WorldWideScience

Sample records for fuel fabrication facilities

  1. 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)

  2. Facility safeguards at an LEU fuel fabrication facility in Japan

    International Nuclear Information System (INIS)

    Kuroi, H.; Osabe, T.

    1984-01-01

    A facility description of a Japanese LEU BWR-type fuel fabrication plant focusing on safeguards viewpoints is presented. Procedures and practices of MC and A plan, measurement program, inventory taking, and the report and record system are described. Procedures and practices of safeguards inspection are discussed and lessons learned from past experiences are reviewed

  3. TRIGA International, a new TRIGA fuel fabrication facility at CERCA

    International Nuclear Information System (INIS)

    Harbonnier, G.

    1997-01-01

    At the time when General Atomics expressed its intention to cease fuel fabrication on its site of San Diego, CERCA has been chosen to carry on the fabrication of TRIGA fuel. After negotiations in 1994 and 1995, a partnership 50%/50% was decided and on July 1995, a new company was founded, with the name TRIGA INTERNATIONAL SAS, head office in Paris and fuel fabrication facility at CERCA in Romans. The intent of this presentation is, after a short reminder about TRIGA fuel design and fabrication to describe the new facility with special emphasis on the safety features associated with the modification of existing fabrication buildings. (author)

  4. Advanced Safeguards Approaches for New TRU Fuel Fabrication Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Durst, Philip C.; Ehinger, Michael H.; Boyer, Brian; Therios, Ike; Bean, Robert; Dougan, A.; Tolk, K.

    2007-12-15

    This second report in a series of three reviews possible safeguards approaches for the new transuranic (TRU) fuel fabrication processes to be deployed at AFCF – specifically, the ceramic TRU (MOX) fuel fabrication line and the metallic (pyroprocessing) line. The most common TRU fuel has been fuel composed of mixed plutonium and uranium dioxide, referred to as “MOX”. However, under the Advanced Fuel Cycle projects custom-made fuels with higher contents of neptunium, americium, and curium may also be produced to evaluate if these “minor actinides” can be effectively burned and transmuted through irradiation in the ABR. A third and final report in this series will evaluate and review the advanced safeguards approach options for the ABR. In reviewing and developing the advanced safeguards approach for the new TRU fuel fabrication processes envisioned for AFCF, the existing international (IAEA) safeguards approach at the Plutonium Fuel Production Facility (PFPF) and the conceptual approach planned for the new J-MOX facility in Japan have been considered as a starting point of reference. The pyro-metallurgical reprocessing and fuel fabrication process at EBR-II near Idaho Falls also provided insight for safeguarding the additional metallic pyroprocessing fuel fabrication line planned for AFCF.

  5. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  7. Fast Flux Test Facility metal fuel pin fabrication

    International Nuclear Information System (INIS)

    Benecke, M.W.; Dittmer, J.O.; Feigenbutz, L.V.

    1989-05-01

    A new initiative to develop, irradiate, and qualify a binary uranium/zirconium metal fuel system in the Fast Flux Test Facility (FFTF) has been implemented by the Westinghouse Hanford Company for the US Department of Energy. Metal fuel test assemblies have been designed and fabricated and are now being irradiated in FFTF to provide the data needed to support the potential use of binary fuels in FFTF and other liquid metal reactors. These development efforts support licensing activities for metal fuel use in near-term advanced liquid metal reactors. New metal fuel pin design features, fabrication development, and fabrication processes for three metal fuel tests will be described and their irradiation status reported in this paper. 11 figs

  8. Greenfield Alternative Study LEU-Mo Fuel Fabrication Facility

    Energy Technology Data Exchange (ETDEWEB)

    Washington Division of URS

    2008-07-01

    This report provides the initial “first look” of the design of the Greenfield Alternative of the Fuel Fabrication Capability (FFC); a facility to be built at a Greenfield DOE National Laboratory site. The FFC is designed to fabricate LEU-Mo monolithic fuel for the 5 US High Performance Research Reactors (HPRRs). This report provides a pre-conceptual design of the site, facility, process and equipment systems of the FFC; along with a preliminary hazards evaluation, risk assessment as well as the ROM cost and schedule estimate.

  9. 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

  10. 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

  11. 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

  12. Facility effluent monitoring plan for the 300 Area Fuels Fabrication Facility

    International Nuclear Information System (INIS)

    Nickels, J.M.; Brendel, D.F.

    1991-11-01

    A facility effluent monitoring plan is required by the US Department of Energy in DOE Order 5400.1 for any operations that involve hazardous materials and radioactive substances that could impact employee or public safety or the environment. This document is prepared using the specific guidelines identified in A Guide for Preparing Hanford Site Facility Effluent Monitoring Plans, WHC-EP- 0438. This facility effluent monitoring plan assesses effluent monitoring systems and evaluates whether they are adequate to ensure the public health and safety as specified in applicable federal, state, and local requirements. This facility effluent monitoring plan is the first annual report. It shall ensure long-range integrity of the effluent monitoring system by requiring an update whenever a new process or operation introduces new hazardous materials or significant radioactive materials. This document must be reviewed annually even if there are no operational changes, and it must be updated as a minimum every three years. The Fuel Fabrication Facility in the Hanford 300 Area supported the production reactors from the 1940's until they were shut down in 1987. Prior to 1987 the Fuel Fabrication Facility released both airborne and liquid radioactive effluents. In January 1987 the emission of airborne radioactive effluents ceased with the shutdown of the fuels facility. The release of liquid radioactive effluents have continued although decreasing significantly from 1987 to 1990

  13. Application of the fuel-nondestructive analysis to plutonium fuel fabricating facilities

    International Nuclear Information System (INIS)

    Tsutsumi, Masayori

    1976-01-01

    The nondestructive measuring instruments for nuclear fuel employed in nuclear fuel facilities are shown in five pictures. Picture 1 illustrates a calorimeter provided for measuring the plutomium contained in fuel pellets as well as the plutonium powder existing in facilities. Picture 2 shows an active neutron device, to which the neutron interrogation method is applied. Cf-252 is used as a neutron source. A sample holder moves while rotating to the detector end. Four plastic scintillators are built in the detector end to detect neutron beam and gamma-ray. The pulses from two or three detector ends are counted. This instrument is suitable particularly for the determination of a minute amount of plutonium and uranium in scraps. Picture 3 shows an active neutron device usually employed in uranium fuel fabricating plants. Picture 4 indicates a fuel rod scanner, to which gamma spectrometry is applied. The fuel rods having the different degree of enrichment of plutonium can be identified, and plutonium in one fuel rod is determined. Picture 5 indicates a passive neutron device which can determine the total amount of plutonium contained in a fuel assembly. More accurate and more reliable nondestructive measuring instruments are necessary to automate nuclear fuel processing facilities. (Iwakiri, K.)

  14. Fabrication Facilities

    Data.gov (United States)

    Federal Laboratory Consortium — The Fabrication Facilities are a direct result of years of testing support. Through years of experience, the three fabrication facilities (Fort Hood, Fort Lewis, and...

  15. Detailed description of an SSAC at the facility level for mixed oxide fuel fabrication facilities

    International Nuclear Information System (INIS)

    Jones, R.J.

    1985-09-01

    The purpose of this document is to provide a detailed description of a system for the accounting for and control of nuclear material in a mixed oxide fuel fabrication facility which can be used by a facility operator to establish his own system to comply with a national system for nuclear material accounting and control and to facilitate application of IAEA safeguards. The scope of this document is limited to descriptions of the following SSAC elements: (1) Nuclear Material Measurements; (2) Measurement Quality; (3) Records and Reports; (4) Physical Inventory Taking; (5) Material Balance Closing

  16. Modern methods of material accounting for mixed oxide fuel fabrication facility

    International Nuclear Information System (INIS)

    Eggers, R.F.; Pindak, J.L.; Brouns, R.J.; Williams, R.C.; Brite, D.W.; Kinnison, R.R.; Fager, J.E.

    1981-01-01

    The generic requirements loss detection, and response to alarms of a contemporary material control and accounting (MCandA) philosophy have been applied to a mixed oxide fuel fabrication plant to produce a detailed preliminary MCandA system design that is generally applicable to facilities of this type. This paper summarizes and discusses detailed results of the mixed oxide fuel fabrication plant study

  17. Evaluation of existing United States' facilities for use as a mixed-oxide (MOX) fuel fabrication facility for plutonium disposition

    International Nuclear Information System (INIS)

    Beard, C.A.; Buksa, J.J.; Chidester, K.; Eaton, S.L.; Motley, F.E.; Siebe, D.A.

    1995-01-01

    A number of existing US facilities were evaluated for use as a mixed-oxide fuel fabrication facility for plutonium disposition. These facilities include the Fuels Material Examination Facility (FMEF) at Hanford, the Washington Power Supply Unit 1 (WNP-1) facility at Hanford, the Barnwell Nuclear Fuel Plant (BNFP) at Barnwell, SC, the Fuel Processing Facility (FPF) at Idaho National Engineering Laboratory (INEL), the Device Assembly Facility (DAF) at the Nevada Test Site (NTS), and the P-reactor at the Savannah River Site (SRS). The study consisted of evaluating each facility in terms of available process space, available building support systems (i.e., HVAC, security systems, existing process equipment, etc.), available regional infrastructure (i.e., emergency response teams, protective force teams, available transportation routes, etc.), and ability to integrate the MOX fabrication process into the facility in an operationally-sound manner that requires a minimum amount of structural modifications

  18. Development of an engineered safeguards system concept for a mixed-oxide fuel fabrication facility

    International Nuclear Information System (INIS)

    Chapman, L.D.; de Montmollin, J.M.; Deveney, J.E.; Fienning, W.C.; Hickman, J.W.; Watkins, L.D.; Winblad, A.E.

    1976-08-01

    An initial concept of an Engineered Safeguards System for a representative commercial mixed-oxide fuel fabrication facility is presented. Computer simulation techniques for evaluation and further development of the concept are described. An outline of future activity is included

  19. Advanced accountability techniques for breeder fuel fabrication facilities

    International Nuclear Information System (INIS)

    Bennion, S.I.; Carlson, R.L.; DeMerschman, A.W.; Sheely, W.F.

    1978-01-01

    The United States Department of Energy (DOE) has assigned the Hanford Engineering Development Laboratory (HEDL), operated by the Westinghouse Hanford Company, the project lead in developing a uniform nuclear materials reporting system for all contractors on the Hanford Reservation. The Hanford Nuclear Inventory System (HANISY) is based upon HEDL's real-time accountability system, originally developed in 1968. The HANISY system will receive accountability data either from entry by process operators at remote terminals or from nondestructive assay instruments connected to the computer network. Nuclear materials will be traced from entry, through processing to final shipment through the use of minicomputer technology. Reports to DOE will be formed directly from the realtime files. In addition, HEDL has established a measurement program that will complement the HANISY system, providing direct interface to the computer files with a minimum of operator intervention. This technology is being developed to support the High Performance Fuels Laboratory (HPFL) which is being designed to assess fuel fabrication techniques for proliferation-resistant fuels

  20. Operation of N Reactor and Fuels Fabrication Facilities, Hanford Reservation, Richland, Benton County, Washington: Environmental assessment

    Energy Technology Data Exchange (ETDEWEB)

    1980-08-01

    Environmental data, calculations and analyses show no significant adverse radiological or nonradiological impacts from current or projected future operations resulting from N Reactor, Fuels Fabrication and Spent Fuel Storage Facilities. Nonoccupational radiation exposures resulting from 1978 N Reactor operations are summarized and compared to allowable exposure limits.

  1. Development of ISA procedure for uranium fuel fabrication and enrichment facilities

    International Nuclear Information System (INIS)

    Yamate, Kazuki; Arakawa, Tomoyuki; Yamashita, Masahiro; Sasaki, Noriaki; Hirano, Mitsumasa

    2011-01-01

    The integrated safety analysis (ISA) procedure has been developed to apply risk-informed regulation to uranium fuel fabrication and enrichment facilities. The major development efforts are as follows: (a) preparing the risk level matrix as an index for items-relied-on-for-safety (IROFS) identification, (b) defining requirements of IROFS, and (c) determining methods of IROFS importance based on the results of risk- and scenario-based analyses. For the risk level matrix, the consequence and likelihood categories have been defined by taking into account the Japanese regulatory laws, rules, and safety standards. The trial analyses using the developed procedure have been performed for several representative processes of the reference uranium fuel fabrication and enrichment facilities. This paper presents the results of the ISA for the sintering process of the reference fabrication facility. The results of the trial analyses have demonstrated the applicability of the procedure to the risk-informed regulation of these facilities. (author)

  2. 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

  3. Simulated physical inventory verification exercise at a mixed-oxide fuel fabrication facility

    International Nuclear Information System (INIS)

    Reilly, D.; Augustson, R.

    1985-01-01

    A physical inventory verification (PIV) was simulated at a mixed-oxide fuel fabrication facility. Safeguards inspectors from the International Atomic Energy Agency (IAEA) conducted the PIV exercise to test inspection procedures under ''realistic but relaxed'' conditions. Nondestructive assay instrumentation was used to verify the plutonium content of samples covering the range of material types from input powders to final fuel assemblies. This paper describes the activities included in the exercise and discusses the results obtained. 5 refs., 1 fig., 6 tabs

  4. 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

  5. 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

  6. Decommissioning of a mixed oxide fuel fabrication facility

    International Nuclear Information System (INIS)

    Buck, S.; Colquhoun, A.

    1990-01-01

    Decommissioning of the coprecipitation plant, which made plutonium/uranium oxide fuel, is a lead project in the BNFL Sellafield decommissioning programme. The overall programme has the objectives of gaining data and experience in a wide range of decommissioning operations and hence in this specific project to pilot the decommissioning of plant heavily contaminated with plutonium and other actinides. Consequently the operations have been used to test improvements in temporary containment, contamination control and decontamination methods and also to develop in situ plutonium assay, plutonium recovery and size-reduction methods. Finally the project is also yielding data on manpower requirements, personnel radiation uptake and waste arisings to help in the planning of future decommissioning projects

  7. Development of probabilistic safety assessment method for mixed oxide fuel fabrication facilities

    International Nuclear Information System (INIS)

    Tamaki, Hitoshi; Yoshida, Kazuo; Watanabe, Norio; Muramatsu, Ken

    2006-01-01

    A Probabilistic Safety Assessment (PSA) procedure for Mixed Oxide (MOX) fuel fabrication facilities was developed. The procedure is a 'two-part five-step' approach which takes characteristics of MOX fuel fabrication facilities into consideration. In the first part, so-called preliminary PSA, the hazard analysis approach was applied, which consists of two analysis steps: Functional Failure Modes and Effects Analysis (FFMEA) and Risk Matrix Analysis. The FFMEA analyzes a variety of functions of equipment composing the facility to identify potential abnormal events exhaustively. In the Risk Matrix Analysis, these potential events are screened to select abnormal events as candidates to be analyzed in the second part, using two-dimensional matrix based on the likelihood evaluated by probabilistic index method and maximum unmitigated radioactive release calculated by the Five-Factor Formula. For the selected abnormal events, in the second part, so-called detailed PSA, accident sequences, their occurrence frequencies and consequences are analyzed. These three analysis steps correspond to PSA procedure for nuclear power plant. The applicability of the PSA procedure was demonstrated through the trial application to model plant of MOX fuel fabrication facility. (author)

  8. Evaluation of environmental control technologies for commercial uranium nuclear fuel fabrication facilities

    International Nuclear Information System (INIS)

    Perkins, B.L.

    1983-01-01

    At present in the United States, there are seven commercial light-water reactor uranium fuel fabrication facilities. Effluent wastes from these facilities include uranium, nitrogen, fluorine, and organic-containing compounds. These effluents may be either discharged to the ambient environment, treated and recycled internally, stored or disposed of on-site, sent off-site for treatment and/or recovery, or sent off-site for disposal (including disposal in low-level waste burial sites). Quantities of waste generated and treatment techniques vary greatly depending on the facility and circuits used internally at the facility, though in general all the fluorine entering the facility as UF 6 is discharged as waste. Further studies to determine techniques and procedures that might minimize dose (ALARA) and to give data on possible long-term effects of effluent discharge and waste disposal are needed

  9. Fire criticality probability analysis for 300 Area N Reactor fuel fabrication and storage facility. Revision 1

    International Nuclear Information System (INIS)

    Kelly, J.E.

    1995-01-01

    Uranium fuel assemblies and other uranium associated with the shutdown N Reactor are stored in the 300 Area N Reactor Fuel Fabrication and Storage Facility (Facility). The 3712 Building, where the majority of the fuel assemblies and other uranium is stored, is where there could be a potential for a criticality bounding case. The purpose of this study is to evaluate the probability of potential fires in the Facility 3712 Building that could lead to criticality. This study has been done to support the criticality update. For criticality to occur, the wooden fuel assembly containers would have to burn such that the fuel inside would slump into a critical geometry configuration, a sufficient number of containers burn to form an infinite wide configuration, and sufficient water (about a 17 inch depth) be placed onto the slump. To obtain the appropriate geometric configuration, enough fuel assembly containers to form an infinite array on the floor would have to be stacked at least three high. Administrative controls require the stacks to be limited to two high for 1.25 wt% enriched finished fuel. This is not sufficient to allow for a critical mass regardless of the fire and accompanying water moderation. It should be noted that 0.95 wt% enriched fuel and billets are stacked higher than only two high. In this analysis, two initiating events will be considered. The first is a random fire that is hot enough and sufficiently long enough to burn away the containers and fuel separators such that the fuel can establish a critical mass. The second is a seismically induced fire with the same results

  10. Fire criticality probability analysis for 300 Area N Reactor fuel fabrication and storage facility. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, J.E.

    1995-02-08

    Uranium fuel assemblies and other uranium associated with the shutdown N Reactor are stored in the 300 Area N Reactor Fuel Fabrication and Storage Facility (Facility). The 3712 Building, where the majority of the fuel assemblies and other uranium is stored, is where there could be a potential for a criticality bounding case. The purpose of this study is to evaluate the probability of potential fires in the Facility 3712 Building that could lead to criticality. This study has been done to support the criticality update. For criticality to occur, the wooden fuel assembly containers would have to burn such that the fuel inside would slump into a critical geometry configuration, a sufficient number of containers burn to form an infinite wide configuration, and sufficient water (about a 17 inch depth) be placed onto the slump. To obtain the appropriate geometric configuration, enough fuel assembly containers to form an infinite array on the floor would have to be stacked at least three high. Administrative controls require the stacks to be limited to two high for 1.25 wt% enriched finished fuel. This is not sufficient to allow for a critical mass regardless of the fire and accompanying water moderation. It should be noted that 0.95 wt% enriched fuel and billets are stacked higher than only two high. In this analysis, two initiating events will be considered. The first is a random fire that is hot enough and sufficiently long enough to burn away the containers and fuel separators such that the fuel can establish a critical mass. The second is a seismically induced fire with the same results.

  11. International safeguards for a modern MOX (mixed-oxide) fuel fabrication facility

    Energy Technology Data Exchange (ETDEWEB)

    Pillay, K.K.S.; Stirpe, D.; Picard, R.R.

    1987-03-01

    Bulk-handling facilities that process plutonium for commercial fuel cycles offer considerable challenges to nuclear materials safeguards. Modern fuel fabrication facilities that handle mixed oxides of plutonium and uranium (MOX) often have large inventories of special nuclear materials in their process lines and in storage areas for feed and product materials. In addition, the remote automated processing prevalent at new MOX facilities, which is necessary to minimize radiation exposures to personnel, tends to limit access for measurements and inspections. The facility design considered in this study incorporates all these features as well as state-of-the-art measurement technologies for materials accounting. Key elements of International Atomic Energy Agency (IAEA) safeguards for such a fuel-cycle facility have been identified in this report, and several issues of primary importance to materials accountancy and IAEA verifications have been examined. We have calculated detection sensitivities for abrupt and protracted diversions of plutonium assuming a single materials balance area for all processing areas. To help achieve optimal use of limited IAEA inspection resources, we have calculated sampling plans for attributes/variables verification. In addition, we have demonstrated the usefulness of calculating sigma/sub (MUF-D)/ and detection probabilities corresponding to specified material-loss scenarios and resource allocations. The data developed and the analyses performed during this study can assist both the facility operator and the IAEA in formulating necessary safeguards approaches and verification procedures to implement international safeguards for special nuclear materials.

  12. International safeguards for a modern MOX [mixed-oxide] fuel fabrication facility

    International Nuclear Information System (INIS)

    Pillay, K.K.S.; Stirpe, D.; Picard, R.R.

    1987-03-01

    Bulk-handling facilities that process plutonium for commercial fuel cycles offer considerable challenges to nuclear materials safeguards. Modern fuel fabrication facilities that handle mixed oxides of plutonium and uranium (MOX) often have large inventories of special nuclear materials in their process lines and in storage areas for feed and product materials. In addition, the remote automated processing prevalent at new MOX facilities, which is necessary to minimize radiation exposures to personnel, tends to limit access for measurements and inspections. The facility design considered in this study incorporates all these features as well as state-of-the-art measurement technologies for materials accounting. Key elements of International Atomic Energy Agency (IAEA) safeguards for such a fuel-cycle facility have been identified in this report, and several issues of primary importance to materials accountancy and IAEA verifications have been examined. We have calculated detection sensitivities for abrupt and protracted diversions of plutonium assuming a single materials balance area for all processing areas. To help achieve optimal use of limited IAEA inspection resources, we have calculated sampling plans for attributes/variables verification. In addition, we have demonstrated the usefulness of calculating σ/sub (MUF-D)/ and detection probabilities corresponding to specified material-loss scenarios and resource allocations. The data developed and the analyses performed during this study can assist both the facility operator and the IAEA in formulating necessary safeguards approaches and verification procedures to implement international safeguards for special nuclear materials

  13. Development of likelihood estimation method for criticality accidents of mixed oxide fuel fabrication facilities

    International Nuclear Information System (INIS)

    Tamaki, Hitoshi; Yoshida, Kazuo; Kimoto, Tatsuya; Hamaguchi, Yoshikane

    2010-01-01

    A criticality accident in a MOX fuel fabrication facility may occur depending on several parameters, such as mass inventory and plutonium enrichment. MOX handling units in the facility are designed and operated based on the double contingency principle to prevent criticality accidents. Control failures of at least two parameters are needed for the occurrence of criticality accident. To evaluate the probability of such control failures, the criticality conditions of each parameter for a specific handling unit are necessary for accident scenario analysis to be clarified quantitatively with a criticality analysis computer code. In addition to this issue, a computer-based control system for mass inventory is planned to be installed into MOX handling equipment in a commercial MOX fuel fabrication plant. The reliability analysis is another important issue in evaluating the likelihood of control failure caused by software malfunction. A likelihood estimation method for criticality accident has been developed with these issues been taken into consideration. In this paper, an example of analysis with the proposed method and the applicability of the method are also shown through a trial application to a model MOX fabrication facility. (author)

  14. Rough order of magnitude cost estimate for immobilization of 18.2 MT of plutonium sharing existing facilities at Hanford with MOX fuel fabrication facility: alternative 4B

    International Nuclear Information System (INIS)

    DiSabatino, A.

