostatic pressure up to 1300 psi after the seepage of groundwater has occurred in the emplacement tunnel. A suggested container is a hollow plain concrete cylinder with both ends capped. 7 refs

LLNL/YMP Waste Container Fabrication and Closure Project; GFY technical activity summary

Energy Technology Data Exchange (ETDEWEB)

NONE

1990-10-01

The Department of Energy`s Office of Civilian Radioactive Waste Management (OCRWM) Program is studying Yucca Mountain, Nevada as a suitable site for the first US high-level nuclear waste repository. Lawrence Livermore National Laboratory (LLNL) has the responsibility for designing and developing the waste package for the permanent storage of high-level nuclear waste. This report is a summary of the technical activities for the LLNL/YMP Nuclear Waste Disposal Container Fabrication and Closure Development Project. Candidate welding closure processes were identified in the Phase 1 report. This report discusses Phase 2. Phase 2 of this effort involved laboratory studies to determine the optimum fabrication and closure processes. Because of budget limitations, LLNL narrowed the materials for evaluation in Phase 2 from the original six to four: Alloy 825, CDA 715, CDA 102 (or CDA 122) and CDA 952. Phase 2 studies focused on evaluation of candidate material in conjunction with fabrication and closure processes.

Energy, electricity and nuclear power

International Nuclear Information System (INIS)

Reuss, P.; Naudet, G.

2008-01-01

After an introduction recalling what energy is, the first part of this book presents the present day energy production and consumption and details more particularly the electricity 'vector' which is an almost perfect form of energy despite the fact that it is not a primary energy source: it must be generated from another energy source and no large scale storage of this energy is possible. The second part of the book is devoted to nuclear energy principles and to the related technologies. Content: 1 - What does energy mean?: the occurrence of the energy concept, the classical notion of energy, energy notion in modern physics, energy transformations, energy conservation, irreversibility of energy transformations, data and units used in the energy domain; 2 - energy production and consumption: energy systems, energy counting, reserves and potentialities of energy resources, production of primary energies, transport and storage of primary energies, energy consumption, energy saving, energy markets and prices, energy indicators; 3 - electric power: specificity of electricity and the electric system, power networks, power generation, electricity storage, power consumption and demand, power generation economics, electricity prices and market; 4 - physical principles of nuclear energy: nuclei structure and binding energy, radioactivity and nuclear reactions, nuclear reactions used in energy generation, basics of fission reactors physics; 5 - nuclear techniques: historical overview, main reactor types used today, perspectives; 6 - fuel cycle: general considerations, uranium mining, conversion, enrichment, fuel fabrication, back-end of the cycle, plutonium recycle in water cooled reactors; 7 - health and environmental aspects of nuclear energy: effects on ionizing radiations, basics of radiation protection, environmental impacts of nuclear energy, the nuclear wastes problem, specific risks; 8 - conclusion; 9 - appendixes (units, physics constants etc..)

China's nuclear safety regulatory body: The national nuclear safety administration

International Nuclear Information System (INIS)

Zhang Shiguan

1991-04-01

The establishment of an independent nuclear safety regulatory body is necessary for ensuring the safety of nuclear installations and nuclear fuel. Therefore the National Nuclear Safety Administration was established by the state. The aim, purpose, organization structure and main tasks of the Administration are presented. At the same time the practical examples, such as nuclear safety regulation on the Qinshan Nuclear Power Plant, safety review and inspections for the Daya Bay Nuclear Power Plant during the construction, and nuclear material accounting and management system in the nuclear fuel fabrication plant in China, are given in order to demonstrate the important roles having been played on nuclear safety by the Administration after its founding

Comparison of methods applicable to evaluation of nuclear power plant technical specifications

International Nuclear Information System (INIS)

Cho, N.Z.; Bozoki, G.E.; Youngblood, R.W.

1986-01-01

This study compares three probabilistic methods based on the static fault tree analysis, time-dependent unavailability analysis, and Markov analysis, which can be used to evaluate technical specifications in nuclear power plants. They are tested on a sample problem which was devised to closely represent the important and essential characteristics that should be addressed in determination and evaluation of the technical specifications

Waste management issues and their potential impact on technical specifications of CANDU fuel materials

International Nuclear Information System (INIS)

Tait, J.C.; Johnson, L.H.

1997-01-01

The technical specifications for the composition of nuclear fuels and materials used in Canada's CANDU reactors have been developed by AECL and materials manufacturers, taking into account considerations specific to their manufacture and the effect of minor impurities on fuel behaviour in reactor. Nitrogen and chlorine are examples of UO 2 impurities, however, where there is no technical specification limit. These impurities are present in the source materials or introduced in the fabrication process and are neutron activated to 14 C and 36 C1, which after 129 I , are the two most significant contributors to dose in safety assessments for the disposal of used fuel. For certain impurities, environmental factors, particularly the safety of the disposal of used fuels, should be taken into consideration when deriving 'allowable' impurity limits for nuclear fuel materials. (author)

Russian nuclear survey

International Nuclear Information System (INIS)

2002-07-01

This document gives a broad overview of the organization of nuclear activities in the Russian federation: Minatom activities, nuclear park and availability (reactors, performances, export activity), perspectives of development (improvement of safety, age of reactors, new realizations); fuel cycle (uranium production, conversion and enrichment, fuel fabrication, spent fuel reprocessing); wastes management (storage and disposal sites); R and D activities (organizations) and nuclear safety authority. (J.S.)

Selective nuclear export of specific classes of mRNA from mammalian nuclei is promoted by GANP

Science.gov (United States)

Wickramasinghe, Vihandha O.; Andrews, Robert; Ellis, Peter; Langford, Cordelia; Gurdon, John B.; Stewart, Murray; Venkitaraman, Ashok R.; Laskey, Ronald A.

2014-01-01

The nuclear phase of the gene expression pathway culminates in the export of mature messenger RNAs (mRNAs) to the cytoplasm through nuclear pore complexes. GANP (germinal- centre associated nuclear protein) promotes the transfer of mRNAs bound to the transport factor NXF1 to nuclear pore complexes. Here, we demonstrate that GANP, subunit of the TRanscription-EXport-2 (TREX-2) mRNA export complex, promotes selective nuclear export of a specific subset of mRNAs whose transport depends on NXF1. Genome-wide gene expression profiling showed that half of the transcripts whose nuclear export was impaired following NXF1 depletion also showed reduced export when GANP was depleted. GANP-dependent transcripts were highly expressed, yet short-lived, and were highly enriched in those encoding central components of the gene expression machinery such as RNA synthesis and processing factors. After injection into Xenopus oocyte nuclei, representative GANP-dependent transcripts showed faster nuclear export kinetics than representative transcripts that were not influenced by GANP depletion. We propose that GANP promotes the nuclear export of specific classes of mRNAs that may facilitate rapid changes in gene expression. PMID:24510098

Y-12 product improvements expected to reduce metal production costs and decrease fabrication losses

International Nuclear Information System (INIS)

Hassler, Morris E.

2005-01-01

The Y-12 National Security Complex (Y-12) supplies uranium metal and uranium oxide feed material for fabrication into fuel for research reactors around the world. Over the past few years, Y-12 has continued to improve its Low Enriched Uranium (LEU) product. The LEU is produced by taking U.S. surplus Highly Enriched Uranium (HEU) and blending it with depleted or natural uranium. The surplus HEU comes from dismantled U.S. weapons parts. Those research reactors that use LEU from Y-12 are making important contributions to international nuclear nonproliferation by using LEU rather than HEU, and helping to disposition former U.S. weapons material. It is clearly understood that the research reactor community must keep fuel costs as low as possible and Y-12 is making every effort to improve efficiencies in producing the uranium through standardizing the chemical specifications as well as the product mass and dimensional qualities. These production cost reductions allows for the U.S. to keep the LEU product price low even with the dramatic increase in the uranium enrichment and feed component market prices in the last few years. This paper will discuss a new standard specification that has been proposed to existing LEU metal customers and fuel fabricators. It will also cover Y-12's progress on a new mold-design that will result in a more uniform, higher quality product and eliminates two steps of the production process. This new product is expected to decrease fabrication losses by 5-10%, depending on the fabricator's process. The paper will include planned activities and the schedule associated with implementation of the new specification and product form. (author)

Design of the MOX fuel fabrication facility

International Nuclear Information System (INIS)

Johnson, J.V.; Brabazon, E.J.

2001-01-01

A consortium of Duke Engineering and Services, Inc., COGEMA, Inc. and Stone and Webster (DCS) are designing a mixed oxide fuel fabrication facility (MFFF) for the U.S. Department of Energy (DOE) to convert surplus plutonium to mixed oxide (MOX) fuel to be irradiated in commercial nuclear power plants based on the proven European technology of COGEMA and BELGONUCLEAIRE. This paper describes the MFFF processes, and how the proven MOX fuel fabrication technology is being adapted as required to comply with U.S. requirements. (author)

Design of the MOX fuel fabrication facility

Energy Technology Data Exchange (ETDEWEB)

Johnson, J.V. [MFFF Technical Manager, U.S. dept. of Energy, Washington, DC (United States); Brabazon, E.J. [MFFF Engineering Manager, Duke Cogema Stone and Webster, Charlotte, NC (United States)

2001-07-01

A consortium of Duke Engineering and Services, Inc., COGEMA, Inc. and Stone and Webster (DCS) are designing a mixed oxide fuel fabrication facility (MFFF) for the U.S. Department of Energy (DOE) to convert surplus plutonium to mixed oxide (MOX) fuel to be irradiated in commercial nuclear power plants based on the proven European technology of COGEMA and BELGONUCLEAIRE. This paper describes the MFFF processes, and how the proven MOX fuel fabrication technology is being adapted as required to comply with U.S. requirements. (author)

Metal finishing and vacuum processes groups, Materials Fabrication Division progress report, March-May 1984

International Nuclear Information System (INIS)

Dini, J.W.; Romo, J.G.; Jones, L.M.

1984-01-01

Progress is reported in fabrication and coating activities being conducted for the weapons program, nuclear test program, nuclear design program, magnetic fusion program, and miscellaneous applications

Optics fabrication technical challenges

International Nuclear Information System (INIS)

Chabassier, G.; Ferriou, N.; Lavastre, E.; Maunier, C.; Neauport, J.; Taroux, D.; Balla, D.; Fornerod, J.C.

2004-01-01

Before the production of all the LMJ (MEGAJOULE laser) optics, the CEA had to proceed with the fabrication of about 300 large optics for the LIL (laser integration line) laser. Thanks to a fruitful collaboration with high-tech optics companies in Europe, this challenge has been successfully hit. In order to achieve the very tight requirements for cleanliness, laser damage threshold and all the other high demanding fabrication specifications, it has been necessary to develop and to set completely new fabrication process going and to build special outsize fabrication equipment. Through a couple of examples, this paper gives an overview of the work which has been done and shows some of the results which have been obtained: continuous laser glass melting, fabrication of the laser slabs, rapid-growth KDP (potassium dihydrogen phosphate) technology, large diffractive transmission gratings engraving and characterization. (authors)

Decontamination chamber for the maintenance of DUPIC nuclear fuel fabrication and process equipment

International Nuclear Information System (INIS)

Kim, K. H.; Park, J. J.; Yang, M. S.; Lee, H. H.; Shin, J. M.

2000-10-01

This report presents the decontamination chamber of being capable of decontaminating and maintaining DUPIC nuclear fuel fabrication equipment contaminated in use. The decontamination chamber is a closed room in which contaminated equipment can be isolated from a hot-cell, be decontaminated and be reparired. This chamber can prevent contamination from spreading over the hot-cell, and it can also be utilized as a part of the hot-cell after maintenance work. The developed decontamination chamber has mainly five sub-modules - a horizontal module for opening and closing a ceil of the chamber, a vertical module for opening and closing a side of the chamber, a subsidiary door module for enforcing the vertical opening/closing module, a rotary module for rotating contaminated equipment, and a grasping module for holding a decontamination device. Such sub-modules were integrated and installed in the M6 hot-cell of the IMEF at the KAERI. The mechanical design considerations of each modules and the arrangement with hot-cell facility, remote operation and manipulation of the decontamination chamber are also described

Decontamination chamber for the maintenance of DUPIC nuclear fuel fabrication and process equipment

Energy Technology Data Exchange (ETDEWEB)

Kim, K. H.; Park, J. J.; Yang, M. S.; Lee, H. H.; Shin, J. M

2000-10-01

This report presents the decontamination chamber of being capable of decontaminating and maintaining DUPIC nuclear fuel fabrication equipment contaminated in use. The decontamination chamber is a closed room in which contaminated equipment can be isolated from a hot-cell, be decontaminated and be reparired. This chamber can prevent contamination from spreading over the hot-cell, and it can also be utilized as a part of the hot-cell after maintenance work. The developed decontamination chamber has mainly five sub-modules - a horizontal module for opening and closing a ceil of the chamber, a vertical module for opening and closing a side of the chamber, a subsidiary door module for enforcing the vertical opening/closing module, a rotary module for rotating contaminated equipment, and a grasping module for holding a decontamination device. Such sub-modules were integrated and installed in the M6 hot-cell of the IMEF at the KAERI. The mechanical design considerations of each modules and the arrangement with hot-cell facility, remote operation and manipulation of the decontamination chamber are also described.

Fracture toughness of fabrication welds investigated by metallographic methods

International Nuclear Information System (INIS)

Canonico, D.A.; Crouse, R.S.

1978-01-01

The intermediate scale test vessels (ITV) were fabricated to provide test specimens that have sufficient wall thickness and simulate light water reactor pressure vessels. They were fabricated from grades of steel that are similar to those used for nuclear pressure vessels, having a wall thickness of 150mm and the same welded construction. They are, however, considerably smaller in height and diameter than actual vessels. To date, ten vessels have been fabricated and eight have been tested. In preparation for testing the eighth vessel (ITV-8), an extensive investigation was conducted of the toughness properties of the fabrication weld. It was thoroughly characterized and the fracture specimens used in this metallographic investigation were taken from that weld metal

Facile fabrication of cobalt oxalate nanostructures with superior specific capacitance and super-long cycling stability

Science.gov (United States)

Cheng, Guanhua; Si, Conghui; Zhang, Jie; Wang, Ying; Yang, Wanfeng; Dong, Chaoqun; Zhang, Zhonghua

2016-04-01

Transition metal oxalate materials have shown huge competitive advantages for applications in supercapacitors. Herein, nanostructured cobalt oxalate supported on cobalt foils has been facilely fabricated by anodization, and could directly serve as additive/binder-free electrodes for supercapacitors. The as-prepared cobalt oxalate electrodes present superior specific capacitance of 1269 F g-1 at the current density of 6 A g-1 in the galvanostatic charge/discharge test. Moreover, the retained capacitance is as high as 87.2% as the current density increases from 6 A g-1 to 30 A g-1. More importantly, the specific capacitance of cobalt oxalate retains 91.9% even after super-long cycling of 100,000 cycles. In addition, an asymmetric supercapacitor assembled with cobalt oxalate (positive electrode) and activated carbon (negative electrode) demonstrates excellent capacitive performance with high energy density and power density.

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

Specific Methods of Information Security for Nuclear Materials Control and Accounting Automate Systems

Directory of Open Access Journals (Sweden)

Konstantin Vyacheslavovich Ivanov

2013-02-01

Full Text Available The paper is devoted to specific methods of information security for nuclear materials control and accounting automate systems which is not required of OS and DBMS certifications and allowed to programs modification for clients specific without defenses modification. System ACCORD-2005 demonstrates the realization of this method.

Nuclear fuel element leak detection system

International Nuclear Information System (INIS)

John, C.D. Jr.

1978-01-01

Disclosed is a leak detection system integral with a wall of a building used to fabricate nuclear fuel elements for detecting radiation leakage from the nuclear fuel elements as the fuel elements exit the building. The leak detecting system comprises a shielded compartment constructed to withstand environmental hazards extending into a similarly constructed building and having sealed doors on both ends along with leak detecting apparatus connected to the compartment. The leak detecting system provides a system for removing a nuclear fuel element from its fabrication building while testing for radiation leaks in the fuel element

Regulations concerning the fabricating business of nuclear fuel materials

International Nuclear Information System (INIS)

1978-01-01

The Regulation is revised on the basis of ''The law for the regulations of nuclear source materials, nuclear fuel materials and reactors'' and the ''Provisions concerning the enterprises processing nuclear fuel materials'' in the Enforcement Ordinance for the Law, to enforce such provisions. This is the complete revision of the regulation of the same name in 1957. Terms are explained, such as exposure radiation dose, cumulative dose, control area, surrounding inspection area, persons engaged in works, radioactive wastes, area for incoming and outgoing of materials, fluctuation of stocks, batch, real stocks, effective value and main measuring points. For the applications for the permission of the enterprises processing nuclear fuel materials, the location of an enterprise, the construction of buildings and the construction of and the equipments for facilities of chemical processing, forming, coating, assembling, storage of nuclear fuel materials, disposal of radioactive wastes and radiation control must be written. Records shall be made and maintained for the periods specified on the inspection of processing facilities, nuclear fuel materials, radiation control, operation, maintainance, accidents of processing facilities and weather. Limit to entrance into the control area, measures for exposure radiation dose, patrol and inspection, operation of processing facilities, transport of materials, disposal of radioactive wastes, safety regulations are provided for. Reports to be filed by the persons engaging in the enterprises processing nuclear fuel materials are prescribed. (Okada, K.)

Fuel fabrication and post-irradiation examination

Energy Technology Data Exchange (ETDEWEB)

Venter, P J; Aspeling, J C [Atomic Energy Corporation of South Africa Ltd., Pretoria (South Africa)

1990-06-01

This paper provides an overview of the A/c's Bevan and Eldopar facilities for the fabrication of nuclear fuel. It also describes the sophisticated Hot Cell Complex, which is capable of accommodating pressurised water reactor fuel and various other irradiated samples. Some interesting problems and their solutions are discussed. (author)

Fuel fabrication and post-irradiation examination

International Nuclear Information System (INIS)

Venter, P.J.; Aspeling, J.C.

1990-01-01

This paper provides an overview of the A/c's Bevan and Eldopar facilities for the fabrication of nuclear fuel. It also describes the sophisticated Hot Cell Complex, which is capable of accommodating pressurised water reactor fuel and various other irradiated samples. Some interesting problems and their solutions are discussed. (author)

Minimization of waste from uranium purification, enrichment and fuel fabrication

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-10-01

As any industry, nuclear industry generates a diverse range of waste which has to be managed in a safe manner to be acceptable to the public and the environment. The cost of waste management, the risks to the public and employees, and the detriment to the environment are dependent on the quantity and radioactive content of the waste generated. Waste minimization is a necessary activity needed to reduce the impact from nuclear fuel cycle operations and it is included in the national policy of some countries. In recognition of the importance of the subject, the IAEA has decided to review the current status of the work aimed at waste minimization in the nuclear fuel cycle. The waste minimization issues related to the back end of the nuclear fuel cycle are covered in Technical Reports Series No. 377 'Minimization of Radioactive Waste from Nuclear Power Plants and the Back End of the Nuclear Fuel Cycle' published in 1995. The present report deals with the front end of the nuclear fuel cycle, including existing options, approaches, developments and some specific considerations to be taken into account in decision making on waste minimization. It has been recognized that, in comparison with the back end of the nuclear fuel cycle, much less information is available, and this report should be considered as a first attempt to analyse waste minimization practices and opportunities in uranium purification, conversion, enrichment and fuel fabrication. Although mining and milling is an important part of the front end of the nuclear fuel cycle, these activities are excluded from consideration since relevant activities are covered in other IAEA publications.

Minimization of waste from uranium purification, enrichment and fuel fabrication

International Nuclear Information System (INIS)

1999-10-01

As any industry, nuclear industry generates a diverse range of waste which has to be managed in a safe manner to be acceptable to the public and the environment. The cost of waste management, the risks to the public and employees, and the detriment to the environment are dependent on the quantity and radioactive content of the waste generated. Waste minimization is a necessary activity needed to reduce the impact from nuclear fuel cycle operations and it is included in the national policy of some countries. In recognition of the importance of the subject, the IAEA has decided to review the current status of the work aimed at waste minimization in the nuclear fuel cycle. The waste minimization issues related to the back end of the nuclear fuel cycle are covered in Technical Reports Series No. 377 'Minimization of Radioactive Waste from Nuclear Power Plants and the Back End of the Nuclear Fuel Cycle' published in 1995. The present report deals with the front end of the nuclear fuel cycle, including existing options, approaches, developments and some specific considerations to be taken into account in decision making on waste minimization. It has been recognized that, in comparison with the back end of the nuclear fuel cycle, much less information is available, and this report should be considered as a first attempt to analyse waste minimization practices and opportunities in uranium purification, conversion, enrichment and fuel fabrication. Although mining and milling is an important part of the front end of the nuclear fuel cycle, these activities are excluded from consideration since relevant activities are covered in other IAEA publications

Evaluation of fuel fabrication and the back end of the fuel cycle for light-water- and heavy-water-cooled nuclear power reactors

International Nuclear Information System (INIS)

Carter, W.L.; Olsen, A.R.