    1998-01-01

    The purpose of this Cost Estimate Report is to identify preliminary capital and operating costs for a facility to immobilize 18.2 metric tons (nominal) of plutonium as a ceramic in an existing facility at Hanford, the Fuels and Materials Examination Facility (FMEF). The MOX Fuel Fabrication Facility (MFFF), which is being costed in a separate report, will also be located in the FMEF in this co-location option

  15. Fabrication of an improved tube-to-pipe header heat exchanger for the Fuel Failure Mockup (FFM) Facility

    International Nuclear Information System (INIS)

    Prislinger, J.J.; Jones, R.H.

    1977-05-01

    The procedure used in fabricating an improved tube-to-pipe header heat exchanger for the Fuel Failure Mockup (FFM) Facility is described. Superior performance is accomplished at reduced cost with adherence to the ASME Boiler and Pressure Vessel Code. The techniques used and the method of fabrication are described in detail

  16. Review of Sodium and Plutonium related Technical Standards in Trans-Uranium Fuel Fabrication Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Misuk; Jeon, Jong Seon; Kang, Hyun Sik; Kim, Seoung Rae [NESS, Daejeon (Korea, Republic of)

    2016-10-15

    In this paper, we would introduce and review technical standards related to sodium fire and plutonium criticality safety. This paper may be helpful to identify considerations in the development of equipment, standards, and etc., to meet the safety requirements in the design, construction and operating of TFFF, KAPF and SFR. The feasibility and conceptual designs are being examined on related facilities, for example, TRU Fuel Fabrication Facilities (TFFF), Korea Advanced Pyro-process Facility (KAPF), and Sodium Cooled Fast Reactor (SFR), in Korea. However, the safety concerns of these facilities have been controversial in part because of the Sodium fire accident and Plutonium related radiation safety caused by transport and handling accident. Thus, many researches have been performed to ensure safety and various documents including safety requirements have been developed. In separating and reducing the long-lived radioactive transuranic(TRU) in the spent nuclear fuel, reusing as the potential energy of uranium fuel resources and reducing the high level wastes, TFFF would be receiving the attention of many people. Thus, people would wonder whether compliance with technical standards that ensures safety. For new facility design, one of the important tasks is to review of technical standards, especially for sodium and Plutonium because of water related highly reactive characteristics and criticality hazard respectively. We have introduced and reviewed two important technical standards for TFFF, which are sodium fire and plutonium criticality safety, in this paper. This paper would provide a brief guidance, about how to start and what is important, to people who are responsible for the initial design to operation of TFFF.

  17. Detailed description of an SSAC at the facility level for a low-enriched uranium conversion and fuel fabrication facility

    International Nuclear Information System (INIS)

    Jones, R.J.

    1984-09-01

    Some States have expressed a need for more detailed guidance with regard to the technical elements in the design and operation of SSACs for both the national and the international objectives. To meet this need the present document has been prepared, describing the technical elements of an SSAC in considerable detail. The purpose of this document is therefore, to provide a detailed description of a system for the accounting for and control of nuclear material in a model low enriched uranium conversion and fuel fabrication facility which can be used by a facility operator to establish his own system in a way which will provide the necessary information for compliance with a national system for nuclear material accounting and control and for the IAEA to carry out its safeguards responsibilities

  18. Glovebox glove deterioration in the Hanford Engineering Development Laboratory fuel fabrication facility

    International Nuclear Information System (INIS)

    Greenhalgh, W.O.; Smith, R.C.; Powell, D.L.

    1979-07-01

    Neoprene glovebox gloves have been found susceptible to periodic rapid deterioration under normal operating conditions in fuel fabrication facilities. Examinations of glove failure histories and measurements of the atmospheres in inert atmosphere dry-boxes indicated ozone at low concentrations of 100 to 500 ppB was probably the most important factor in rapid glove deterioration. Testing of a varity of new glove materials indicated that Hypalon and ethylene-propylene-diamine monomer (EDPM) gloves have greater than 30 times the longevity of neoprene in low-level ozone concentration atmospheres. comparative tests over a 30-month period have also confirmed that the two glove candidates have a significantly longer operative life. 14 figures

  19. Feedback on FBR fuel fabrication at ATPu facility, as a support to the design of a future facility

    International Nuclear Information System (INIS)

    Geneves, T.; Favet, D.; Audubert, F.; Paret, L.

    2013-01-01

    Conclusions: → Within 40 years, AREVA and CEA gained a large experience in the domain of Fast Breeder fuel fabrication: – Standard and experimental fuels from different designs; – Throughput from 500 kgHM/year to 18 tHM/year; – Technological improvements all along the 120tHM manufactured. → MELOX plant operates industrial methods and technologies since 17 years: – 2000 tHM LWR fuels produced; → The basic design of the workshop for ASTRID fuel fabrication is based on this large experience

  20. Options for converting excess plutonium to feed for the MOX fuel fabrication facility

    Energy Technology Data Exchange (ETDEWEB)

    Watts, Joe A [Los Alamos National Laboratory; Smith, Paul H [Los Alamos National Laboratory; Psaras, John D [Los Alamos National Laboratory; Jarvinen, Gordon D [Los Alamos National Laboratory; Costa, David A [Los Alamos National Laboratory; Joyce, Jr., Edward L [Los Alamos National Laboratory

    2009-01-01

    The storage and safekeeping of excess plutonium in the United States represents a multibillion-dollar lifecycle cost to the taxpayers and poses challenges to National Security and Nuclear Non-Proliferation. Los Alamos National Laboratory is considering options for converting some portion of the 13 metric tons of excess plutonium that was previously destined for long-term waste disposition into feed for the MOX Fuel Fabrication Facility (MFFF). This approach could reduce storage costs and security ri sks, and produce fuel for nuclear energy at the same time. Over the course of 30 years of weapons related plutonium production, Los Alamos has developed a number of flow sheets aimed at separation and purification of plutonium. Flow sheets for converting metal to oxide and for removing chloride and fluoride from plutonium residues have been developed and withstood the test oftime. This presentation will address some potential options for utilizing processes and infrastructure developed by Defense Programs to transform a large variety of highly impure plutonium into feedstock for the MFFF.

  1. 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

  2. Supplemental information for a notice of construction for the Fueled Clad Fabrication System, the Radioisotope Power Systems Facility, and the Fuel Assembly Area

    Energy Technology Data Exchange (ETDEWEB)

    1989-08-01

    This ''Notice of Construction'' has been submitted by the US Department of Energy-Richland Operations Office (P.O. Box 550, Richland, Washington 99352), pursuant to WAC 402-80-070, for three new sources of radionuclide emissions at the Hanford Site in Washington State (Figure 1). The three new sources, the Fueled Clad Fabrication System (FCFS) the Radioisotope Power Systems Facility (RPSF) and the Fuel Assembly Area (FAA) will be located in one facility, the Fuels and materials Examination Facility (FMEF) of the 400 Area. The FMEF was originally designed to provide for post- irradiation examination and fabrication of breeder reactor fuels. These FMEF missions were cancelled before the introduction of any fuel materials or any irradiated material. The current plans are to use the facility to fabricate power supplies to be used in space applications and to produce Fast Flux Test Facility (FFTF) fuel and target assemblies. The FCFS and the RPSF will produce materials and assemblies for application in space. The FAA project will produce FFTF fuel and target assemblies. The FCFS and the RPSF will share the same building, stack, and, in certain cases, the same floor space. Given this relationship, to the extent possible, these systems will be dealt with separately. The FAA is a comparatively independent operation though it will share the FMEF complex.

  3. Prevention of significant deterioration permit application for the Fueled Clad Fabrication System, the Radioisotope Power Systems Facility, and the Fuel Assembly Area

    Energy Technology Data Exchange (ETDEWEB)

    1989-08-01

    This New Source Review'' has been submitted by the US Department of Energy-Richland Operations Office (PO Box 550, Richland, Washington 99352), pursuant to WAC 173-403-050 and in compliance with the Department of Ecology Guide to Processing A Prevention Of Significant Deterioration (PSD) Permit'' for three new sources of radionuclide emissions at the Hanford Site in Washington State. The three new sources, the Fueled Clad Fabrication System (FCFS), the Radioisotope Power Systems Facility (RPSF), and the Fuel Assembly Area (FAA), will be located in one facility, the Fuels and Materials Examination Facility (FMEF) of the 400 Area. The FMEF was originally designed to provide for post-irradiation examination and fabrication of breeder reactor fuels. These FMEF missions were cancelled before the introduction of any fuel materials or any irradiated material. The current plans are to use the facility to fabricate power supplies for use in space applications and to produce Fast Flux Test Facility (FFTF) fuel and target assemblies. The FCFS and the RPSF will produce materials and assemblies for application in space. The FAA project will produce FFTF fuel and target assemblies. The FCFS and the RPSF will share the same building, stack, and, in certain cases, the same floor space. Given this relationship, these systems will be dealt with separately to the extent possible. The FAA is a comparatively independent operation though it will share the FMEF complex.

  4. Fabrication of nuclear fuel

    International Nuclear Information System (INIS)

    Ion, S.E.; Watson, R.H.; Loch, E.P.

    1989-01-01

    Commercial nuclear fuel fabrication is focused on providing quality products. Throughout manufacturing attention is placed on the precise control of processes, procedures and tooling to reduce product variability. This paper describes the processes used to manufacture fuel for the advanced gas cooled reactor, light water reactor, fast breeder reactor and Magnox reactor systems at the BNFL, Springfields, and Westinghouse Columbia, USA, facilities. It covers the chemical purification and conversion stages of fuel production before describing the fabrication of both oxide and metallic fuel and gives a general overview of the technology used. Although the principles of fuel fabrication in terms of providing a cladding designed to be the primary envelope to encase the fuel are similar for each system, there are detailed differences in manufacturing route and cladding type which arise from differing reactor requirements. These, together with developments made in response to customer needs, are reviewed and the need to maintain a high purity reliable product with low variability at all stages in manufacture is emphasized throughout. (author)

  5. Automated breeder fuel fabrication

    International Nuclear Information System (INIS)

    Goldmann, L.H.; Frederickson, J.R.

    1983-01-01

    The objective of the Secure Automated Fabrication (SAF) Project is to develop remotely operated equipment for the processing and manufacturing of breeder reactor fuel pins. The SAF line will be installed in the Fuels and Materials Examination Facility (FMEF). The FMEF is presently under construction at the Department of Energy's (DOE) Hanford site near Richland, Washington, and is operated by the Westinghouse Hanford Company (WHC). The fabrication and support systems of the SAF line are designed for computer-controlled operation from a centralized control room. Remote and automated fuel fabriction operations will result in: reduced radiation exposure to workers; enhanced safeguards; improved product quality; near real-time accountability, and increased productivity. The present schedule calls for installation of SAF line equipment in the FMEF beginning in 1984, with qualifying runs starting in 1986 and production commencing in 1987. 5 figures

  6. 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

  7. MOX fuel fabrication at AECL

    International Nuclear Information System (INIS)

    Dimayuga, F.C.; Jeffs, A.T.

    1995-01-01

    Atomic Energy of Canada Limited's mixed-oxide (MOX) fuel fabrication activities are conducted in the Recycle Fuel Fabrication Laboratories (RFFL) at the Chalk River Laboratories. The RFFL facility is designed to produce experimental quantities of CANDU MOX fuel for reactor physics tests or demonstration irradiations. From 1979 to 1987, several MOX fuel fabrication campaigns were run in the RFFL, producing various quantities of fuel with different compositions. About 150 bundles, containing over three tonnes of MOX, were fabricated in the RFFL before operations in the facility were suspended. In late 1987, the RFFL was placed in a state of active standby, a condition where no fuel fabrication activities are conducted, but the monitoring and ventilation systems in the facility are maintained. Currently, a project to rehabilitate the RFFL and resume MOX fuel fabrication is nearing completion. This project is funded by the CANDU Owners' Group (COG). The initial fabrication campaign will consist of the production of thirty-eight 37-element (U,Pu)O 2 bundles containing 0.2 wt% Pu in Heavy Element (H.E.) destined for physics tests in the zero-power ZED-2 reactor. An overview of the Rehabilitation Project will be given. (author)

  8. Atomics International fuel fabrication facility and low enrichment program [contributed by T.A. Moss, AI

    International Nuclear Information System (INIS)

    Moss, T.A.

    1993-01-01

    The AI facility is approximately 30,000 square feet in area and consists of four general areas. One area is devoted to the production of UAl x powder. It consists of a series of arc melting furnaces, crushing lines, glove boxes, and compacting presses. The second area is used for the rolling of fuel plates. The third area is used for the machining of the plates to final size and also the machining of the fuel elements. In the fourth area the fuel plates are swaged into assemblies, and all welding and inspection operations are performed. As part of the lower enrichment program we are scheduled to put a second UAl x powder line into operation and we have had to expand some of our storage area. Under the low enrichment program the AI fuel facility will be modified to accommodate a separate low enrichment Al x production line and compacting line. This facility modification should be done by the end of the fiscal year. We anticipate producing fuel with an enrichment slightly less than 20% We anticipate powder being available for plate production shortly after the facility is completed. Atomics International is scheduled to conduct plate LEU verification work using fully enriched material in the June-July time period, at which time we will investigate what level of uranium loadings we can go to using the current process. It is anticipated that 55 volume percent uranium compound in our fuel form can be achieved

  9. Advanced fuel fabrication

    International Nuclear Information System (INIS)

    Bernard, H.

    1989-01-01

    This paper deals with the fabrication of advanced fuels, such as mixed oxides for Pressurized Water Reactors or mixed nitrides for Fast Breeder Reactors. Although an extensive production experience exists for the mixed oxides used in the FBR, important work is still needed to improve the theoretical and technical knowledge of the production route which will be introduced in the future European facility, named Melox, at Marcoule. Recently, the feasibility of nitride fuel fabrication in existing commercial oxide facilities was demonstrated in France. The process, based on carbothermic reduction of oxides with subsequent comminution of the reaction product, cold pressing and sintering provides (U, Pu)N pellets with characteristics suitable for irradiation testing. Two experiments named NIMPHE 1 and 2 fabricated in collaboration with ITU, Karlsruhe, involve 16 nitride and 2 carbide pins, operating at a linear power of 45 and 73 kW/m with a smear density of 75-80% TD and a high burn-up target of 15 at%. These experiments are currently being irradiated in Phenix, at Marcoule. (orig.)

  10. Cost effectiveness of robotics and remote tooling for occupational risk reduction at a nuclear fuel fabrication facility

    International Nuclear Information System (INIS)

    Lochard, Jacques

    1989-01-01

    This case study, related to the design stage of a fuel fabrication facility, presents the evaluation of alternative options to manipulate mixed oxide fuel rods in a quality control shop. It is based on a study performed in the framework of the 'MELOX project' developed by COGEMA in France. The methodology for evaluating robotic actions is resulting from a research work part funded by the IAEA under the co-ordinated research programme on 'Comparison of cost-effectiveness of risk reduction among different energy systems', and by the commission of the European Communities under the research and training programme on radiation protection

  11. Operational experience in the non-destructive assay of fissile material in General Electric's nuclear fuel fabrication facility

    International Nuclear Information System (INIS)

    Stewart, J.P.

    1976-01-01

    Operational experience in the non-destructive assay of fissile material in a variety of forms and containers and incorporation of the assay devices into the accountability measurement system for General Electric's Wilmington Fuel Fabrication Facility measurement control programme is detailed. Description of the purpose and related operational requirements of each non-destructive assay system is also included. In addition, the accountability data acquisition and processing system is described in relation to its interaction with the various non-destructive assay devices and scales used for accountability purposes within the facility. (author)

  12. Atomics international fuel fabrication facility and low enrichment program [contributed by H.W. Hassel, NUKEM

    International Nuclear Information System (INIS)

    Hassel, H.W.

    1993-01-01

    NUKEM handles around almost two tons of highly enriched uranium a year and it was necessary to satisfy all the new physical protection philosophies. That means that we have to install storage and safe fabrication sites for a lot of money, 25 meter thick concrete walls, and different alarm systems. So just to demonstrate how silly this business is, we have just overcome this for highly enriched uranium, and now we speak about low enriched uranium for which we don't need all of these investments to make this business safe. I would-just like to concentrate my words on the status of fabrication and considerations in my company concerning the medium enriched uranium and low enriched uranium. In the table are the different fuel types (see column) and then we have the fabrication in column 2 the experience of my comp any in kg. In column 3 is the irradiation experience of these fuels types. Column 4 shows the studies and calculations made in our company for lower and medium enriched fuels. The preliminary fabrication tests and calculations are in column 5, and in column 6 we have the delivery time for a prototype core in months after UF 6 supply. Column 7 shows the time for the development of specifications including irradiation time in years for 6 and 7 and column is the estimated cost of 6 and 7 There is just one fuel that is not in this summary and that is U-Zr. We now see how complex and sophisticated this business is. I have told you already that we have installed for a lot of millions of Deutsche Mark the physical protection, storage vaults and things like that. Now we have to investigate all these different types of fuels for, as you see, a lot of money. Maybe these are a lot of optimistic figures; anyway the question is, does this make all the overall nuclear situation worldwide easier or not. One cannot answer for the moment, but anyway we have a lot of problems

  13. Coordinated safeguards for materials management in chemical separation, conversion, and fuel fabrication facilities

    International Nuclear Information System (INIS)

    Dayem, H.A.; Cobb, D.; Dietz, R.J.; Hakkila, E.A.; Shipley, J.P.; Smith, D.B.

    1978-01-01

    The benefits of dynamic materials accounting are demonstrated by comparing the timeliness, spatial specificity, and sensitivity with that of current materials accounting requirements. Examples of conceptual dynamic accountability systems are given for proposed domestic fuel-cycle facilities. Modeling and simulation results indicate that such systems can detect diversion in hours or days with improved diversion sensitivity and can localize diversion to a single unit process area. 17 refs

  14. Atomics International fuel fabrication facility and low enrichment program. Part 1

    International Nuclear Information System (INIS)

    Moss, T.A.

    1993-01-01

    The Al facility is approximately 30,000 square feet in area and consists of four general areas. One area is devoted to the production of UAl x powder. It consists of a series of arc melting furnaces, crushing lines, glove boxes, and compacting presses. The second area is used for the rolling of fuel plates. The third area is used for the machining of the plates to final size and also the machining of the fuel elements. In the fourth area the fuel plates are swaged into assemblies, and all welding and inspection operations are performed. As part of the lower enrichment program we are scheduled to put a second UAl x powder line into operation and we have had to expand some of our storage area

  15. Atomics International fuel fabrication facility and low enrichment program. Part 2

    International Nuclear Information System (INIS)

    Hassel, H.W.

    1993-01-01

    Most of you know our company from the last meeting in May in Vienna, so I won't steal your time with explaining and demonstrating the same techniques that we have heard this morning f rom the other speakers. I would just take some words to explain the order of business with highly enriched uranium. NUKEM handles around almost two tons of highly enriched uranium a year and it was necessary to satisfy all the new physical protection philosophies. That means that we have to install storage and safe fabrication sites for a lot of money, 2.5 meter thick concrete walls, and different alarm systems. So just to demonstrate how silly this business is, we have just overcome this for highly enriched uranium, and now we speak about low enriched uranium for which we don't need all of these investments to make this business safe. I would just like to concentrate my words on the status of fabrication and considerations in my company concerning the medium enriched uranium and low enriched uranium. In TABLE I are the different fuel types (see column 1) and then we have the fabrication in column 2; (The reason that I use the blackboard this morning is that I try to demonstrate all the techniques. However, all the speakers before me did this and in theory we are not so far away from each other.) the experience of my company in kg. In column 3 is the irradiation experience of these fuels types. Column 4 shows the studies and calculations made in our company for lower and medium enriched fuels. The preliminary fabrication tests and calculations are in column 5, and in column 6 we have the delivery time for a prototype core in months after UF 6 supply. Column 7 shows the time for the development of specifications including irradiation time in years for 6 and 7, and column 8 is the estimated cost of 6 and 7. There is just one fuel that is not in this summary and that is U-Zr

  16. FFTF metal fuel pin fabrication

    International Nuclear Information System (INIS)

    Dittmer, J.O.; Benecke, M.W.; Feigenbutz, L.V.

    1989-01-01

    A major new initiative to develop, irradiate, and qualify a binary uranium/zirconium metal-fuel system in the Fast Flux Test Facility (FFTF) has been implemented by the Westinghouse Hanford Company for the US Department of Energy. Metal-fuel test assemblies have been designed and fabricated, and are now being irradiated in FFTF to provide the data needed to support the potential use of binary metal fuels in FFTF and other liquid-metal reactors. These development efforts support licensing activities for metal-fuel use in near-term advanced liquid-metal reactors

  17. LEU fuel fabrication in Argentina

    International Nuclear Information System (INIS)

    Giorsetti, D.R.; Gomez, J.O.; Marajofsky, A.; Kohut, C.

    1985-01-01

    As an Institution, aiming to meet with its own needs, CNEA has been intensively developing reduced enriched fuel to use in its own research and test reactors. Development of the fabrication technology as well as the design, installation and operation of the manufacturing plant, have been carried out with its own funds. Irradiation and post-irradiation of test miniplates have been taking place within the framework of the RERTR program. During the last years, CNEA has developed three LEU fuel types. In the previous RERTR meetings, we presented the technological results obtained with these fuel types. This paper focuses on CNEA LEU fuel element manufacturing status and the trained personnel we can offer in design and manufacture fuel capability. CNEA has its own fuel manufacturing technology; the necessary facilities to start the fuel fabrication; qualified technicians and professionals for: fuel design and behaviour analysis; fuel manufacturing and QA; international recognition of its fuel development and manufacturing capability through its ORR miniplate irradiation; its own natural uranium and the future possibility to enrich up to 20% U 235 ; the probability to offer a competitive fuel manufacturing cost in the international market; the disposition to cooperate with all countries that wish to take part and aim to reach an self-sufficiency in their own fuel supply needs

  18. Criticality safety evaluation report for 300 Area N Reactor fuel fabrication and storage facility

    International Nuclear Information System (INIS)

    Schwinkendorf, K.N.

    1993-12-01

    Criticality safety support calculations for the 300 Area Fuels Manufacturing Facility have been performed to update values currently found in the criticality prevention specifications. The old safe masses and associated dimensions table has been updated with new values, which were produced using more modern computer codes that comply with Software Quality Assurance (SQA) requirements. Monte Carlo calculations, using MCNP (Carter 1991), were performed to spot check these results, obtained using the WIMS-E code (Gubbins et. al. 1982). In addition, the criticality considerations associated with certain accident, or upset, conditions were analyzed. These scenarios include fire, the bringing together of multiple safe masses into one neutronically coupled system, misstacking, and accidental interspersed moderation

  19. On-line item control at a high enriched nuclear fuel fabrication facility

    International Nuclear Information System (INIS)

    Lewis, T.W.; Lewis, H.M.

    1984-01-01

    The on-line item control system at Nuclear Fuel Services, Inc., is a near-real time method capable of tracking uniquely identified items from creation through disposition. The system provides for improved control, timeliness, accuracy and usability of company information and the necessary data required to support the regulatory program for the protection against diversion of Special Nuclear Materials. The system consists of software applications (approximately 150 programs) with man/machine interface controls which provide facilities for correct data entry and for the protection of data integrity. This system went into stand-alone operation in September, 1983 after a twenty month parallel test run with the previous keybatched (manual forms) item control system

  20. Development of ISA procedure for uranium fuel fabrication and enrichment facilities: overview of ISA procedure and its application

    International Nuclear Information System (INIS)

    Yamate, Kazuki; Yamada, Takashi; Takanashi, Mitsuhiro; Sasaki, Noriaki

    2013-01-01

    Integrated Safety Analysis (ISA) procedure for uranium fuel fabrication and enrichment facilities has been developed for aiming at applying risk-informed regulation to these uranium facilities. The development has carried out referring to the ISA (NUREG-1520) by the Nuclear Regulatory Commission (NRC). The paper presents purpose, principles and activities for the development of the ISA procedure, including Risk Level (RL) matrix and grading evaluation method of IROFS (Items Relied on for Safety), as well as general description and features of the procedure. Also described in the paper is current status in application of risk information from the ISA. Japanese four licensees of the uranium facilities have been conducting ISA for their representative processes using the developed procedure as their voluntary safety activities. They have been accumulating experiences and knowledge on the ISA procedure and risk information through the field activities. NISA (Nuclear and Industrial Safety Agency) and JNES (Japan Nuclear Energy Safety Organization) are studying how to use such risk information for the safety regulation of the uranium facilities, taking into account the licensees' experiences and knowledge. (authors)

  1. Catalytic Fuel Conversion Facility

    Data.gov (United States)

    Federal Laboratory Consortium — This facility enables unique catalysis research related to power and energy applications using military jet fuels and alternative fuels. It is equipped with research...

  2. Status of plutonium recycle from mixed oxide fuel fabrication wastes (U,Pu)O2 facility activities

    International Nuclear Information System (INIS)

    Quesada, Calixto A.; Adelfang, Pablo; Greiner, G.; Orlando, Oscar S.; Mathot, Sergio R.

    1999-01-01

    Within the specific subject of mixed oxides corresponding to the Fuel Cycle activities performed at CNEA, the recovery of plutonium from wastes originated during tests and pre-fabrication stages is performed. (author)

  3. Introduction of a new programme-device facilities ASU TP in the frameworks of creation of experimental-technological complex for MOX fuel fabrication

    International Nuclear Information System (INIS)

    Vol'skij, A.S.; Kurbatov, A.A.

    2005-01-01

    New facilities are developed and introduced for automation of technological process for control and maintaining of technological parameters of the process of MOX fuel fabrication. All automated control systems are individual distributed subsystems and are combined by computer networks for storage and treatment of technological and accounting information [ru

  4. 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)

  5. Integrated software package for nuclear material safeguards in a MOX fuel fabrication facility

    International Nuclear Information System (INIS)

    Schreiber, H.J.; Piana, M.; Moussalli, G.; Saukkonen, H.

    2000-01-01

    Since computerized data processing was introduced to Safeguards at large bulk handling facilities, a large number of individual software applications have been developed for nuclear material Safeguards implementation. Facility inventory and flow data are provided in computerized format for performing stratification, sample size calculation and selection of samples for destructive and non-destructive assay. Data is collected from nuclear measurement systems running in attended, unattended mode and more recently from remote monitoring systems controlled. Data sets from various sources have to be evaluated for Safeguards purposes, such as raw data, processed data and conclusions drawn from data evaluation results. They are reported in computerized format at the International Atomic Energy Agency headquarters and feedback from the Agency's mainframe computer system is used to prepare and support Safeguards inspection activities. The integration of all such data originating from various sources cannot be ensured without the existence of a common data format and a database system. This paper describes the fundamental relations between data streams, individual data processing tools, data evaluation results and requirements for an integrated software solution to facilitate nuclear material Safeguards at a bulk handling facility. The paper also explains the basis for designing a software package to manage data streams from various data sources and for incorporating diverse data processing tools that until now have been used independently from each other and under different computer operating systems. (author)

  6. 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)

  7. Dispersion of long-lived radionuclides from uranium mining, milling and fuel fabrication facilities

    International Nuclear Information System (INIS)

    Pettersson, H.B.L.

    1990-11-01

    The principal aim of the study was to gain further insight into the environmental dispersion of long-lived U series radionuclides from selected part of the nuclear fuel cycle and to assess the resulting exposure of members of the public. The specific objectives of this study were: 1. To determine the levels of natural radioactivity in the vicinity of two U deposits in Sweden and to establish whether U prospecting had generated significant radiological impact on man. 2. To investigate the spatial distributions of long-lived U series radionuclides caused by the dispersion of dust from the Ranger open-pit U mine in Australia. 3. To study the uptakes of long-lived U and T series radionuclides by the waterlily in order to facilitate assessment of natural exposures to the public and predictions of exposures arising from consumption of the plant due to any subsequent discharges of water from the Ranger U mine. 4. To investigate the spatial distributions of U isotopes in environmental air as a result of the release of radionuclides from the ABB-ATOM nuclear fuel factory at Vaesteraas in Sweden. In these investigations special emphasis was given to - activity ratio techniques suitable for distinguishing between natural and operation-related concentrations and for facilitating determination of the source of radionuclide uptake in the waterlily, and - the use of passive air samplers such as 'sticky vinyl' and bioindicators in investigating the aerial dispersion of radionuclides. (author)

  8. Improvements in fabrication of metallic fuels

    International Nuclear Information System (INIS)

    Tracy, D.B.; Henslee, S.P.; Dodds, N.E.; Longua, K.J.

    1989-12-01

    Argonne National Laboratory is currently developing a new liquid- metal cooled breeder reactor known as the Integral Fast Reactor (IFR). IFR fuels represent the state-of-the-art in metal-fueled reactor technology. Improvements in the fabrication of metal fuel, to be discussed below, will support the fully remote fuel cycle facility that as an integral part of the IFR concept will be demonstrated at the EBR-II site. 3 refs

  9. Safeguards implementation practices for a model mixed oxide recycle fuel fabrication facility

    International Nuclear Information System (INIS)

    Glancy, J.E.

    1976-05-01

    Conclusions on the magnitude of the inspection effort and its effectiveness in meeting IAEA objectives are summarized: the minimum goal quantity that can be detected with 95 percent probability is 6 kg Pu. eight to nine resident inspectors are necessary to monitor all operations (3 shifts a day) and to perform verification measurements (23 per day). The facility must also perform inventories monthly if the goal of 6.0 kg is to be attained. Bi-monthly inventory and continual on-site resident inspection can achieve a detection goal quantity of 8.0 kg Pu with 95 percent detection probability. For annual inventories, the detection goal quantity for a 95 percent detection probability is 40 kg Pu. The probability for detecting 8 kg is less than 20 percent. The total inspection effort for a 200 MT/year RFP is estimated to exceed 15 man-years if IAEA headquarters support and analytical laboratory time are added to on-site manpower requirements. Surveillance and containment in the form of seals must be applied to prevent double inventorying and to assure that samples are extracted from the vessel or container being measured. These assurances will probably be achieved by on-line instrumentation installed by the IAEA prior to plant start-up. For assessment, detailed knowledge of plant equipment and floor layout including piping used to transport material is needed. Such information should be requested from the facility prior to preparation of the SIP. Present NDA appears capable of measuring all material in the RFP; however, verification of residual hold-up will require development of new procedures. In-process inventory verification will require special procedures but can be accomplished with present technology. Alternatives to accountability such as perimeter surveillance and containment of declared material within the perimeter should be studied in order to improve detection sensitivity or decrease the effort

  10. Facile in-situ fabrication of graphene/riboflavin electrode for microbial fuel cells

    International Nuclear Information System (INIS)

    Wang, Qian-Qian; Wu, Xia-Yuan; Yu, Yang-Yang; Sun, De-Zhen; Jia, Hong-Hua; Yong, Yang-Chun

    2017-01-01

    A novel graphene/riboflavin (RF) composite electrode was developed and its potential application as microbial fuel cell (MFC) anode was demonstrated. Graphene layers were first grown on the surface of graphite electrode by a one-step in-situ electrochemical exfoliation approach. Then, noncovalent functionalization of the graphene layers with RF was achieved by a simple spontaneous adsorption process. The graphene/RF electrode was extensively characterized by transmission electron microscopy, Fourier transform infrared spectroscopy, Raman analysis, and cyclic voltammetry analysis. Remarkably, when applied as the anode of Shewanella oneidensis MR-1 inoculated MFCs, the graphene/RF electrode significantly decreased charge transfer over-potential and enhanced cell attachment, which in turn delivered about 5.3- and 2.5-fold higher power output, when compared with that produced by the bare graphite paper electrode and graphene electrode, respectively. These results demonstrated that electron shuttle immobilization on the electrode surface could be a promising and practical strategy for improving the performance of microbial electrochemical systems.

  11. Results of current fabrication technology developments and MOX fuel fabrication for ''JOYO'' MK-III initial load fuel

    International Nuclear Information System (INIS)

    Kayano, Masashi

    2003-01-01

    In the Plutonium Fuel Production Facility (PFPF), MOX fuel fabrication technologies have been developed and demonstrated through MOX fuels fabrication for Experimental Fast Reactor ''JOYO'' and Prototype FBR MONJU since 1988. From 1995 to 2000, replacement, modification and repair works for process equipment were conducted to improve performance of the MOX pellet fabrication process in PFPF as scheduled shut-down maintenance. Because the MOX fuel fabrication for ''JOYO'' MK-III initial load fuels was the first fuel fabrication after these major maintenance works, the performance of the MOX pellet fabrication process in PFPF was evaluated though this fuel fabrication experience. This MOX fuel fabrication was completed within the scheduled period and showed higher yield of product MOX pellets than before. Therefore, the performance of the MOX pellet fabrication process in PFPF was improved by this maintenance work. Furthermore, the quality assurance system for MOX fuel fabrication was strengthened by acquisition of ISO9001 certificate in 2002. (author)

  12. 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)

  13. 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)

  14. Conceptual design study of advanced fuel fabrication systems

    International Nuclear Information System (INIS)

    Ken-ya, Tanaka; Shusaku, Kono; Kiyoshi, Ono

    2001-01-01

    The fuel fabrication plant images based on the advanced equipment with availability to operate in hot-cell facility are constructed. The characteristics of each fuel fabrication system for economical and environmental are evaluated roughly. The advanced fuel fabrication routes such as simplified pelletizing, vibration compaction and casting process would have the potential for reducing plant construction cost and minimizing the radioactive waste generated from fuel fabrication process. (author)

  15. Improvements in the fabrication of metallic fuels

    International Nuclear Information System (INIS)

    Tracy, D.B.; Henslee, S.P.; Dodds, N.E.; Longua, K.J.

    1989-01-01

    Argonne National Laboratory (ANL) is currently developing a new liquid-metal-cooled breeder reactor known as the Integral Fast Reactor (IFR). The IFR represents the state of the art in metal-fueled reactor technology. Improvements in the fabrication of metal fuel, discussed in this paper, will support ANL-West's (ANL-W) fully remote fuel cycle facility, which is an integral part of the IFR concept

  16. Nuclear fuel storage facility

    International Nuclear Information System (INIS)

    Matsumoto, Takashi; Isaka, Shinji.

    1987-01-01

    Purpose: To increase the spent fuel storage capacity and reduce the installation cost in a nuclear fuel storage facility. Constitution: Fuels handled in the nuclear fuel storage device of the present invention include the following four types: (1) fresh fuels, (2) 100 % reactor core charged fuels, (3) spent fuels just after taking out and (4) fuels after a certain period (for example one half-year) from taking out of the reactor. Reactivity is high for the fuels (1), and some of fuels (2), while low in the fuels (3) (4), Source intensity is strong for the fuels (3) and some of the fuels (2), while it is low for the fuels (1) and (4). Taking notice of the fact that the reactivity, radioactive source intensity and generated after heat are different in the respective fuels, the size of the pool and the storage capacity are increased by the divided storage control. While on the other hand, since the division is made in one identical pool, the control method becomes important, and the working range is restricted by means of a template, interlock, etc., the operation mode of the handling machine is divided into four, etc. for preventing errors. (Kamimura, M.)

  17. Report of an investigation into deterioration of the Plutonium Fuel Form Fabrication Facility (PuFF) at the DOE Savannah River Site

    International Nuclear Information System (INIS)

    1991-10-01

    This investigations of the Savannah River Site's Plutonium Fuel Form fabrication facility located in Building 235-F was initiated in April 1991. The purpose of the investigation was to determine whether, as has been alleged, operation of the facility's argon inert gas system was terminated with the knowledge that continued inoperability of the argon system would cause accelerated corrosion damage to the equipment in the plutonium 238 processing cells. The investigation quickly established that the decision to discontinue operation of the argon system, by not repairing it, was merely one of the measures, and not the most important one, which led to the current deteriorated state of the facility. As a result, the scope of the investigation was broadened to more identify and assess those factors which contributed to the facility's current condition. This document discusses the backgrounds, results, and recommendations of this investigation

  18. Report of an investigation into deterioration of the Plutonium Fuel Form Fabrication Facility (PuFF) at the DOE Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    1991-10-01

    This investigations of the Savannah River Site's Plutonium Fuel Form fabrication facility located in Building 235-F was initiated in April 1991. The purpose of the investigation was to determine whether, as has been alleged, operation of the facility's argon inert gas system was terminated with the knowledge that continued inoperability of the argon system would cause accelerated corrosion damage to the equipment in the plutonium 238 processing cells. The investigation quickly established that the decision to discontinue operation of the argon system, by not repairing it, was merely one of the measures, and not the most important one, which led to the current deteriorated state of the facility. As a result, the scope of the investigation was broadened to more identify and assess those factors which contributed to the facility's current condition. This document discusses the backgrounds, results, and recommendations of this investigation.

  19. Fabrication of CANDU dupic fuel

    International Nuclear Information System (INIS)

    Sullivan, J.D.; Ryz, M.A.; Lee, J.W.

    1999-01-01

    An important new fuel cycle that exploits the synergism between CANDU and pressurized-water reactors (PWRs) is the Direct Use of spent PWR Fuel in CANDU (DUPIC). In this fuel cycle, spent PWR fuel is reconfigured, using only dry processing techniques, to make it compatible with a CANDU reactor. The dry processing technique is inherently simpler than wet chemical processing techniques used for recycling spent fuel. Actinides and fission products are retained in the fuel, so that DUPIC fuel is highly radioactive, affording the fuel cycle a high degree of proliferation resistance. AECL's project to develop the fuel cycle has now progressed to the stage of fabricating DUPIC fuel elements for irradiation testing in a research reactor. The goal of this phase of the project is to demonstrate that the DUPIC fuel cycle is technically feasible. A major part of the technical feasibility study is demonstration of the irradiation performance of DUPIC fuel under CANDU conditions. Spent PWR fuel has been subjected to the oxidation and reduction of oxide fuels (OREOX) process, and the resulting powder has been fabricated into CANDU-quality pellets. The DUPIC pellets have been loaded into fuel elements for irradiation testing in the NRU research reactor at the Chalk River Laboratories. The fabrication stages included spent fuel decladding, powder production using the OREOX process, powder milling (to improve sinterability), pellet pressing, sintering, centreless grinding, element loading and element welding. This paper details the fabrication of the DUPIC pellets and elements and initial results of their characterization. The equipment used for fabrication of the DUPIC fuel elements is described, and the irradiation plan for these elements is also outlined. (author)

  20. The fabrication of nuclear fuel

    International Nuclear Information System (INIS)

    D'Amore, Mead

    1987-01-01

    The chronology of fuel product and core management development over the past 25 years in the USA is explained. Nuclear fuel for Westinghouse reactors is made by converting enriched uranium hexafluoride (UF 6 ) into uranium dioxide (UO 2 ) powder. The powder is pressed into pellets which are loaded into zircalloy fuel tubes (typically over 14 million pellets in 50,952 rods). The fuel rods are arranged in fuel assemblies which are shipped to the reactor site (typically 193 fuel assemblies are needed for one 1000MWe reactor). Each stage of the fuel fabrication cycle (cladding manufacture, chemical conversion UF 6 - UO 2 , pellet production, fuel rod fabrication, grid assembly, skeleton assembly, fuel assembly) is described, with particular reference to the Westinghouse process and plant. (UK)

  1. 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

  2. Experimental Fabrication Facility

    Data.gov (United States)

    Federal Laboratory Consortium — Provides aviation fabrication support to special operations aircraft residing at Fort Eustis and other bases in the United States. Support is also provided to AATD...

  3. Automated fabrication of reactor fuel

    International Nuclear Information System (INIS)

    Nyman, D.H.; McLemore, D.R.; Bennett, D.W.; Yatabe, J.M.

    1980-01-01

    The development of Secure Automated Fabrication (SAF) methods is the goal of the United States breeder reactor program and will form the technological basis for the future breeder reactor fuel supply. A major factor in achieving this goal is the development of remotely operated fuel fabrication equipment. The unit operations are being designed for microprocessor control connected to a central control center. The program schedule is to demonstrate the feasibility of automated pellet fuel fabrication and remote maintenance techniques by the mid-1980's. Development of major ceramic unit operations and the required computer control system is currently in the engineering testing stage

  4. 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

  5. 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.

  6. Neutronic fuel element fabrication

    Science.gov (United States)

    Korton, George

    2004-02-24

    This disclosure describes a method for metallurgically bonding a complete leak-tight enclosure to a matrix-type fuel element penetrated longitudinally by a multiplicity of coolant channels. Coolant tubes containing solid filler pins are disposed in the coolant channels. A leak-tight metal enclosure is then formed about the entire assembly of fuel matrix, coolant tubes and pins. The completely enclosed and sealed assembly is exposed to a high temperature and pressure gas environment to effect a metallurgical bond between all contacting surfaces therein. The ends of the assembly are then machined away to expose the pin ends which are chemically leached from the coolant tubes to leave the coolant tubes with internal coolant passageways. The invention described herein was made in the course of, or under, a contract with the U.S. Atomic Energy Commission. It relates generally to fuel elements for neutronic reactors and more particularly to a method for providing a leak-tight metal enclosure for a high-performance matrix-type fuel element penetrated longitudinally by a multiplicity of coolant tubes. The planned utilization of nuclear energy in high-performance, compact-propulsion and mobile power-generation systems has necessitated the development of fuel elements capable of operating at high power densities. High power densities in turn require fuel elements having high thermal conductivities and good fuel retention capabilities at high temperatures. A metal clad fuel element containing a ceramic phase of fuel intimately mixed with and bonded to a continuous refractory metal matrix has been found to satisfy the above requirements. Metal coolant tubes penetrate the matrix to afford internal cooling to the fuel element while providing positive fuel retention and containment of fission products generated within the fuel matrix. Metal header plates are bonded to the coolant tubes at each end of the fuel element and a metal cladding or can completes the fuel-matrix enclosure

  7. 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

  8. HEU Measurements of Holdup and Recovered Residue in the Deactivation and Decommissioning Activities of the 321-M Reactor Fuel Fabrication Facility at the Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    DEWBERRY, RAYMOND; SALAYMEH, SALEEM R.; CASELLA, VITO R.; MOORE, FRANK S.

    2005-03-11

    This paper contains a summary of the holdup and material control and accountability (MC&A) assays conducted for the determination of highly enriched uranium (HEU) in the deactivation and decommissioning (D&D) of Building 321-M at the Savannah River Site (SRS). The 321-M facility was the Reactor Fuel Fabrication Facility at SRS and was used to fabricate HEU fuel assemblies, lithium-aluminum target tubes, neptunium assemblies, and miscellaneous components for the SRS production reactors. The facility operated for more than 35 years. During this time thousands of uranium-aluminum-alloy (U-Al) production reactor fuel tubes were produced. After the facility ceased operations in 1995, all of the easily accessible U-Al was removed from the building, and only residual amounts remained. The bulk of this residue was located in the equipment that generated and handled small U-Al particles and in the exhaust systems for this equipment (e.g., Chip compactor, casting furnaces, log saw, lathes A & B, cyclone separator, Freon{trademark} cart, riser crusher, ...etc). The D&D project is likely to represent an important example for D&D activities across SRS and across the Department of Energy weapons complex. The Savannah River National Laboratory was tasked to conduct holdup assays to quantify the amount of HEU on all components removed from the facility prior to placing in solid waste containers. The U-235 holdup in any single component of process equipment must not exceed 50 g in order to meet the container limit. This limit was imposed to meet criticality requirements of the low level solid waste storage vaults. Thus the holdup measurements were used as guidance to determine if further decontamination of equipment was needed to ensure that the quantity of U-235 did not exceed the 50 g limit and to ensure that the waste met the Waste Acceptance Criteria (WAC) of the solid waste storage vaults. Since HEU is an accountable nuclear material, the holdup assays and assays of recovered

  9. Waste management in MOX fuel fabrication plants

    International Nuclear Information System (INIS)

    Schneider, V.

    1982-01-01

    After a short description of a MOX fuel fabrication plant's activities the waste arisings in such a plant are discussed according to nature, composition, Pu-content. Experience has shown that proper recording leads to a reduction of waste arisings by waste awareness. Aspects of the treatment of α-waste are given and a number of treatment processes are reviewed. Finally, the current waste management practice and the α-waste treatment facility under construction at ALKEM are outlined. (orig./RW)

  10. Automated fabrication of thoria based fuel: challenges and initiatives

    International Nuclear Information System (INIS)

    Jayarajan, K.

    2015-01-01

    The processes and techniques for fabrication of uranium-based fuel and plutonium-based fuel are well established. However, fabrication of thorium-based fuel has an additional challenge, due to high gamma radiation. 233 U produced from thorium in reactor is contaminated with 232 U, whose daughter products are hard gamma emitters. Therefore, 233 U fuel can be fabricated only in shielded cells. We have limited experience of remote fuel fabrication in shielded facility. Therefore, a mock up facility is being set-up at BARC to demonstrate fully remotised and fully automated fuel fabrication. In the facility, the automated handling of mixed oxide powder, fuel pellets, fuel pins and fuel cluster will be demonstrated. In addition to various processing, these components are subjected to a many inspections, such as dimension inspection, weight measurement, visual inspection and various NDTs. Mock-up systems for automated fabrication of pellets and pins are already made. Many novel tools and techniques were developed for automated material transfer. Another outcome of the development is the layout of hot cell and equipments, optimising the hot cell space, the number of material transfers and the number of equipments. The layout also provides access for remote maintenance

  11. Nuclear fuel elements design, fabrication and performance

    CERN Document Server

    Frost, Brian R T

    1982-01-01

    Nuclear Fuel Elements: Design, Fabrication and Performance is concerned with the design, fabrication, and performance of nuclear fuel elements, with emphasis on fast reactor fuel elements. Topics range from fuel types and the irradiation behavior of fuels to cladding and duct materials, fuel element design and modeling, fuel element performance testing and qualification, and the performance of water reactor fuels. Fast reactor fuel elements, research and test reactor fuel elements, and unconventional fuel elements are also covered. This volume consists of 12 chapters and begins with an overvie

  12. Research on plant of metal fuel fabrication using casting process

    International Nuclear Information System (INIS)

    Senda, Yasuhide; Mori, Yukihide

    2003-12-01

    This document presents the plant concept of metal fuel fabrication system (38tHM/y) using casting process in electrolytic recycle, which based on recent studies of its equipment design and quality control system. And we estimate the cost of its construction and operation, including costs of maintenance, consumed hardware and management of waste. The content of this work is as follows. (1) Designing of fuel fabrication equipment: We make material flow diagrams of the fuel fabrication plant and rough designs of the injection casting furnace, demolder and inspection equipment. (2) Designing of resolution system of liquid waste, which comes from analytical process facility. Increased analytical items, we rearrange analytical process facility, estimate its chemicals and amount of waste. (3) Arrangement of equipments: We made a arrangement diagram of the metal fuel fabrication equipments in cells. (4) Estimation of cost data: We estimated cost to construct the facility and to operate it. (author)

  13. Descriptions of reference LWR facilities for analysis of nuclear fuel cycles. Appendixes

    International Nuclear Information System (INIS)

    Schneider, K.J.; Kabele, T.J.