1979-06-01

The classification of water-cooled nuclear reactors offers a number of fuel cycles that present inherently low risk of weapons proliferation while making power available to the international community. Eight fuel cycles in light water reactor (LWR), heavy water reactor (HWR), and the spectral shift controlled reactor (SSCR) systems have been proposed to promote these objectives in the International Fuel Cycle Evaluation (INFCE) program. Each was examined in an effort to provide technical and economic data to INFCE on fuel fabrication, refabrication, and reprocessing for an initial comparison of alternate cycles. The fuel cycles include three once-through cycles that require only fresh fuel fabrication, shipping, and spent fuel storage; four cycles that utilize denatured uranium--thorium and require all recycle operations; and one cycle that considers the LWR--HWR tandem operation requiring refabrication but no reprocessing

Evaluation of fuel fabrication and the back end of the fuel cycle for light-water- and heavy-water-cooled nuclear power reactors

Energy Technology Data Exchange (ETDEWEB)

Carter, W.L.; Olsen, A.R.

1979-06-01

The classification of water-cooled nuclear reactors offers a number of fuel cycles that present inherently low risk of weapons proliferation while making power available to the international community. Eight fuel cycles in light water reactor (LWR), heavy water reactor (HWR), and the spectral shift controlled reactor (SSCR) systems have been proposed to promote these objectives in the International Fuel Cycle Evaluation (INFCE) program. Each was examined in an effort to provide technical and economic data to INFCE on fuel fabrication, refabrication, and reprocessing for an initial comparison of alternate cycles. The fuel cycles include three once-through cycles that require only fresh fuel fabrication, shipping, and spent fuel storage; four cycles that utilize denatured uranium--thorium and require all recycle operations; and one cycle that considers the LWR--HWR tandem operation requiring refabrication but no reprocessing.

Radiological and environmental safety aspects of uranium fuel fabrication plants at Nuclear Fuel Complex, Hyderabad

International Nuclear Information System (INIS)

Viswanathan, S.; Surya Rao, B.; Lakshmanan, A.R.; Krishna Rao, T.

1991-01-01

Nuclear Fuel Complex, Hyderabad manufactures uranium dioxide fuel assemblies for PHWRs and BWRs operating in India. Starting materials are magnesium diuranate received from UCIL, Jaduguda and imported UF. Both of these are converted to UO 2 pellets by identical chemical processes and mechanical compacting. Since the uranium handled here is free of daughter product activities, external radiation is not a problem. Inhalation of airborne U compounds is one of the main source of exposure. Engineered protective measures like enclosures around U bearing powder handling equipment and local exhausts reduce worker's exposure. Installation of pre-filters, wet rotoclones and electrostatic precipitators in the ventillation system reduces the release of U into the environment. The criticality hazard in handling slightly enriched uranium is very low due to the built-in control based on geometry and inventory. Where airborne uranium is significant, workers are provided with protective respirators. The workers are regularly monitored for external exposure and also for internal exposure. The environmental releases from the NFC facility is well controlled. Soil, water and air from the NFC environment are routinely collected and analysed for all the possible pollutants. The paper describes the Health Physics experience during the last five years on occupational exposures and on environmental surveillance which reveals the high quality of safety observed in our nuclear fuel fabricating installations. (author). 4 refs., 6 tabs

Fabrication and testing of a 4-node micro-pocket fission detector array for the Kansas State University TRIGA Mk. II research nuclear reactor

Science.gov (United States)

Reichenberger, Michael A.; Nichols, Daniel M.; Stevenson, Sarah R.; Swope, Tanner M.; Hilger, Caden W.; Unruh, Troy C.; McGregor, Douglas S.; Roberts, Jeremy A.

2017-08-01

Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Micro-Pocket Fission Detectors (MPFDs) have been fabricated and tested previously, but successful testing of these prior detectors was limited to single-node operation with specialized designs. Described in this work is a modular, four-node MPFD array fabricated and tested at Kansas State University (KSU). The four sensor nodes were equally spaced to span the length of the fuel-region of the KSU TRIGA Mk. II research nuclear reactor core. The encapsulated array was filled with argon gas, serving as an ionization medium in the small cavities of the MPFDs. The unified design improved device ruggedness and simplified construction over previous designs. A 0.315-in. (8-mm) penetration in the upper grid plate of the KSU TRIGA Mk. II research nuclear reactor was used to deploy the array between fuel elements in the core. The MPFD array was coupled to an electronic support system which has been developed to support pulse-mode operation. Neutron-induced pulses were observed on all four sensor channels. Stable device operation was confirmed by testing under steady-state reactor conditions. Each of the four sensors in the array responded to changes in reactor power between 10 kWth and full power (750 kWth). Reactor power transients were observed in real-time including positive transients with periods of 5, 15, and 30 s. Finally, manual reactor power oscillations were observed in real-time.

Nuclear power and its fuel cycle

International Nuclear Information System (INIS)

Wymer, R.G.

1986-01-01

A series of viewgraphs describes the nuclear fuel cycle and nuclear power, covering reactor types, sources of uranium, enrichment of uranium, fuel fabrication, transportation, fuel reprocessing, and radioactive wastes

Steel structures for nuclear facilities

International Nuclear Information System (INIS)

1993-01-01

In the guide the requirements concerning design and fabrication of steel structures for nuclear facilities and documents to be submitted to the Finnish Centre for Radiation and Nuclear Safety (STUK) are presented. Furthermore, regulations concerning inspection of steel structures during construction of nuclear facilities and during their operation are set forth

Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Chinese Ed.)

International Nuclear Information System (INIS)

2012-01-01

On the basis of the principles included in the Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, this Safety Requirements publication establishes requirements applicable to the design of nuclear power plants. It covers the design phase and provides input for the safe operation of the power plant. It elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

Safety of Nuclear Power Plants: Design. Specific Safety Requirements (French Ed.)

International Nuclear Information System (INIS)

2012-01-01

On the basis of the principles included in the Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, this Safety Requirements publication establishes requirements applicable to the design of nuclear power plants. It covers the design phase and provides input for the safe operation of the power plant. It elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Arabic Ed.)

International Nuclear Information System (INIS)

2012-01-01

On the basis of the principles included in the Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, this Safety Requirements publication establishes requirements applicable to the design of nuclear power plants. It covers the design phase and provides input for the safe operation of the power plant. It elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Spanish Edition); Seguridad de las centrales nucleares: Diseno. Requisitos de seguridad especificos

Energy Technology Data Exchange (ETDEWEB)

NONE

2012-04-15

This publication is a revision of Safety Requirements No. NS-R-1, Safety of Nuclear Power Plants: Design. It establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in the design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

Method for the fabrication of nuclear fuel bodies

International Nuclear Information System (INIS)

Davis, D.E.; Leary, D.F.

1976-01-01

According to the method, graphite particles are treated with a liquid impregnating agent containing heat-hardenable resin components; the resulting particles are mixed with nuclear fuel particles, and a nuclear fuel body is formed by binding the mixture of particles into a cohesive mass by means of a carbon-contained binder. The claim concerns the details of the process. (UA) [de

Design and fabrication of stainless steel components for long life of spent fuel reprocessing plants

International Nuclear Information System (INIS)

Natarajan, R.; Ramkumar, P.; Sundararaman, V.; Kamachi Mudali, U.; Baldev Raj; Shanmugam, K.

2010-01-01

Reprocessing of spent nuclear fuels based on the PUREX process is the proven process with many commercial plants operating satisfactorily worldwide. The process medium being nitric acid, austenitic stainless steel is the material of construction as it is the best commercially available material for meeting the conditions in the reprocessing plants. Because of the high radiation fields, contact maintenance of equipment and systems of these plants are very time consuming and costly unlike other chemical process plants. Though the plants constructed in the early years required extensive shut downs for replacement of equipment and systems within the first fifteen years of operation itself, development in the field of stainless steel metallurgy and fabrication techniques have made it possible to design the present day plants for an operating life period of forty years. A review of the operational experience of the PUREX process based aqueous reprocessing plants has been made in this paper and reveals that life limiting failures of equipment and systems are mainly due to corrosion while a few are due to stresses. Presently there are no standards for design specification of materials and fabrication of reprocessing plants like the nuclear power plants, where well laid down ASTM and ASME codes and standards are available which are based on the large scale operational feedbacks on pressure vessels for conventional and nuclear industries. (author)

ITER nuclear components, preparing for the construction and R&D results

Science.gov (United States)

Ioki, K.; Akiba, M.; Barabaschi, P.; Barabash, V.; Chiocchio, S.; Daenner, W.; Elio, F.; Enoeda, M.; Ezato, K.; Federici, G.; Gervash, A.; Grebennikov, D.; Jones, L.; Kajiura, S.; Krylov, V.; Kuroda, T.; Lorenzetto, P.; Maruyama, S.; Merola, M.; Miki, N.; Morimoto, M.; Nakahira, M.; Ohmori, J.; Onozuka, M.; Rozov, V.; Sato, K.; Strebkov, Yu; Suzuki, S.; Tanchuk, V.; Tivey, R.; Utin, Yu

2004-08-01

Progress has been made in the preparation of the procurement specifications for key nuclear components of ITER. Detailed design of the vacuum vessel (VV) and in-vessel components is being performed to consider fabrication methods and non-destructive tests (NDT). R&D activities are being carried out on vacuum vessel UT inspection with waves launched at an angle of 20° or 30°, on flow distribution tests of a two-channel model, on fabrication and testing of FW mock-ups and panels, on the blanket flexible support as a complete system including the housing, on the blanket co-axial pipe connection with guard vacuum for leak detection, and on divertor vertical target prototypes. The results give confidence in the validity of the design and identify possibilities of attractive alternate fabrication methods.

Recommendations for the specification of thermocouples for nuclear applications

International Nuclear Information System (INIS)

1977-05-01

This Code of Practice is a guide for use in the preparation of individual specifications to cover, as fully as possible the conditions governing the supply of raw materials and the ordering, manufacture, testing, inspection, handling and installation of thermocouples for use in nuclear environments in order that reliable, consistent and generally acceptable results can be obtained. The insulation resistance values quoted in this document apply to magnesium oxide. If other insulants are called for, appropriate values must be specified. (author)

Development of advanced nuclear materials - Fabrication of Zr-Nb alloy used in PHWRs

Energy Technology Data Exchange (ETDEWEB)

Lee, Kang In; Kim, Won Baek; Lee, Chul Kyung; Choi, Kuk Sun; Kang, Dae Kyu; Seo, Chang Ryul [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-07-01

The following conclusions can be made from the second year research: 1. Easy control for alloying elements can be made for the following adding metals like Nb, V, Sn, Mo, Fe due to low vapor pressure. In case of Cr and Te= known to have high vapor pressure, they are controlled by adding master alloy(Zr-Cr) or quite excess of aimed composition. However, Bi was found to be very difficult to charging the certain amount into the melt. 2. Oxygen content can be adjusted by adding the Zr-10%O master alloy considering the inherent amount of oxygen in sponge zirconium. 3. The charging rod of 38 mm in diameter, 96 mm in length was made by a series of button melting, casting and vacuum welding, from this, Zr-2.5Nb ingot of 50 mm in diameter and 550 mm in length was fabricated by EB drip melting process. 4. The amount of Nb can be successfully adjusted at 2.8% with charging 15% excess. Nb as adding element is easily controlled due to high-melting -point metal and its low vapor pressure. 5. Oxygen content is not varied during remelting, casting, and drip melting, only slight change was observed in button melting stage due to uptake the desorbed gases during the melting operation. Nuclear materials in domestic nuclear power plants depend on import and this amount reaches 100 million dollars per year. The increase in demand for the development of new zirconium based alloys are expecting. All the results involving this research can be applied for the melting of reactive metals, vacuum refining and alloy design. 13 refs., 6 tabs., 10 figs., 10 ills. (author)

Supplier responsibility for nuclear material quality

International Nuclear Information System (INIS)

Stuart, P.S.; Dohna, A.E.

1976-01-01

Nuclear materials must be delivered by either the manufacturer or the distributor with objective, documented evidence that the material was manufactured, inspected, and tested by proven techniques performed by qualified personnel working to documented procedures. Measurement devices used for acceptance must be of proven accuracy. The material and all records must be identified for positive traceability as part of the quality history of the nuclear components, system, or structure in which the material was used. In conclusion, the nuclear material supplier must join the fabricator, the installer, and the user in effective implementation of the total systems approach to the application of quality assurance principles to all phases of procurement, fabrication, installation, and use of the safety-related components, systems, and structures in a nuclear power plant

Waste management issues and their potential impact on technical specifications of CANDU fuel materials

Energy Technology Data Exchange (ETDEWEB)

Tait, J.C.; Johnson, L.H. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada)

1997-07-01

The technical specifications for the composition of nuclear fuels and materials used in Canada's CANDU reactors have been developed by AECL and materials manufacturers, taking into account considerations specific to their manufacture and the effect of minor impurities on fuel behaviour in reactor. Nitrogen and chlorine are examples of UO{sub 2} impurities, however, where there is no technical specification limit. These impurities are present in the source materials or introduced in the fabrication process and are neutron activated to {sup 14}C and {sup 36}C1, which after {sup 129}I , are the two most significant contributors to dose in safety assessments for the disposal of used fuel. For certain impurities, environmental factors, particularly the safety of the disposal of used fuels, should be taken into consideration when deriving 'allowable' impurity limits for nuclear fuel materials. (author)

Administrative controls and quality assurance for the operational phase of nuclear power plants

International Nuclear Information System (INIS)

Anon.

1983-01-01

This standard provides requirements and recommendations for an adminstrative controls and quality assurance program to help ensure that activities associated with nuclear power plant operation are carried out without undue risk to the health and safety of the public. This standard applies to all activities affecting those functions important to the safety of nuclear power plant structures, systems, and components. Activities included are: design changes, purchasing, fabricating, handling, shipping, storing, cleaning and decontaminating, erecting, installing, inspecting, testing, operating, maintaining, repairing, refueling, modifying and decomissioning. This standard is not specifically intended for application to test, mobile or experimental reactors, nor reactors not subject to U.S. Nuclear Regulatory Commission licensing

Technical specification for fabrication of HANARO pool cover

Energy Technology Data Exchange (ETDEWEB)

Ryu, Jeong Soo; Woo, Sang Ik

2001-06-01

This technical specification details the requirements and the acceptance criteria for design, seismic analysis, function test, installation and quality assurance for HANARO pool cover which will be installed at the top of reactor pool. The pool cover is classified as non-nuclear safety, seismic category II and quality class T. The basic design of the pool cover for increasing HANARO applications has been carried out for supporting the driving devices which can load, unload and rotate the irradiation targets in the in-core and out-core vertical irradiation holes under on-power operation. The comments of HANARO user group related with irradiation tests have optimally considered in the process of design. The interference between fuel handling and control absorber units in the reactor pool and activities to load, unload and rotate the irradiation targets at the top of the reactor pool have been minimized. The pool cover can be moved for maintenance and can protect the reactor pool from unexpected drop of foreign materials. It provides the space to vertical access of driving devices for NTD, CT/IR and OR4/OR5 under on-power operation. And the pool cover assembly must maintain its structural integrity under seismic load. Based on the above design concept, the HANARO pool cover has been proposed as supporting structure of driving devices for NTD, fission moly and RI production under on-power operation.

Application of RBS and NRA in the fabrication of carbon based devices

International Nuclear Information System (INIS)

Ila, D.; Zimmerman, R.L.; Maleki, H.; Evelyn, A.L.; Poker, D.B.

1995-06-01

We have used Nuclear Reaction Analysis (NRA) and Rutherford Backscattering Spectrometry (RBS) as well as resonant backscattering as analytical tools in fabricating carbon based drug delivery bio-implants, electrodes for batteries, and devices to entrap or filter specific toxins. Precursor is resol C 7 H 8 O 2 liquid, which converts to fully cured phenolic resin C 7 H 6 O (sp gr 1.25) on heating at 170 C. This resin further transforms with no change in shape to glassy carbon (sp gr 1.45) on heating to 1000 C. Final product consists of long ribbon-like molecules of sp2 carbon atoms aggregated locally to form subcrystalline domains arranged randomly in space. RBS and NRA were used in measuring the porosity before and after activation, in concentration profiling of stored drugs before and after leaching, in detecting low level light element impurities, and in detecting changes in the structure of the device due to fabrication

FPGA fabric specific optimization for RLT design

International Nuclear Information System (INIS)

Perwaiz, A.; Khan, S.A.

2010-01-01

This paper proposes a technique custom to the optimization requirements suited for a particular family of Field Programmable Gate Arrays (FPGAs). As FPGAs have introduced re configurable black boxes there is a need to perform optimization across FPGAs slice fabric in order to achieve optimum performance. Though the Register Transfer Level (RTL) Hardware Descriptive Language (HDL) code should be technology independent but in many design instances it is imperative to understand the target technology especially once the target device embeds dedicated arithmetic blocks. No matter what the degree of optimization of the algorithm is, the configuration of target device plays an important role as far as the device utilization and path delays are concerned Index Terms: Field Programmable Gate Arrays (FPGA), Compression Tree, Bit Width Reduction, Look Ahead Pipelining. (author)

Fuel Fabrication and Nuclear Reactors

Energy Technology Data Exchange (ETDEWEB)

Karpius, Peter Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-02-02

The uranium from the enrichment plant is still in the form of UF₆. UF₆ is not suitable for use in a reactor due to its highly corrosive chemistry as well as its phase diagram. UF₆ is converted into UO₂ fuel pellets, which are in turn placed in fuel rods and assemblies. Reactor designs are variable in moderators, coolants, fuel, performance etc.The dream of energy ‘too-cheap to meter’ is no more, and now the nuclear power industry is pushing ahead with advanced reactor designs.

Nuclear Forensics and Attribution for Improved Energy Security: The Use of Taggants in Nuclear Fuel

International Nuclear Information System (INIS)

Kristo, M J; Robel, M; Hutcheon, I D

2007-01-01

The Global Nuclear Energy Partnership (GNEP), recently announced by DOE Secretary Bodman, poses significant new challenges with regard to securing, safeguarding, monitoring and tracking nuclear materials. In order to reduce the risk of nuclear proliferation, new technologies must be developed to reduce the risk that nuclear material can be diverted from its intended use. Regardless of the specific nature of the fuel cycle, nuclear forensics and attribution will play key roles to ensure the effectiveness of nonproliferation controls and to deter the likelihood of illicit activities. As the leader of the DHS nuclear and radiological pre-detonation attribution program, LLNL is uniquely positioned to play a national leadership role in this effort. Ensuring that individuals or organizations engaged in illicit trafficking are rapidly identified and apprehended following theft or diversion of nuclear material provides a strong deterrent against unlawful activities. Key to establishing this deterrent is developing the ability to rapidly and accurately determine the identity, source and prior use history of any interdicted nuclear material. Taggants offer one potentially effective means for positively identifying lost or stolen nuclear fuels. Taggants are materials that can be encoded with a unique signature and introduced into nuclear fuel during fuel fabrication. During a nuclear forensics investigation, the taggant signature can be recovered and the nuclear material identified through comparison with information stored in an appropriate database. Unlike serial numbers or barcodes, microtaggants can provide positive identification with only partial recovery, providing extreme resistance to any attempt to delete or alter them

Nuclear power plant piping prefabrication and assembly

International Nuclear Information System (INIS)

Schmidt, H.

1990-01-01

The piping design for nuclear power plants projects reveals, at the beginning, a modification through the application of new fabrication techniques for prefabrication and assembly. This report presents a fabrication methodology which aims to minimize the fabrication and assembly costs as well as to improve and assure quality. (Author) [es

Evolution of stainless steels in nuclear industry

International Nuclear Information System (INIS)

Tavassoli, Farhad

2010-01-01

Starting with the stainless steels used in the conventional industry, their adoption and successive evolutions in the nuclear industry, from one generation of nuclear reactors to another, is presented. Specific examples for several steels are given, covering fabrication procedures, qualification methods, property databases and design allowable stresses, to show how the ever-increasing demands for better performance and reliability, in particular under neutron irradiation, have been met. Particular attention is paid to the austenitic stainless steels types 304L, 316L, 316L(N), 316L(N)-IG, titanium stabilized grade 321, precipitation strengthened alloy 800, conventional and low activation ferritic/martensitic steels and their oxygen dispersion strengthening (ODS) derivatives. For each material, the evolution of the associated filler metal and welding techniques are also presented. (author)

Fabrication process of expanded cooling jackets

International Nuclear Information System (INIS)

Weber, C.M.