    1979-09-01

    The appendixes present the calculations that were used to derive the release factors discussed for each fuel cycle facility in Volume I. Appendix A presents release factor calculations for a surface mine, underground mine, milling facility, conversion facility, diffusion enrichment facility, fuel fabrication facility, PWR, BWR, and reprocessing facility. Appendix B contains additional release factors calculated for a BWR, PWR, and a reprocessing facility. Appendix C presents release factors for a UO 2 fuel fabrication facility

  14. Fuel pellets of various shapes- fabrication experience

    International Nuclear Information System (INIS)

    Ramachandran, R.; Nair, M.R.; Majumdar, S.; Purushotham, D.S.C.

    1996-10-01

    Sintered uranium oxide and mixed oxide pellets are extensively used as nuclear reactor fuel. The shape of the fuel-pellets influence greatly their in-reactor performance. Fuel pellets of various shapes were prepared in Radiometallurgy Division to study their fabricability and in-reactor performance. This paper presents the experience in fabricating these fuel pellets. (author)

  15. Fuel Fabrication Capability Research and Development Plan

    Energy Technology Data Exchange (ETDEWEB)

    Senor, David J.; Burkes, Douglas

    2014-04-17

    is between now and 2016 when the candidate processes are down-selected in preparation for the MP-1, FSP-1, and MP-2 plate manufacturing campaigns. A number of key risks identified by the FFC are discussed in this plan, with recommended mitigating actions for those activities within FFC, and identification of risks that are impacted by activities in other areas of the Convert Program. The R&D Plan does not include discussion of FFC initiatives related to production-scale manufacturing of fuel (e.g., establishment of the Pilot Line Production Facility), rather, the goal of this plan is to document the R&D activities needed ultimately to enable high-quality and cost-effective production of the fuel by the commercial fuel fabricator. The intent is for this R&D Plan to be a living document that will be reviewed and updated on a regular basis (e.g., annually) to ensure that FFC R&D activities remain properly aligned to the needs of the Convert Program. This version of the R&D Plan represents the first annual review and revision.

  16. Experiences in the fabrication of aluminium clad metallic uranium fuel

    International Nuclear Information System (INIS)

    Vijayaraghavan, R.

    1989-01-01

    With a view to achieve self-sufficiency and self-reliance in the fabrication of metallic natural uranium fuel, a full fledged fuel fabrication facility was set up in 1958. Based on the then technical information available and the development work carried out, a flow-sheet for the fabrication of metallic uranium fuel, starting from uranium ingots, was worked out and the first fuel element was successfully fabricated in June 1959. More than half the first charge for the initial criticality of CIRUS, a 40 MWt research reactor at Trombay, was fabricated and supplied. Since then, this facility has been regularly catering to the replacement fuel and component requirements for CIRUS. The fuel for Dhruva, a 100 MWt research reactor at Trombay, is in the form of a cluster consisting of 7 fuel pins as compared to the rigid single fuel element for CIRUS. The fabrication process chosen for making this fuel is more or less on the same lines as that has been followed for CIRUS fuel. However, because of the smaller diameter of uranium metal rod, higher length to diameter ratio, configurations of fins on aluminium sheaths, cluster assembly etc., extensive development work was required to be undertaken for optimising various production parameters. Several prototype fuel clusters of different designs were made and subjected to rigorous out-of-pile and in-pile testing. Based on the reliable satisfactory performance, type II-B SPT cluster design was finally frozen for production. Stringent quality control is of prime importance in ensuring good performance of the fuel in the reactor. Hence, appropriate quality control measures have been adopted at various stages of fuel fabrication to ensure conformance with the specifications. (author) 9 refs., 11 figs., 1 tab

  17. ATLAS facility fabrication and assembly

    CERN Document Server

    Ballard, E O; Davis, H A; Nielsen, K E; Parker, G V; Parsons, W M

    2001-01-01

    Summary form only given. Atlas is a pulsed-power facility recently completed at Los Alamos National Laboratory to drive hydrodynamic experiments. This new generation pulsed-power machine consists of a radial array of 24, 240-kV Marx modules and transmission lines supplying current to the load region at the machine center. The transmission lines, powered by the Marx modules, consist of cable headers, load protection switches and tri-plates interfacing to the center transition section through detachable current joints. A conical power-flow-channel attaches to the transition section providing an elevated interface to attach the experimental loads for diagnostic access. Fabrication and assembly of all components for the Atlas machine was completed in August 2000. The machine has also progressed through a test phase where the Marx module/transmission line units were fired, individually, into a test load. Progression continued with eight and sixteen lines being fired. Subsequently, an overall machine test was condu...

  18. Environmental concerns in regarding a materials test reactor fuel fabrication facility at the Nuclear and Energy Research Institute - IPEN

    International Nuclear Information System (INIS)

    Santos, Glaucia R.T.; Durazzo, Michelangelo; Carvalho, Elita F.U.; Riella, Humberto G.

    2008-01-01

    The aim of the industrial activities success, front to a more and more informed and demanding society and to a more and more competitive market demands an environmental administration policy which doesn't limit itself to assist the legislation but anticipate and prevent, in a responsible way, possible damages to the environment. One of the main programs of the Institute of Energetic and Nuclear Research of the national Commission of Nuclear Energy located in Brazil, through the Center of Nuclear Fuel -CCN- is to manufacture MTR-type fuel elements using low-enrichment uranium (20 wt % 235 U), to supply its IEA-R1 research reactor. Integrated in this program, this work aims at well developing and assuring a methodology to implant an environment, health and safety policy, foreseeing its management with the use of detailed data reports and through the adoption of new tools for improving the management, in order to fulfil the applicable legislation and accomplish all the environmental, operational and works aspects. The applied methodology for the effluents management comprises different aspects, including the specific environmental legislation of a country, main available effluents treatment techniques, process flow analyses from raw materials and intakes to products, generated effluents, residuals and emissions. Data collections were accomplished for points gathering and tests characterization, classification and compatibility of the generated effluents and their eventual environmental impacts.This study aims to implant the Sustainability Concept in order to guarantee access to financial resources, allowing cost reduction, maximizing long-term profits, preventing and reducing environmental accident risks and stimulating both the attraction and the keeping of a motivated manpower. Work on this project has already started and, even though many technical actions have not still ended, the results have being extremely valuable. These results can already give to CCN

  19. Update On Monolithic Fuel Fabrication Development

    Energy Technology Data Exchange (ETDEWEB)

    C. R Clark; J. M. Wight; G. C. Knighton; G. A. Moore; J. F. Jue

    2005-11-01

    Efforts to develop a viable monolithic research reactor fuel plate have continued at Idaho National Laboratory. These efforts have concentrated on both fabrication process refinement and scale-up to produce full sized fuel plates. Advancements have been made in the production of U-Mo foil including full sized foils. Progress has also been made in the friction stir welding and transient liquid phase bonding fabrication processes resulting in better bonding, more stable processes and the ability to fabricate larger fuel plates.

  20. Hot Fuel Examination Facility (HFEF)

    Data.gov (United States)

    Federal Laboratory Consortium — The Hot Fuel Examination Facility (HFEF) is one of the largest hot cells dedicated to radioactive materials research at Idaho National Laboratory (INL). The nation's...

  1. Regulations concerning the fabricating business of nuclear fuel materials

    International Nuclear Information System (INIS)

    1985-01-01

    In the Law for the Regulations of Nuclear Source Material, Nuclear Fuel Material and Reactors, the regulations have all been revised on the fabrication business of nuclear fuel materials. The revised regulations are given : application for permission of the fabrication business, application for permission of the alteration, application for approval of the design and the construction methods, application for approval of the alteration, application for the facilities inspection, facilities inspection, recordings, entry limitations etc. for controlled areas, measures concerning exposure radiation doses etc., operation of the fabrication facilities, transport within the site of the business, storage, disposal within the site of the business, security regulations, designation etc. of the licensed engineer of nuclear fuels, collection of reports, etc. (Mori, K.)

  2. 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)

  3. Fuels and Lubricants Facility

    Data.gov (United States)

    Federal Laboratory Consortium — Modern naval aircraft and turbine-powered craft require reliable and high-quality fuels and lubricants to satisfy the demands imposed upon them for top performance...

  4. 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.

  5. UPDATE ON MONOLITHIC FUEL FABRICATION METHODS

    Energy Technology Data Exchange (ETDEWEB)

    C. R. Clark; J. F. Jue; G. A. Moore; N. P. Hallinan; B. H. Park; D. E. Burkes

    2006-10-01

    Efforts to develop a viable monolithic research reactor fuel plate have continued at Idaho National Laboratory. These efforts have concentrated on both fabrication process refinement and scale-up to produce full sized fuel plates. Progress at INL has led to fabrication of hot isostatic pressed uranium-molybdenum bearing monolithic fuel plates. These miniplates are part of the RERTR-8 miniplate irradiation test. Further progress has also been made on friction stir weld processing which has been used to fabricate full size fuel plates which will be irradiated in the ATR and OSIRIS reactors.

  6. 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

  7. 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

  8. Development of fabrication technology for CANDU advanced fuel -Development of the advanced CANDU technology-

    International Nuclear Information System (INIS)

    Choi, Chang Beom; Kim, Hyeong Soo; Kim, Sang Won; Seok, Ho Cheon; Shim, Ki Seop; Byeon, Taek Sang; Jang, Ho Il; Kim, Sang Sik; Choi, Il Kwon; Cho, Dae Sik; Sheo, Seung Won; Lee, Soo Cheol; Kim, Yoon Hoi; Park, Choon Ho; Jeong, Seong Hoon; Kang, Myeong Soo; Park, Kwang Seok; Oh, Hee Kwan; Jang, Hong Seop; Kim, Yang Kon; Shin, Won Cheol; Lee, Do Yeon; Beon, Yeong Cheol; Lee, Sang Uh; Sho, Dal Yeong; Han, Eun Deok; Kim, Bong Soon; Park, Cheol Joo; Lee, Kyu Am; Yeon, Jin Yeong; Choi, Seok Mo; Shon, Jae Moon

    1994-07-01

    The present study is to develop the advanced CANDU fuel fabrication technologies by means of applying the R and D results and experiences gained from localization of mass production technologies of CANDU fuels. The annual portion of this year study includes following: 1. manufacturing of demo-fuel bundles for out-of-pile testing 2. development of technologies for the fabrication and inspection of advanced fuels 3. design and munufacturing of fuel fabrication facilities 4. performance of fundamental studies related to the development of advanced fuel fabrication technology

  9. The development of PHWR fuel fabrication in Korea

    International Nuclear Information System (INIS)

    Kyung-soo Suh; Myung-seung Yang; Dong-hoon Kim; Chang-saeng Rim

    1987-01-01

    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-irrradiation 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 marking use of experience accumulated during reaearch. A quality assurance program was also establisned, and quality inspection technology was reviewed and improved to fit the mass production. After mid-1987 when the expansion of fuel manufacturing capacities and establishment of quality assurance system are to be completed, all the nuclearr fuel for the Wolsung power reactor will be supplied by KAERI. This paper deals with the wherewithals and development experience so far obtained with the design and fabrication of the Korean PHWR fuel. (author)

  10. 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

  11. Chemical and Radiochemical Composition of Thermally Stabilized Plutonium Oxide from the Plutonium Finishing Plant Considered as Alternate Feedstock for the Mixed Oxide Fuel Fabrication Facility

    International Nuclear Information System (INIS)

    Tingey, Joel M.; Jones, Susan A.

    2005-01-01

    Eighteen plutonium oxide samples originating from the Plutonium Finishing Plant (PFP) on the Hanford Site were analyzed to provide additional data on the suitability of PFP thermally stabilized plutonium oxides and Rocky Flats oxides as alternate feedstock to the Mixed Oxide Fuel Fabrication Facility (MFFF). Radiochemical and chemical analyses were performed on fusions, acid leaches, and water leaches of these 18 samples. The results from these destructive analyses were compared with nondestructive analyses (NDA) performed at PFP and the acceptance criteria for the alternate feedstock. The plutonium oxide materials considered as alternate feedstock at Hanford originated from several different sources including Rocky Flats oxide, scrap from the Remote Mechanical C-Line (RMC) and the Plutonium Reclamation Facility (PRF), and materials from other plutonium conversion processes at Hanford. These materials were received at PFP as metals, oxides, and solutions. All of the material considered as alternate feedstock was converted to PuO2 and thermally stabilized by heating the PuO2 powder at 950 C in an oxidizing environment. The two samples from solutions were converted to PuO2 by precipitation with Mg(OH)2. The 18 plutonium oxide samples were grouped into four categories based on their origin. The Rocky Flats oxide was divided into two categories, low- and high-chloride Rocky Flats oxides. The other two categories were PRF/RMC scrap oxides, which included scrap from both process lines and oxides produced from solutions. The two solution samples came from samples that were being tested at Pacific Northwest National Laboratory because all of the plutonium oxide from solutions at PFP had already been processed and placed in 3013 containers. These samples originated at the PFP and are from plutonium nitrate product and double-pass filtrate solutions after they had been thermally stabilized. The other 16 samples originated from thermal stabilization batches before canning at

  12. Progress on KMRR fuel fabrication

    International Nuclear Information System (INIS)

    Kuk, I.H.; Lee, J.B.; Rhee, C.K.; Kim, K.W.

    1991-01-01

    In order to increase the practical applicability of powder heat-treatment in KMRR fuel fabrication, efforts were made to reduce the critical size. Primary U 3 Si 2 particle size was reviewed in terms of cooling rate. Temperature dependence of peritectoid reaction was reviewed as well. (1) Cooling rate of U 3 Si molten alloy was calculated by ADINA program. In practice, particle size of the primary U 3 Si 2 varies radially. U 3 Si 2 size increases as it goes deeper from the surface. As cooling rate increases, primary U 3 Si 2 size decreases. (2) Peritectoid reaction occurs in two unique groups of temperature; one is below 790 C where β-U and U 3 Si 2 reaction occurs, and the other above 790 C where γ-U and U 3 Si 2 reaction occurs. 780 C is most completely reacting temperature in β-U region, and 810 C is so in γ-U region. Reaction is completed more perfectly in γ-U region than in β-U region. 810 C is found to be the optimized heat-treatment temperature, but it is desirable not to approach to 790 C in heat-treatment. (3) The critical powder size in powder heat-treatment is dependent on the primary U 3 Si 2 particle size. The smaller the primary U 3 Si 2 particle size, the smaller the critical particle size of the powder. At present, the primary U 3 Si 2 particle size can be reduced to 3∝5 μm at 4∝5 mm deep from surface in Cu mold. This may be reduced further by rapid solidification process. (orig.)

  13. Reprocessing and fuel fabrication systems

    International Nuclear Information System (INIS)

    Field, F.R.; Tooper, F.E.

    1978-01-01

    The study of alternative fuel cycles was initiated to identify a fuel cycle with inherent technical resistance to proliferation; however, other key features such as resource use, cost, and development status are major elements in a sound fuel cycle strategy if there is no significant difference in proliferation resistance. Special fuel reprocessing techniques such as coprocessing or spiking provide limited resistance to diversion. The nuclear fuel cycle system that will be most effective may be more dependent on the institutional agreements that can be implemented to supplement the technical controls of fuel cycle materials

  14. Fuel Handling Facility Description Document

    International Nuclear Information System (INIS)

    M.A. LaFountain

    2005-01-01

    The purpose of the facility description document (FDD) is to establish the requirements and their bases that drive the design of the Fuel Handling Facility (FHF) to allow the design effort to proceed to license application. This FDD is a living document that will be revised at strategic points as the design matures. It identifies the requirements and describes the facility design as it currently exists, with emphasis on design attributes provided to meet the requirements. This FDD was developed as an engineering tool for design control. Accordingly, the primary audience and users are design engineers. It leads the design process with regard to the flow down of upper tier requirements onto the facility. Knowledge of these requirements is essential to performing the design process. It trails the design with regard to the description of the facility. This description is a reflection of the results of the design process to date

  15. IFR fuel cycle demonstration in the EBR-II Fuel Cycle Facility

    International Nuclear Information System (INIS)

    Lineberry, M.J.; Phipps, R.D.; Rigg, R.H.; Benedict, R.W.; Carnes, M.D.; Herceg, J.E.; Holtz, R.E.

    1991-01-01

    The next major milestone of the IFR (Integral Fast Reactor) program is engineering-scale demonstration of the pyroprocess fuel cycle. The EBR-II Fuel Cycle Facility has just entered a startup phase which includes completion of facility modifications, and installation and cold checkout of process equipment. This paper reviews the design and construction of the facility, the design and fabrication of the process equipment, and the schedule and initial plan for its operation. (author)

  16. 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.)

  17. 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

  18. Fuel conditioning facility material accountancy

    International Nuclear Information System (INIS)

    Yacout, A.M.; Bucher, R.G.; Orechwa, Y.

    1995-01-01

    The operation of the Fuel conditioning Facility (FCF) is based on the electrometallurgical processing of spent metallic reactor fuel. It differs significantly, therefore, from traditional PUREX process facilities in both processing technology and safeguards implications. For example, the fissile material is processed in FCF only in batches and is transferred within the facility only as solid, well-characterized items; there are no liquid steams containing fissile material within the facility, nor entering or leaving the facility. The analysis of a single batch lends itself also to an analytical relationship between the safeguards criteria, such as alarm limit, detection probability, and maximum significant amount of fissile material, and the accounting system's performance, as it is reflected in the variance associated with the estimate of the inventory difference. This relation, together with the sensitivity of the inventory difference to the uncertainties in the measurements, allows a thorough evaluation of the power of the accounting system. The system for the accountancy of the fissile material in the FCF has two main components: a system to gather and store information during the operation of the facility, and a system to interpret this information with regard to meeting safeguards criteria. These are described and the precision of the inventory closure over one batch evaluated

  19. Nuclear facilities in the fuel cycle - Federal Republic of Germany

    International Nuclear Information System (INIS)

    1984-11-01

    The map of the Federal Republic of Germany (scale 1:800000) presents the sites of nuclear facilities. These include uranium mines and mills, fuel element fabrication and reprocessing plants, nuclear power plants (in operation or under construction), fuel transfer storage, and ultimate storage facilities. As a geographical extension, the PAMELA vitrification plant in Belgium is also shown. The map is of November 1984. (UA) [de

  20. Development of technology of high density LEU dispersion fuel fabrication

    International Nuclear Information System (INIS)

    Wiencek, T.; Totev, T.

    2007-01-01

    Advanced Materials Fabrication Facilities at Argonne National Laboratory have been involved in development of LEU dispersion fuel for research and test reactors from the beginning of RERTR program. This paper presents development of technology of high density LEU dispersion fuel fabrication for full size plate type fuel elements. A brief description of Advanced Materials Fabrication Facilities where development of the technology was carried out is given. A flow diagram of the manufacturing process is presented. U-Mo powder was manufactured by the rotating electrode process. The atomization produced a U-Mo alloy powder with a relatively uniform size distribution and a nearly spherical shape. Test plates were fabricated using tungsten and depleted U-7 wt.% Mo alloy, 4043 Al and Al-2 wt% Si matrices with Al 6061 aluminum alloy for the cladding. During the development of the technology of manufacturing of full size high density LEU dispersion fuel plates special attention was paid to meet the required homogeneity, bonding, dimensions, fuel out of zone and other mechanical characteristics of the plates.

  1. 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.)

  2. Design and construction of the Fuels and Materials Examination Facility

    International Nuclear Information System (INIS)

    Burgess, C.A.

    1979-01-01

    Final design is more than 85 percent complete on the Fuels and Materials Examination Facility, the facility for post-irradiation examination of the fuels and materials tests irradiated in the FFTF and for fuel process development, experimental test pin fabrication and supporting storage, assay, and analytical chemistry functions. The overall facility is generally described with specific information given on some of the design features. Construction has been initiated and more than 10% of the construction contracts have been awarded on a fixed price basis

  3. Fabrication and performance testing of CANDU mixed-oxide fuel

    International Nuclear Information System (INIS)

    Dimayuga, F.C.; Floyd, M.R.; Cox, D.S.

    2000-01-01

    AECL's mixed-oxide fuel fabrication activities are performed in the Recycle Fuel Fabrication Laboratories (RFFL) at the Chalk River Laboratories. Since the start-up of the RFFL in the mid-1970s, several fabrication campaigns have been conducted in the facility, producing various types of mixed-oxide (MOX) fuel, which were used for both irradiation and physics testing. More recently, CANDU fuel bundles containing 0.5 w-t % plutonium in natural uranium, produced in the RFFL, were successfully irradiated in the NRU reactor at powers up to 65 kW/m and to burnups ranging from 13 to 23 MW·d/kg HE. Two of the bundles had power histories that bound the normal powers and burnups of natural UO 2 CANDU fuel ( 2 fuel. Significantly more grain growth was observed than that expected for UO 2 fuel; however, this increase in grain growth had no effect on the overall performance of the fuel. Two other bundles operated to extended burnups of 19 to 23 MW·d/kg HE. Burnup extension above 15 MW·d/kg HE only had a small effect on FGR. (author)

  4. Experimental Fuels Facility Re-categorization Based on Facility Segmentation

    Energy Technology Data Exchange (ETDEWEB)

    Reiss, Troy P.; Andrus, Jason

    2016-07-01

    The Experimental Fuels Facility (EFF) (MFC-794) at the Materials and Fuels Complex (MFC) located on the Idaho National Laboratory (INL) Site was originally constructed to provide controlled-access, indoor storage for radiological contaminated equipment. Use of the facility was expanded to provide a controlled environment for repairing contaminated equipment and characterizing, repackaging, and treating waste. The EFF facility is also used for research and development services, including fuel fabrication. EFF was originally categorized as a LTHC-3 radiological facility based on facility operations and facility radiological inventories. Newly planned program activities identified the need to receive quantities of fissionable materials in excess of the single parameter subcritical limit in ANSI/ANS-8.1, “Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors” (identified as “criticality list” quantities in DOE-STD-1027-92, “Hazard Categorization and Accident Analysis Techniques for Compliance with DOE Order 5480.23, Nuclear Safety Analysis Reports,” Attachment 1, Table A.1). Since the proposed inventory of fissionable materials inside EFF may be greater than the single parameter sub-critical limit of 700 g of U-235 equivalent, the initial re-categorization is Hazard Category (HC) 2 based upon a potential criticality hazard. This paper details the facility hazard categorization performed for the EFF. The categorization was necessary to determine (a) the need for further safety analysis in accordance with LWP-10802, “INL Facility Categorization,” and (b) compliance with 10 Code of Federal Regulations (CFR) 830, Subpart B, “Safety Basis Requirements.” Based on the segmentation argument presented in this paper, the final hazard categorization for the facility is LTHC-3. Department of Energy Idaho (DOE-ID) approval of the final hazard categorization determined by this hazard assessment document (HAD) was required per the

  5. Fabrication of cermet fuel for fast reactor

    International Nuclear Information System (INIS)

    Mishra, Sudhir; Kumar, Arun; Kutty, T.R.G.; Kamath, H.S.