1980-01-01

The present invention concerns the fabrication process of heat exchangers and in particular, the fabrication and assembly process of cooling jackets of the system driving the control rods used in nuclear reactors. The cooling jackets are assembled for cooling the stator of a tubular motor displacing the control rods. The fabrication and assembling of the cooling jacket is made up by the following operations: - a sleeve has an inner fluid inlet and outlet ways, - an external socket is fitted to the sleeve, - on the external socket a continuous welding is realized, which join the socket to the sleeve, and define a serie of parallel welded turns, - a pressure is established between the sleeve and the socket by a fluid through the outlet or inlet ways of the sleeve. When the other way is sealed up, the socket expands between the welded turns, and the fluid can pass through the jacket [fr

IAEA safeguards to prevent nuclear matrials diversion for fabrication of nuclear explosives

International Nuclear Information System (INIS)

Preuschen von und zu Liebenstein, R.

1982-01-01

The IAEA precautionary measures in accordance with the Non-Proliferation Treaty can be characterized as measures creating confidence. They constitute at present the essential basis for peaceful use of atomic energy. Even though there is a lot of criticism concerning the efficiency of the precautionary measures, and all justified calls for the elaboration of further legal instruments against nuclear materials diversion must not be neglected, the IAEA precautionary measures have already in a credible way contributed to contain the proliferation of nuclear weapons. (orig./HSCH) [de

Development of inspection data collection and evaluation system for large scale MOX fuel fabrication plant safeguards (3)

International Nuclear Information System (INIS)

Kumakura, Shinichi; Masuda, Shoichiro; Iso, Shoko; Hisamatsu, Yoshinori; Kurobe, Hiroko; Nakajima, Shinji

2015-01-01

Inspection Data Collection and Evaluation System is the system to store inspection data and operator declaration data collected from various measurement equipment, which is installed in fuel fabrication processes of the large-scale MOX fuel fabrication plant, and to make safeguards evaluation based on Near Real Time Accountancy (NRTA) using these data. Nuclear Material Control Center developed the simulator to simulate fuel fabrication process, in-process material inventory/flow data and the measurement data and the adequacy/impact to the uncertainty of the material balance using the simulation results, such as the facility operation and the operational status, has been reviewed. Following the 34th INMM Japan chapter presentation, the model similar to the real nuclear material accountancy during the fuel fabrication process was simulated and the nuclear material accountancy and its uncertainty (Sigma MUF) have been reviewed. Some findings have been obtained, such as regarding evaluation related indicators for verification under a more realistic accountancy which could be applied by operator. (author)

Standard specification for uranium metal enriched to more than 15 % and less Than 20 % 235U

CERN Document Server

American Society for Testing and Materials. Philadelphia

2000-01-01

1.1 This specification covers nuclear grade uranium metal that has either been processed through an enrichment plant, or has been produced by the blending of highly enriched uranium with other uranium, to obtain uranium of any 235U concentration below 20 % (and greater than 15 %) and that is intended for research reactor fuel fabrication. The scope of this specification includes specifications for enriched uranium metal derived from commercial natural uranium, recovered uranium, or highly enriched uranium. Commercial natural uranium, recovered uranium and highly enriched uranium are defined in Section 3. The objectives of this specification are to define the impurity and uranium isotope limits for commercial grade enriched uranium metal. 1.2 This specification is intended to provide the nuclear industry with a standard for enriched uranium metal which is to be used in the production of research reactor fuel. In addition to this specification, the parties concerned may agree to other appropriate conditions. ...

Nuclear materials for fission reactors

International Nuclear Information System (INIS)

Matzke, H.; Schumacher, G.

1992-01-01

This volume brings together 47 papers from scientists involved in the fabrication of new nuclear fuels, in basic research of nuclear materials, their application and technology as well as in computer codes and modelling of fuel behaviour. The main emphasis is on progress in the development of non -oxide fuels besides reporting advances in the more conventional oxide fuels. The two currently performed large reactor safety programmes CORA and PHEBUS-FP are described in invited lectures. The contributions review basic property measurements, as well as the present state of fuel performance modelling. The performance of today's nuclear fuel, hence UO 2 , at high burnup is also reviewed with particular emphasis on the recently observed phenomenon of grain subdivision in the cold part of the oxide fuel at high burnup, the so-called 'rim' effect. Similar phenomena can be simulated by ion implantation in order to better elucidate the underlying mechanism and reviews on high resolution electron microscopy provide further information. The papers will provide a useful treatise of views, ideas and new results for all those scientists and engineers involved in the specific questions of current nuclear waste management

ITER nuclear components, preparing for the construction and R and D results

Energy Technology Data Exchange (ETDEWEB)

Ioki, K. E-mail: iokik@itereu.de; Akiba, M.; Barabaschi, P.; Barabash, V.; Chiocchio, S.; Daenner, W.; Elio, F.; Enoeda, M.; Ezato, K.; Federici, G.; Gervash, A.; Grebennikov, D.; Jones, L.; Kajiura, S.; Krylov, V.; Kuroda, T.; Lorenzetto, P.; Maruyama, S.; Merola, M.; Miki, N.; Morimoto, M.; Nakahira, M.; Ohmori, J.; Onozuka, M.; Rozov, V.; Sato, K.; Strebkov, Yu.; Suzuki, S.; Tanchuk, V.; Tivey, R.; Utin, Yu

2004-08-01

Progress has been made in the preparation of the procurement specifications for key nuclear components of ITER. Detailed design of the vacuum vessel (VV) and in-vessel components is being performed to consider fabrication methods and non-destructive tests (NDT). R and D activities are being carried out on vacuum vessel UT inspection with waves launched at an angle of 20 deg. or 30 deg. , on flow distribution tests of a two-channel model, on fabrication and testing of FW mock-ups and panels, on the blanket flexible support as a complete system including the housing, on the blanket co-axial pipe connection with guard vacuum for leak detection, and on divertor vertical target prototypes. The results give confidence in the validity of the design and identify possibilities of attractive alternate fabrication methods.

MOX fuel fabrication technology in J-MOX

International Nuclear Information System (INIS)

Osaka, Shuichi; Yoshida, Ryouichi; Yamazaki, Yukiko; Ikeda, Hiroyuki

2014-01-01

Japan Nuclear Fuel Ltd. (JNFL) has constructed JNFL MOX Fuel Fabrication Plant (J-MOX) since 2010. The MIMAS process has been introduced in the powder mixing process from AREVA NC considering a lot of MOX fuel fabrication experiences at MELOX plant in France. The feed material of Pu for J-MOX is MH-MOX powder from Rokkasho Reprocessing Plant (RRP) in Japan. The compatibility of the MH-MOX powder with the MIMAS process was positively evaluated and confirmed in our previous study. This paper describes the influences of the UO2 powder and the recycled scrap powder on the MOX pellet density. (author)

Risk assessment for nuclear processes at the Savannah River Site

International Nuclear Information System (INIS)

Durant, W.S.

1992-01-01

The Savannah River Site, one of the US Department of Energy's nuclear materials processing facilities, has for many years conducted risk-based safety analyses for the nuclear processes conducted at the facilities. This approach has allowed comparisons of risks to established criteria for acceptability. When the risk-based program was begun, it was evident that its success would depend upon having a compilation of data that was site specific. The decision was made to create a data bank of undesirable events that had occurred at the site's nuclear fuel reprocessing facilities. From this modest beginning, five data banks have been created for nuclear fuel reprocessing, waste management, nuclear fuel fabrication, tritium operations, and the Savannah River Technology Center. In addition to the primary purpose of providing a sound basis for risk-based safety analyses, these highly versatile data banks are routinely used for equipment breakdown histories, incident investigations, design studies, project justifications, reliability studies, process problem solving, training, and audits

Serotype-specific Differences in Dengue Virus Non-structural Protein 5 Nuclear Localization*

Science.gov (United States)

Hannemann, Holger; Sung, Po-Yu; Chiu, Han-Chen; Yousuf, Amjad; Bird, Jim; Lim, Siew Pheng; Davidson, Andrew D.

2013-01-01

The four serotypes of dengue virus (DENV-1 to -4) cause the most important arthropod-borne viral disease of humans. DENV non-structural protein 5 (NS5) contains enzymatic activities required for capping and replication of the viral RNA genome that occurs in the host cytoplasm. However, previous studies have shown that DENV-2 NS5 accumulates in the nucleus during infection. In this study, we examined the nuclear localization of NS5 for all four DENV serotypes. We demonstrate for the first time that there are serotypic differences in NS5 nuclear localization. Whereas the DENV-2 and -3 proteins accumulate in the nucleus, DENV-1 and -4 NS5 are predominantly if not exclusively localized to the cytoplasm. Comparative studies on the DENV-2 and -4 NS5 proteins revealed that the difference in DENV-4 NS5 nuclear localization was not due to rapid nuclear export but rather the lack of a functional nuclear localization sequence. Interaction studies using DENV-2 and -4 NS5 and human importin-α isoforms failed to identify an interaction that supported the differential nuclear localization of NS5. siRNA knockdown of the human importin-α isoform KPNA2, corresponding to the murine importin-α isoform previously shown to bind to DENV-2 NS5, did not substantially affect DENV-2 NS5 nuclear localization, whereas knockdown of importin-β did. The serotypic differences in NS5 nuclear localization did not correlate with differences in IL-8 gene expression. The results show that NS5 nuclear localization is not strictly required for virus replication but is more likely to have an auxiliary function in the life cycle of specific DENV serotypes. PMID:23770669

Serotype-specific differences in dengue virus non-structural protein 5 nuclear localization.

Science.gov (United States)

Hannemann, Holger; Sung, Po-Yu; Chiu, Han-Chen; Yousuf, Amjad; Bird, Jim; Lim, Siew Pheng; Davidson, Andrew D

2013-08-02

The four serotypes of dengue virus (DENV-1 to -4) cause the most important arthropod-borne viral disease of humans. DENV non-structural protein 5 (NS5) contains enzymatic activities required for capping and replication of the viral RNA genome that occurs in the host cytoplasm. However, previous studies have shown that DENV-2 NS5 accumulates in the nucleus during infection. In this study, we examined the nuclear localization of NS5 for all four DENV serotypes. We demonstrate for the first time that there are serotypic differences in NS5 nuclear localization. Whereas the DENV-2 and -3 proteins accumulate in the nucleus, DENV-1 and -4 NS5 are predominantly if not exclusively localized to the cytoplasm. Comparative studies on the DENV-2 and -4 NS5 proteins revealed that the difference in DENV-4 NS5 nuclear localization was not due to rapid nuclear export but rather the lack of a functional nuclear localization sequence. Interaction studies using DENV-2 and -4 NS5 and human importin-α isoforms failed to identify an interaction that supported the differential nuclear localization of NS5. siRNA knockdown of the human importin-α isoform KPNA2, corresponding to the murine importin-α isoform previously shown to bind to DENV-2 NS5, did not substantially affect DENV-2 NS5 nuclear localization, whereas knockdown of importin-β did. The serotypic differences in NS5 nuclear localization did not correlate with differences in IL-8 gene expression. The results show that NS5 nuclear localization is not strictly required for virus replication but is more likely to have an auxiliary function in the life cycle of specific DENV serotypes.

Fabrication of preliminary fuel rods for SFR

International Nuclear Information System (INIS)

Kim, Sun Ki; Oh, Seok Jin; Ko, Young Mo; Woo, Youn Myung; Kim, Ki Hwan

2012-01-01

Metal fuels was selected for fueling many of the first reactors in the US, including the Experimental Breeder Reactor-I (EBR-I) and the Experimental Breeder Reactor-II (EBR-II) in Idaho, the FERMI-I reactor, and the Dounreay Fast Reactor (DFR) in the UK. Metallic U.Pu.Zr alloys were the reference fuel for the US Integral Fast Reactor (IFR) program. Metallic fuel has advantages such as simple fabrication procedures, good neutron economy, high thermal conductivity, excellent compatibility with a Na coolant and inherent passive safety. U-Zr-Pu alloy fuels have been used for SFR (sodium-cooled fast reactor) related to the closed fuel cycle for managing minor actinides and reducing a high radioactivity levels since the 1980s. Fabrication technology of metallic fuel for SFR has been in development in Korea as a national nuclear R and D program since 2007. For the final goal of SFR fuel rod fabrication with good performance, recently, three preliminary fuel rods were fabricated. In this paper, the preliminary fuel rods were fabricated, and then the inspection for QC(quality control) of the fuel rods was performed

Technical study report on fuel fabrication system

International Nuclear Information System (INIS)

Kono, Shusaku; Tanaka, Kenya; Ono, Kiyoshi; Iwasa, Katsuyoshi; Hoshino, Yasushi; Shinkai, Yasuo

2000-07-01

The feasibility study of FBR and related fuel cycle is performed for developing the FBR recycle system which ensures safety, economic competitiveness, efficient utilization of resources, reduction of environmental burden and enhancement of nuclear non-proliferation under consistency of FBR reactor and fuel cycle systems. In this study, a conceptual design study and system characteristics evaluation are conducted for fuel fabrication systems of pellet process, vibropack process for oxide and nitride fuel and casting process for metal fuel. Technical issues in each process are also extracted. In 1999 fiscal year, a conceptual design study were conducted for the fuel fabrication plants adopting (1) the short pellet process which simplifies the conventional MOX pellet fabrication processes, (2) vibropack processes of aqueous gelation process, improved RIAR process, improved ANL process and fluoride volatility process, (3) casting processes of injection process, centrifuging process. As a result, attainable perspective was obtained for each fuel fabrication system through the evaluation of apparatuses, layout and facility volume, etc. In each fuel fabrication system, technical issues for practical use were made clear. Hereafter, more detailed study will be performed for each system, and research programs for phase II study will be planned. (author)

Nuclear-specific AR-V7 Protein Localization is Necessary to Guide Treatment Selection in Metastatic Castration-resistant Prostate Cancer.

Science.gov (United States)

Scher, Howard I; Graf, Ryon P; Schreiber, Nicole A; McLaughlin, Brigit; Lu, David; Louw, Jessica; Danila, Daniel C; Dugan, Lyndsey; Johnson, Ann; Heller, Glenn; Fleisher, Martin; Dittamore, Ryan

2017-06-01

Circulating tumor cells (CTCs) expressing AR-V7 protein localized to the nucleus (nuclear-specific) identify metastatic castration-resistant prostate cancer (mCRPC) patients with improved overall survival (OS) on taxane therapy relative to the androgen receptor signaling inhibitors (ARSi) abiraterone acetate, enzalutamide, and apalutamide. To evaluate if expanding the positivity criteria to include both nuclear and cytoplasmic AR-V7 localization ("nuclear-agnostic") identifies more patients who would benefit from a taxane over an ARSi. The study used a cross-sectional cohort. Between December 2012 and March 2015, 193 pretherapy blood samples, 191 of which were evaluable, were collected and processed from 161 unique mCRPC patients before starting a new line of systemic therapy for disease progression at the Memorial Sloan Kettering Cancer Center. The association between two AR-V7 scoring criteria, post-therapy prostate-specific antigen (PSA) change (PTPC) and OS following ARSi or taxane treatment, was explored. One criterion required nuclear-specific AR-V7 localization, and the other required an AR-V7 signal but was agnostic to protein localization in CTCs. Correlation of AR-V7 status to PTPC and OS was investigated. Relationships with survival were analyzed using multivariable Cox regression and log-rank analyses. A total of 34 (18%) samples were AR-V7-positive using nuclear-specific criteria, and 56 (29%) were AR-V7-positive using nuclear-agnostic criteria. Following ARSi treatment, none of the 16 nuclear-specific AR-V7-positive samples and six of the 32 (19%) nuclear-agnostic AR-V7-positive samples had ≥50% PTPC at 12 weeks. The strongest baseline factor influencing OS was the interaction between the presence of nuclear-specific AR-V7-positive CTCs and treatment with a taxane (hazard ratio 0.24, 95% confidence interval 0.078-0.79; p=0.019). This interaction was not significant when nuclear-agnostic criteria were used. To reliably inform treatment selection

Application specific integrated circuits and hybrid micro circuits for nuclear instrumentation

International Nuclear Information System (INIS)

Chandratre, V.B.; Sukhwani, Menka; Mukhopadhyay, P.K.; Shastrakar, R.S.; Sudheer, M.; Shedam, V.; Keni, Anubha

2009-01-01

Rapid development in semiconductor technology, sensors, detectors and requirements of high energy physics experiments as well as advances in commercially available nuclear instruments have lead to challenges for instrumentation. These challenges are met with development of Application Specific Integrated Circuits and Hybrid Micro Circuits. This paper discusses various activities in ASIC and HMC development in Bhabha Atomic Research Centre. (author)

Interpretation of the results from individual monitoring of workers at the Nuclear Fuel Fabrication Facility, Brazil

International Nuclear Information System (INIS)

Castro, Marcelo Xavier de

2005-01-01

In nuclear fuel fabrication facilities, workers are exposed to different compounds of enriched uranium. Although in this kind of facility the main route of intake is inhalation, ingestion may occur in some situations, and also a mixture of both. The interpretation of the bioassay data is very complex, since it is necessary taking into account all the different parameters, which is a big challenge. Due to the high cost of the individual monitoring programme for internal dose assessment in the routine monitoring programmes, usually only one type of measurement is assigned. In complex situations like the one described in this study, where several parameters can compromise the accuracy of the bioassay interpretation it is need to have a combination of techniques to evaluate the internal dose. According to ICRP 78 (1997), the general order of preference of measurement methodologies in terms of accuracy of interpretation is: body activity measurement, excreta analysis and personal air sampling. Results of monitoring of working environment may provide information that assists in the interpretation on particle size, chemical form, solubility and date of intake. A group of fifteen workers from controlled area of the studied nuclear fuel fabrication facility was selected to evaluate the internal dose using all different available techniques during a certain period. The workers were monitored for determination of uranium content in the daily urinary and faecal excretion (collected over a period of 3 consecutive days), chest counting and personal air sampling. The results have shown that at least two types of sensitivity techniques must be used, since there are some sources of uncertainties on the bioassay interpretation, like mixture of uranium compounds intake and different routes of intake. The combination of urine and faeces analysis has shown to be the more appropriate methodology for assessing internal dose in this situation. The chest counting methodology has not shown

Fabric quality issues related to apparel merchandising

CSIR Research Space (South Africa)

Das, Sonali

2015-02-01

Full Text Available The objectives of this study are to develop an understanding of fabric quality related issues and research gaps relevant to apparel manufacturing and merchandising within the South African context. The specific focus is on fabric objective...

Impact of inservice inspection on the reliability of nuclear piping

International Nuclear Information System (INIS)

Woo, H.H.

1983-12-01

The reliability of nuclear piping is a function of piping quality as fabricated, service loadings and environments, plus programs of continuing inspection during operation. This report presents the results of a study of the impact of inservice inspection (ISI) programs on the reliability of specific nuclear piping systems that have actually failed in service. Two major factors are considered in the ISI programs: one is the capability of detecting flaws; the other is the frequency of performing ISI. A probabilistic fracture mechanics model issued to estimate the reliability of two nuclear piping lines over the plant life as functions of the ISI programs. Examples chosen for the study are the PWR feedwater steam generator nozzle cracking incident and the BWR recirculation reactor vessel nozzle safe-end cracking incident

Chilean fuel elements fabrication progress report

International Nuclear Information System (INIS)

Baeza, J.; Contreras, H.; Chavez, J.; Klein, J.; Mansilla, R.; Marin, J.; Medina, R.

1993-01-01

Due to HEU-LEU core conversion necessity for the Chilean MTR reactors, the Fuel Elements Plant is being implemented to LEU nuclear fuel elements fabrication. A glove box line for powder-compact processing designed at CCHEN, which supposed to operate under an automatic control system, is at present under initial tests. Results of first natural uranium fuel plates manufacturing runs are shown

Computerized information system for inventory-taking and verification at a nuclear fuel fabrication plant with closed production lines

International Nuclear Information System (INIS)

Bahm, W.; Brueckner, C.; Hartmann, G.

1976-01-01

By means of a model the use of electronic data processing is studied for preparing inventory listings and for inventory verification in a fabrication plant for Pu-U mixed-oxide fuel pins. It is postulated that interruptions in operation should be avoided as much as possible. Closed-Line production is assumed so that access to nuclear material calls for special withdrawal via locks. The production line is subdivided into sections with measuring points placed in between to record the nuclear material flow. The measured results are fed to a central data acquisition and reporting system capable of calculating on-line from these results the book inventories present in the individual sections. Inventory-taking and verification are carried out simultaneously in the sections of the production line using the EDP system. The production is not interrupted for this purpose. The production stream is tagged prior to reaching a section to be measured and is subsequently measured when entering the respective section until the tag has reached the end of the section. The measurement can be verified by inspectors. Movements of nuclear materials in and from other plant areas such as the storage area are likewise fed into the central data processing system so that inventory lists can be recalled at any moment. By this means the inventory can be taken quickly and at any time. The inventory is verified in the conventional way. (author)

Industrias Nucleares do Brasil in the context of the Brazilian nuclear program

International Nuclear Information System (INIS)

1996-10-01

The activities carried out by Industrias Nucleares Brasileiras (INB) related to the nuclear fuel cycle are described. These activities comprise presently uranium prospecting and processing and fuel elements assembly.Starting in 1997,INB will also perform the reconversion of enriched uranium hexafluoride and the fabrication of fuel pellets.Furthermore, INB produces as well rare earth oxides

Fabrication of metallic fuel for fast breeder reactor

International Nuclear Information System (INIS)

Saify, M.T.; Jha, S.K.; Abdulla, K.K.; Kumar, Arbind; Mittal, R.K.; Prasad, R.S.; Mahule, N.; Kumar, Arun; Prasad, G.J.

2012-01-01

Natural uranium oxide fuelled PHWRs comprises of first stage of Indian nuclear power programme. Liquid metal fast breeder reactors fuelled by Pu (from PHWR's) form the second stage. A shorter reactor doubling time is essential in order to accelerate the nuclear power growth in India. Metallic fuels are known to provide shorter doubling times, necessitating to be used as driver fuel for fast breeder reactors. One of the fabrication routes for metallic fuels having random grain orientation, is injection casting technique. The technique finds its basis in an elementary physical concept - the possibility of supporting a liquid column within a tube, by the application of a pressure difference across the liquid interface inside and outside the tube. At AFD, BARC a facility has been set-up for injection casting of uranium rods in quartz tube moulds, demoulding of cast rods, end-shearing of rods and an automated inspection system for inspection of fuel rods with respect to mass, length, diameter and diameter variation along the length and internal and external porosities/voids. All the above facilities have been set-up in glove boxes and have successfully been used for fabrication of uranium bearing fuel rods. The facility has been designed for fabrication and inspection of Pu-bearing metallic fuels also, if required

Fabricating Copper Nanotubes by Electrodeposition

Science.gov (United States)

Yang, E. H.; Ramsey, Christopher; Bae, Youngsam; Choi, Daniel

2009-01-01

Copper tubes having diameters between about 100 and about 200 nm have been fabricated by electrodeposition of copper into the pores of alumina nanopore membranes. Copper nanotubes are under consideration as alternatives to copper nanorods and nanowires for applications involving thermal and/or electrical contacts, wherein the greater specific areas of nanotubes could afford lower effective thermal and/or electrical resistivities. Heretofore, copper nanorods and nanowires have been fabricated by a combination of electrodeposition and a conventional expensive lithographic process. The present electrodeposition-based process for fabricating copper nanotubes costs less and enables production of copper nanotubes at greater rate.

MODELLING OF NUCLEAR FUEL CLADDING TUBES CORROSION

Directory of Open Access Journals (Sweden)

Miroslav Cech

2016-12-01

Full Text Available This paper describes materials made of zirconium-based alloys used for nuclear fuel cladding fabrication. It is focused on corrosion problems their theoretical description and modeling in nuclear engineering.

The fuel of nuclear reactors

International Nuclear Information System (INIS)

1995-03-01

This booklet is a presentation of the different steps of the preparation of nuclear fuels performed by Cogema. The documents starts with a presentation of the different French reactor types: graphite moderated reactors, PWRs using MOX fuel, fast breeder reactors and research reactors. The second part describes the fuel manufacturing process: conditioning of nuclear materials and fabrication of fuel assemblies. The third part lists the different companies involved in the French nuclear fuel industry while part 4 gives a short presentation of the two Cogema's fuel fabrication plants at Cadarache and Marcoule. Part 5 and 6 concern the quality assurance, the safety and reliability aspects of fuel elements and the R and D programs. The last part presents some aspects of the environmental and personnel protection performed by Cogema. (J.S.)

Development of Hi-Tech ceramics fabrication technologies - Development of advanced nuclear materials

Energy Technology Data Exchange (ETDEWEB)

Kang, Thae Kap; Park, Ji Youn; Kim, Sun Jae; Kim, Kyong Ho; Jung, Choong Hwan; Oh, Seok Jin [Korea Atomic Energy Res. Inst., Taejon (Korea, Republic of)

1994-07-15

The objective of the present work is to prepare the foundation of hi-tech ceramics fabrication technologies through developing important processes i.e., tape casting, sol-gel, single crystal growing, compacting and sintering, and grinding and machining processes. Tape casting process is essential to manufacture hard and functional thin plates and structural elements for some composite materials. For the fabrication of spherical mono-sized micropowders of oxides, sol-gel process has widely been used. Piezoelectric elements that are the core parts of the sensors of LPMS (loose part monitoring system) and ALMS (acoustic leakage monitoring system) are used in single crystal forms. Compacting and sintering processes are general methods for fabricating structural parts using powders. Grinding and machining processes are important to achieve the final dimensions and surface properties of the parts. (Author).

Manufacturing of nuclear power components in CDM

International Nuclear Information System (INIS)

Krishnan, J.; Jawale, S.B.

2002-01-01

Full text: In the nuclear research programme in India, Dr. H.J. Bhabha, the architecture of the Indian Nuclear programme felt a need for proto-type development and precision manufacturing facility to fulfill the requirements of mechanical components in establishing the manufacturing capability for the successful and self sustained nuclear programme. Centre for Design and Manufacture (CDM) hitherto known as CWS was established in 1964 to cater to the specific requirements of DAE and other associated units like ISRO, DRDO. Since then CDM has made multiple technological achievements and changes towards high quality products. The acquisition of up-to-date machines during High-Tech facility under VIII Plan project and Advance Precision Fabrication facility under IX Plan project has changed the capability of CDM towards CAD, CAM, CAE and CNC machining centres. Considering the rapid growth in the design and manufacturing, it was renamed as Centre for Design and Manufacture in March 2002, with the mission of quality output through group effort and team work

The nuclear fuel cycle

International Nuclear Information System (INIS)

1998-05-01

After a short introduction about nuclear power in the world, fission physics and the French nuclear power plants, this brochure describes in a digest way the different steps of the nuclear fuel cycle: uranium prospecting, mining activity, processing of uranium ores and production of uranium concentrates (yellow cake), uranium chemistry (conversion of the yellow cake into uranium hexafluoride), fabrication of nuclear fuels, use of fuels, reprocessing of spent fuels (uranium, plutonium and fission products), recycling of energetic materials, and storage of radioactive wastes. (J.S.)

An ISO response to the needs of the nuclear industry

International Nuclear Information System (INIS)

Buyers, R.

1982-01-01

The history and development of the International Standards Organisation's standard DIS6215, Nuclear Power Plants - Quality Assurance, is presented. The aim is to produce quality assurance standards covering the design, procurement, fabrication, construction, commissioning, operation, maintenance and decommissioning of nuclear power plants. The remit was later extended to include fuel fabrication and reprocessing plants. (U.K.)

Fuel Fabrication Capability Research and Development Plan

Energy Technology Data Exchange (ETDEWEB)

Senor, David J.; Burkes, Douglas

2014-04-17

The purpose of this document is to provide a comprehensive review of the mission of the Fuel Fabrication Capability (FFC) within the Global Threat Reduction Initiative Convert Program, along with research and development (R&D) needs that have been identified as necessary to ensuring mission success. The design and fabrication of successful nuclear fuels must be closely linked endeavors. Therefore, the overriding motivation behind the FFC R&D program described in this plan is to foster closer integration between fuel design and fabrication to reduce programmatic risk. These motivating factors are all interrelated, and progress addressing one will aid understanding of the others. The FFC R&D needs fall into two principal categories, 1) baseline process optimization, to refine the existing fabrication technologies, and 2) manufacturing process alternatives, to evaluate new fabrication technologies that could provide improvements in quality, repeatability, material utilization, or cost. The FFC R&D Plan examines efforts currently under way in regard to coupon, foil, plate, and fuel element manufacturing, and provides recommendations for a number of R&D topics that are of high priority but not currently funded (i.e., knowledge gaps). The plan ties all FFC R&D efforts into a unified vision that supports the overall Convert Program schedule in general, and the fabrication schedule leading up to the MP-1 and FSP-1 irradiation experiments specifically. The fabrication technology decision gates and down-selection logic and schedules are tied to the schedule for fabricating the MP-1 fuel plates, which will provide the necessary data to make a final fuel fabrication process down-selection. Because of the short turnaround between MP-1 and the follow-on FSP-1 and MP-2 experiments, the suite of specimen types that will be available for MP-1 will be the same as those available for FSP-1 and MP-2. Therefore, the only opportunity to explore parameter space and alternative processing

A muscle-specific knockout implicates nuclear receptor coactivator MED1 in the regulation of glucose and energy metabolism.

Science.gov (United States)

Chen, Wei; Zhang, Xiaoting; Birsoy, Kivanc; Roeder, Robert G

2010-06-01

As conventional transcriptional factors that are activated in diverse signaling pathways, nuclear receptors play important roles in many physiological processes that include energy homeostasis. The MED1 subunit of the Mediator coactivator complex plays a broad role in nuclear receptor-mediated transcription by anchoring the Mediator complex to diverse promoter-bound nuclear receptors. Given the significant role of skeletal muscle, in part through the action of nuclear receptors, in glucose and fatty acid metabolism, we generated skeletal muscle-specific Med1 knockout mice. Importantly, these mice show enhanced insulin sensitivity and improved glucose tolerance as well as resistance to high-fat diet-induced obesity. Furthermore, the white muscle of these mice exhibits increased mitochondrial density and expression of genes specific to type I and type IIA fibers, indicating a fast-to-slow fiber switch, as well as markedly increased expression of the brown adipose tissue-specific UCP-1 and Cidea genes that are involved in respiratory uncoupling. These dramatic results implicate MED1 as a powerful suppressor in skeletal muscle of genetic programs implicated in energy expenditure and raise the significant possibility of therapeutical approaches for metabolic syndromes and muscle diseases through modulation of MED1-nuclear receptor interactions.

Nucleus-specific expression in the multinuclear mushroom-forming fungus Agaricus bisporus reveals different nuclear regulatory programs.

Science.gov (United States)

Gehrmann, Thies; Pelkmans, Jordi F; Ohm, Robin A; Vos, Aurin M; Sonnenberg, Anton S M; Baars, Johan J P; Wösten, Han A B; Reinders, Marcel J T; Abeel, Thomas

2018-04-24

Many fungi are polykaryotic, containing multiple nuclei per cell. In the case of heterokaryons, there are different nuclear types within a single cell. It is unknown what the different nuclear types contribute in terms of mRNA expression levels in fungal heterokaryons. Each cell of the mushroom Agaricus bisporus contains two to 25 nuclei of two nuclear types originating from two parental strains. Using RNA-sequencing data, we assess the differential mRNA contribution of individual nuclear types and its functional impact. We studied differential expression between genes of the two nuclear types, P1 and P2, throughout mushroom development in various tissue types. P1 and P2 produced specific mRNA profiles that changed through mushroom development. Differential regulation occurred at the gene level, rather than at the locus, chromosomal, or nuclear level. P1 dominated mRNA production throughout development, and P2 showed more differentially up-regulated genes in important functional groups. In the vegetative mycelium, P2 up-regulated almost threefold more metabolism genes and carbohydrate active enzymes (cazymes) than P1, suggesting phenotypic differences in growth. We identified widespread transcriptomic variation between the nuclear types of A. bisporus Our method enables studying nucleus-specific expression, which likely influences the phenotype of a fungus in a polykaryotic stage. Our findings have a wider impact to better understand gene regulation in fungi in a heterokaryotic state. This work provides insight into the transcriptomic variation introduced by genomic nuclear separation. Copyright © 2018 the Author(s). Published by PNAS.

Fabrication and Testing of a Modular Micro-Pocket Fission Detector Instrumentation System for Test Nuclear Reactors

Science.gov (United States)

Reichenberger, Michael A.; Nichols, Daniel M.; Stevenson, Sarah R.; Swope, Tanner M.; Hilger, Caden W.; Roberts, Jeremy A.; Unruh, Troy C.; McGregor, Douglas S.

2018-01-01

Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Measurement of the neutron-flux distribution within the reactor core provides a more complete understanding of the operating conditions in the reactor than typical ex-core sensors. Micro-Pocket Fission Detectors have been developed and tested previously but have been limited to single-node operation and have utilized highly specialized designs. The development of a widely deployable, multi-node Micro-Pocket Fission Detector assembly will enhance nuclear research capabilities. A modular, four-node Micro-Pocket Fission Detector array was designed, fabricated, and tested at Kansas State University. The array was constructed from materials that do not significantly perturb the neutron flux in the reactor core. All four sensor nodes were equally spaced axially in the array to span the fuel-region of the reactor core. The array was filled with neon gas, serving as an ionization medium in the small cavities of the Micro-Pocket Fission Detectors. The modular design of the instrument facilitates the testing and deployment of numerous sensor arrays. The unified design drastically improved device ruggedness and simplified construction from previous designs. Five 8-mm penetrations in the upper grid plate of the Kansas State University TRIGA Mk. II research nuclear reactor were utilized to deploy the array between fuel elements in the core. The Micro-Pocket Fission Detector array was coupled to an electronic support system which has been specially developed to support pulse-mode operation. The Micro-Pocket Fission Detector array composed of four sensors was used to monitor local neutron flux at a constant reactor power of 100 kWth at different axial locations simultaneously. The array was positioned at five different radial locations within the core to emulate the deployment of multiple arrays and develop a 2-dimensional measurement of

Nuclear power generation and nuclear nonproliferation

International Nuclear Information System (INIS)

Walske, C.

1978-01-01

In the future outlook around year 2000 of nuclear power, thought must be given to fuel reprocessing and plutonium utilization. The adverse utilization of plutonium may be prevented by the means balanced with its economical value. As the method of less cost with lower effect of nonproliferation, combination of fuel reprocessing and fuel fabrication facilities and mixed plutonium/uranium processing are possible. As the method of more cost with higher effect of nonproliferation the maintenance of high radioactivity and inaccessibility of plutonium is conceivable. As for the agreeable methods in 2000, seven principles may be mentioned, such as the dependence upon the agreements among major nations and upon nuclear exporting countries. These are still inadequate, however. What is important is to provide with the sufficient safeguards to countries concerned to negate the need for nuclear weapons. Efforts are then necessary for leading nuclear countries to extend aids to other nuclear-oriented countries. (Mori, K.)

Industrias Nucleares do Brasil in the context of the Brazilian nuclear program; A INB no contexto do programa nuclear brasileiro

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-10-01

The activities carried out by Industrias Nucleares Brasileiras (INB) related to the nuclear fuel cycle are described. These activities comprise presently uranium prospecting and processing and fuel elements assembly.Starting in 1997,INB will also perform the reconversion of enriched uranium hexafluoride and the fabrication of fuel pellets.Furthermore, INB produces as well rare earth oxides 2 figs., 1 tab.

Specific regulation of thermosensitive lipid droplet fusion by a nuclear hormone receptor pathway.

Science.gov (United States)

Li, Shiwei; Li, Qi; Kong, Yuanyuan; Wu, Shuang; Cui, Qingpo; Zhang, Mingming; Zhang, Shaobing O

2017-08-15

Nuclear receptors play important roles in regulating fat metabolism and energy production in humans. The regulatory functions and endogenous ligands of many nuclear receptors are still unidentified, however. Here, we report that CYP-37A1 (ortholog of human cytochrome P450 CYP4V2), EMB-8 (ortholog of human P450 oxidoreductase POR), and DAF-12 (homolog of human nuclear receptors VDR/LXR) constitute a hormone synthesis and nuclear receptor pathway in Caenorhabditis elegans This pathway specifically regulates the thermosensitive fusion of fat-storing lipid droplets. CYP-37A1, together with EMB-8, synthesizes a lipophilic hormone not identical to Δ7-dafachronic acid, which represses the fusion-promoting function of DAF-12. CYP-37A1 also negatively regulates thermotolerance and lifespan at high temperature in a DAF-12-dependent manner. Human CYP4V2 can substitute for CYP-37A1 in C. elegans This finding suggests the existence of a conserved CYP4V2-POR-nuclear receptor pathway that functions in converting multilocular lipid droplets to unilocular ones in human cells; misregulation of this pathway may lead to pathogenic fat storage.

The nuclear higher-order structure defined by the set of topological relationships between DNA and the nuclear matrix is species-specific in hepatocytes.

Science.gov (United States)

Silva-Santiago, Evangelina; Pardo, Juan Pablo; Hernández-Muñoz, Rolando; Aranda-Anzaldo, Armando

2017-01-15

During the interphase the nuclear DNA of metazoan cells is organized in supercoiled loops anchored to constituents of a nuclear substructure or compartment known as the nuclear matrix. The stable interactions between DNA and the nuclear matrix (NM) correspond to a set of topological relationships that define a nuclear higher-order structure (NHOS). Current evidence suggests that the NHOS is cell-type-specific. Biophysical evidence and theoretical models suggest that thermodynamic and structural constraints drive the actualization of DNA-NM interactions. However, if the topological relationships between DNA and the NM were the subject of any biological constraint with functional significance then they must be adaptive and thus be positively selected by natural selection and they should be reasonably conserved, at least within closely related species. We carried out a coarse-grained, comparative evaluation of the DNA-NM topological relationships in primary hepatocytes from two closely related mammals: rat and mouse, by determining the relative position to the NM of a limited set of target sequences corresponding to highly-conserved genomic regions that also represent a sample of distinct chromosome territories within the interphase nucleus. Our results indicate that the pattern of topological relationships between DNA and the NM is not conserved between the hepatocytes of the two closely related species, suggesting that the NHOS, like the karyotype, is species-specific. Copyright © 2016 Elsevier B.V. All rights reserved.

Silica Treatments: A Fire Retardant Strategy for Hemp Fabric/Epoxy Composites

Directory of Open Access Journals (Sweden)

Francesco Branda

2016-08-01

Full Text Available In this paper, for the first time, inexpensive waterglass solutions are exploited as a new, simple and ecofriendly chemical approach for promoting the formation of a silica-based coating on hemp fabrics, able to act as a thermal shield and to protect the latter from heat sources. Fourier Transform Infrared (FTIR and solid-state Nuclear Magnetic Resonance (NMR analysis confirm the formation of –C–O–Si– covalent bonds between the coating and the cellulosic substrate. The proposed waterglass treatment, which is resistant to washing, seems to be very effective for improving the fire behavior of hemp fabric/epoxy composites, also in combination with ammonium polyphosphate. In particular, the exploitation of hemp surface treatment and Ammonium Polyphosphate (APP addition to epoxy favors a remarkable decrease of the Heat Release Rate (HRR, Total Heat Release (THR, Total Smoke Release (TSR and Specific Extinction Area (SEA (respectively by 83%, 35%, 45% and 44% as compared to untreated hemp/epoxy composites, favoring the formation of a very stable char, as also assessed by Thermogravimetric Analysis (TGA. Because of the low interfacial adhesion between the fabrics and the epoxy matrix, the obtained composites show low strength and stiffness; however, the energy absorbed by the material is higher when using treated hemp. The presence of APP in the epoxy matrix does not affect the mechanical behavior of the composites.

Process for fabricating mixed-oxide powders

International Nuclear Information System (INIS)

Elmaleh, D.; Giraudel, A.

1975-01-01

A physical-chemical process for fabricating homogeneous powders suitable for sintering is described. It can be applied to the synthesis of all mixed oxides having mutually compatible and water soluble salts. As a specific example, the fabrication of lead titanate-zirconate powders used to make hot pressed ceramics is described. These ceramics show improved piezoelectric properties [fr

Conceptual Design of On-line Based Licensing Review and Assesment System of Nuclear Installations and Nuclear Materials ('PRIBEN')

International Nuclear Information System (INIS)

Melani, Ai; Chang, Soon Heung

2008-01-01

At the present Indonesia has no nuclear power plant in operation yet, although it is expected that the first nuclear power plant will be operated and commercially available in around the year of 2016 to 2017 in Muria Peninsula. There are only three research reactors, one nuclear fuel fabrication plant for research reactors, and one experimental fuel fabrication plant for nuclear power, one isotope production facility and some other research facilities. All the facility is under Nuclear Energy Regulatory Agency (BAPETEN) controlling through regulation, licensing and inspection. The organizations operation submits licensing application to BAPETEN before utilizing the facility. According to the regulation before BAPETEN give license they perform review and assessment for the utility application. Based on the review and assessment result, BAPETEN may stipulate, reject, delay or terminate the license. In anticipation of expansion of the nuclear program in Indonesia, BAPETEN should have an integrated and updated system for review and asses the licensing application. For this reason, an expert system for the review and asses the licensing application, so-called PRIBEN (Perizinan Reaktor, Instalasi dan Bahan Nuklir/Licensing of Reactor, Nuclear Installations and Nuclear Materials), is developed which runs on the online-based reality environment

USHPRR FUEL FABRICATION PILLAR: FABRICATION STATUS, PROCESS OPTIMIZATIONS, AND FUTURE PLANS

Energy Technology Data Exchange (ETDEWEB)

Wight, Jared M.; Joshi, Vineet V.; Lavender, Curt A.