    2011-01-01

    Mixed oxide (MOX) (U,Pu)O 2 , and metallic (U,Pu ,Zr) fuels are considered promising fuels for the fast reactor. The fuel cycle of MOX is well established. The advantages of the oxide fuel are its easy fabricability, good performance in the reactor and a well established reprocessing technology. However the problems lie in low thermal conductivity , low density of the fuel leading to low breeding ratio and consequently longer doubling time. The metallic fuel has the advantages of high thermal conductivity, higher metal density and higher coefficient of linear expansion. The higher coefficient of linear expansion is good from the safety consideration (negative reactivity factor). Because of higher metal density it offers highest breeding ratio and shortest doubling time. Metallic fuel disadvantages comprise large swelling at high burnup, fuel cladding interaction and lower margin between operating and melting temperature. The optimal solution may lie in cermet fuel (U, PuO 2 ), where PuO 2 is dispersed in U metal matrix and combines the favorable features of both the fuel types. The advantages of this fuel include high thermal conductivity, larger margin between melting and operating temperature, ability to retain fission product etc. The matrix being of high density metal the advantage of high breeding ratio is also maintained. In this report some results of fabrication of cermet pellet comprising of UO 2 /PuO 2 dispersed in U metal powder through classical powder metallurgy route and characterization are presented. (author)

  6. Fuel-cycle facilities: preliminary safety and environmental information document. Volume VII

    Energy Technology Data Exchange (ETDEWEB)

    1980-01-01

    Information is presented concerning the mining and milling of uranium and thorium; uranium hexafluoride conversion; enrichment; fuel fabrication; reprocessing; storage options; waste disposal options; transportation; heavy-water-production facilities; and international fuel service centers.

  7. Fuel-cycle facilities: preliminary safety and environmental information document. Volume VII

    International Nuclear Information System (INIS)

    1980-01-01

    Information is presented concerning the mining and milling of uranium and thorium; uranium hexafluoride conversion; enrichment; fuel fabrication; reprocessing; storage options; waste disposal options; transportation; heavy-water-production facilities; and international fuel service centers

  8. 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.

  9. Fuel Fabrication and Nuclear Reactors

    International Nuclear Information System (INIS)

    Karpius, Peter Joseph

    2017-01-01

    The uranium from the enrichment plant is still in the form of UF 6 . UF 6 is not suitable for use in a reactor due to its highly corrosive chemistry as well as its phase diagram. UF 6 is converted into UO 2 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.

  10. Development of metallic fuel fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Young Ho; Lee, Chong Yak; Lee, Myung Ho and others

    1999-03-01

    With the vacuum melting and casting of the U-10wt%Zr alloy which is metallic fuel for liquid metal fast breeder reactor, we studied the microstructure of the alloy and the parameters of the melting and casting for the fuel rods. Internal defects of the U-10wt%Zr fuel by gravity casting, were inspected by non-destructive test. U-10wt%Zr alloy has been prepared for the thermal stability test in order to estimate the decomposition of the lamellar structure with relation to swelling under irradiation condition. (author)

  11. Fuel reprocessing/fabrication interface

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  12. A facile method to fabricate superhydrophobic cotton fabrics

    Science.gov (United States)

    Zhang, Ming; Wang, Shuliang; Wang, Chengyu; Li, Jian

    2012-11-01

    A facile and novel method for fabricating superhydrophobic cotton fabrics is described in the present work. The superhydrophobic surface has been prepared by utilizing cationic poly (dimethyldiallylammonium chloride) and silica particles together with subsequent modification of (heptadecafluoro-1,1,2,2-tetradecyl) trimethoxysilane. The size distribution of silica particles was measured by Particle Size Analyzer. The cotton textiles before and after treatment were characterized by using scanning electron microscope (SEM) and X-ray photoelectron spectroscopy (XPS). The wetting behavior of cotton samples was investigated by water contact angle measurement. Moreover, the superhydrophobic durability of coated cotton textiles has been evaluated by exposure, immersion and washing tests. The results show that the treated cotton fabrics exhibited excellent chemical stability and outstanding non-wettability with the WCA of 155 ± 2°, which offers an opportunity to accelerate the large-scale production of superhydrophobic textiles materials for new industrial applications.

  13. 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.)

  14. Process development and fabrication for sphere-pac fuel rods

    International Nuclear Information System (INIS)

    Welty, R.K.; Campbell, M.H.

    1981-06-01

    Uranium fuel rods containing sphere-pac fuel have been fabricated for in-reactor tests and demonstrations. A process for the development, qualification, and fabrication of acceptable sphere-pac fuel rods is described. Special equipment to control fuel contamination with moisture or air and the equipment layout needed for rod fabrication is described and tests for assuring the uniformity of the fuel column are discussed. Fuel retainers required for sphere-pac fuel column stability and instrumentation to measure fuel column smear density are described. Results of sphere-pac fuel rod fabrication campaigns are reviewed and recommended improvements for high throughput production are noted

  15. Microencapsulation and fabrication of fuel pellets for inertial confinement fusion

    International Nuclear Information System (INIS)

    Nolen, R.L. Jr.; Kool, L.B.

    1981-01-01

    Various microencapsulation techniques were evaluated for fabrication of thermonuclear fuel pellets for use in existing experimental facilities studying inertial confinement fusion and in future fusion-power reactors. Coacervation, spray drying, in situ polymerization, and physical microencapsulation methods were employed. Highly spherical, hollow polymeric shells were fabricated ranging in size from 20 to 7000 micron. In situ polymerization microencapsulation with poly(methyl methacrylate) provided large shells, but problems with local wall defects still must be solved. Extension to other polymeric systems met with limited success. Requirements for inertial confinement fusion targets are described, as are the methods that were used

  16. Fast facility spent-fuel and waste assay instrument. [Fluorinel Dissolution and Fuel Storage (FAST) Facility

    Energy Technology Data Exchange (ETDEWEB)

    Eccleston, G.W.; Johnson, S.S.; Menlove, H.O.; Van Lyssel, T.; Black, D.; Carlson, B.; Decker, L.; Echo, M.W.

    1983-01-01

    A delayed-neutron assay instrument was installed in the Fluorinel Dissolution and Fuel Storage Facility at Idaho National Engineering Laboratory. The dual-assay instrument is designed to measure both spent fuel and waste solids that are produced from fuel processing. A set of waste standards, fabricated by Los Alamos using uranium supplied by Exxon Nuclear Idaho Company, was used to calibrate the small-sample assay region of the instrument. Performance testing was completed before installation of the instrument to determine the effects of uranium enrichment, hydrogenous materials, and neutron poisons on assays. The unit was designed to measure high-enriched uranium samples in the presence of large neutron backgrounds. Measurements indicate that the system can assay low-enriched uranium samples with moderate backgrounds if calibrated with proper standards.

  17. Integrated five station nondestructive assay system for the support of decontamination and decommissioning of a former plutonium mixed oxide fuel fabrication facility

    International Nuclear Information System (INIS)

    Caldwell, J.T.; Bieri, J.M.; Hastings, R.D.; Horton, W.S.; Kuckertz, T.H.; Kunz, W.E.; Plettenberg, K.; Smith, L.D.

    1990-01-01

    The goal of a safe, efficient and economic decontamination and decommissioning of plutonium facilities can be greatly enhanced through the intelligent use of an integrated system of nondestructive assay equipment. We have designed and fabricated such a system utilizing five separate NDA stations integrated through a single data acquisition and management personal computer-based controller. The initial station utilizes a passive neutron measurement to determine item Pu inventory to the 0.1 gm level prior to insertion into the decontamination cell. A large active neutron station integrated into the cell is used to measure decontamination effectiveness at the 10 nci/gm level. Cell Pu buildup at critical points is monitored with passive neutron detectors. An active neutron station having better than 1 mg Pu assay sensitivity is used to quantify final compacted waste pucks outside the cell. Bulk Pu in various forms and isotopic enrichments is quantified in a combined passive neutron coincidence and high resolution gamma ray spectrometer station outside the cell. Item control and Pu inventory are managed with bar code labeling and a station integrating algorithm. Overall economy is achieved by multiple station use of the same expensive hardware such as the neutron generator

  18. MOX fuel fabrication: Technical and industrial developments

    International Nuclear Information System (INIS)

    Lebastard, G.; Bairiot, H.

    1990-01-01

    The plutonium available in the near future is generally estimated rather precisely on the basis of the reprocessing contracts and the performance of the reprocessing plants. A few years ago, decision makers were convinced that a significant share of this fissile material would be used as the feed material for fast breeder reactors (FBRs) or other advanced reactors. The facts today are that large reprocessing plants are coming into commercial operations: UP3 and soon UP2-800 and THORP, but that FBR deployment is delayed worldwide. As a consequence, large quantities of plutonium will be recycled in light water reactors as mixed oxide (MOX) fuels. MOX fuel technology has been properly demonstrated in the past 25 years. All specific problems have been addressed, efficient fabrication processes and engineering background have been implemented to a level of maturity which makes MOX fuel behaving as well as Uranium fuel. The paper concentrates on todays MOX fabrication expertise and presents the technical and industrial developments prepared by the MOX fuel fabrication industry for this last decade of the century

  19. 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)

  20. SuperCDMS Underground Detector Fabrication Facility

    Energy Technology Data Exchange (ETDEWEB)

    Platt, M.; Mahapatra, R.; Bunker, Raymond A.; Orrell, John L.

    2018-03-01

    The SuperCDMS SNOLAB dark matter experiment processes Ge and Si crystals into fully tested phonon and ionization detectors at surface fabrication and test facilities. If not mitigated, it is anticipated that trace-level production of radioisotopes in the crystals due to exposure to cosmic rays at (or above) sea level will result in the dominant source of background events in future dark matter searches using the current SuperCDMS detector technology. Fabrication and testing of detectors in underground facilities shielded from cosmic radiation is one way to directly reduce production of trace levels of radioisotopes, thereby improving experimental sensitivity for the discovery of dark matter beyond the level of the current experiment. In this report, we investigate the cost and feasibility to establish a complete detector fabrication processing chain in an underground location to mitigate cosmogenic activation of the Ge and Si detector substrates. For a specific and concrete evaluation, we explore options for such a facility located at SNOLAB, an underground laboratory in Sudbury, Canada hosting the current and future experimental phases of SuperCDMS.

  1. Assesment On The Possibility To Modify Fabrication Equipment For Fabrication Of HWR And LWR Fuel Elements

    International Nuclear Information System (INIS)

    Tri-Yulianto

    1996-01-01

    Based on TOR BATAN for PELITA VI. On of BATAN program in the fuel element production technology section is the acquisition of the fuel element fabrication technology for research reactor as well as power reactor. The acquisition can be achieved using different strategies, e.g. by utilizing the facility owned for research and development of the technology desired or by transferring the technology directly from the source. With regards to the above, PEBN through its facility in BEBE has started the acquisition of the fuel element fabrication technology for power reactor by developing the existing equipment initially designed to fabricate HWR Cinere fuel element. The development, by way of modifying the equipment, is intended for the production of HWR (Candu) and LWR (PWR and BWR) fuel elements. To achieve above objective, at the early stage of activity, an assesment on the fabrication equipment for pelletizing, component production and assembly. The assesment was made by comparing the shape and the size of the existing fuel element with those used in the operating reactors such as Candu reactors, PWR and BWR. Equipment having the potential to be modified for the production of HWR fuel elements are as followed: For the pelletizing equipment, the punch and dies can be used of the pressing machine for making green pellet can be modified so that different sizes of punch and dies can be used, depending upon the size of the HWR and LWR pellets. The equipment for component production has good potential for modification to produce the HWR Candu fuel element, which has similar shape and size with those of the existing fuel element, while the possibility of producing the LWR fuel element component is small because only a limited number of the required component can be made with the existing equipment. The assembly equipment has similar situation whit that of the component production, that is, to assemble the HWR fuel element modification of few assembly units very probable

  2. Fabrication of Sintered Annular Fuel Pellet

    International Nuclear Information System (INIS)

    Rhee, Young Woo; Kim, Dong Joo; Kim, Jong Hun; Yang, Jae Ho; Kim, Keon Sik; Kang, Ki Won; Song, Kun Woo

    2010-01-01

    A dual cooled annular fuel has been seriously considered as a favorable option for uprating the power density of a Pressurized Water Reactor fuel assembly. An annular fuel has a geometrically inherent advantage such as an increased heat transfer area and a thin pellet thickness. It results in a lot of advantages from the point of a fuel safety and its economy. In order to actualize the dual cooled fuel, an essential element is the annular pellet with precisely controlled diametric tolerance. However, the unique shape of annular fuel pellet causes challenging difficulties to satisfy a diametric tolerance specification. Because of an inhomogeneous green density distribution along the compact height, an hour-glassing usually occurred in a sintered cylindrical PWR fuel pellet fabricated by a conventional doubleacting press. Thus, a sintered pellet usually undergoes a centerless grinding process in order to secure a pellet's specifications. In the case of an annular pellet fabrication using a conventional double-acting press, the same hour-glass shape would probably occur. The outer diameter tolerance of an annular pellet can be controlled easily similar to that of a conventional cylindrical PWR pellet through a centerless grinding. However, it appears not to be simple in the case of an inner surface grinding. It would be the best way to satisfy the specifications for the inner diameter in an as-fabricated pellet. In the present study, we are trying to find a way to minimize the diametric tolerance of the sintered annular pellet without inner surface grinding. This paper deals with a new approach that we have tried to reduce the diametric tolerance of the sintered annular pellet

  3. SAF-BRET-FMEF: a developmental LMR fuel cycle facility

    International Nuclear Information System (INIS)

    Stradley, J.G.; Yook, H.R.; Gerber, E.W.; Lerch, R.E.; Rice, L.H.

    1985-01-01

    The SAF-BRET-FMEF complex represents a versatile fuel cycle facility for processing LMR fuel. While originally conceived for processing FFTF and CRBRP fuel, it represents a facility where LMR fuel from the first generation of innovative LMRs could be processed. The cost of transporting fuel from the LMR to the Hanford site would have to be assessed when the LMR site is identified. The throughput of BRET was set at 15 MTHM/yr during conceptual design of the facility, a rate which was adequate to process all of the fuel from FFTF and fuel and blanket material from CRBRP. The design is currently being reevaluated to see if BRET could be expanded to approx.35 MTHM/yr to process fuel and blanket material from approx.1300 MWe generating capacity of the innovative LMRs. This expanded throughput is possible by designing the equipment for an instantaneous throughput of 0.2 MTHM/d, and by selected additional modifications to the facility (e.g., expansion of shipping and receiving area, and addition of a second entry tunnel transporter), and by the fact that the LMR fuel assemblies contain more fuel than the FFTF assemblies (therefore, fewer assemblies must be handled for the same throughput). The estimated cost of such an expansion is also being assessed. As stated previously, the throughput of SAF and Fuel Assembly could be made to support typical LMRs at little additional cost. The throughput could be increased to support the fuel fabrication requirements for 1300 MWe generating capacity of the innovative LMRs. This added capacity may be achieved by increasing the number of operating shifts, and is affected by variables such as fuel design, fuel enrichment, and plutonium isotopic composition

  4. Construction and engineering report for advanced nuclear fuel development facility

    Energy Technology Data Exchange (ETDEWEB)

    Cho, S. W.; Park, J. S.; Kwon, S.J.; Lee, K. W.; Kim, I. J.; Yu, C. H

    2003-09-01

    The design and construction of the fuel technology development facility was aimed to accommodate general nuclear fuel research and development for the HANARO fuel fabrication and advanced fuel researches. 1. Building size and room function 1) Building total area : approx. 3,618m{sup 2}, basement 1st floor, ground 3th floor 2) Room function : basement floor(machine room, electrical room, radioactive waste tank room), 1st floor(research reactor fuel fabrication facility, pyroprocess lab., metal fuel lab., nondestructive lab., pellet processing lab., access control room, sintering lab., etc), 2nd floor(thermal properties measurement lab., pellet characterization lab., powder analysis lab., microstructure analysis lab., etc), 3rd floor(AHU and ACU Room) 2. Special facility equipment 1) Environmental pollution protection equipment : ACU(2sets), 2) Emergency operating system : diesel generator(1set), 3) Nuclear material handle, storage and transport system : overhead crane(3sets), monorail hoist(1set), jib crane(2sets), tank(1set) 4) Air conditioning unit facility : AHU(3sets), packaged air conditioning unit(5sets), 5) Automatic control system and fire protection system : central control equipment(1set), lon device(1set), fire hose cabinet(3sets), fire pump(3sets) etc.

  5. History and present status of nuclear fuel fabrication in Japan

    International Nuclear Information System (INIS)

    Kiyosawa, Nobuhito

    1985-01-01

    Fabrication of the nuclear fuel for LWRs (PWR, BWR) represents an area where domestic production advanced foremost in the Japanese nuclear fuel cycle. Until the middle of the 1970s, part of the initial load was made up of imported fuels. From that time onward, however, all fuel needs for initial or reload, have been fulfilled by three domestic fuel fabricators; Japan Nuclear Fuel Co., Ltd. (JNF) and Mitsubishi Nuclear Fuel Co., Ltd (MNF) for both initial and reload fuel and Nuclear Fuel Industries, Ltd (NFl) for reload fuel. Basic research and development work on the LWR fuel fabrication technology began in the late part of 1950s in Japan among such companies as Toshiba Corp., Hitachi Ltd., Mitsubishi Metal Corp., Mitsubishi Atomic Power Industries Inc., Sumitomo Electric Industries Ltd., The Furukawa Electric Company Ltd., etc. and the expertise was built up thereafter through a series of irradiation test in the JPDR (Japan Power Demonstration Reactor) and fabrication of reloaded fuel of the reactor. As construction of commercial light water reactor got under way, JNF was established in 1967, the MNF in 1971 and NFl in 1972 respectively from those parent companies mentioned above setting up the framework for domestic production of LWR fuel. Those fuel fabricators introduced commercial scale fabrication technologies from abroad fuel fabricators and mixed up with the basic domestic technology nurtured by them theretofore and established the LWR fuel fabrication technology which meets the rigorous requirements for LWRs fuel in Japan. In accordance with growing domestic demands, the fabrication capacity was increased while utmost efforts were made to improve productivity, product quality and reliability to set up a system for mass producing LWR fuel of world highest quality. The leak rate of Japanese domestic fuel shows one of the lowest records among other foreign countries. The amount of fuel fabricated by three fabricators to date is approximately 6,900 tons in UO

  6. Fuel fabrication instrumentation and control system overview

    International Nuclear Information System (INIS)

    Bennett, D.W.; Fritz, R.L.

    1980-10-01

    A process instrumentation and control system is being developed for automated fabrication of breeder reactor fuel at the Hanford Engineering Development Laboratory (HEDL) in Richland, Washington. The basic elements of the control system are a centralized computer system linked to distributed local computers, which direct individual process applications. The control philosophy developed for the equipment automation program stresses system flexibility and inherent levels of redundant control capabilities. Four different control points have been developed for each unit process operation

  7. Utility industry evaluation of the metal fuel facility and metal fuel performance for liquid metal reactors

    International Nuclear Information System (INIS)

    Burstein, S.; Gibbons, J.P.; High, M.D.; O'Boyle, D.R.; Pickens, T.A.; Pilmer, D.F.; Tomonto, J.R.; Weinberg, C.J.

    1990-02-01

    A team of utility industry representatives evaluated the liquid metal reactor metal fuel process and facility conceptual design being developed by Argonne National Laboratory (ANL) under Department of Energy sponsorship. The utility team concluded that a highly competent ANL team was making impressive progress in developing high performance advanced metal fuel and an economic processing and fabrication technology. The utility team concluded that the potential benefits of advanced metal fuel justified the development program, but that, at this early stage, there are considerable uncertainties in predicting the net overall economic benefit of metal fuel. Specific comments and recommendations are provided as a contribution towards enhancing the development program. 6 refs

  8. Interpretation of the results from individual monitoring of workers at the Nuclear Fuel Fabrication Facility, Brazil; Interpretacao de resultados de monitoracao individual interna da Fabrica de Combustivel Nuclear - FCN

    Energy Technology Data Exchange (ETDEWEB)

    Castro, Marcelo Xavier de

    2005-07-01

    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

  9. Advanced high throughput MOX fuel fabrication technology and sustainable development

    International Nuclear Information System (INIS)

    Krellmann, Juergen

    2005-01-01

    The MELOX plant in the south of France together with the La Hague reprocessing plant, are part of the two industrial facilities in charge of closing the nuclear fuel cycle in France. Started up in 1995, MELOX has since accumulated a solid know-how in recycling plutonium recovered from spent uranium fuel into MOX: a fuel blend comprised of both uranium and plutonium oxides. Converting recovered Pu into a proliferation-resistant material that can readily be used to power a civil nuclear reactor, MOX fabrication offers a sustainable solution to safely take advantage of the plutonium's high energy content. Being the first large-capacity industrial facility dedicated to MOX fuel fabrication, MELOX distinguishes itself from the first generation MOX plants with high capacity (around 200 tHM versus around 40 tHM) and several unique operational features designed to improve productivity, reliability and flexibility while maintaining high safety standards. Providing an exemplary reference for high throughput MOX fabrication with 1,000 tHM produced since start-up, the unique process and technologies implemented at MELOX are currently inspiring other MOX plant construction projects (in Japan with the J-MOX plant, in the US and in Russia as part of the weapon-grade plutonium inventory reduction). Spurred by the growing international demand, MELOX has embarked upon an ambitious production development and diversification plan. Starting from an annual level of 100 tons of heavy metal (tHM), MELOX demonstrated production capacity is continuously increasing: MELOX is now aiming for a minimum of 140 tHM by the end of 2005, with the ultimate ambition of reaching the full capacity of the plant (around 200 tHM) in the near future. With regards to its activity, MELOX also remains deeply committed to sustainable development in a consolidated involvement within AREVA group. The French minister of Industry, on August 26th 2005, acknowledged the benefits of MOX fuel production at MELOX: 'In

  10. Consideration of Regulatory System for Fuel Cycle Facilities in Foreign Countries

    International Nuclear Information System (INIS)

    Ahn, Sang-Kyu; Ahn, Hyung-Joon; Oh, Kyu-Myung; Lee, Hoon-Joo

    2006-01-01

    Three units of fuel fabrication plants are operating now in Korea. These fuel fabrication plants produce fuels for PWR, CANDU and HANARO research reactor, respectively. To enhance our regulatory systems for fuel cycle facilities including fuel fabrication plants, the related system of foreign countries is surveyed. In this study, the overall regulatory systems of the United States of America, France, Germany, the United Kingdom and Japan are surveyed. The survey items consist of laws and regulations, regulatory organization, licensing procedure, documents to be submitted with a license application, regulatory inspection, and etc

  11. Licensed fuel facility status report. Volume 5, No. 2. Inventory difference data, July 1984-December 1984

    International Nuclear Information System (INIS)

    1985-10-01

    NRC is committed to the periodic publication of licensed fuel facilities' inventory difference data, following agency review of the information and completion of any related investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233

  12. Licensed fuel facility status report. Inventory difference data, July-December 1985. Volume 6, No. 2

    International Nuclear Information System (INIS)

    1986-08-01

    NRC is committed to the periodic publication of licensed fuel facilities' inventory difference data, following agency review of the information and completion of any related investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233

  13. Licensed fuel facility status report. Inventory difference data, January-June 1985. Volume 6, No. 1

    International Nuclear Information System (INIS)

    1986-02-01

    NRC is committed to the periodic publication of licensed fuel facilities' inventory difference data, following agency review of the information and completion of any related investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233

  14. Licensed fuel facility status report. Inventory difference data, July 1983-December 1983. Volume 4, No. 2

    International Nuclear Information System (INIS)

    1984-08-01

    NRC is committed to the periodic publication of licensed fuel facilities inventory difference data, following agency review of the information and completion of any related investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233

  15. Licensed fuel facility status report: Inventory difference data, July 1, 1994--June 30, 1995. Volume 15

    International Nuclear Information System (INIS)

    Joy, D.R.