2018-03-12

The Fuel Fabrication (FF) Pillar, a project within the U.S. High Performance Research Reactor Conversion program of the National Nuclear Security Administration’s Office of Material Management and Minimization, is tasked with the scale-up and commercialization of high-density monolithic U-Mo fuel for the conversion of appropriate research reactors to use of low-enriched fuel. The FF Pillar has made significant steps to demonstrate and optimize the baseline co-rolling process using commercial-scale equipment at both the Y-12 National Security Complex (Y-12) and BWX Technologies (BWXT). These demonstrations include the fabrication of the next irradiation experiment, Mini-Plate 1 (MP-1), and casting optimizations at Y-12. The FF Pillar uses a detailed process flow diagram to identify potential gaps in processing knowledge or demonstration, which helps direct the strategic research agenda of the FF Pillar. This paper describes the significant progress made toward understanding the fuel characteristics, and models developed to make informed decisions, increase process yield, and decrease lifecycle waste and costs.

Application of new technology to the fabrication and installation of stainless steel pipework

International Nuclear Information System (INIS)

Halford, P.; Carrick, L.

1989-01-01

BNFL has been constructing new reprocessing plant on a continuous basis since the late seventies. Initially the productivity achieved when fabricating and welding stainless steel to nuclear standards was poor. In order to complete projects to programme and cost BNFL developed a Total Fabrication System that is now applied to all of their construction projects and has resulted in overall productivity increases by a factor of 2.4 with major quality and cost benefits. The development of the Total Fabrication System is described

Approaches to characterization of nuclear material for establishment of nuclear forensics

International Nuclear Information System (INIS)

Okazaki, Hiro; Sumi, Mika; Sato, Mitsuhiro; Kayano, Masashi; Kageyama, Tomio; Shinohara, Nobuo; Martinez, Patrick; Xu, Ning; Thomas, Mariam; Porterfield, Donivan; Colletti, Lisa; Schwartz, Dan; Tandon, Lav

2014-01-01

The Plutonium Fuel Development Center (PFDC) of Japan Atomic Energy Agency has been analyzing isotopic compositions and contents of plutonium and uranium as well as trace impurities and physics in the nuclear fuel from MOX fuel fabrication process for accountancy and process control purpose. These analytical techniques are also effective for nuclear forensics to identify such as source, history, and route of the material by determining a composition and characterization of nuclear material. Therefore, PFDC cooperates with Los Alamos National Laboratory which has broad experience and established measurement skill for nuclear forensics, and evaluates the each method, procedure and analytical data toward R and D of characterizing a nuclear material for forensic purposes. This paper describes the approaches to develop characterization techniques of nuclear material for nuclear forensics purposes at PFDC. (author)

Spent Nuclear Fuel project, project management plan

International Nuclear Information System (INIS)

Fuquay, B.J.

1995-01-01

The Hanford Spent Nuclear Fuel Project has been established to safely store spent nuclear fuel at the Hanford Site. This Project Management Plan sets forth the management basis for the Spent Nuclear Fuel Project. The plan applies to all fabrication and construction projects, operation of the Spent Nuclear Fuel Project facilities, and necessary engineering and management functions within the scope of the project

Specification and acceptance testing of nuclear medicine equipment

International Nuclear Information System (INIS)

Wegst, A.V.; Erickson, J.J.

1984-01-01

The purchase of nuclear medicine equipment is of prime importance in the operation of a clinical service. Failure to properly evaluate the potential uses of the instrumentation and the various operational characteristics of the equipment can often result in the purchase of inappropriate or inferior instruments. The magnitude of the purchase in terms of time and financial investments make it imperative that the purchase be approached in a systematic manner. Consideration of both the intended clinical functions and personnel requirements is important. It is necessary also to evaluate the ability of the equipment vendor to support the instrumentation after the purchase has been completed and the equipment installed in the clinical site. The desired specifications of the instrument characteristics should be stated in terms that can be verified by acceptance testing. The complexity of modern instrumentation and the sensitivity of it to the environment require the buyer to take into account the potential problems of controlling the temperature, humidity, and electrical power of the installation site. If properly and systematically approached, the purchase of new nuclear medicine instrumentation can result in the acquisition of a powerful diagnostic tool which will have a useful lifetime of many years. If not so approached, it may result in the expenditure of a large amount of money and personnel time without the concomitant return in useful clinical service. (author)

Homogeneity of blended nuclear fuel powders after pneumatic transport

International Nuclear Information System (INIS)

Smeltzer, E.E.; Skriba, M.C.; Lyon, W.L.

1982-01-01

A study of the pneumatic transport of fine (approx. 1μm) cohesive nuclear fuel powders was conducted for the U.S. Department of Energy to demonstrate the feasibility of this method of transport and to develop a design data base for use in a large scale nuclear fuel production facility. As part of this program, a considerable effort was directed at following the homogeneity of blended powders. Since different reactors require different enrichments, blending and subsequent transport are critical parts of the fabrication sequence. The various materials used represented analogs of a wide range of powders and blends that could be expected in a commercial mixed oxide fabrication facility. All UO 2 powders used were depleted and a co-precipitated master mix of (U, Th)O 2 was made specifically for this program, using thorium as an analog for plutonium. In order to determine the effect of pneumatic transport on a blended powder, samples were taken from a feeder vessel before each test, and from a receiver vessel and a few line sections after each transfer test. The average difference between the before and after degree of non-homogeneity was < 1%, for the 21 tests considered. This shows that overall, the pneumatic transport of blended, fine nuclear fuel powders is possible, with only minor unblending occurring

The future supply of and demand for candidate materials for the fabrication of nuclear fuel waste disposal containers

International Nuclear Information System (INIS)

Grover, L.K.

1990-01-01

This report summarizes the findings of a literature survey carried out to assess the future world supply of and demand for titanium, copper and lead. These metals are candidate materials for the fabrication of containers for the immobilization and disposal of Canada's nuclear used-fuel waste for a reference Used-fuel Disposal Centre. Such a facility may begin operation by approximately 2020, and continue for about 40 years. The survey shows that the world has abundant supplies of titanium minerals (mostly in the form of ilmenite), which are expected to last up to at least 2110. However, for copper and lead the balance between supply and demand may warrant increased monitoring beyond the year 2000. A number of factors that can influence future supply and demand are discussed in the report

Report on fabrication of pin components for fuel fabrication in FUJI project (Co-operation in the research and development of advanced sphere-pac fuel among PSI, JNC, and NRG)

International Nuclear Information System (INIS)

Suzuki, Masahiro; Hinai, Hiroshi; Shigetome, Yoshiaki; Kono, Shusaku; Matsuzaki, Masaaki

2003-03-01

Japan Nuclear Cycle Development Institute (JNC) has conducted the co-operation concerning vibro-packed fuels with Paul Scherrer Institut (PSI) in Switzerland and Nuclear Research and consultancy Group (NRG) in the Netherlands. The project 'Research and Development of advanced Sphere-pac Fuel' is called FUJI (FUel irradiations for JNC and PSI) Project. In this project, three types of fuels that are sphere-pac fuels, vipac fuels, and pellet fuels will be irradiated in the High Flux Reactor (HFR) to compare their performance. Based on the drawing which has been agreed among three parties, fabrication of the pin components and welding of the upper and lower connection end plugs were performed in accordance with ISO9001 in JNC. This report describes data of the fabricated pin components, results of welding qualification tests, and quality assurance of the welded components. The fabrication of pin components was successfully completed and they were delivered to PSI in October 2002. (author)

Maintenance system for immersed seals, specifically for nuclear reactors

International Nuclear Information System (INIS)

Poindexter, A.M.; Ricks, H.E.

1977-01-01

The invention concerns the immersed seals of nuclear reactors and specifically a maintenance system for the immersed seals of the revolving closing plugs of liquid metal breeder nuclear reactors. A liquid sodium immersed joint may be located at a given place or be surrounded by heating elements so that the sodium stays liquid whilst the reactor is working. In other cases, the sodium in the immersed seal is allowed to solidify whilst the reactor is working, thereby increasing the efficiency of the seal. At all events, the sodium must be in a liquid state during reloading with fuel to enable the plug to turn. The invention consists in fitting an ultrasonic transducer to the closure head of the reactor vessel so that the vibration emitting surface directs these vibrations towards the immersed seals so as to detach the deposits of impurities on them and ensure the wetting of the metal surfaces of which they are formed. Additionally, an envelope that can be placed around the ultrasonic transducer in conjunction with a suction appliance provides a mechanism through which the impurities can be removed from the area of the immersed seal [fr

Competition still fierce in the US fuel fabrication market

International Nuclear Information System (INIS)

Schwartz, M.H.

1990-01-01

The US market for nuclear fuel fabrication services is characterized by an annual production capacity significantly in excess of both current and anticipated demand. The trends toward longer operating cycle lengths and higher burnup fuel continue in the United States. This, together with the lack of any prospects for new light water reactors coming on line in the US during the next ten years, is expected to hold the annual demand for fuel fabrication services from US LWRs at around 2000t of uranium into the next century. (author)

LEU fuel fabrication program for the RECH-1 reactor. Status report

International Nuclear Information System (INIS)

Chavez, J.C.; Barrera, M.; Jimenez, O.; Lisboa, J.; Marin, J.

2000-01-01

In 1995 a 50 LEU U 3 Si 2 fuel elements fabrication program for the RECH-1 research reactor was established at the Comision Chilena de Energia Nuclear, CCHEN. After a fabrication process qualification stage, in 1998, four elements were early delivered to the reactor in order to start an irradiation qualification stage. The irradiation has reached an estimated 10% burn-up and no fabrication problems have been detected up to this burn-up level. During 1999 and up to the first quarter of 2000, 19 fuel elements were produced and 7 fuel elements are expected for the end of 2000. This report presents an updated summary of the main results obtained in this fuel fabrication program. A summary of other activities generated by this program, such as in core follow-up of the four leader fuel elements, ISO 9001 implementation for the fabrication process and a fabrication and qualification optimization planning, is also presented here. (author)

Site-specific parameter values for the Nuclear Regulatory Commission's food pathway dose model

International Nuclear Information System (INIS)

Hamby, D.M.

1992-01-01

Routine operations at the Savannah River Site (SRS) in Western South Carolina result in radionuclide releases to the atmosphere and to the Savannah River. The resulting radiation doses to the off-site maximum individual and the off-site population within 80 km of the SRS are estimated on a yearly basis. These estimates are currently generated using dose models prescribed for the commercial nuclear power industry by the Nuclear Regulatory Commission (NRC). The NRC provides default values for dose-model parameters for facilities without resources to develop site-specific values. A survey of land- and water-use characteristics for the Savannah River area has been conducted to determine site-specific values for water recreation, consumption, and agricultural parameters used in the NRC Regulatory Guide 1.109 (1977) dosimetric models. These site parameters include local characteristics of meat, milk, and vegetable production; recreational and commercial activities on the Savannah River; and meat, milk, vegetable, and seafood consumption rates. This paper describes how parameter data were obtained at the Savannah River Site and the impacts of such data on off-site dose. Dose estimates using site-specific parameter values are compared to estimates using the NRC default values

Chemical aspects of nuclear fuel fabrication processes

Energy Technology Data Exchange (ETDEWEB)

Naylor, A; Ellis, J F; Watson, R H

1986-04-01

Processes used by British Nuclear Fuels plc for the conversion of uranium ore concentrates to uranium metal and uranium hexafluoride, are reviewed. Means of converting the latter compound, after enrichment, to sintered UO/sub 2/ fuel bodies are also described. An overview is given of the associated chemical engineering technology.

Pilot-scale demonstration of the modified direct denitration process to prepare uranium oxide for fuel fabrication evaluation

International Nuclear Information System (INIS)

Kitts, F.G.

1994-04-01

The Uranium-Atomic Vapor Laser Isotope Separation (U-AVLIS) Program has the objective of developing a cost-competitive enrichment process that will ultimately replace the gaseous diffusion process used in the United States. Current nuclear fuel fabricators are set up to process only the UF 6 product from gaseous diffusion enrichment. Enriched uranium-iron alloy from the U-AVLIS separator system must be chemically converted into an oxide form acceptable to these fabricators to make fuel pellets that meet American Society for Testing and Materials (ASTM) and utility company specifications. A critical step in this conversion is the modified direct denitration (MDD) that has been selected and presented in the AVLIS Conceptual Design for converting purified uranyl nitrate to UO 3 to be shipped to fabricators for making UO 2 pellets for power reactor fuel. This report describes the MDD process, the equipment used, and the experimental work done to demonstrate the conversion of AVLIS product to ceramic-grade UO 3 suitable for making reactor-grade fuel pellets

Design Evolutuion of Hot Isotatic Press Cans for NTP Cermet Fuel Fabrication

Science.gov (United States)

Mireles, O. R.; Broadway, J.; Hickman, R.

2014-01-01

Nuclear Thermal Propulsion (NTP) is under consideration for potential use in deep space exploration missions due to desirable performance properties such as a high specific impulse (> 850 seconds). Tungsten (W)-60vol%UO2 cermet fuel elements are under development, with efforts emphasizing fabrication, performance testing and process optimization to meet NTP service life requirements [1]. Fuel elements incorporate design features that provide redundant protection from crack initiation, crack propagation potentially resulting in hot hydrogen (H2) reduction of UO2 kernels. Fuel erosion and fission product retention barriers include W coated UO2 fuel kernels, W clad internal flow channels and fuel element external W clad resulting in a fully encapsulated fuel element design as shown.

Construction for Nuclear Installations. Specific Safety Guide

International Nuclear Information System (INIS)

2015-01-01

This Safety Guide provides recommendations and guidance based on international good practices in the construction of nuclear installations, which will enable construction to proceed with high quality. It can be applied to support the development, implementation and assessment of construction methods and procedures and the identification of good practices for ensuring the quality of the construction to meet the design intent and ensure safety. It will be a useful tool for regulatory bodies, licensees and new entrant countries for nuclear power plants and other nuclear installations

The design, fabrication and maintenance of semi-trailers employed in the highway transport of weight-concentrated radioactive loads

Energy Technology Data Exchange (ETDEWEB)

Huffman, D.S. [Allied-Signal Inc., Metropolis, IL (United States)

1991-12-31

Transportation of weight-concentrated radioactive loads by truck is an essential part of a safe and economical nuclear industry. This proposed standard presents guidance and performance criteria for the safe transport of these weight-concentrated radioactive loads. ANSI N14.30 will detail specific requirements for the design, fabrication, testing, in-service inspections, maintenance and certification of the semi-trailers to be employed in said service. Furthermore, guidelines for a quality assurance program are also enumerated. This standard would apply to any semi-trailer that may or may not be specifically designed to carry weight-concentrated loads. Equipment not suitable per the criteria established in the standard would be removed from service. The nature of the nuclear industry and the need for a positive public perception of the various processes and players, mandates that the highway transportation of weight-concentrated radioactive loads be standardized and made inherently safe. This proposed standard takes a giant step in that direction.

Specification of steam generator, condenser and regenerative heat exchanger materials for nuclear applications

International Nuclear Information System (INIS)

Jovasevic, J.V.; Stefanovic, V.M.; Spasic, Z.LJ.

1977-01-01

The basic standards specifications of materials for nuclear applications are selected. Seamless Ni-Cr-Fe alloy Tubes (Inconel-600) for steam generators, condensers and other heat exchangers can be employed instead of austenitic stainless steal or copper alloys tubes; supplementary requirements for these materials are given. Specifications of Ni-Cr-Fe alloy plate, sheet and strip for steam generator lower sub-assembly, U-bend seamless copper-alloy tubes for heat exchanger and condensers are also presented. At the end, steam generator channel head material is proposed in the specification for carbon-steel castings suitable for welding

Romanian nuclear fuel fabrication and in-reactor fuel operational experience

International Nuclear Information System (INIS)

Budan, O.

2003-01-01

A review of the Romanian nuclear program since mid 60's is made. After 1990, the new Romanian nuclear power authority, RENEL-GEN, elaborated a realistic Nuclear Fuel Program. This program went through the Romanian nuclear fuel plant qualification with the Canadian (AECL and ZPI) support, restarting in January 1995 of the industrial nuclear fuel production, quality evaluation of the fuel produced before 1990 and the recovery of this fuel. This new policy produced good results. FCN is since 1995 the only CANDU fuel supplier from outside Canada recognised by AECL as an authorised CANDU fuel manufacturer. The in-reactor performances and behaviour of the fuel manufactured by FCN after its qualification have been excellent. Very low - more then five times lesser than the design value - fuel defect rate has been recorded up to now and the average discharge of this fuel was with about 9% greater than the design value. Since mid 1998 when SNN took charge of the production of nuclear generated electricity, FCN made significant progresses in development and procurement of new and more efficient equipment and is now very close to double its fuel production capacity. After the completion of the recovery of the fuel produced before June 1990, FCN is already prepared to shift its fuel production to the so-called 'heavy' bundle containing about 19.3 kg of Uranium per bundle

Standard specification for uranium hexafluoride enriched to less than 5 % 235U

CERN Document Server

American Society for Testing and Materials. Philadelphia

2010-01-01

1.1 This specification covers nuclear grade uranium hexafluoride (UF6) that either has been processed through an enrichment plant, or has been produced by the blending of Highly Enriched Uranium with other uranium to obtain uranium of any 235U concentration below 5 % and that is intended for fuel fabrication. The objectives of this specification are twofold: (1) To define the impurity and uranium isotope limits for Enriched Commercial Grade UF6 so that, with respect to fuel design and manufacture, it is essentially equivalent to enriched uranium made from natural UF6; and (2) To define limits for Enriched Reprocessed UF6 to be expected if Reprocessed UF6 is to be enriched without dilution with Commercial Natural UF6. For such UF6, special provisions, not defined herein, may be needed to ensure fuel performance and to protect the work force, process equipment, and the environment. 1.2 This specification is intended to provide the nuclear industry with a standard for enriched UF6 that is to be used in the pro...

An Ethology of Urban Fabric(s)

DEFF Research Database (Denmark)

Fritsch, Jonas; Thomsen, Bodil Marie Stavning

2014-01-01

The article explores a non-metaphorical understanding of urban fabric(s), shifting the attention from a bird’s eye perspective to the actual, textural manifestations of a variety of urban fabric(s) to be studied in their real, processual, ecological and ethological complexity within urban life. We...... effectuate this move by bringing into resonance a range of intersecting fields that all deal with urban fabric(s) in complementary ways (interaction design and urban design activism, fashion, cultural theory, philosophy, urban computing)....

EDF's approach to determine specifications for nuclear power plant bulk chemicals

International Nuclear Information System (INIS)

Basile, Alix; Dijoux, Michel; Le-Calvar, Marc; Gressier, Frederic; Mole, Didier

2012-09-01

Chemical impurities in the primary, secondary and auxiliary nuclear power plants circuits generate risks of corrosion of the fuel cladding, steel and nickel based alloys. The PMUC (Products and Materials Used in plants) organization established by EDF intends to limit this risk by specifying maximum levels of impurities in products and materials used for the operation and maintenance of Nuclear Power Plants (NPPs). Bulk chemicals specifications, applied on primary and secondary circuit chemicals and hydrogen and nitrogen gases, are particularly important to prevent chemical species to be involved in the corrosion of the NPPs materials. The application of EDF specifications should lead to reasonably exclude any risk of degradation of the first and second containment barriers and auxiliary circuits Important to Safety (IPS) by limiting the concentrations of chlorides, fluorides, sulfates... The risk of metal embrittlement by elements with low melting point (mercury, lead...) is also included. For the primary circuit, the specifications intend to exclude the risk of activation of impurities introduced by the bulk chemicals. For the first containment barrier, to reduce the risk of deposits like zeolites, PMUC products specifications set limit values for calcium, magnesium, aluminum and silica. EDF's approach for establishing specifications for bulk chemicals is taking also into account the capacity of industrial production, as well as costs, limitations of analytical control methods (detection limits) and environmental releases issues. This paper aims to explain EDF's approach relative to specifications of impurities in bulk chemicals. Also presented are the various parameters taken into account to determine the maximum pollution levels in the chemicals, the theoretical hypothesis to set the specifications and the calculation method used to verify that the specifications are suitable. (authors)

Water chemistry control in thermal and nuclear power plants. 9. Nuclear fuel management

International Nuclear Information System (INIS)

2008-01-01

The chemical management of fuels in nuclear power plants aims at maintenance of the soundness of nuclear fuels and at reduction of the radiation exposure of the working employees. With regard to the former, particular attention should be paid to the fabrication process of fuel assembly, mainly for chemical management for fuel cladding tubes together with fuel pellet-clad chemical interactions, and to the outer tubes in the power plants. With regard to the latter, the fabrication process should be carefully controlled to prevent radioactive impurity increase in primary cooling water systems by maintaining cleaning level and decreasing surface contamination. Reactions of zircalloy with water or hydrogen forming ZrH 2 , sintered density of UO 2 pellet controlling water content, pellet-clad interactions, stress corrosion cracking, crud induced fuel failure, behaviors of such fission products as I, Xe, Kr, and Cs in plants are also important to water and chemical management of nuclear fuels. (S. Ohno)

Nuclear fuels

International Nuclear Information System (INIS)

Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F.

2009-01-01

Fuel is one of the essential components in a reactor. It is within that fuel that nuclear reactions take place, i.e. fission of heavy atoms, uranium and plutonium. Fuel is at the core of the reactor, but equally at the core of the nuclear system as a whole. Fuel design and properties influence reactor behavior, performance, and safety. Even though it only accounts for a small part of the cost per kilowatt-hour of power provided by current nuclear power plants, good utilization of fuel is a major economic issue. Major advances have yet to be achieved, to ensure longer in-reactor dwell-time, thus enabling fuel to yield more energy; and improve ruggedness. Aside from economics, and safety, such strategic issues as use of plutonium, conservation of resources, and nuclear waste management have to be addressed, and true technological challenges arise. This Monograph surveys current knowledge regarding in-reactor behavior, operating limits, and avenues for R and D. It also provides illustrations of ongoing research work, setting out a few noteworthy results recently achieved. Content: 1 - Introduction; 2 - Water reactor fuel: What are the features of water reactor fuel? 9 (What is the purpose of a nuclear fuel?, Ceramic fuel, Fuel rods, PWR fuel assemblies, BWR fuel assemblies); Fabrication of water reactor fuels (Fabrication of UO 2 pellets, Fabrication of MOX (mixed uranium-plutonium oxide) pellets, Fabrication of claddings); In-reactor behavior of UO 2 and MOX fuels (Irradiation conditions during nominal operation, Heat generation, and removal, The processes involved at the start of irradiation, Fission gas behavior, Microstructural changes); Water reactor fuel behavior in loss of tightness conditions (Cladding, the first containment barrier, Causes of failure, Consequences of a failure); Microscopic morphology of fuel ceramic and its evolution under irradiation; Migration and localization of fission products in UOX and MOX matrices (The ceramic under irradiation

Nuclear fuels

Energy Technology Data Exchange (ETDEWEB)

Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F

2009-07-01

nature of spent nuclear fuel, Anticipated evolution of fuel in dry storage, Anticipated evolution of fuel in deep geological disposal); Boiling-water reactor fuel (Similarities, and differences with PWR fuel, Axial and radial zoning, Rod and channel box sizes, Poisoning and reactivity control, Cladding specific characteristics, Trends in fuel evolution); 3 - Liquid-metal-cooled fast reactor fuel: Fast-neutron irradiation damage in structural materials (Fast-neutron-induced damage in metals, What materials should be used?); Fuels and targets for fast-reactor transmutation (Fast reactors: reactors affording the ability to carry out effective actinide transmutation, Recycling: homogeneous, or heterogeneous?); 4 - gas-cooled reactor fuel: Particle fuel (From the initial concept to the advanced TRISO particle concept, Kernel fabrication processes, Particle coating by chemical vapor deposition, Fuel element fabrication: particle compaction, Characterization of fuel particles, and elements, From HTR fuel to VHTR and GFR fuels: the GAIA facility at CEA/Cadarache); Irradiation behavior of particle fuels (Particle fuel: a variety of failure modes for a high-strength object, The amoeba effect, Fission product behavior, and diffusion in particle fuels); Mechanical modeling of particle fuel; Very-high-temperature reactor (VHTR) fuel; Gas-cooled fast reactor (GFR) fuel (The specifications for GFR fuel, GFR fissile material, First containment baffler materials, GFR fuel element concepts); 5 - Research reactor fuels (A considerable feedback from experience, Conversion of French reactors to low-enriched ({<=}20% U-235)U{sub 3}Si{sub 2} fuel, Conversion of all reactors: R and D requirements for high-performance reactors, An 'advanced' research reactor fuel: UMo, The startup fuel for the Jules Horowitz Reactor (JHR) will still be U{sub 3}Si{sub 2}-Al; 6 - An instrument for future fuel research: the Jules Horowitz Reactor (JHR): Fuel irradiation experiments in JHR, JHR: a flexible

Fabrication and testing of U–7Mo monolithic plate fuel with Zircaloy cladding

Energy Technology Data Exchange (ETDEWEB)

Pasqualini, E.E. [Laboratorio de Nanotecnología Nuclear, Centro Atómico Constituyentes, Comisión Nacional de Energía Atómica, Av. General Paz 1499, B1650KNA, San Martín, Prov. Buenos Aires (Argentina); Robinson, A.B. [Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID, 83415-6188 (United States); Porter, D.L., E-mail: Douglas.Porter@inl.gov [Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID, 83415-6188 (United States); Wachs, D.M. [Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID, 83415-6188 (United States); Finlay, M.R. [Australian Nuclear Science and Technology Organisation, PMB 1, Menai, NSW, 2234 (Australia)

2016-10-15

Nuclear fuel designs are being developed to replace highly enriched fuel used in research and test reactors with fuels of low enrichment. In the most challenging cases, U–(7–10 wt%)Mo monolithic plate fuels are proposed. One of the considered designs includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction during service. Zircaloy cladding, specifically Zry–4, was investigated as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry–4 clad U–7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry–4 and U–(7–10)Mo have similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch, which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly during or between roll passes. The final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction—either from fabrication or in-reactor testing—and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.7E+21 (average) fissions/cm{sup 3}, 3.8E+21 (peak).

Reference Neutron Radiographs of Nuclear Reactor Fuel

DEFF Research Database (Denmark)

Domanus, Joseph Czeslaw

1986-01-01

Reference neutron radiographs of nuclear reactor fuel were produced by the Euraton Neutron Radiography Working Group and published in 1984 by the Reidel Publishing Company. In this collection a classification is given of the various neutron radiographic findings, that can occur in different parts...... of pelletized, annular and vibro-conpacted nuclear fuel pins. Those parts of the pins are shown where changes of appearance differ from those for the parts as fabricated. Also radiographs of those as fabricated parts are included. The collection contains 158 neutron radiographs, reproduced on photographic paper...

Fabricated Elastin.

Science.gov (United States)

Yeo, Giselle C; Aghaei-Ghareh-Bolagh, Behnaz; Brackenreg, Edwin P; Hiob, Matti A; Lee, Pearl; Weiss, Anthony S

2015-11-18

The mechanical stability, elasticity, inherent bioactivity, and self-assembly properties of elastin make it a highly attractive candidate for the fabrication of versatile biomaterials. The ability to engineer specific peptide sequences derived from elastin allows the precise control of these physicochemical and organizational characteristics, and further broadens the diversity of elastin-based applications. Elastin and elastin-like peptides can also be modified or blended with other natural or synthetic moieties, including peptides, proteins, polysaccharides, and polymers, to augment existing capabilities or confer additional architectural and biofunctional features to compositionally pure materials. Elastin and elastin-based composites have been subjected to diverse fabrication processes, including heating, electrospinning, wet spinning, solvent casting, freeze-drying, and cross-linking, for the manufacture of particles, fibers, gels, tubes, sheets and films. The resulting materials can be tailored to possess specific strength, elasticity, morphology, topography, porosity, wettability, surface charge, and bioactivity. This extraordinary tunability of elastin-based constructs enables their use in a range of biomedical and tissue engineering applications such as targeted drug delivery, cell encapsulation, vascular repair, nerve regeneration, wound healing, and dermal, cartilage, bone, and dental replacement. © 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

San Rafael mining and fabrication complex today

International Nuclear Information System (INIS)

Navarra, Pablo; Aldebert, Sergio R.

2005-01-01

In Mendoza province, 35 km West San Rafael city, is located a CNEA installation for uranium ore extraction and concentration: the San Rafael Mining and Fabrication Complex. By the middle of the nineties, as a consequence of the very low prices of uranium concentrate in the international market and of the high internal production costs, uranium extraction was stopped. To day, the international price of the concentrate had a very important increase and the Government has decided the completion of the Atucha II Nuclear Power Station construction. Moreover, studies have been started for new nuclear power plants. In such circumstances the reactivation of the Complex will make sure the uranium supply for our nuclear power stations, contributing to the improvement of the energy generation mix in our country. (author) [es

Study of developing nuclear fabrication facility's integrated emergency response manual

International Nuclear Information System (INIS)

Kim, Taeh Yeong; Cho, Nam Chan; Han, Seung Hoon; Moon, Jong Han; Lee, Jin Hang; Min, Guem Young; Han, Ji Ah

2016-01-01

Public begin to pay attention to emergency management. Thus, public's consensus on having high level of emergency management system up to advanced country's is reached. In this social atmosphere, manual is considered as key factor to prevent accident or secure business continuity. Therefore, we first define possible crisis at KEPCO Nuclear Fuel (hereinafter KNF) and also make a 'Reaction List' for each crisis situation at the view of information-design. To achieve it, we analyze several country's crisis response manual and then derive component, indicate duties and roles at the information-design point of view. From this, we suggested guideline to make 'Integrated emergency response manual(IERM)'. The manual we used before have following few problems; difficult to applicate at the site, difficult to deliver information. To complement these problems, we searched manual elements from the view of information-design. As a result, we develop administrative manual. Although, this manual could be thought as fragmentary manual because it confined specific several agency/organization and disaster type

Design, fabrication and installation of irradiation facilities

International Nuclear Information System (INIS)

Kim, Yong Sung; Lee, C. Y.; Kim, J. Y.; Chi, D. Y.; Kim, S. H.; Ahn, S. H.; Kim, S. J.; Kim, J. K.; Yang, S. H.; Yang, S. Y.; Kim, H. R.; Kim, H.; Lee, K. H.; Lee, B. C.; Park, C.; Lee, C. T.; Cho, S. W.; Kwak, K. K.; Suk, H. C.

1997-07-01

The principle contents of this project are to design, fabricate and install the steady-state fuel test loop and non-instrumented capsule in HANARO for nuclear technology development. This project will be completed in 1999, the basic and detail design, safety analysis, and procurement of main equipment for fuel test loop have been performed and also the piping in gallery and the support for IPS piping in reactor pool have been installed in 1994. In the area of non-instrumented capsule for material irradiation test, the fabrication of capsule has been completed. Procurement, fabrication and installation of the fuel test loop will be implemented continuously till 1999. As besides, as these irradiation facilities will be installed in HANARO, review of safety concern, discussion with KINS for licensing and safety analysis report has been submitted to KINS to get a license and review of HANARO interface have been performed respectively. (author). 39 refs., 28 tabs., 21 figs

Design, fabrication and installation of irradiation facilities

Energy Technology Data Exchange (ETDEWEB)

Kim, Yong Sung; Lee, C. Y.; Kim, J. Y.; Chi, D. Y.; Kim, S. H.; Ahn, S. H.; Kim, S. J.; Kim, J. K.; Yang, S. H.; Yang, S. Y.; Kim, H. R.; Kim, H.; Lee, K. H.; Lee, B. C.; Park, C.; Lee, C. T.; Cho, S. W.; Kwak, K. K.; Suk, H. C. [and others

1997-07-01

The principle contents of this project are to design, fabricate and install the steady-state fuel test loop and non-instrumented capsule in HANARO for nuclear technology development. This project will be completed in 1999, the basic and detail design, safety analysis, and procurement of main equipment for fuel test loop have been performed and also the piping in gallery and the support for IPS piping in reactor pool have been installed in 1994. In the area of non-instrumented capsule for material irradiation test, the fabrication of capsule has been completed. Procurement, fabrication and installation of the fuel test loop will be implemented continuously till 1999. As besides, as these irradiation facilities will be installed in HANARO, review of safety concern, discussion with KINS for licensing and safety analysis report has been submitted to KINS to get a license and review of HANARO interface have been performed respectively. (author). 39 refs., 28 tabs., 21 figs.

Recent advances in fuel fabrication techniques and prospects for the nineties

International Nuclear Information System (INIS)

Frain, R.G.; Caudill, H.L.; Faulhaber, R.

1987-01-01

Advanced Nuclear Fuels Corporation's approach and experience with the application of a flexible, just-in-time manufacturing philosophy to the production of customized nuclear fuel is described. Automation approaches to improve productivity are described. The transfer of technology across product lines is discussed as well as the challenges presented by a multiple product fabrication facility which produces a wide variety of BWR and PWR designs. This paper also describes the method of managing vendor quality control programs in support of standardization and clarity of documentation. Process simplification and the ensuing experience are discussed. Prospects for fabrication process advancements in the nineties are given with emphasis on the benefits of dry conversion of UF 6 to UO 2 powder, and increased use of automated and computerized inspection techniques. (author)

A review on the development of the MOX fuel fabrication technology

Energy Technology Data Exchange (ETDEWEB)

Kim, See Hyung; Lee, Yung Woo; Sohn, Dong Sung; Yang, Myung Seung; Bae, Kee Kwang; Nah, Sang Hoh; Kim, Han Soo; Lee, Jung Won; Kim, Bong Koo; Song, Keun Woo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-07-01

Development of the Mixed Oxide(MOX) fuel fabrication technology was reviewed in this study. Firstly, the feasibility of Pu utilization for nuclear fuel was analyzed by comparison of nuclear characteristics between U and Pu. Secondly, the feature and problem of processes developed so far was revealed and analyzed by reviewing each process in terms of technical difficulties and in connection with the pellet characteristics. Also, fabrication facilities currently existing were analyzed to understand particularities and circumstances in view of Pu handling, and finally, in-reactor behaviors of MOX fuel was compared with those of U fuel to understand how the Pu has an effect on fuel was compared with those of U fuel to understand how the Pu has an effect on fuel pellet structure and fuel rod. 73 figs., 15 tabs., 58 refs. (Author).

Nuclear

International Nuclear Information System (INIS)

2014-01-01

This document proposes a presentation and discussion of the main notions, issues, principles, or characteristics related to nuclear energy: radioactivity (presence in the environment, explanation, measurement, periods and activities, low doses, applications), fuel cycle (front end, mining and ore concentration, refining and conversion, fuel fabrication, in the reactor, back end with reprocessing and recycling, transport), the future of the thorium-based fuel cycle (motivations, benefits and drawbacks), nuclear reactors (principles of fission reactors, reactor types, PWR reactors, BWR, heavy-water reactor, high temperature reactor of HTR, future reactors), nuclear wastes (classification, packaging and storage, legal aspects, vitrification, choice of a deep storage option, quantities and costs, foreign practices), radioactive releases of nuclear installations (main released radio-elements, radioactive releases by nuclear reactors and by La Hague plant, gaseous and liquid effluents, impact of releases, regulation), the OSPAR Convention, management and safety of nuclear activities (from control to quality insurance, to quality management and to sustainable development), national safety bodies (mission, means, organisation and activities of ASN, IRSN, HCTISN), international bodies, nuclear and medicine (applications of radioactivity, medical imagery, radiotherapy, doses in nuclear medicine, implementation, the accident in Epinal), nuclear and R and D (past R and D programmes and expenses, main actors in France and present funding, main R and D axis, international cooperation)

Alternative Fabrication of Recycling Fast Reactor Metal Fuel

International Nuclear Information System (INIS)

Kim, Ki-Hwan; Kim, Jong Hwan; Song, Hoon; Kim, Hyung-Tae; Lee, Chan-Bock

2015-01-01

Metal fuels such as U-Zr/U-Pu-Zr alloys have been considered as a nuclear fuel for a sodium-cooled fast reactor (SFR) related to the closed fuel cycle for managing minor actinides and reducing a high radioactivity levels since the 1980s. In order to develop innovative fabrication method of metal fuel for preventing the evaporation of volatile elements such as Am, modified casting under inert atmosphere has been applied for metal fuel slugs for SFR. Alternative fabrication method of fuel slugs has been introduced to develop an improved fabrication process of metal fuel for preventing the evaporation of volatile elements. In this study, metal fuel slugs for SFR have been fabricated by modified casting method, and characterized to evaluate the feasibility of the alternative fabrication method. In order to prevent evaporation of volatile elements such as Am and improve quality of fuel slugs, alternative fabrication methods of metal fuel slugs have been studied in KAERI. U-10Zr-5Mn fuel slug containing volatile surrogate element Mn was soundly cast by modified injection casting under modest pressure. Evaporation of Mn during alternative casting could not be detected by chemical analysis. Mn element was most recovered with prevention of evaporation by alternative casting. Modified injection casting has been selected as an alternative fabrication method in KAERI, considering evaporation prevention, and proven benefits of high productivity, high yield, and good remote control

Reviewing PSA-based analyses to modify technical specifications at nuclear power plants

International Nuclear Information System (INIS)

Samanta, P.K.; Martinez-Guridi, G.; Vesely, W.E.

1995-12-01

Changes to Technical Specifications (TSs) at nuclear power plants (NPPs) require review and approval by the United States Nuclear Regulatory Commission (USNRC). Currently, many requests for changes to TSs use analyses that are based on a plant's probabilistic safety assessment (PSA). This report presents an approach to reviewing such PSA-based submittals for changes to TSs. We discuss the basic objectives of reviewing a PSA-based submittal to modify NPP TSs; the methodology of reviewing a TS submittal, and the differing roles of a PSA review, a PSA Computer Code review, and a review of a TS submittal. To illustrate this approach, we discuss our review of changes to allowed outage time (AOT) and surveillance test interval (STI) in the TS for the South Texas Project Nuclear Generating Station. Based on this experience gained, a check-list of items is given for future reviewers; it can be used to verify that the submittal contains sufficient information, and also that the review has addressed the relevant issues. Finally, recommended steps in the review process and the expected findings of each step are discussed

Specific aspects of insurance of nuclear risks

International Nuclear Information System (INIS)

Angelici, C.

1980-03-01

The following questions are discussed in connection with the insurance of nuclear risks: insurance techniques, the nuclear operator's limitation of liability in amount and in time, its channelling, the principle of sole liability and exonerations, the insurers' position, the cover provided and state intervention beyond that amount. (NEA) [fr

Guidance for the application of an assessment methodology for innovative nuclear energy systems. INPRO manual - Safety of nuclear fuel cycle facilities. Vol. 9 of the final report of phase 1 of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)

International Nuclear Information System (INIS)

2008-11-01

assessment (including the safety analysis). This chapter also discusses the timing of the INPRO assessment. Chapter 3 identifies the different types of facilities that can form part of the nuclear fuel cycle. This chapter also provides an overview of the safety aspects of these facilities, discusses the defence in depth approach and discusses the differences between nuclear reactors and nuclear fuel cycle facilities. Chapter 4 presents the basis for the INPRO safety basic principles. This starts from the IAEA Safety Fundamentals, identifies the specific requirements applicable to innovative nuclear fuel cycle facilities and derives the INPRO safety basic principles, user requirements and criteria that need to be applied in the assessment process. Chapter 5 discusses specific fuel cycle operations from the point of view of safety indicating possible innovations. This relates to INPRO safety criteria (with indicators and acceptance limits) as they apply to the following fuel cycle activities: mining and milling, enrichment, fuel fabrication, and fuel reprocessing. Chapter 6 provides a summary and also recommends future activities, which could be undertaken by IAEA, which would enable a robust assessment of the safety of nuclear fuel cycle facilities in the framework of INPRO. In Annex A, application of INPRO methodology to a hypothetical fuel fabrication facility using sol-gel process is given as an illustration. In Annex B a few ideas are laid out how to modify the INPRO methodology in the area of safety to increase consistency. In Annex C a table is presented that could be used by an assessor to summarize his assessment

Using existing European MOX fabrication plants for the disposal of plutonium from dismantled Russian warheads

International Nuclear Information System (INIS)

Schaper, A.

1995-01-01

One of the disposition options for excess weapons plutonium which is favored by the study of the American National Academy of Sciences is the fabrication and use as fuel, without reprocessing, in existing or modified nuclear reactors. An important criterion for reducing the proliferation risks is minimizing the time during which the plutonium is stored in forms readily usable for nuclear weapons. The study recommends to either modify an almost completed facility for experimental fast reactors or to construct a new fuel fabrication capability. The estimated time for siting, building, and licensing is a decade or more

Automation of technical specification monitoring for nuclear power plants

International Nuclear Information System (INIS)

Lin, J.C.; Abbott, E.C.; Hubbard, F.R.