    1996-05-01

    The Nuclear Regulatory Commission (NRC) is committed to the periodic publication of licensed fuel facility inventory difference data, following agency review of the information and completion of any related NRC investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of special nuclear material

  16. Licensed fuel facility status report: Inventory difference data, July 1986-December 1986

    International Nuclear Information System (INIS)

    1987-08-01

    NRC is committed to the periodic publication of licensed fuel facilities' inventory difference data, following agency review of the information and completion of any related investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233

  17. Licensed fuel facility status report. Inventory difference data, January-June 1983. Volume 4, No. 1

    International Nuclear Information System (INIS)

    1984-03-01

    NRC is committed to the periodic publication of licensed fuel facilities inventory difference data, following agency review of the information and completion of any related investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, or uranium-233

  18. Licensed fuel facility status report: Inventory difference data, July 1987-December 1987

    International Nuclear Information System (INIS)

    1988-09-01

    NRC is committed to the periodic publication of licensed fuel facilities' inventory difference data, following agency review of the information and completion of any related investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of high enriched uranium, low enriched uranium, plutonium, or uranium-233

  19. Recent Progress on the DUPIC Fuel Fabrication Technology at KAERI

    International Nuclear Information System (INIS)

    Jung-Won Lee; Ho-Jin Ryu; Geun-Il Park; Kee-Chan Song

    2008-01-01

    Since 1991, KAERI has been developing the DUPIC fuel cycle technology. The concept of a direct use of spent PWR fuel in Candu reactors (DUPIC) is based on a dry processing method to re-fabricate Candu fuel from spent PWR fuel without any intentional separation of the fissile materials and fission products. A DUPIC fuel pellet was successfully fabricated and the DUPIC fuel element fabrication processes were qualified on the basis of a Quality Assurance program. Consequently, the DUPIC fuel fabrication technology was verified and demonstrated on a laboratory-scale. Recently, the fuel discharge burn-up of PWRs has been extended to reduce the amount of spent fuel and the fuel cycle costs. Considering this trend of extending the fuel burn-up in PWRs, the DUPIC fuel fabrication technology should be improved to process high burn-up spent fuels. Particularly the release behavior of cesium from the pellet prepared with a high burn-up spent fuel was assessed. an improved DUPIC fuel fabrication technology was experimentally established with a fuel burn-up of 65,000 MWd/tU. (authors)

  20. Property-process relationships in nuclear fuel fabrication

    International Nuclear Information System (INIS)

    Tikare, V.

    2015-01-01

    Nuclear fuels are fabricated using many different techniques as they come in a large variety of shapes and compositions. The design and composition of nuclear fuels are predominantly dictated by the engineering requirements necessary for their function in reactors of various designs. Other engineering properties requirements originate from safety and security concerns, and the easy of handling, storing, transporting and disposing of the radioactive materials. In this chapter, the more common of these fuels will be briefly reviewed and the methods used to fabricate them will be presented. The fuels considered in this paper are oxide fuels used in LWRs and FRs, metal fuels in FRs and particulate fuels used in HTGRs. Fabrication of alternative fuel forms and use of standard fuels in alternative reactors will be discussed briefly. The primary motivation to advance fuel fabrication is to improve performance, reduce cost, reduce waste or enhance safety and security of the fuels. To achieve optimal performance, developing models to advance fuel fabrication has to be done in concert with developing fuel performance models. The specific properties and microstructures necessary for improved fuel performance must be identified using fuel performance models, while fuel fabrication models that can determine processing variables to give the desired microstructure and materials properties must be developed. (author)

  1. Nuclear fuel cycle facility accident analysis handbook

    Energy Technology Data Exchange (ETDEWEB)

    Ayer, J E; Clark, A T; Loysen, P; Ballinger, M Y; Mishima, J; Owczarski, P C; Gregory, W S; Nichols, B D

    1988-05-01

    The Accident Analysis Handbook (AAH) covers four generic facilities: fuel manufacturing, fuel reprocessing, waste storage/solidification, and spent fuel storage; and six accident types: fire, explosion, tornado, criticality, spill, and equipment failure. These are the accident types considered to make major contributions to the radiological risk from accidents in nuclear fuel cycle facility operations. The AAH will enable the user to calculate source term releases from accident scenarios manually or by computer. A major feature of the AAH is development of accident sample problems to provide input to source term analysis methods and transport computer codes. Sample problems and illustrative examples for different accident types are included in the AAH.

  2. Nuclear fuel cycle facility accident analysis handbook

    International Nuclear Information System (INIS)

    Ayer, J.E.; Clark, A.T.; Loysen, P.; Ballinger, M.Y.; Mishima, J.; Owczarski, P.C.; Gregory, W.S.; Nichols, B.D.

    1988-05-01

    The Accident Analysis Handbook (AAH) covers four generic facilities: fuel manufacturing, fuel reprocessing, waste storage/solidification, and spent fuel storage; and six accident types: fire, explosion, tornado, criticality, spill, and equipment failure. These are the accident types considered to make major contributions to the radiological risk from accidents in nuclear fuel cycle facility operations. The AAH will enable the user to calculate source term releases from accident scenarios manually or by computer. A major feature of the AAH is development of accident sample problems to provide input to source term analysis methods and transport computer codes. Sample problems and illustrative examples for different accident types are included in the AAH

  3. Hazard Classification for Fuel Supply Shutdown Facility

    International Nuclear Information System (INIS)

    BENECKE, M.W.

    2000-01-01

    Final hazard classification for the 300 Area N Reactor fuel storage facility resulted in the assignment of Nuclear Facility Hazard Category 3 for the uranium metal fuel and feed material storage buildings (303-A, 303-B, 303-G, 3712, and 3716). Radiological for the residual uranium and thorium oxide storage building and an empty former fuel storage building that may be used for limited radioactive material storage in the future (303-K/3707-G, and 303-E), and Industrial for the remainder of the Fuel Supply Shutdown buildings (303-F/311 Tank Farm, 303-M, 313-S, 333, 334 and Tank Farm, 334-A, and MO-052)

  4. Structure, conduct, and sustainability of the international low-enriched fuel fabrication industry

    International Nuclear Information System (INIS)

    Rothwell, Geoffrey

    2008-01-01

    This paper examines the cost structures of fabricating Low-Enriched Uranium fuel (LEU, enriched to 5% enrichment) light water reactor fuels. The LEU industry is decades old, and (except for high entry cost, i.e., the cost of designing and licensing a fuel fabrication facility and its fuel), labor and additional fabrication lines can be added by industry incumbents at Nth-of-a-Kind cost to the maximum capacity allowed by the license. On the other hand, new entrants face higher First-of-a-Kind costs and high new-facility licensing costs, increasing the scale required for entry thus discouraging small scale entry by countries with only a few nuclear power plants. Therefore, the industry appears to be competitive with sustainable investment in fuel-cycle states, and structural barriers-to-entry increase its proliferation resistance. (author)

  5. Design study and evaluation of fuel fabrication systems for FR fuel cycle

    International Nuclear Information System (INIS)

    Namekawa, Takashi; Tanaka, Kenya; Kawaguchi, Koichi; Koike, Kazuhiro; Shimuta, Hiroshi; Suzuki, Yoshihiro

    2004-01-01

    The plant concept for each FBR fuel fabrication system has been constructed and evaluated, which achieves economical improvement, decrease in the environmental burden, better resource utilization, and proliferation resistance by the various innovative techniques employed. The results are as follows: (1) For oxide fuels, the simplified pelletizing method has a high technical feasibility, and it is possible to apply this method to practical process at early stage, because this method is based on wealth results of a conventional method. (2) For oxide fuels, the sphere packing fuel fabrication system by gelation and vibro-compaction processes has the advantage of lesser dispersion of the fine powder due to the use of solution and granule in the process. However this system shoulders additional cost for the liquid waste treatment process to dispose a large bulk of process liquid waste. (3) For the metal fuel, the casting system is generally expected to have high economical efficiency even for small-scale facilities, although verification for fabrication of the TRU alloy slug is required. (author)

  6. Improvement on fabrication process of CANDU type reactor fuels

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Chang Bum; Chung, Sang Tae; Kim, Hyung Soo; Park, Choon Ho

    1993-05-01

    The study on fabrication of the nuclear fuel for heavy water cooled reactor is performed. Among the fabrication processes of nuclear fuel, welding thickness between fuel rod and end cap can occur a serious error in total fuel length. Therefore, for nuclear fuel design, the thickness of end cap and changed weight of zircaloy-4 tube must be measured accurately. For welding performance, microstructure of welding point is investigated successfully. The result of the study shows the possibility of cost reduction and quality improvement by simplification of fabrication process of nuclear fuel. (Author).

  7. Facilities of fuel transfer for nuclear reactors

    International Nuclear Information System (INIS)

    Wade, E.E.

    1977-01-01

    This invention relates to sodium cooled fast breeder reactors. It particularly concerns facilities for the transfer of fuel assemblies between the reactor core and a fuel transfer area. The installation is simple in construction and enables a relatively small vessel to be used. In greater detail, the invention includes a vessel with a head, fuel assemblies housed in this vessel, and an inlet and outlet for the coolant covering these fuel assemblies. The reactor has a fuel transfer area in communication with this vessel and gear inside the vessel for the transfer of these fuel assemblies. These facilities are borne by the vessel head and serve to transfer the fuel assemblies from the vessel to the transfer area; whilst leaving the fuel assemblies completely immersed in a continuous mass of coolant. A passageway is provided between the vessel and this transfer area for the fuel assemblies. Facilities are provided for closing off this passageway so that the inside of the reactor vessel may be isolated as desired from this fuel transfer area whilst the reactor is operating [fr

  8. History of research reactor fuel fabrication at Babcock and Wilcox

    International Nuclear Information System (INIS)

    Freim, James B.

    1983-01-01

    B and W Research Reactor Fuel Element facility at Lynchburg, Virginia now produces national laboratory and university fuel assemblies. The Company's 201000 square foot facility is devoted entirely to supplying research fuel and related products. B and W re-entered the research reactor fuel market in 1981

  9. Characterization of the 309 fuel examination facility

    International Nuclear Information System (INIS)

    Greenhalgh, W.O.; Cornwell, B.C.

    1997-01-01

    This document identifies radiological, chemical and physical conditions inside the Fuel Examination Facility. It is located inside the Plutonium Recycle Test Reactor containment structure (309 Building.) The facility was a hot cell used for examination of PRTR fuel and equipment during the 1960's. Located inside the cell is a PRTR shim rod assembly, reported are radiological conditions of the sample. The conditions were assessed as part of overall 309 Building transition

  10. Characterization of the 309 fuel examination facility

    Energy Technology Data Exchange (ETDEWEB)

    Greenhalgh, W.O.; Cornwell, B.C.

    1997-07-09

    This document identifies radiological, chemical and physical conditions inside the Fuel Examination Facility. It is located inside the Plutonium Recycle Test Reactor containment structure (309 Building.) The facility was a hot cell used for examination of PRTR fuel and equipment during the 1960`s. Located inside the cell is a PRTR shim rod assembly, reported are radiological conditions of the sample. The conditions were assessed as part of overall 309 Building transition.

  11. PND fuel handling decontamination: facilities and techniques

    International Nuclear Information System (INIS)

    Pan, R.Y.

    1996-01-01

    The use of various decontamination techniques and equipment has become a critical part of Fuel Handling maintenance work at Ontario Hydro's Pickering Nuclear Division. This paper presents an overview of the set up and techniques used for decontamination in the PND Fuel Handling Maintenance Facility and the effectiveness of each. (author). 1 tab., 9 figs

  12. MELOX fuel fabrication plant: Operational feedback and future prospects

    International Nuclear Information System (INIS)

    Hugelmann, D.; Greneche, D.

    2000-01-01

    inspection equipment). In such conditions, plant safety is not affected and operation remains the same. To cope with the growing MOX fuel demand, some countries have equipped themselves (or should equip themselves in the near future) with the state-of-art MOX industrial capabilities. This growing demand is obviously linked with a higher diversity in fuel designs requirements. The empowerment of the MELOX plant, the first high-throughput MOX fuel fabrication facility in operation in the world, is in keeping with this situation: the MELOX West Fitting Building (MWFB) resulting from an optimized design know-how, is the demonstration of the COGEMA Group high capability of adaptation. With the MWFB, the completion of a versatile fabrication plant adapted to the international fuel market is reached. (author)

  13. Design and fuel fabrication processes for the AC-3 mixed-carbide irradiation test

    International Nuclear Information System (INIS)

    Latimer, T.W.; Chidester, K.M.; Stratton, R.W.; Ledergerber, G.; Ingold, F.

    1992-01-01

    The AC-3 test was a cooperative U.S./Swiss irradiation test of 91 wire-wrapped helium-bonded U-20% Pu carbide fuel pins irradiated to 8.3 at % peak burnup in the Fast Flux Test Facility. The test consisted of 25 pins that contained spherepac fuel fabricated by the Paul Scherrer Institute (PSI) and 66 pins that contained pelletized fuel fabricated by the Los Alamos National Laboratory. Design of AC-3 by LANL and PSI was begun in 1981, the fuel pins were fabricated from 1983 to 1985, and the test was irradiated from 1986 to 1988. The principal objective of the AC-3 test was to compare the irradiation performance of mixed-carbide fuel pins that contained either pelletized or sphere-pac fuel at prototypic fluence and burnup levels for a fast breeder reactor

  14. Remote fabrication of pellet fuels for United States breeder reactors

    International Nuclear Information System (INIS)

    Dahl, R.E.; Nyman, D.H.; Yatabe, J.M.

    1979-04-01

    Goal of the program is to demonstrate the feasibility of fabricating breeder fuel in a remotely operated and maintained mode by 1985. Development for pellet fuel fabrication is in the engineering stage with much of the equipment for ceramic unit operations in final design or currently under testing. Results to date confirm that remote fabrication of pellet fuels is feasible. Several of the processes and equipment items are described in this report

  15. Development of the fabrication technology of the simulated DUPIC fuel

    International Nuclear Information System (INIS)

    Kang, Kweon Ho; Yang, M. S.; Bae, K. K. and others

    2000-06-01

    It is important to get basic data to analysis physical properties, behavior in reactor and performance of the DUPIC fuel because physical properties of the DUPIC fuel is different from the commercial UO 2 fuel. But what directly measures physical properties et al. of DUPIC fuel being resinterred simulated spent fuel through OREOX process is very difficult in laboratory owing to its high level radiation. Then fabrication of simulated DUPIC fuel is needed to measure its properties. In this study, processes on powder treatment, OREOX, compaction and sintering to fabricate simulated DUPIC fuel using simulated spent fuel are discribed. To fabricate simulated DUPIC fuel, the powder from 3 times OREOX and 5 times attrition milling simulated spent fuel is compacted with 1.3 ton/cm 2 . Pellets are sintered in 100% H 2 atmosphere over 10 h at 1800 deg C. Sintered densities of pellets are 10.2-10.5 g/cm 3

  16. Regional spent fuel storage facility (RSFSF)

    International Nuclear Information System (INIS)

    Dyck, H.P.

    1999-01-01

    The paper gives an overview of the meetings held on the technology and safety aspects of regional spent fuel storage facilities. The questions of technique, economy and key public and political issues will be covered as well as the aspects to be considered for implementation of a regional facility. (author)

  17. One-Pot and Facile Fabrication of Hierarchical Branched Pt-Cu Nanoparticles as Excellent Electrocatalysts for Direct Methanol Fuel Cells.

    Science.gov (United States)

    Cao, Yanqin; Yang, Yong; Shan, Yufeng; Huang, Zhengren

    2016-03-09

    Hierarchical branched nanoparticles are one promising nanostructure with three-dimensional open porous structure composed of integrated branches for superior catalysis. We have successfully synthesized Pt-Cu hierarchical branched nanoparticles (HBNDs) with small size of about 30 nm and composed of integrated ultrathin branches by using a modified polyol process with introduction of poly(vinylpyrrolidone) and HCl. This strategy is expected to be a general strategy to prepare various metallic nanostructures for catalysis. Because of the special open porous structure, the as-prepared Pt-Cu HBNDs exhibit greatly enhanced specific activity toward the methanol oxidation reaction as much as 2.5 and 1.7 times compared with that of the commercial Pt-Ru and Pt-Ru/C catalysts, respectively. Therefore, they are potentially applicable as electrocatalysts for direct methanol fuel cells.

  18. Shore Facility Fuel Cell Demonstration

    National Research Council Canada - National Science Library

    Lund, Christian; Weaver, Elizabeth

    2004-01-01

    ...) Air Station Cape Cod. The purpose of this evaluation was to determine whether fuel cell technology could be a reliable, more environmentally friendly, and cost-effective alternative to commercially procured electricity...

  19. 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

  20. Environmental concerns regarding a materials test reactor fuel fabrication facility at the Nuclear and Energy Research Institute - IPEN; Atomos para el desarrollo de Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Santos, G. R. T.; Durazzo, M.; Carvalho, E. F. U. [IPEN, CNEN-SP, P.O. Box 11049, CEP 05422-970, Sao Paulo (Brazil); Riella, H. G. [Universidade Federal de Santa Catarina, Departamento de Engenharia Quimica, Campus Universitario, Florianopolis, CEP 88040-900 (Brazil)]. e-mail: grsantos@ipen.br

    2008-07-01

    The aim of the industrial activities success, front to a more and more informed and demanding society and to a more and more competitive market demands an environmental administration policy which doesn't limit itself to assist the legislation but anticipate and prevent, in a responsible way, possible damages to the environment. One of the maim programs of the Institute of Energetic and Nuclear Research of the national Commission of Nuclear Energy located in Brazil, through the Center of Nuclear Fuel - CCN - is to manufacture MTR-type fuel elements using low-enrichment uranium (20 wt% {sup 2}35U), to supply its IEA-RI research reactor. Integrated in this program, this work aims at well developing and assuring a methodology to implant an environment, health and safety policy, foreseeing its management with the use of detailed data reports and through the adoption of new tools for improving the management, in order to fulfil the applicable legislation and accomplish all the environmental, operational and works aspects. The applied methodology for the effluents management comprises different aspects, including the specific environmental legislation of a country, main available effluents treatment techniques, process flow analyses from raw materials and intakes to products, generated effluents, residuals and emissions. Data collections were accomplished for points gathering and tests characterization, classification and compatibility of the generated effluents and their eventual environmental impacts. This study aims to implant the Sustainable Concept in order to guarantee access to financial resources, allowing cost reduction, maximizing long-term profits, preventing and reducing environmental accident risks and stimulating both the attraction and the keeping of a motivated manpower. Work on this project has already started and, even though many technical actions have not still ended, the results have being extremely valuable. These results can already give to

  1. Characterization of candidate DOE sites for fabricating MOX fuel for lead assemblies

    International Nuclear Information System (INIS)

    Holdaway, R.F.; Miller, J.W.; Sease, J.D.; Moses, R.J.; O'Connor, D.G.; Carrell, R.D.; Jaeger, C.D.; Thompson, M.L.; Strasser, A.A.

    1998-03-01

    The Office of Fissile Materials Disposition (MD) of the Department of Energy (DOE) is directing the program to disposition US surplus weapons-usable plutonium. For the reactor option for disposition of this surplus plutonium, MD is seeking to contract with a consortium, which would include a mixed-oxide (MOX) fuel fabricator and a commercial US reactor operator, to fabricate and burn MOX fuel in existing commercial nuclear reactors. This option would entail establishing a MOX fuel fabrication facility under the direction of the consortium on an existing DOE site. Because of the lead time required to establish a MOX fuel fabrication facility and the need to qualify the MOX fuel for use in a commercial reactor, MD is considering the early fabrication of lead assemblies (LAs) in existing DOE facilities under the technical direction of the consortium. The LA facility would be expected to produce a minimum of 1 metric ton heavy metal per year and must be operational by June 2003. DOE operations offices were asked to identify candidate sites and facilities to be evaluated for suitability to fabricate MOX fuel LAs. Savannah River Site, Argonne National Laboratory-West, Hanford, Lawrence Livermore National Laboratory, and Los Alamos National Laboratory were identified as final candidates to host the LA project. A Site Evaluation Team (SET) worked with each site to develop viable plans for the LA project. SET then characterized the suitability of each of the five plans for fabricating MOX LAs using 28 attributes and documented the characterization to aid DOE and the consortium in selecting the site for the LA project. SET concluded that each option has relative advantages and disadvantages in comparison with other options; however, each could meet the requirements of the LA project as outlined by MD and SET

  2. Characterization of candidate DOE sites for fabricating MOX fuel for lead assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Holdaway, R.F.; Miller, J.W.; Sease, J.D.; Moses, R.J.; O`Connor, D.G. [Oak Ridge National Lab., TN (United States); Carrell, R.D. [Technical Resources International, Inc., Richland, WA (United States); Jaeger, C.D. [Sandia National Labs., Albuquerque, NM (United States); Thompson, M.L.; Strasser, A.A. [Delta-21 Resources, Inc., Oak Ridge, TN (United States)

    1998-03-01

    The Office of Fissile Materials Disposition (MD) of the Department of Energy (DOE) is directing the program to disposition US surplus weapons-usable plutonium. For the reactor option for disposition of this surplus plutonium, MD is seeking to contract with a consortium, which would include a mixed-oxide (MOX) fuel fabricator and a commercial US reactor operator, to fabricate and burn MOX fuel in existing commercial nuclear reactors. This option would entail establishing a MOX fuel fabrication facility under the direction of the consortium on an existing DOE site. Because of the lead time required to establish a MOX fuel fabrication facility and the need to qualify the MOX fuel for use in a commercial reactor, MD is considering the early fabrication of lead assemblies (LAs) in existing DOE facilities under the technical direction of the consortium. The LA facility would be expected to produce a minimum of 1 metric ton heavy metal per year and must be operational by June 2003. DOE operations offices were asked to identify candidate sites and facilities to be evaluated for suitability to fabricate MOX fuel LAs. Savannah River Site, Argonne National Laboratory-West, Hanford, Lawrence Livermore National Laboratory, and Los Alamos National Laboratory were identified as final candidates to host the LA project. A Site Evaluation Team (SET) worked with each site to develop viable plans for the LA project. SET then characterized the suitability of each of the five plans for fabricating MOX LAs using 28 attributes and documented the characterization to aid DOE and the consortium in selecting the site for the LA project. SET concluded that each option has relative advantages and disadvantages in comparison with other options; however, each could meet the requirements of the LA project as outlined by MD and SET.