1986-01-01

The complexity of today's nuclear power plants combined with an equally detailed regulatory process makes it necessary for the plant staff to have access to an automated system capable of monitoring the status of limiting conditions for operation (LCO). Pickard, Lowe and Garrick, Inc. (PLG), has developed the first of such a system, called Limiting Conditions for Operation Monitor (LIMCOM). LIMCOM provides members of the operating staff with an up-to-date comparison of currently operable equipment and plant operating conditions with what is required in the technical specifications. LIMCOM also provides an effective method of screening tagout requests by evaluating their impact on the LCOs. Finally, LIMCOM provides an accurate method of tracking and scheduling routine surveillance. (author)

A Specific Long-Term Plan for Management of U.S. Nuclear Spent Fuel

International Nuclear Information System (INIS)

Levy, Salomon

2006-01-01

A specific plan consisting of six different steps is proposed to accelerate and improve the long-term management of U.S. Light Water Reactor (LWR) spent nuclear fuel. The first step is to construct additional, centralized, engineered (dry cask) spent fuel facilities to have a backup solution to Yucca Mountain (YM) delays or lack of capacity. The second step is to restart the development of the Integral Fast Reactor (IFR), in a burner mode, because of its inherent safety characteristics and its extensive past development in contrast to Acceleration Driven Systems (ADS). The IFR and an improved non-proliferation version of its pyro-processing technology can burn the plutonium (Pu) and minor actinides (MA) obtained by reprocessing LWR spent fuel. The remaining IFR and LWR fission products will be treated for storage at YM. The radiotoxicity of that high level waste (HLW) will fall below that of natural uranium in less than one thousand years. Due to anticipated increased capital, maintenance, and research costs for IFR, the third step is to reduce the required number of IFRs and their potential delays by implementing multiple recycles of Pu and Neptunium (Np) MA in LWR. That strategy is to use an advanced separation process, UREX+, and the MIX Pu option where the role and degradation of Pu is limited by uranium enrichment. UREX+ will decrease proliferation risks by avoiding Pu separation while the MIX fuel will lead to an equilibrium fuel recycle mode in LWR which will reduce U. S. Pu inventory and deliver much smaller volumes of less radioactive HLW to YM. In both steps two and three, Research and Development (R and D) is to emphasize the demonstration of multiple fuel reprocessing and fabrication, while improving HLW treatment, increasing proliferation resistance, and reducing losses of fissile material. The fourth step is to license and construct YM because it is needed for the disposal of defense wastes and the HLW to be generated under the proposed plan. The

Characterization of electro-conductive fabrics prepared by in situ chemical and electrochemical polymerization of pyrrole onto polyester fabric

Energy Technology Data Exchange (ETDEWEB)

Maiti, Syamal; Das, Dipayan; Sen, Kushal, E-mail: kushal@textile.iitd.ernet.in

2014-09-15

Highlights: • Surface resistivity of the fabrics decreased rapidly with an increase in add-on. • Add-on and resistivity were not correlated below a resistivity value of about 200 Ω. • Higher add-on but lower surface roughness resulted in lower surface resistivity. • The voltage–current and voltage–temperature behaviours were found to be non-linear. • Electro-conductive fabric exhibited 98% electromagnetic shielding efficiency. - Abstract: This paper reports a study on electro-conductive fabrics prepared by a combined in situ chemical and electrochemical polymerization of pyrrole. Specific observations are made to establish the roles of add-on and surface roughness on the surface resistivity of the electro-conductive fabrics. The performance characteristics of the fabrics are reported in terms of electrical conductivity, voltage–current and voltage–temperature characteristics and electromagnetic interference (EMI) shielding capability. The surface resistivity of the fabric was found to be as low as 11.79 Ω. The voltage–current profile of the fabric is observed to be non-ohmic as well as the voltage–temperature curve is found to be exponential. The EMI shielding efficiency of the fabric was found to be about 98%.

Characterization of electro-conductive fabrics prepared by in situ chemical and electrochemical polymerization of pyrrole onto polyester fabric

International Nuclear Information System (INIS)

Maiti, Syamal; Das, Dipayan; Sen, Kushal

2014-01-01

Highlights: • Surface resistivity of the fabrics decreased rapidly with an increase in add-on. • Add-on and resistivity were not correlated below a resistivity value of about 200 Ω. • Higher add-on but lower surface roughness resulted in lower surface resistivity. • The voltage–current and voltage–temperature behaviours were found to be non-linear. • Electro-conductive fabric exhibited 98% electromagnetic shielding efficiency. - Abstract: This paper reports a study on electro-conductive fabrics prepared by a combined in situ chemical and electrochemical polymerization of pyrrole. Specific observations are made to establish the roles of add-on and surface roughness on the surface resistivity of the electro-conductive fabrics. The performance characteristics of the fabrics are reported in terms of electrical conductivity, voltage–current and voltage–temperature characteristics and electromagnetic interference (EMI) shielding capability. The surface resistivity of the fabric was found to be as low as 11.79 Ω. The voltage–current profile of the fabric is observed to be non-ohmic as well as the voltage–temperature curve is found to be exponential. The EMI shielding efficiency of the fabric was found to be about 98%

Comparative assessment of out-of-core nuclear thermionic power systems

International Nuclear Information System (INIS)

Estabrook, W.C.; Koenig, D.R.; Prickett, W.Z.

1975-01-01

The hardware selections available for fabrication of a nuclear electric propulsion stage for planetary exploration were explored. The investigation was centered around a heat-pipe-cooled, fast-spectrum nuclear reactor for an out-of-core power conversion system with sufficient detail for comparison with the in-core system studies completed previously. A survey of competing power conversion systems still indicated that the modular reliability of thermionic converters makes them the desirable choice to provide the 240-kWe end-of-life power for at least 20,000 full power hours. The electrical energy will be used to operate a number of mercury ion bombardment thrusters with a specific impulse in the range of about 4,000-5,000 seconds. (Author)

Project Plan Remote Target Fabrication Refurbishment Project

International Nuclear Information System (INIS)

Bell, Gary L.; Taylor, Robin D.

2009-01-01

In early FY2009, the DOE Office of Science - Nuclear Physics Program reinstated a program for continued production of 252 Cf and other transcurium isotopes at the Radiochemical Engineering Development Center (REDC) at Oak Ridge National Laboratory (ORNL). The FY2009 major elements of the workscope are as follows: (1) Recovery and processing of seven transuranium element targets undergoing irradiation at the High Flux Isotope Reactor (HFIR) at ORNL; (2) Development of a plan to manufacture new targets for irradiation beginning in early- to mid-FY10 to supply irradiated targets for processing Campaign 75 (TRU75); and (3) Refurbishment of the target manufacturing equipment to allow new target manufacture in early FY10 The 252 Cf product from processing Campaign 74 (recently processed and currently shipping to customers) is expected to supply the domestic demands for a period of approximately two years. Therefore it is essential that new targets be introduced for irradiation by the second quarter of FY10 (HFIR cycle 427) to maintain supply of 252 Cf; the average irradiation period is ∼10 HFIR cycles, requiring about 1.5 calendar years. The strategy for continued production of 252 Cf depends upon repairing and refurbishing the existing pellet and target fabrication equipment for one additional target production campaign. This equipment dates from the mid-1960s to the late 1980s, and during the last target fabrication campaign in 2005- 2006, a number of component failures and operations difficulties were encountered. It is expected that following the target fabrication and acceptance testing of the targets that will supply material for processing Campaign 75 a comprehensive upgrade and replacement of the remote hot-cell equipment will be required prior to subsequent campaigns. Such a major refit could start in early FY 2011 and would take about 2 years to complete. Scope and cost estimates for the repairs described herein were developed, and authorization for the work

Test Operation of Oxygen-Enriched Incinerator for Wastes From Nuclear Fuel Fabrication Facility

International Nuclear Information System (INIS)

Kim, J.-G.; Yang, H.cC.; Park, G.-I.; Kim, I.-T.; Kim, J.-K.

2002-01-01

The oxygen-enriched combustion concept, which can minimize off-gas production, has been applied to the incineration of combustible uranium-containing wastes from a nuclear fuel fabrication facility. A simulation for oxygen combustion shows the off-gas production can be reduced by a factor of 6.7 theoretically, compared with conventional air combustion. The laboratory-scale oxygen enriched incineration (OEI) process with a thermal capacity of 350 MJ/h is composed of an oxygen feeding and control system, a combustion chamber, a quencher, a ceramic filter, an induced draft fan, a condenser, a stack, an off-gas recycle path, and a measurement and control system. Test burning with cleaning paper and office paper in this OEI process shows that the thermal capacity is about 320 MJ/h, 90 % of design value and the off-gas reduces by a factor of 3.5, compared with air combustion. The CO concentration for oxygen combustion is lower than that of air combustion, while the O2 concentration in off-gas is kept above 25 vol % for a simple incineration process without any grate. The NOx concentration in an off-gas stream does not reduce significantly due to air incoming by leakage, and the volume and weight reduction factors are not changed significantly, which suggests a need for an improvement in sealing

Nuclear fuel activities in Belgium

Energy Technology Data Exchange (ETDEWEB)

Bairiot, H

1997-12-01

In his presentation on nuclear fuel activities in belgium the author considers the following directions of this work: fuel fabrication, NPP operation, fuel performance, research and development programmes.

India's nuclear program

International Nuclear Information System (INIS)

Anon.

1992-01-01

India made an early commitment to being as self-sufficient as possible in nuclear energy and has largely achieved that goal. The country operates eight nuclear reactors with a total capacity of 1,304 MWe, and it remains committed to an aggressive growth plan for its nuclear industry, with six reactors currently under construction, and as many as twelve more planned. India also operates several heavy water production facilities, fabrication facilities, reprocessing works, and uranium mines and mills. Due to India's decision not to sign the Treaty on the Non-Proliferation of Nuclear Weapons (NPT), the country has had to develop nearly all of its nuclear industry and infrastructure domestically. Overall, India's nuclear power program is self-contained and well integrated, with plans to expand to provide up to ten percent of the country's electrical generating capacity

Proceedings of a NEA/CSNI-UNIPEDE specialist meeting on improving technical specifications for nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

NONE

1987-07-01

This CSNI specialist meeting on improving technical specifications for nuclear power plants is sponsored by the OECD Nuclear Energy Agency jointly with UNIPEDE. Technical specifications for nuclear power plants are in a way a prescription which has a direct bearing on the success or failure of the particular installation, and on the success or failure of fission energy around the world. It is therefore highly important that these prescriptions are made as clear and as concise as possible and that it distinguishes requirements which are essential for public health and safety, from the many others which are less important accordingly. The conference was held in september 1987 in madrid (Spain); it is composed of about 40 papers grouped into 8 sessions: invited papers (6 papers), international survey results (1 paper), limiting conditions for operation (8 papers), maintenance and testing (4 papers), actions statements and allowed outage times (8 papers), methodology and technical justification (8 papers), future trends and alternative approaches (4 papers), and a final panel

Maintenance and fabrication of nuclear electronic equipment

International Nuclear Information System (INIS)

Hong, Seok Boong; Chung, Chong Eun; Hwang, In Koo; Koo, In Soo; Park, Bum; Kim, Soo Hee; Lee, Seong Joo; Kim, Min Seok; Choi, Wha Lim

2011-12-01

- process equipment at PIEF, Chemical Analysis Team and RWFTF have been calibrated. - The electronic equipment and radiation equipment at RWTF and PIEF have been prepared. - Development and installation of integrated RMS software for Hanaro Cold Neutron Laboratory Building(CNLB) RMS, and development and performance upgrade of a portal monitor for CNLB. - Performance test of the Hardware/Software of digital unit controller has been performed, and functional upgrade of the Hardware/Software of stimulator for SMART MMIS performance test facility has also been performed. - A controller of high voltage power supply for a small electron beam generator and a controller for razer pinning and a remote monitoring apparatus of cathode power supply for a 0.2 Mev. small electron beam generator have been designed and fabricated. - Database construction for effective maintenance for the process equipment and radiation instruments are designed and constructed

Drip bloodstain appearance on inclined apparel fabrics: Effect of prior-laundering, fibre content and fabric structure.

Science.gov (United States)

de Castro, Therese C; Carr, Debra J; Taylor, Michael C; Kieser, Jules A; Duncan, Warwick

2016-09-01

and fabric structure for both impact angles investigated. It is therefore necessary to consider the age of the fabric (which is fabric specific), the fibre type (including blends) and the fabric structure, before interpreting bloodstain patterns. An understanding of this simplified inclined drip stain interaction has been investigated to generate a basis for more complex interactions, such as spatter bloodstains. Copyright © 2016 Elsevier Ireland Ltd. All rights reserved.

Specific nuclear localizing sequence directs two myosin isoforms to the cell nucleus in calmodulin-sensitive manner.

Science.gov (United States)

Dzijak, Rastislav; Yildirim, Sukriye; Kahle, Michal; Novák, Petr; Hnilicová, Jarmila; Venit, Tomáš; Hozák, Pavel

2012-01-01

Nuclear myosin I (NM1) was the first molecular motor identified in the cell nucleus. Together with nuclear actin, they participate in crucial nuclear events such as transcription, chromatin movements, and chromatin remodeling. NM1 is an isoform of myosin 1c (Myo1c) that was identified earlier and is known to act in the cytoplasm. NM1 differs from the "cytoplasmic" myosin 1c only by additional 16 amino acids at the N-terminus of the molecule. This amino acid stretch was therefore suggested to direct NM1 into the nucleus. We investigated the mechanism of nuclear import of NM1 in detail. Using over-expressed GFP chimeras encoding for truncated NM1 mutants, we identified a specific sequence that is necessary for its import to the nucleus. This novel nuclear localization sequence is placed within calmodulin-binding motif of NM1, thus it is present also in the Myo1c. We confirmed the presence of both isoforms in the nucleus by transfection of tagged NM1 and Myo1c constructs into cultured cells, and also by showing the presence of the endogenous Myo1c in purified nuclei of cells derived from knock-out mice lacking NM1. Using pull-down and co-immunoprecipitation assays we identified importin beta, importin 5 and importin 7 as nuclear transport receptors that bind NM1. Since the NLS sequence of NM1 lies within the region that also binds calmodulin we tested the influence of calmodulin on the localization of NM1. The presence of elevated levels of calmodulin interfered with nuclear localization of tagged NM1. We have shown that the novel specific NLS brings to the cell nucleus not only the "nuclear" isoform of myosin I (NM1 protein) but also its "cytoplasmic" isoform (Myo1c protein). This opens a new field for exploring functions of this molecular motor in nuclear processes, and for exploring the signals between cytoplasm and the nucleus.

Reprocessing of nuclear fuels at the Savannah River Plant

International Nuclear Information System (INIS)

Gray, L.W.

1986-01-01

For more than 30 years, the Savannah River Plant (SRP) has been a major supplier of nuclear materials such as plutonium-239 and tritium-3 for nuclear and thermonuclear weapons, plutonium-238 for space exploration, and isotopes of americium, curium, and californium for use in the nuclear research community. SRP is a complete nuclear park, providing most of the processes in the nuclear fuel cycle. Key processes involve fabrication and cladding of the nuclear fuel, target, and control assemblies; rework of heavy water for use as reactor moderator; reactor loading, operation, and unloading; chemical recovery of the reactor transmutation products and spent fuels; and management of the gaseous, liquid, and solid nuclear and chemical wastes; plus a host of support operations. The site's history and the key processes from fabrication of reactor fuels and targets to finishing of virgin plutonium for use in the nuclear weapons complex are reviewed. Emphasis has been given to the chemistry of the recovery and purification of weapons grade plutonium from irradiated reactor targets

Sustainable and safe nuclear fission energy technology and safety of fast and thermal nuclear reactors

CERN Document Server

Kessler, Günter

2012-01-01

Unlike existing books of nuclear reactor physics, nuclear engineering and nuclear chemical engineering this book covers a complete description and evaluation of nuclear fission power generation. It covers the whole nuclear fuel cycle, from the extraction of natural uranium from ore mines, uranium conversion and enrichment up to the fabrication of fuel elements for the cores of various types of fission reactors. This is followed by the description of the different fuel cycle options and the final storage in nuclear waste repositories. In addition the release of radioactivity under normal and possible accidental conditions is given for all parts of the nuclear fuel cycle and especially for the different fission reactor types.

AM600: A new look at the nuclear steam cycle

International Nuclear Information System (INIS)

Field, Robert M.

2017-01-01

Many developing countries considering the introduction of nuclear power find that large-scale reactor plants in the range of 1,000 MWe to 1,600 MWe are not grid appropriate for their current circumstance. By contrast, small modular reactors are generally too small to make significant contributions toward rapidly growing electricity demand and to date have not been demonstrated. This paper proposes a radically simplified re-design for the nuclear steam cycle for a medium-sized reactor plant in the range of 600 MWe. Historically, balance of plant designs for units of this size have emphasized reliability and efficiency. It will be demonstrated here that advances over the past 50 years in component design, materials, and fabrication techniques allow both of these goals to be met with a less complex design. A disciplined approach to reduce component count will result in substantial benefits in the life cycle cost of the units. Specifically, fabrication, transportation, construction, operations, and maintenance costs and expenses can all see significant reductions. In addition, the design described here can also be expected to significantly reduce both construction duration and operational requirements for maintenance and inspections

AM600: A new look at the nuclear steam cycle

Energy Technology Data Exchange (ETDEWEB)

Field, Robert M. [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2017-04-15

Many developing countries considering the introduction of nuclear power find that large-scale reactor plants in the range of 1,000 MWe to 1,600 MWe are not grid appropriate for their current circumstance. By contrast, small modular reactors are generally too small to make significant contributions toward rapidly growing electricity demand and to date have not been demonstrated. This paper proposes a radically simplified re-design for the nuclear steam cycle for a medium-sized reactor plant in the range of 600 MWe. Historically, balance of plant designs for units of this size have emphasized reliability and efficiency. It will be demonstrated here that advances over the past 50 years in component design, materials, and fabrication techniques allow both of these goals to be met with a less complex design. A disciplined approach to reduce component count will result in substantial benefits in the life cycle cost of the units. Specifically, fabrication, transportation, construction, operations, and maintenance costs and expenses can all see significant reductions. In addition, the design described here can also be expected to significantly reduce both construction duration and operational requirements for maintenance and inspections.

AM600: A New Look at the Nuclear Steam Cycle

Directory of Open Access Journals (Sweden)

Robert M. Field

2017-04-01

Full Text Available Many developing countries considering the introduction of nuclear power find that large-scale reactor plants in the range of 1,000 MWe to 1,600 MWe are not grid appropriate for their current circumstance. By contrast, small modular reactors are generally too small to make significant contributions toward rapidly growing electricity demand and to date have not been demonstrated. This paper proposes a radically simplified re-design for the nuclear steam cycle for a medium-sized reactor plant in the range of 600 MWe. Historically, balance of plant designs for units of this size have emphasized reliability and efficiency. It will be demonstrated here that advances over the past 50 years in component design, materials, and fabrication techniques allow both of these goals to be met with a less complex design. A disciplined approach to reduce component count will result in substantial benefits in the life cycle cost of the units. Specifically, fabrication, transportation, construction, operations, and maintenance costs and expenses can all see significant reductions. In addition, the design described here can also be expected to significantly reduce both construction duration and operational requirements for maintenance and inspections.

Development of DIPRES feed for the fabrication of mixed-oxide fuels for fast breeder reactors

International Nuclear Information System (INIS)

Griffin, C.W.; Rasmussen, D.E.; Lloyd, M.H.

1983-01-01

The DIrect PREss Spheroidized feed process combines the conversion of uranium-plutonium solutions into spheres by internal gelation with conventional pellet fabrication techniques. In this manner, gel spheres could replace conventional powders as the feed material for pellet fabrication of nuclear fuels. Objective of the DIPRES feed program is to develop and qualify a process to produce mixed-oxide fuel pellets from gel spheres for fast breeder reactors. This process development includes both conversion and fabrication activities

Advanced accounting techniques in automated fuel fabrication facilities

International Nuclear Information System (INIS)

Carlson, R.L.; DeMerschman, A.W.; Engel, D.W.

1977-01-01

The accountability system being designed for automated fuel fabrication facilities will provide real-time information on all Special Nuclear Material (SNM) located in the facility. It will utilize a distributed network of microprocessors and minicomputers to monitor material movement and obtain nuclear materials measurements directly from remote, in-line Nondestructive Assay instrumentation. As SNM crosses an accounting boundary, the accountability computer will update the master files and generate audit trail records. Mass balance accounting techniques will be used around each unit process step, while item control will be used to account for encapsulated material, and SNM in transit

Risk-based analysis methods applied to nuclear power plant technical specifications

International Nuclear Information System (INIS)

Wagner, D.P.; Minton, L.A.; Gaertner, J.P.