  3. Safety of fuel cycle facilities. Topical issues paper no. 3

    International Nuclear Information System (INIS)

    Ranguelova, V.; Niehaus, F.; Delattre, D.

    2001-01-01

    A wide range of nuclear fuel cycle facilities are in operation. These installations process, use, store and dispose of radioactive material and cover: mining and milling, conversion, enrichment, fuel fabrication (including mixed oxide fuel), reactor, interim spent fuel storage, reprocessing, waste treatment and waste disposal facilities. For the purposes of this paper, reactors and waste disposal facilities are not considered. The term 'fuel cycle facilities' covers only the remainder of the installations listed above. The IAEA Secretariat maintains a database of fuel cycle facilities in its Member States. Known as the Nuclear Fuel Cycle Information System (NFCIS), it is available as an on-line service through the Internet. More than 500 such facilities have been reported under this system. The facilities are listed by facility type and operating status. Approximately one third of all of the facilities are located in developing States. About half of all facilities are reported to be operating, of which approximately 40% are operating in developing States. In addition, some 60 facilities are either in the design stage or under construction. Although the radioactive source term for most fuel cycle facilities is lower than the source term for reactors, which results in less severe consequences to the public from potential accidents at these fuel cycle installations, recent events at some fuel cycle facilities have given rise to public concern which has to be addressed adequately by national regulatory bodies and at the international level. Worldwide, operational experience feedback warrants improvements in the safety of these facilities. Some of the hazards are similar for reactor and non-reactor facilities. However, the differences between these installations give rise to specific safety concerns at fuel cycle facilities. In particular, these concerns include: criticality, radiation protection of workers, chemical hazards, fire and explosion hazards. It is recognized

  4. Hazard analysis for 300 Area N Reactor Fuel Fabrication and Storage Facilty

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, D.J.; Brehm, J.R.

    1994-01-25

    This hazard analysis (HA) has been prepared for the 300 Area N Reactor Fuel Fabrication and Storage Facility (Facility), in compliance with the requirements of Westinghouse Hanford Company (Westinghouse Hanford) controlled manual WHC-CM-4-46, Nonreactor Facility Safety Analysis Manual, and to the direction of WHC-IP-0690, Safety Analysis and Regulation Desk Instructions, (WHC 1992). An HA identifies potentially hazardous conditions in a facility and the associated potential accident scenarios. Unlike the Facility hazard classification documented in WHC-SD-NR-HC-004, Hazard Classification for 300 Area N Reactor Fuel Fabrication and Storage Facility, (Huang 1993), which is based on unmitigated consequences, credit is taken in an HA for administrative controls or engineered safety features planned or in place. The HA is the foundation for the accident analysis. The significant event scenarios identified by this HA will be further evaluated in a subsequent accident analysis.

  5. Radiation protection at nuclear fuel cycle facilities.

    Science.gov (United States)

    Endo, Kuniaki; Momose, Takumaro; Furuta, Sadaaki

    2011-07-01

    Radiation protection methodologies concerning individual monitoring, workplace monitoring and environmental monitoring in nuclear fuel facilities have been developed and applied to facilities in the Nuclear Fuel Cycle Engineering Laboratories (NCL) of Japan Atomic Energy Agency (JAEA) for over 40 y. External exposure to photon, beta ray and neutron and internal exposure to alpha emitter are important issues for radiation protection at these facilities. Monitoring of airborne and surface contamination by alpha and beta/photon emitters at workplace is also essential to avoid internal exposure. A critical accident alarm system developed by JAEA has been proved through application at the facilities for a long time. A centralised area monitoring system is effective for emergency situations. Air and liquid effluents from facilities are monitored by continuous monitors or sampling methods to comply with regulations. Effluent monitoring has been carried out for 40 y to assess the radiological impacts on the public and the environment due to plant operation.

  6. Technology, safety and costs of decommissioning a reference small mixed oxide fuel fabrication plant. Volume 2. Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Jenkins, C. E.; Murphy, E. S.; Schneider, K. J.

    1979-01-01

    Volume 2 contains appendixes on small MOX fuel fabrication facility description, site description, residual radionuclide inventory estimates, decommissioning, financing, radiation dose methodology, general considerations, packaging and shipping of radioactive materials, cost assessment, and safety (JRD)

  7. Radiation shielding calculation for the MOX fuel fabrication plant Melox

    International Nuclear Information System (INIS)

    Lee, Y.K.; Nimal, J.C.; Chiron, M.

    1994-01-01

    Radiation shielding calculation is an important engineering work in the design of the MOX fuel fabrication plant MELOX. Due to the recycle of plutonium and uranium from UO2 spent fuel reprocessing and the large capacity of production (120t HM/yr.), the shielding design requires more attention in this LWR fuel plant. In MELOX, besides several temporary storage facilities of massive fissile material, about one thousand radioactive sources with different geometries, forms, densities, quantities and Pu concentrations, are distributed through different workshops from the PuO 2 powder reception unit to the fuel assembly packing room. These sources, with or without close shield, stay temporarily in different locations, containers and glove boxes. In order to optimize the dimensions, the material and the cost of shield as well as to limit the calculation work in a reasonable engineer-hours, a calculation scheme for shielding design of MELOX is developed. This calculation scheme has been proved to be useful in consideration of the feedback from the evolutionary design and construction. The validated shielding calculations give a predictive but reliable radiation doses information. (authors). 2 figs., 10 refs

  8. Modernization of RTC for fabrication of MOX fuel, Vibropac fuel pins and BN-600 FA with weapon grade plutonium

    International Nuclear Information System (INIS)

    Grachyov, A.F.; Kalygin, V.V.; Skiba, O.V.; Mayorshin, A. A.; Bychkov, A.V.; Kisly, V.A.; Ovsyannikov, Y.F.; Bobrov, D.A.; Mamontov, S.I.; Tsyganov, A.N.; Churutkin, E.I.; Davydov, P.I.; Samosenko, E.A; Shalak, A.R.; Ojima, Hisao

    2004-01-01

    Since mid 70's RIAR has been performing activities on plutonium involvement in fuel cycle. These activities are considered a stage within the framework of the closed fuel cycle development. Developed at RIAR fuel cycle is based on two technologies: 'dry' process of fuel reprocessing and vibro-packing method for fuel pin fabrication. Due to the available scientific capabilities and a gained experience in operating the technological facilities (ORYOL, SIC) for plutonium (various grade) blending into fuel for fast reactors, RIAR is a participant of the activities aimed at solving these tasks. Under international program RIAR with financial support of JNC (Japan) is modernizing the facility for granulated fuel production, vibro-pac fuel pins and FA fabrication to provide the BN-600 'hybrid' core. In order to provide 'hybrid' core it is necessary to produce (per year): - 1775 kg of granulated MOX-fuel, 6500 fuel pins, 50 fuel assemblies. Potential output of the facility under construction is as follows: - 1800 kg of granulated MOX-fuel per year, 40 fuel pins per shift, 200 FAs for the BN-600 reactor per year. Taking into account domestic and foreign experience in MOX-fuel production, different options were discussed of the equipment layouts in the available premises of chemical technological division of RIAR: - in the shielded manipulator boxes, in the existing hot cells. During construction of the facility in the building under operation the following requirements should be met: - facility must meet all standards and regulations set for nuclear facilities, installation work at the facility must not influence other production programs implemented in the building, engineering supply lines of the facility must be connected to the existing service lines of the building, cost of the activities must not exceed amount of JNC funding. The paper presents results of comparison between two options of the process equipment layout: in boxes and hot cells. This equipment is intended

  9. 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)

  10. Significant incidents in nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    1996-03-01

    In contrast to nuclear power plants, events in nuclear fuel cycle facilities are not well documented. The INES database covers all the nuclear fuel cycle facilities; however, it was developed in the early 1990s and does not contain information on events prior to that. The purpose of the present report is to collect significant events and analyze them in order to give a safety related overview of nuclear fuel cycle facilities. Significant incidents were selected using the following criteria: release of radioactive material or exposure to radiation; degradation of items important to safety; and deficiencies in design, quality assurance, etc. which include criticality incidents, fire, explosion, radioactive release and contamination. This report includes an explanation, where possible, of root causes, lessons learned and action taken. 4 refs, 4 tabs

  11. FAST FLUX TEST FACILITY DRIVER FUEL MEETING

    Energy Technology Data Exchange (ETDEWEB)

    None,

    1966-06-01

    The Pacific Northwest Laboratory has convened this meeting to enlist the best talents of our laboratories and industry in soliciting factual, technical information pertinent to the Pacific Northwest's Laboratory's evaluation of the potential fuel systems for the Fast Flux Test Facility. The particular factors emphasized for these fuel systems are those associated with safety, ability to meet testing objectives, and economics. The proceedings includes twenty-three presentations, along with a transcript of the discussion following each, as well as a summary discussion.

  12. Hot Fuel Examination Facility/North Facility safety report

    Energy Technology Data Exchange (ETDEWEB)

    Adams, R.M.; Hampson, D.C.; Ferguson, K.R.; Hylsky, E.

    1975-02-01

    Design and safety-related construction features of the Hot Fuel Examination Facility/North, located on the Argonne--West site at the Idaho National Engineering Laboratory are described. The proposed operations, the organizational structure, and emergency plans are given. Evaluations of potential accident situations are presented and it is concluded that HFEF/N can be operated safely and without undue hazard.

  13. Operation of spent fuel storage facilities

    International Nuclear Information System (INIS)

    1994-01-01

    This Safety Guide was prepared as part of the IAEA's programme on safety of spent fuel storage. This is for interim spent fuel storage facilities that are not integral part of an operating nuclear power plant. Following the introduction, Section 2 describes key activities in the operation of spent fuel storage facilities. Section 3 lists the basic safety considerations for storage facility operation, the fundamental safety objectives being subcriticality, heat removal and radiation protection. Recommendations for organizing the management of a facility are contained in Section 4. Section 5 deals with aspects of training and qualification; Section 6 describes the phases of the commissioning of a spent fuel storage facility. Section 7 describes operational limits and conditions, while Section 8 deals with operating procedures and instructions. Section 9 deals with maintenance, testing, examination and inspection. Section 10 presents recommendations for radiation and environmental protection. Recommendations for the quality assurance (QA) system are presented in Section 11. Section 12 describes the aspects of safeguards and physical protection to be taken into account during operations; Section 13 gives guidance for decommissioning. 15 refs, 5 tabs

  14. Design of spent fuel storage facilities

    International Nuclear Information System (INIS)

    1994-01-01

    This Safety Guide is for interim spent fuel storage facilities that are not integral part of an operating nuclear power plant. Following the introduction, Section 2 describes the general safety requirements applicable to the design of both wet and dry spent fuel storage facilities; Section 3 deals with the design requirements specific to either wet or dry storage. Recommendations for the auxiliary systems of any storage facility are contained in Section 4; these are necessary to ensure the safety of the system and its safe operation. Section 5 provides recommendations for establishing the quality assurance system for a storage facility. Section 6 discusses the requirements for inspection and maintenance that must be considered during the design. Finally, Section 7 provides guidance on design features to be considered to facilitate eventual decommissioning. 18 refs

  15. Current status of vibro-packed fuel fabrication process development

    International Nuclear Information System (INIS)

    Kihara, Yoshiyuki

    2004-01-01

    In the feasibility study of FBR cycle system, the concept of future commercialized fuel is 'low-decontaminated MA(minor actinide) fuel', which contains high radioactive rare earth elements and MAs. This fuel will optimize a positive feature of the fast reactor, which allows high impurities for the fuel. Therefore, this fuel should be fabricated not in a conventional glove-box but in a cell with remote handling. A vibro-packed fuel, like sphere-pac fuel or vipac fuel, is well-suited, and the development of fabrication procedure is in progress as a candidate for the commercialized fuel fabrication process. Results of the three topics are reported: 1) results of uranium dioxide particle fabrication with external gelation process containing samarium (Sm) as a representative element of rare earth impurities. 2) results of vibro-packing experiments using several simulated materials for the prediction of filling behavior during sphere-pac fuel fabrication and 3) results of evaluation on filling behavior of vipac fuel particles with getter (uranium metal) particles. (author)

  16. Reduced enrichment fuels for Canadian research reactors - Fabrication and performance

    International Nuclear Information System (INIS)

    Wood, J.C.; Foo, M.T.; Berthiaume, L.C.; Herbert, L.N.; Schaefer, J.D.; Hawley, D.

    1985-01-01

    Our facilities have been upgraded to manufacture fuel rods comprising dispersions of U 3 Si in aluminum, to complement the dispersions of U 3 Si alloyed with 1.5 and 3.0 wt% Al fabricated and tested previously. Further advances have been made in process optimization particularly in core extrusion where production rate has been doubled while maintaining high quality standards. Our mini-element irradiations of Al-61.5 wt% (U,3.5 wt% Si, 1.5 wt% Al) and Al-62.4 wt% (U,3.2 wt% Si, 30 wt% Al) have been completed successfully up to the terminal burnup of 93 atomic percent. Fuel core swelling remained marginally below 1% per 10 atomic percent burnup over the whole irradiation. Also mini-elements containing Al-72.4 wt% USiAl and Al-73.4 wt% USi*Al have been irradiated to 82 atomic percent burnup, their swelling rate marginally exceeding 1% per 10 atomic percent burnup. Three full-size 12-element NRU assemblies containing Al-62.4 wt% USi*Al have been fabricated and installed in the NRU reactor where they have performed normally without problems. The cores for four more full-size 12-element NRU assemblies containing Al-61.0 wt% U 3 Si have been manufactured. (author)

  17. Hot Fuel Examination Facility/South

    Energy Technology Data Exchange (ETDEWEB)

    1990-05-01

    This document describes the potential environmental impacts associated with proposed modifications to the Hot Fuel Examination Facility/South (HFEF/S). The proposed action, to modify the existing HFEF/S at the Argonne National Laboratory-West (ANL-W) on the Idaho National Engineering Laboratory (INEL) in southeastern Idaho, would allow important aspects of the Integral Fast Reactor (IFR) concept, offering potential advantages in nuclear safety and economics, to be demonstrated. It would support fuel cycle experiments and would supply fresh fuel to the Experimental Breeder Reactor-II (EBR-II) at the INEL. 35 refs., 12 figs., 13 tabs.

  18. Hot Fuel Examination Facility/South

    International Nuclear Information System (INIS)

    1990-05-01

    This document describes the potential environmental impacts associated with proposed modifications to the Hot Fuel Examination Facility/South (HFEF/S). The proposed action, to modify the existing HFEF/S at the Argonne National Laboratory-West (ANL-W) on the Idaho National Engineering Laboratory (INEL) in southeastern Idaho, would allow important aspects of the Integral Fast Reactor (IFR) concept, offering potential advantages in nuclear safety and economics, to be demonstrated. It would support fuel cycle experiments and would supply fresh fuel to the Experimental Breeder Reactor-II (EBR-II) at the INEL. 35 refs., 12 figs., 13 tabs

  19. Safety issues in fabricating mixed oxide fuel using surplus weapons plutonium

    International Nuclear Information System (INIS)

    Buksa, J.; Badwan, F.; Barr, M.; Motley, F.

    1998-07-01

    This paper presents an assessment of the safety issues and implications of fabricating mixed oxide (MOX) fuel using surplus weapons plutonium. The basis for this assessment is the research done at Los Alamos National Laboratory (LANL) in identifying and resolving the technical issues surrounding the production of PuO 2 feed, removal of gallium from the PuO 2 feed, the fabrication of test fuel, and the work done at the LANL plutonium processing facility. The use of plutonium in MOX fuel has been successfully demonstrated in Europe, where the experience has been almost exclusively with plutonium separated from commercial spent nuclear fuel. This experience in safely operating MOX fuel fabrication facilities directly applies to the fabrication and irradiation of MOX fuel made from surplus weapons plutonium. Consequently, this paper focuses on the technical difference between plutonium from surplus weapons, and light-water reactor recycled plutonium. Preliminary assessments and research lead to the conclusion that no new process or product safety concerns will arise from using surplus weapons plutonium in MOX fuel

  20. Report of the AD HOC Study Group on integrated versus dispersed fuel cycle facilities

    International Nuclear Information System (INIS)

    Kreiter, M.R.; Platt, A.M.

    1975-04-01

    To provide isolation of strategic materials and confinement of nuclear wastes, the basic facilities considered in assessing the DFCF and IFCF were mixed plutonium and uranium oxide and HTGR fuel fabrication, fuel reprocessing, high-enrichment isotopic separation and interim waste storage. Reactors, low-enrichment isotopic separation, and low-enrichment uranium facilities were excluded. It is expected that the IFCF would attract uranium fuel fabrication and possibly reactors. An assumption was made for the study that the choice of either IFCF or DFCF would not alter the nuclear power generation pattern postulated to exist up to the year 2000. The advantages of IFCF are seen to outweigh disadvantages

  1. Estimates of Canadian fuel fabrication costs for alternative fuel cycles and systems

    International Nuclear Information System (INIS)

    Blahnik, C.

    1979-04-01

    Unit fuel fabrication costs are estimated for alternate fuel cycles and systems that may be of interest in Ontario Hydro's strategy analyses. A method is proposed for deriving the unit fuel fabrication price to be paid by a Canadian utility as a function of time (i.e. the price that reflects the changing demand/supply situation in the particular scenario considered). (auth)

  2. Concept for a small, colocated fuel cycle facility for oxide breeder fuels

    International Nuclear Information System (INIS)

    Burch, W.D.; Lerch, R.E.; Stradley, J.G.

    1987-01-01

    As part of a United States Department of Energy (USDOE) program to examine innovative liquid-metal reactor (LMR) system designs over the past three years, the Oak Ridge National Laboratory (ORNL) and the Westinghouse Hanford Company (WHC) collaborated on studies of mixed oxide fuel cycle options. A principal effort was an advanced concept for a small integrated fuel cycle colocated with a 1300-MW(e) reactor station. The study provided a scoping design, capital and operating cost estimates, and a basis on which to proceed with implementation of such a facility if future plans so dictate. The facility integrated reprocessing, waste management, and refabrication functions in a single facility of nominal 35-t/year capacity utilizing the latest technology developed in fabrication programs at WHC and in reprocessing at ORNL. The concept was based on many years of work at both sites and extensive design studies of prior years

  3. Breeder fuel pellet fabrication from gel-sphere conversion

    International Nuclear Information System (INIS)

    McLemore, D.R.; Bennett, D.W.; Hart, P.E.; Norman, R.E.

    1980-01-01

    The development of automated nuclear fuel fabrication methods is the goal of the United States breeder reactor program and will form the technological basis for the future breeder reactor fuel supply. A major factor in achieving this goal is the development of remotely operated fuel fabrication equipment. The program schedule is to demonstrate the feasibility of automated pellet fuel fabrication and remote maintenance techniques by the mid-1980's. Development of major ceramic unit operations and the required computer control system is currently in the engineering testing stage. This report describes the planned program activities for evaluating the gel-sphere conversion process and demonstrating the fabrication of breeder fuel pellets from these gel-derived spheres

  4. Estimation of costs for fabrication of pressurized-water reactor fuel

    International Nuclear Information System (INIS)

    Judkins, R.R.; Olsen, A.R.

    1979-01-01

    To provide a reference case on which to base cost estimates of the several fuel cycles to be considered, the facility, equipment, and operating requirements for the fabrication of fuel for current-design pressurized-water reactors were examined. From an analysis of these requirements, the capital and operating costs of a plant with a capacity of two metric tons of heavy metal per day (MTHM/day) were estimated. In a cash flow analysis, the lifetime of the plant was assumed to be 20 y, and the income from the sale of nuclear fuel assemblies over this period was equated to the total capital and operating expenses of the plant, including a specified 15% return on investment. In this way a levelized unit price for the fuel was obtained. The effects of inflation were not considered since the purpose of these estimates and the determination of unit price was to permit comparison of different types of fuels. The capital costs of the fuel fabrication plant were estimated at $32 million for the facility--land, site preparation, building--and $34 million for equipment. Annual operating costs including labor, management, materials, and utilities were estimated to be $36.5 million. From these estimates, the unit price for fabricating the fuel for the reference pressurized-water reactor was determined to be $138/kg of heavy metal or $63,600 per fuel assembly

  5. Fiber Fabrication Facility for Non-Oxide and Specialty Glasses

    Data.gov (United States)

    Federal Laboratory Consortium — FUNCTION: Unique facility for the research, development, and fabrication of non-oxide and specialty glasses and fibers in support of Navy/DoD programs.DESCRIPTION:...

  6. Liquefied Gaseous Fuels Spill Test Facility

    International Nuclear Information System (INIS)

    1993-02-01

    The US Department of Energy's liquefied Gaseous Fuels Spill Test Facility is a research and demonstration facility available on a user-fee basis to private and public sector test and training sponsors concerned with safety aspects of hazardous chemicals. Though initially designed to accommodate large liquefied natural gas releases, the Spill Test Facility (STF) can also accommodate hazardous materials training and safety-related testing of most chemicals in commercial use. The STF is located at DOE's Nevada Test Site near Mercury, Nevada, USA. Utilization of the Spill Test Facility provides a unique opportunity for industry and other users to conduct hazardous materials testing and training. The Spill Test Facility is the only facility of its kind for either large- or small-scale testing of hazardous and toxic fluids including wind tunnel testing under controlled conditions. It is ideally suited for test sponsors to develop verified data on prevention, mitigation, clean-up, and environmental effects of toxic and hazardous gaseous liquids. The facility site also supports structured training for hazardous spills, mitigation, and clean-up. Since 1986, the Spill Test Facility has been utilized for releases to evaluate the patterns of dispersion, mitigation techniques, and combustion characteristics of select materials. Use of the facility can also aid users in developing emergency planning under US P.L 99-499, the Superfund Amendments and Reauthorization Act of 1986 (SARA) and other regulations. The Spill Test Facility Program is managed by the US Department of Energy (DOE), Office of Fossil Energy (FE) with the support and assistance of other divisions of US DOE and the US Government. DOE/FE serves as facilitator and business manager for the Spill Test Facility and site. This brief document is designed to acquaint a potential user of the Spill Test Facility with an outline of the procedures and policies associated with the use of the facility

  7. Advances in AGR fuel fabrication - now and the future

    International Nuclear Information System (INIS)

    Bleasdale, P.A.