1989-01-01

A computer-aided methodology and practical applications of risk-based evaluation of technical specifications are described. The methodology, developed for use by the utility industry, is a part of the overall process of improving nuclear power plant technical specifications. The SOCRATES computer program uses the results of a probabilistic risk assessment or a system-level risk analysis to calculate changes in risk due to changes in the surveillance test interval and/or the allowed outage time stated in the technical specification. The computer program can accommodate various testing strategies (such as staggered or simultaneous testing) to allow modeling of component testing as it is carried out at the plant. The methods and computer program are an integral part of a larger decision process aimed at determining benefits from technical specification changes. These benefits can include cost savings to the utilities by reducing forced shutdowns and decreasing labor requirements for test and maintenance activities, with no adverse impacts on risk. The methodology and the SOCRATES computer program have been used extensively toe valuate several actual technical specifications in case studies demonstrating the methods. Summaries of these applications demonstrate the types of results achieved and the usefulness of the risk-based evaluation in improving the technical specifications

Intracellular lysyl oxidase: Effect of a specific inhibitor on nuclear mass in proliferating cells

Energy Technology Data Exchange (ETDEWEB)

Saad, Fawzy A. [Laboratory for the Study of Skeletal Disorders and Rehabilitation, Department of Orthopedics, Children' s Hospital Boston, 300 Longwood Avenue EN926, Boston, MA 02115 (United States); Harvard Medical School, Boston, MA 02115 (United States); Torres, Marie [Laboratory for the Study of Skeletal Disorders and Rehabilitation, Department of Orthopedics, Children' s Hospital Boston, 300 Longwood Avenue EN926, Boston, MA 02115 (United States); Wang, Hao [Laboratory for the Study of Skeletal Disorders and Rehabilitation, Department of Orthopedics, Children' s Hospital Boston, 300 Longwood Avenue EN926, Boston, MA 02115 (United States); Harvard Medical School, Boston, MA 02115 (United States); Graham, Lila, E-mail: lilagraham@cs.com [Laboratory for the Study of Skeletal Disorders and Rehabilitation, Department of Orthopedics, Children' s Hospital Boston, 300 Longwood Avenue EN926, Boston, MA 02115 (United States); Harvard Medical School, Boston, MA 02115 (United States)

2010-06-11

LOX, the principal enzyme involved in crosslinking of collagen, was the first of several lysyl oxidase isotypes to be characterized. Its active form was believed to be exclusively extracellular. Active LOX was later reported to be present in cell nuclei; its function there is unknown. LOX expression opposes the effect of mutationally activated Ras, which is present in about 30% of human cancers. The mechanism of LOX in countering the action of Ras is also unknown. In the present work, assessment of nuclear protein for possible effects of lysyl oxidase activity led to the discovery that proliferating cells dramatically increase their nuclear protein content when exposed to BAPN ({beta}-aminopropionitrile), a highly specific lysyl oxidase inhibitor that reportedly blocks LOX inhibition of Ras-induced oocyte maturation. In three cell types (PC12 cells, A7r5 smooth muscle cells, and NIH 3T3 fibroblasts), BAPN caused a 1.8-, 1.7-, and 2.1-fold increase in total nuclear protein per cell, respectively, affecting all major components in both nuclear matrix and chromatin fractions. Since nuclear size is correlated with proliferative status, enzyme activity restricting nuclear growth may be involved in the lysyl oxidase tumor suppressive effect. Evidence is also presented for the presence of apparent lysyl oxidase isotype(s) containing a highly conserved LOX active site sequence in the nuclei of PC12 cells, which do not manufacture extracellular lysyl oxidase substrates. Results reported here support the hypothesis that nuclear lysyl oxidase regulates nuclear growth, and thereby modulates cell proliferation.

Vibratory-compacted (vipac/sphere-pac) nuclear fuels - a comparison with pelletized nuclear fuels

Energy Technology Data Exchange (ETDEWEB)

Chidester, K.; Rubin, J. [Los Alamos National Lab., NM (United States); Thompson, M

2001-07-01

In order to achieve the packing densities required for nuclear fuel stability, economy and performance, the fuel material must be densified. This has traditionally been performed by high-temperature sintering. (At one time, fuel densification was investigated using cold/hot swaging. However, this fabrication method has become uncommon.) Alternatively, fuel can be densified by vibratory compaction (VIPAC). During the late 1950's and into the 1970's, in the U.S., vibratory compaction fuel was fabricated and test irradiated to evaluate its applicability compared to the more traditional pelletized fuel for nuclear reactors. These activities were primarily focused on light water reactors (LWR) but some work was performed for fast reactors. This paper attempts to summarize these evaluations and proposes to reconsider VIPAC fuel for future use. (author)

Vibratory-compacted (vipac/sphere-pac) nuclear fuels - a comparison with pelletized nuclear fuels

International Nuclear Information System (INIS)

Chidester, K.; Rubin, J.; Thompson, M.

2001-01-01

In order to achieve the packing densities required for nuclear fuel stability, economy and performance, the fuel material must be densified. This has traditionally been performed by high-temperature sintering. (At one time, fuel densification was investigated using cold/hot swaging. However, this fabrication method has become uncommon.) Alternatively, fuel can be densified by vibratory compaction (VIPAC). During the late 1950's and into the 1970's, in the U.S., vibratory compaction fuel was fabricated and test irradiated to evaluate its applicability compared to the more traditional pelletized fuel for nuclear reactors. These activities were primarily focused on light water reactors (LWR) but some work was performed for fast reactors. This paper attempts to summarize these evaluations and proposes to reconsider VIPAC fuel for future use. (author)

Material development for India's nuclear power programme

Indian Academy of Sciences (India)

rials with emphasis on development of fabrication routes of zirconium alloys for .... nuclear power programme, which envisages design and construction of thermal breeder ... Production of Hf-free nuclear grade zirconium ..... Later on for pressure tubes specified limit for hydrogen content in the as manufactured condition.

Risk based optimization of technical specifications for operation of nuclear power plants

International Nuclear Information System (INIS)

1993-12-01

The objective of the report is to present an overview of the risk and reliability based approaches (using a probabilistic safety assessment (PSA)) for improving nuclear power plant technical specifications (TS). In that case, it will provide an information base to the Member States in seeking PSA based applications to enhance the effectiveness of their technical specifications. To achieve this objective, the report discusses the basic objectives and reasons for seeking TS changes, the methods, data requirements and uses of different types of applications, and an overview of different applications that have been completed, including detailed descriptions of selected applications. Refs, figs and tabs

Re-qualification of MTR-type fuel plates fabrication process

International Nuclear Information System (INIS)

Elseaidy, I.M.; Ghoneim, M.M.

2010-01-01

The fabricability issues with increased uranium loading due to use low enrichment of uranium (LEU), i.e. less than 20 % of U 235 , increase the problems which occur during compact manufacturing, roll bonding of the fuel plates, potential difficulty in forming during rolling process, mechanical integrity of the core during fabrication, potential difficulty in meat homogeneity, and the ability to fabricate plates with thicker core as a means of increasing total uranium loading. To produce MTR- type fuel plates with high uranium loading (HUL) and keep the required quality of these plates, many of qualification process must be done in the commissioning step of fuel fabrication plant. After that any changing of the fabrication parameters, for example changing of any of the raw materials, devises, operators, and etc., a re- qualification process should be done in order to keep the quality of produced plates. Objective of the present work is the general description of the activities to be accomplished for re-qualification of manufacturing MTR- type nuclear fuel plates. For each process to be re-qualified, a detailed of re-qualification process were established. (author)

Fabrication of recyclable superhydrophobic cotton fabrics

Science.gov (United States)

Han, Sang Wook; Park, Eun Ji; Jeong, Myung-Geun; Kim, Il Hee; Seo, Hyun Ook; Kim, Ju Hwan; Kim, Kwang-Dae; Kim, Young Dok

2017-04-01

Commercial cotton fabric was coated with SiO2 nanoparticles wrapped with a polydimethylsiloxane (PDMS) layer, and the resulting material surface showed a water contact angle greater than 160°. The superhydrophobic fabric showed resistance to water-soluble contaminants and maintained its original superhydrophobic properties with almost no alteration even after many times of absorption-washing cycles of oil. Moreover, superhydrophobic fabric can be used as a filter to separate oil from water. We demonstrated a simple method of fabrication of superhydrophobic fabric with potential interest for use in a variety of applications.

Fabrication and application of nanostructured materials for sulfite biosensing

OpenAIRE

Hussain, Shahid

2017-01-01

A biosensor as an integrated miniaturized device, exploits the modern microelectronics with specific sensing probe through signal transduction. The challenge for new generation biosensors is to achieve specific analyte detection at very low concentrations, which is possible by tailoring the materials used for fabrication of these devices based on nanoscience and nanotechnology. The new approach is explored in this thesis for fabrication of novel nanobiosensors for ultrasensitive detection of ...

Radiation exposure of employees in nuclear fuel facilities in fiscal 1982

International Nuclear Information System (INIS)

1984-01-01

The enterprises of nuclear fuel refining, fabrication, reprocessing and usage are obligated by law to keep the radiation exposure dose of the employees below the permissible level. The radiation exposure dose in the respective enterprises in the fiscal year 1982 is summarized in a table as follows: radiation exposure dose distribution, the number of employees, total exposure dose, and average dose. The radiation exposure dose was all well below the permissible level. The enterprises covered were one refining (Power Reactor and Nuclear Fuel Development Corporation), five fabrication (Mitsubishi Nuclear Fuel Co., Ltd., etc.), one reprocessing (Power Reactor and Nuclear Fuel Development Corporation), and ten usage (Power Reactor and Nuclear Fuel Development Corporation, Japan Atomic Energy Research Institute, etc.). (Mori, K.)

Rules specific to nuclear incidence occurring in installations or during transport of nuclear substances

International Nuclear Information System (INIS)

Rocamora, P.

1976-01-01

International nuclear third party liability conventions deal in depth with the liability system governing the transport of nuclear substances. Without appropriate legislation, international transport would be likely to meet very serious legal difficulties. The rule of nuclear conventions apply the same system to transport as to nuclear installations and mainly enable a determination of the operator liable. They also allow the person responsible for transport to assume liability therefor in place of the operator who whould normally have been liable. These nuclear conventions do not affect application of international transport conventions and this provision has been the cause of serious difficulties regarding maritime transport. This resulted in the adoption in 1971 in Brussels of a convention relating to civil liability in the field of maritime carriage of nuclear material. The purpose of this convention is to establish in the field of maritime transport, the priority of the system of absolute, exclusive and limited liability in the nuclear conventions. (NEA) [fr

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.

Evaluation of filter fabrics for use in silt fences.

Science.gov (United States)

1980-01-01

The study reported was initiated to develop tests simulating field conditions that could be used to develop information for the formulation of specifications for use in purchasing filter fabrics to be used to construct silt fences. Fifteen fabrics we...

Water reactor fuel element fabrication, with special emphasis on its effects on fuel performance

International Nuclear Information System (INIS)

1979-01-01

Full text: The performance of nuclear fuel has improved over the years and is now a minor cause of outages and of power limitations in nuclear power plants. On the other hand, an increasing number of countries are in the process of developing or implementing their own capability for manufacturing fuel elements. In this context, the IAEA International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) advised that a symposium be organized devoted to the relationship between fuel fabrication and performance The Czechoslovak Atomic Energy Commission agreed to co-operate in the organization of this symposium and to host it in Prague. Those factors which influence fuel fabrication requirements are now well ascertained: as little reactor primary circuit contamination as possible, the tendency to increased burnups, reactor manoeuverability to match power grid demands, the desirability of an autonomous fabrication capability. It is the general experience of fuel element suppliers that fuel quality and performance has increased over the years, the importance of quality assurance and process monitoring has been decisive in this respect The ever increasing mass-production aspect of nuclear fuel leads to some processing steps being revised and alternatives being developed. The relation between fabrication processes and fuel performance characteristics, although generally well perceived, are still the subject of a large amount of experiment and assessment in most countries, both industrial and developing This evidence is most encouraging; it means indeed that nuclear power, which is already amongst the cheapest and safest sources of energy, will continue to be improved. The performance of Zircaloy fuel cladding - presently the material used in most water reactors - is under particular consideration. Better understanding of this quite recent alloy will pave the way for broader fuel utilization limits in the future. The panel discussion, which noted some

Nuclear Fuel Complex - a landmark of indigenous nuclear technology

Energy Technology Data Exchange (ETDEWEB)

Katiyar, H C [Nuclear Fuel Complex, Hyderabad (India)

1977-01-01

The Nuclear Fuel Complex (NFC) set up in India for manufacturing fuel and related hardware has proved to be a significant step towards self-sufficiency and saving of foreign exchange. The complex is involved in the entire operations starting from processing of raw material concentrates to finishing of fuel assemblies and other zircaloy reactor components. The complex consists of the following units : (1) Zirconium Oxide Plant, (2) Zirconium Sponge Plant, (3) Zircaloy Fabrication Plant, (4) Uranium Oxide Plant, (5) Ceramic Fabrication Plant, (6) Enriched Uranium Oxide Plant, (7) Enriched Fuel Fabrication Plant, (8) Special Materials Plant and (9) Titanium Plant. A brief description of the activities of the various units of the complex are given. The effluent management scheme is outlined. The requirements and cost of fuel and zircaloy components for the power stations at Tarapur, Kota and Kalpakkam are mentioned.

Process and device for fabricating nuclear fuel assembly grids

International Nuclear Information System (INIS)

Thiebaut, B.; Duthoo, D.; Germanaz, J.J.; Angilbert, B.

1991-01-01

The method for fabricating PWR fuel assembly grids consists to place the grid of which the constituent parts are held firmly in place within a frame into a sealed chamber full of inert gas. This chamber can rotate about an axis. The welding on one face at a time is carried out with a laser beam orthogonal to the axis orientation of the device. The laser source is outside of the chamber and the beam penetrates via a transparent view port

Fabrication of nanoporous nuclear track membranes

International Nuclear Information System (INIS)

Peng Liangqiang; Wang Shicheng; Ju Xin; Masaru Yoshida; Yasunari Maekawa

2001-01-01

Polyethylene terephthalate (PET) and polycarbonate (PC) films were irradiated by S, Kr and Xe ions and were illuminated with ultraviolet light. The normalized track etch rate for PET and PC films etched in different conditions were measured by conductometric experiments. It is shown that normalized track etch rate can be over 1000 for PET films, 2000 for PC films under optimized condition. TEM photographs of copper nanowires electroplated into nanoporous nuclear track membranes show that the narrowest wire diameter of copper nanowires is 20 nm and that the pore diameter calculated by conductometric experiments is in agreement with the wire diameter measured by TEM when the pore diameter is over 30 nm

Order for execution of the law concerning regulation of nuclear source materials, nuclear fuel materials and reactors

International Nuclear Information System (INIS)

1977-01-01

The designations according to Item 1, Article 3 and Item 1, Article 13 of the Law must be obtained for each factory or business place where refining and fabrication of nuclear material are to be performed. One who wants to obtain such designation should file an application attached with a business plan and other documents via the director of a regional bureau of international trade and industry having jurisdiction over such factory or business place. When nuclear material refiners and nuclear material fabricators wish to obtain the approval for change stipulated in Item 1, Article 6 and Item 1, Article 16 of the Law, they must file applications to the Prime Minister and the Minister of International Trade and Industry via said directors. Chief handlers of nuclear fuel materials shall be approved among those meeting the strict requirements. One who wishes to install reactors must obtain the approval for each factory or business place where the reactors are to be installed. The permission must be obtained for each nuclear ship entering Japanese waters. The reactors proper and several facilities are subject to periodic inspection. (Rikitake, Y.)

Los Alamos National Laboratory summary plan to fabricate mixed oxide lead assemblies for the fissile material disposition program

International Nuclear Information System (INIS)

Buksa, J.J.; Eaton, S.L.; Trellue, H.R.; Chidester, K.; Bowidowicz, M.; Morley, R.A.; Barr, M.

1997-12-01

This report summarizes an approach for using existing Los Alamos National Laboratory (Laboratory) mixed oxide (MOX) fuel-fabrication and plutonium processing capabilities to expedite and assure progress in the MOX/Reactor Plutonium Disposition Program. Lead Assembly MOX fabrication is required to provide prototypic fuel for testing in support of fuel qualification and licensing requirements. It is also required to provide a bridge for the full utilization of the European fabrication experience. In part, this bridge helps establish, for the first time since the early 1980s, a US experience base for meeting the safety, licensing, safeguards, security, and materials control and accountability requirements of the Department of Energy and Nuclear Regulatory Commission. In addition, a link is needed between the current research and development program and the production of disposition mission fuel. This link would also help provide a knowledge base for US regulators. Early MOX fabrication and irradiation testing in commercial nuclear reactors would provide a positive demonstration to Russia (and to potential vendors, designers, fabricators, and utilities) that the US has serious intent to proceed with plutonium disposition. This report summarizes an approach to fabricating lead assembly MOX fuel using the existing MOX fuel-fabrication infrastructure at the Laboratory

Los Alamos National Laboratory summary plan to fabricate mixed oxide lead assemblies for the fissile material disposition program

Energy Technology Data Exchange (ETDEWEB)

Buksa, J.J.; Eaton, S.L.; Trellue, H.R.; Chidester, K.; Bowidowicz, M.; Morley, R.A.; Barr, M.

1997-12-01

This report summarizes an approach for using existing Los Alamos National Laboratory (Laboratory) mixed oxide (MOX) fuel-fabrication and plutonium processing capabilities to expedite and assure progress in the MOX/Reactor Plutonium Disposition Program. Lead Assembly MOX fabrication is required to provide prototypic fuel for testing in support of fuel qualification and licensing requirements. It is also required to provide a bridge for the full utilization of the European fabrication experience. In part, this bridge helps establish, for the first time since the early 1980s, a US experience base for meeting the safety, licensing, safeguards, security, and materials control and accountability requirements of the Department of Energy and Nuclear Regulatory Commission. In addition, a link is needed between the current research and development program and the production of disposition mission fuel. This link would also help provide a knowledge base for US regulators. Early MOX fabrication and irradiation testing in commercial nuclear reactors would provide a positive demonstration to Russia (and to potential vendors, designers, fabricators, and utilities) that the US has serious intent to proceed with plutonium disposition. This report summarizes an approach to fabricating lead assembly MOX fuel using the existing MOX fuel-fabrication infrastructure at the Laboratory.

The role of a multinational nuclear fuel fabrication supplier

International Nuclear Information System (INIS)

Beard, S.J.

1987-01-01

The author argues that international markets and multinational suppliers provide large benefits to utilities. It represents a long term commitment to the nuclear business that these companies will be able to supply nuclear technology on the long haul. The technology that is available around the world becomes available to everyone through the international markets and multinational suppliers. The increased experience base is seen as valuable in that errors that have been made or have not been made yet can be avoided through the transfer or experience. The security of supply is discussed as important to any utility that is operating a reactor

Nuclear for curious

International Nuclear Information System (INIS)

Fontesse, Max

2015-01-01

The author aims at proposing a popularisation of nuclear science. In a first part which addresses general aspects and uses of radioactivity, he presents and describes the matter structure, proposes a historical overview, introduces the various forms of nuclear activity (α, β, and γ radioactivity) and natural radioactivity. He explains why radiation can be called ionizing radiation (action of ionizing radiations on matter, comparison between α, β, X, and γ radiations). He presents the various radiological units, addresses issues related to radiation protection (main damages on human body, impact of a radiation on the whole body), and describes various uses of radioactivity (nuclear medicine, dating, heritage conservation, and others). In the second part, the author addresses uses of nuclear energy. He describes artificial radioactivity (nuclear fission, criticality), the various types of nuclear reactors (natural uranium graphite gas, heavy water, pressurised water, boiling water, fast breeders), and the nuclear fuel cycle (mining extraction, refinery, isotopic enrichment, assembly fabrication, cooling in pool, reprocessing, nuclear wastes, MOX fuel). The third part addresses perspectives: EPR, future reactors, nuclear fusion

Secure automated fabrication (SAF). Phase I interim report: a systems analysis

International Nuclear Information System (INIS)

1979-01-01

An advanced Secure Automated Fabrication (SAF) System is being developed for mixed uranium and plutonium fuel fabrication. SAF System development will ultimately result in systems which maximize personnel radiation protection, restrict and control access to SNM material, improve containment and detection systems for nuclear materials, provide adequate SNM accountability and improve product uniformity and quality. A systems requirement analysis study was initiated to establish the consistent and objective set of requirements within which the choice among alternatives represents the balanced viewpoints of performance, achievability and risk

Projected developments in the US and European LWR fabrication market

International Nuclear Information System (INIS)

Anderson, C.K.; Varley, G.

1996-01-01

There are several important factors influencing change in the fuel fabrication supply industry in Western Europe today. The most important is the changing supply relationships resulting from lower costs in the USA. With US fabrication prices substantially lower than in Western Europe, a situation which is sustainable based on production costs, a major force for change prevails which will have significant implications for the commercial development of both markets over the next decade. The extent to which supply from the USA will drive the Western Europe market will depend, to some extent, on three other factors: the new US-Euratom agreement on nuclear trade; electricity market deregulation in the European Union (EU); near term MOX fuel fabrication capacity. The main purpose of this paper is to discuss the current situation in the USA and Europe, and the manner in which this will lead to changing fuel fabrication supply dynamics in both markets. (author)