    1995-01-01

    To date, over 3 million AGR fuel pins have been manufactured at Springfields for the UK AGR programme. During this time, AGR fuel design and manufacture has developed and evolved in response to the needs of the reactor operators to enhance fuel reliability and performance. More recently, major advances have been made in the systems and organisational culture which support fuel manufacture at Fuel Division. The introduction of MRP II in 1989 into Fuel Division enabled significant reductions in stock and work-in-progress, together with reductions in manufacturing lead times. Other successful initiatives introduced into Fuel Division have been Just-in-Time (JIT) and AST (Additional Skills Training) which have built on the success of MRP II. All of these initiatives are evidence of Fuel Division's ''Total Quality'' approach to fabricating fuel. Fuel Division is currently in the final stages of commissioning the New Oxide Fuels Complex (NOFC) where both AGR and PWR fuel will be manufactured to the highest standards of quality, safety and environmental protection. NOFC is a totally integrated plant which represents a Pound 200M investment, demonstrating Fuel Division's commitment to building on its 40+ years of fuel fabrication experience and ensuring secure supply of fuel to its customers for years to come. (author)

  8. AERIAL DELIVERY DESIGN AND FABRICATION FACILITY

    Data.gov (United States)

    Federal Laboratory Consortium — Skilled personnel are equipped to design and develop various prototype airdrop items. This facility has all classes of sewing machines, ranging from lightweight to...

  9. Review of Re-fabrication Methods for ATW Fuels and Targets

    International Nuclear Information System (INIS)

    Williams, D.F.; Collins, E.D.; Felker, L.K.; Benker, D.E.; Toth, L.M.

    2002-01-01

    Fabrication of the minor actinides into a fuel and fabrication of the long-lived fission products into a target are key steps in the Accelerator Transmutation of Waste (ATW) fuel cycle. Fabrication is typically preceded by conversion of a purified liquid stream into a form suitable for fuel fabrication. For many fuel recycle flowsheets, these two steps are combined (conversion + fabrication), and the joint operation identified as re-fabrication. The requirements for ATW re-fabrication activities are somewhat similar to those of the U/Th fuel recycle. Because of the high radiation fields, fully remote operations and heavily shielded facilities are required. The high specific activity of the materials will place extreme demands on the chemical and radiation stability of processing media. Many different types of transmutation fuels have been proposed by various researchers from around the world: molten salts, coated particles, metal alloys, ceramic-metal composites (Cermets), and conventional pellet fuels. These forms are largely defined on the basis of fuel performance criteria. However, re-fabrication must also be a safe, practical, and efficient operation that can accept feed materials from the separations complex. Some balancing of priorities and integration is necessary. The Advanced Accelerator Applications (AAA) program of the Department of Energy has been charged with the responsibility to screen potential re-fabrication technologies for ATW. The present baseline fuel form is a Zr-based metal alloy fuel, and Tc metal and NaI are the reference fission product target forms. In this paper we present an initial review of ATW re-fabrication methods. ORNL has prepared minor actinide transmutation targets (Pu, Am, and Cm) for production of heavy elements in the High Flux Isotope Reactor (HFIR) for over 30 years. These targets are made by a resin-loading process followed by a blending with metallic aluminum powder to form a Cermet, then pellet fabrication, and target

  10. 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.)

  11. Hematite nuclear fuel cycle facility decommissioning

    International Nuclear Information System (INIS)

    Hayes, K.

    2004-01-01

    Westinghouse Electric Company LLC ('Westinghouse') acquired a nuclear fuel processing plant at Hematite, Missouri ('Hematite', the 'Facility', or the 'Plant') in April 2000. The plant has subsequently been closed, and its operations have been relocated to a newer, larger facility. Westinghouse has announced plans to complete its clean-up, decommissioning, and license retirement in a safe, socially responsible, and environmentally sound manner as required by internal policies, as well as those of its parent company, British Nuclear Fuels plc. ('BNFL'). Preliminary investigations have revealed the presence of environmental contamination in various areas of the facility and grounds, including both radioactive contamination and various other substances related to the nuclear fuel processing operations. The disparity in regulatory requirements for radiological and nonradiological contaminants, the variety of historic and recent operations, and the number of previous owners working under various contractual arrangements for both governmental and private concerns has resulted in a complex project. This paper discusses Westinghouse's efforts to develop and implement a comprehensive decontamination and decommissioning (D and D) strategy for the facility and grounds. (author)

  12. Alpha Fuels Environmental Test Facility impact gun

    International Nuclear Information System (INIS)

    Anderson, C.G.

    1978-01-01

    The Alpha Fuels Environmental Test Facility (AFETF) impact gun is a unique tool for impact testing 238 PuO 2 -fueled heat sources of up to 178-mm dia at velocities to 300 m/s. An environmentally-sealed vacuum chamber at the muzzle of the gun allows preheating of the projectile to 1,000 0 C. Immediately prior to impact, the heat source projectile is completely sealed in a vacuum-tight catching container to prevent escape of its radioactive contents should rupture occur. The impact velocity delivered by this gas-powered gun can be regulated to within +-2%

  13. Fabrication of simulated mid-burnup CANDU fuel in the RFFL

    International Nuclear Information System (INIS)

    Dimayuga, F.C.

    1997-01-01

    Prediction of the coefficient of void reactivity for CANDU reactor cores is key to modeling postulated loss-of-coolant scenarios to support safety analyses. To reduce the uncertainty in these predictions, computer codes used to generate the predictions have to be well validated for cores at equilibrium burnup. To this end, a COG-funded project was launched to resume mixed oxide (MOX) fuel fabrication operations in the mothballed Recycle Fuel Fabrication Laboratory (RFFL) at CRL, and produce (U,Pu)O 2 fuel simulating mid-burnup CANDU fuel for physics testing in the ZED-2 reactor. In August 1996, rehabilitation of the RFFL was completed, and MOX operations were resumed in the facility. An up-to-date description of the RFFL, including the upgraded safety systems and process equipment, is presented. An overview of the fabrication campaign to produce 37 MOX fuel bundles for ZED-2 tests is given. The fabrication process used to manufacture the fuel from the starting powders to the finished elements and bundles is summarized. Fabrication data including production throughputs and inspection results is discussed. (author)

  14. Impact of fuel fabrication and fuel management technologies on uranium management

    International Nuclear Information System (INIS)

    Arnsberger, P.L.; Stucker, D.L.

    1994-01-01

    Uranium utilization in commercial pressurized water reactors is a complex function of original NSSS design, utility energy requirements, fuel assembly design, fuel fabrication materials and fuel fabrication materials and fuel management optimization. Fuel design and fabrication technologies have reacted to the resulting market forcing functions with a combination of design and material changes. The technologies employed have included ever-increasing fuel discharge burnup, non-parasitic structural materials, burnable absorbers, and fissile material core zoning schemes (both in the axial and radial direction). The result of these technological advances has improved uranium utilization by roughly sixty percent from the infancy days of nuclear power to present fuel management. Fuel management optimization technologies have also been developed in recent years which provide fuel utilization improvements due to core loading pattern optimization. This paper describes the development and impact of technology advances upon uranium utilization in modern pressurized water reactors. 10 refs., 3 tabs., 10 figs

  15. Fabrication of (Th-Pu) MOX fuel pellets using CAP route for experimental irradiation

    International Nuclear Information System (INIS)

    Nehete, Y.G.; Shelke, B.K.; Shotriya, S.K.; Subbarayal, K.; Kumar, Neeraj; Khot, P.M.; Surendra, B.; Yadav, M.K.; Mishra, A.K.; Afzal, Mohd; Panakkal, J.P.

    2011-01-01

    Advanced Fuel Fabrication Facility (AFFF), Tarapur is an industrial scale plant for the fabrication of MOX fuel for thermal and fast reactor. Mixed oxide fuels of different composition have been fabricated in this facility for both thermal and fast reactors. In order to simulate actual condition in Advanced Heavy Water Reactors (AHWR), (Th-Pu) MOX fuel pins of different % of PuO 2 were fabricated at AFFF, BARC Tarapur for irradiation in research reactor. ThO 2 -1% PuO 2 was fabricated by conventional powder pelletization (POP) as well as Coated agglomerate pelletization (CAP) route. Powder pelletization route involves co-milling of milled ThO 2 powder with calcined PuO 2 , precompaction, granulation, final compaction followed by sintering in reducing atmosphere at 1650 deg C. Coated Agglomerate Pelletization (CAP) route involves making spheroids of milled ThO 2 powder using extrusion and spherodization technique then coating of ThO 2 spheroids with desired amount of calcined PuO 2 powder followed by final compaction and sintering in reducing atmosphere. Necessary modifications in process parameters have been made and the tool design was so chosen to get sintered to sized pellet to avoid grinding. Pellets fabricated by CAP route meets all desired specification like POP route. It also has the advantages over POP route like small number of process steps and reduction of man-rem. This paper highlights feasibility of CAP route for fabrication of (Th-Pu)O 2 fuel and characterization of fuel pellets. (author)

  16. Prototype fuel fabrication for nuclear reactors of Laguna Verde

    International Nuclear Information System (INIS)

    Nocetti, C.; Torres, J.; Medrano, A.

    1996-01-01

    Four prototype fuel bundles for the Laguna Verde Nuclear Power Plant have been fabricated. the type of nuclear fuel produced is described and the process used is commented. As an example of the fabrication criteria adopted, the production model to determine the density of the U O 2 pellets for the different batches of ceramic powder is described. the results are evaluated using the statistical indexes C p and C pk . (author)

  17. FUEL HANDLING FACILITY WORKER DOSE ASSESSMENT

    International Nuclear Information System (INIS)

    Achudume, A.

    2004-01-01

    The purpose of this design calculation is to estimate radiation doses received by personnel working in the Fuel Handling Facility (FHF) of the Monitored Geological Repository (MGR). The FHF is a surface facility supporting waste handling operations i.e. receive transportation casks, transfer wastes, prepare waste packages, and ship out loaded waste packages and empty casks. The specific scope of work contained in this calculation covers both collective doses and individual worker group doses on an annual basis, and includes the contributions due to external and internal radiation. The results are also limited to normal operations only. Results of this calculation will be used to support the FHF design and License Application

  18. 77 FR 823 - Guidance for Fuel Cycle Facility Change Processes

    Science.gov (United States)

    2012-01-06

    ... is required before implementing them. Operating experience from nuclear fuel cycle facilities shows... NUCLEAR REGULATORY COMMISSION [NRC-2009-0262] Guidance for Fuel Cycle Facility Change Processes... Fuel Cycle Facility Change Processes.'' This regulatory guide describes the types of changes for which...

  19. Safety assessment for spent fuel storage facilities

    International Nuclear Information System (INIS)

    1994-01-01

    This Safety Practice has been prepared as part of the IAEA's programme on the safety assessment of interim spent fuel storage facilities which are not an integral part of an operating nuclear power plant. This report provides general guidance on the safety assessment process, discussing both deterministic and probabilistic assessment methods. It describes the safety assessment process for normal operation and anticipated operational occurrences and also related to accident conditions. 10 refs, 2 tabs

  20. Plutonium fuel fabrication residues and wastes

    International Nuclear Information System (INIS)

    Arnal, T.; Cousinou, G.; Desille, H.

    1982-04-01

    This paper discusses the current situation in the fabrication plant at Cadarache with an annual plutonium throughput of several tons. Three major fabrication byproduct categories are defined in this plant: 1) scraps, directly recycled at the fabrication input station; 2) residues, byproducts recycled by chemical processes, or processed in washing and incineration stations; 3) wastes, placed in drums and evacuated directly to a waste conditioning station. The borderline between residues and wastes has yet to be precisely determined

  1. World scale fuel methanol facility siting

    International Nuclear Information System (INIS)

    Stapor, M.C.; Hederman, W.F.

    1990-01-01

    Since the Administration announced a clean alternative fuels initiative, industry and government agencies' analyses of the economics of methanol as an alternative motor vehicle fuel have accelerated. In the short run, methanol appears attractive because excess production capacity currently has depressed methanol prices and marginal costs of production are lower than other fuels (current excess capacity). In the long run, however, full costs are the more relevant. To lower average production costs, U.S. policy interest has focused on production from a world-scale, 10,000 tons per day (tpd) methanol plant facility on a foreign site. This paper reviews several important site and financial considerations in a framework to evaluate large scale plant development. These considerations include: risks associated with a large process plant; supply economics of foreign sites; and investment climates and financial incentives for foreign investment at foreign sites

  2. MOX fuel fabrication, in reactor performance and improvement

    International Nuclear Information System (INIS)

    Vliet, J. van; Deramaix, P.; Nigon, J.L.; Fournier, W.

    1998-01-01

    In Europe, MOX fuel for light water reactors (LWRs) has first been manufactured in Belgium and Germany. Belgonucleaire (BN) loaded the first MOX assembly in the BR3 Pressurised Water Reactor (PWR) in 1963. In June 1998, more than 750 tHM LWR MOX fuel assemblies were manufactured on a industrial scale in Europe without any particular difficulty relating to fuel fabrication, reactor operation or fuel behaviour. So, today plutonium recycling through MOX fuel is a mature industry, with successful operational experience and large-scale fabrication plants. In this field, COGEMA and BELGONUCLEAIRE are the main actors by operating simultaneously three complete multidesign fuel production plants: MELOX plant (in Marcoule), CADARACHE plant and P0 plant (in Dessel, Belgium). Present MOX production capacity available to COGEMA and BN fits 175 tHM per year and is to be extended to reach about 325 tHM in the year 2000. This will represent 75% of the total MOX fabrication capacity in Europe. The industrial mastery and the high production level in MOX fabrication assured by high technology processes confer to these companies a large expertise for Pu recycling. This allows COGEMA and BN to be major actors in Pu-based fuels in the coming second nuclear era with advanced fuel cycles. (author)

  3. Hoisting appliances and fuel handling equipment at nuclear facilities

    International Nuclear Information System (INIS)

    1987-01-01

    The guide is followed by the Finnish Centre for Radiation and Nuclear Safety (STUK) in regulating hoisting and handling equipment Class 3 at nuclear facilities. The guide is applied e.g. to the following equipment: reactor building overhead cranes, hoisting appliances at nuclear fuel storages, fuel handling machines, other hoisting appliances, which because of nuclear safety aspects are classified in Safety Class 3, and load-bearing devices connected with the above equipment, such as replaceable hoisting tools and auxiliary lifting devices. The regulating of hoisting and handling equipment comprises the following stages: handling of preliminary and final safety analysis reports, inspection of the construction plan, supervision of fabrication and construction inspection, and supervision of initial start-up and commissioning inspection

  4. FFTF [fast flux tests facility] metal fuel pin sodium bond quality verification

    International Nuclear Information System (INIS)

    Pitner, A.L.

    1989-01-01

    The Fast Flux Tests Facility (FFTF) Series III driver fuel design consists of U-10 Zr fuel slugs contained in a ferritic alloy cladding. A liquid-metal, sodium bond between the fuel and cladding is required to prevent unacceptable temperatures during operation. Excessive voiding or porosity in the sodium thermal bond could result in localized fuel melting during irradiation. It is therefore imperative that bond quality be verified during fabrication of these metal fuel pins prior to irradiation. Cladding stripping of two spare metal fuel pins was successfully performed to inspect the quality of the sodium thermal bond and to compare observed defects in the bond with eddy current measurements. Bond quality was found to be generally good, thereby confirming the fabrication process used to bond the fuel and cladding with liquid-metal sodium to be acceptable. Observed porosity distribution in the sodium bond correlated well with eddy current indications

  5. Fabrication and characterization of fully ceramic microencapsulated fuels

    International Nuclear Information System (INIS)

    Terrani, K.A.; Kiggans, J.O.; Katoh, Y.; Shimoda, K.; Montgomery, F.C.; Armstrong, B.L.; Parish, C.M.; Hinoki, T.; Hunn, J.D.; Snead, L.L.

    2012-01-01

    The current generation of fully ceramic microencapsulated fuels, consisting of Tristructural Isotropic fuel particles embedded in a silicon carbide matrix, is fabricated by hot pressing. Matrix powder feedstock is comprised of alumina–yttria additives thoroughly mixed with silicon carbide nanopowder using polyethyleneimine as a dispersing agent. Fuel compacts are fabricated by hot pressing the powder–fuel particle mixture at a temperature of 1800–1900 °C using compaction pressures of 10–20 MPa. Detailed microstructural characterization of the final fuel compacts shows that oxide additives are limited in extent and are distributed uniformly at silicon carbide grain boundaries, at triple joints between silicon carbide grains, and at the fuel particle–matrix interface.

  6. Quality control in the fabrication of prototype nuclear fuel in Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Nocetti, C.A.; Rodriguez, J.G. (Instituto Nacional de Investigaciones Nucleares, Mexico City (Mexico)); Muehling, G. (Kernforschungszentrum Karlsruhe GmbH (Germany, F.R.))

    1991-02-01

    At the Mexican National Institute of Nuclear Research (ININ) a pilot plant facility was installed for UO{sub 2} fuel and fuel pin fabrication. During various fabrication campaigns different quality control procedures were used starting from a simple inspection and control of the final products up to more detailed quality control procedures, where statistical methods were used to control the entire fabrication steps. The results of diameter and density measurements, received on the basis of two different fabrication campaigns with different quality control procedures, clearly show the advantages of the more developed statistical procedures. Procress capability analyses were also performed confirming a considerable improvement with respect to the values for process performance and distribution and for the process yield also in the same direction. (orig.).

  7. Coated U(Mo) Fuel: As-Fabricated Microstructures

    Energy Technology Data Exchange (ETDEWEB)

    Emmanuel Perez; Dennis D. Keiser, Jr.; Ann Leenaers; Sven Van den Berghe; Tom Wiencek

    2014-04-01

    As part of the development of low-enriched uranium fuels, fuel plates have recently been tested in the BR-2 reactor as part of the SELENIUM experiment. These fuel plates contained fuel particles with either Si or ZrN thin film coating (up to 1 µm thickness) around the U-7Mo fuel particles. In order to best understand irradiation performance, it is important to determine the starting microstructure that can be observed in as-fabricated fuel plates. To this end, detailed microstructural characterization was performed on ZrN and Si-coated U-7Mo powder in samples taken from AA6061-clad fuel plates fabricated at 500°C. Of interest was the condition of the thin film coatings after fabrication at a relatively high temperature. Both scanning electron microscopy and transmission electron microscopy were employed. The ZrN thin film coating was observed to consist of columns comprised of very fine ZrN grains. Relatively large amounts of porosity could be found in some areas of the thin film, along with an enrichment of oxygen around each of the the ZrN columns. In the case of the pure Si thin film coating sample, a (U,Mo,Al,Si) interaction layer was observed around the U-7Mo particles. Apparently, the Si reacted with the U-7Mo and Al matrix during fuel plate fabrication at 500°C to form this layer. The microstructure of the formed layer is very similar to those that form in U-7Mo versus Al-Si alloy diffusion couples annealed at higher temperatures and as-fabricated U-7Mo dispersion fuel plates with Al-Si alloy matrix fabricated at 500°C.

  8. Fuel fabrication activities for power reactors in India - present and future prospects

    International Nuclear Information System (INIS)

    Jayaraj, R.N.

    2010-01-01

    'Full text:' Based on the initial developmental works and pilot plant studies carried out at Bhabha Atomic Research Centre (BARC), Mumbai, the Nuclear Fuel Complex (NFC) was established as an industrial unit during the early 1970s to manufacture fuel assemblies and reactor core structurals for PHWRs and BWRs. The plants consisting of uranium and zirconium streams process raw materials in the form of magnesium di-uranate (MDU) and zircon sand respectively to fabricate finished fuel assemblies. Presently, 17 PHWRs and 2 BWRs are operating in India, generating about 4340 MWe of electricity. The reload fuel assemblies for all these reactors are manufactured at NFC in addition to the reactor structurals like coolant channels and calandria tubes. In tune with the decision of the Department of Atomic Energy to set up twelve PHWRs of 700 MWe capacity, NFC has embarked on an expansion programme to manufacture fuel for all these reactors. In addition, the plans are afoot to set up fuel fabrication facilities for catering to the needs of large number of imported light water reactors (LWRs), in collaboration with the fuel manufacturers abroad. This paper highlights the present scenario and the future plans for fabrication of nuclear fuel assemblies for both indigenously built and imported reactors. (author)

  9. Fabrication of Separator Demonstration Facility process vessel

    International Nuclear Information System (INIS)

    Oberst, E.F.

    1985-01-01

    The process vessel system is the central element in the Separator Development Facility (SDF). It houses the two major process components, i.e., the laser-beam folding optics and the separators pods. This major subsystem is the critical-path procurement for the SDF project. Details of the vaious parts of the process vessel are given

  10. Irradiated fuel reassembling experience at the fuel monitoring facility

    International Nuclear Information System (INIS)

    Aratani, K.; Koizumi, A.; Matsushima, H.

    1989-01-01

    For the first time in the 10-yr hot operation of the fuel monitoring facility (FMF), the reassembling or irradiated fuels was successfully conducted and a reassembled irradiation vehicle was reinserted in Japanese experimental fast reactor JOYO. The FMF, one of the largest hot laboratories in Japan operated by the Power Reactor and Nuclear Fuel Development Corporation (PNC), demonstrated its new capability in remote handling. More than 130 assemblies have already been examined and disassembled at FMF for postirradiation examination and many results have been obtained to evaluate fuel performance. In addition to these once-through examinations, it is becoming more and more important to conduct interim examinations and reinsertion for continuous irradiation. More flexibility for irradiation experiments will thus be provided. Since FMF was originally designed to make the reinsertion possible, there is a path to get the assembly back to the reactor. The main developments needed for the reinsertion of assemblies were as follows: (1) irradiation vehicle, (2) disassembling and interim examination, (3) decontamination of fuel pin surface, and (4) reassembling machine. This paper mainly describes items 2, 3, and 4. The reinsertion program is now planned for two vehicles a year, and several new types of irradiation vehicles for the reinsertion are now being developed. The reassembling machine may be slightly modified so that those new types of rigs can also be handled

  11. Coated fuel particles: requirements and status of fabrication technology

    International Nuclear Information System (INIS)

    Huschka, H.; Vygen, P.

    1977-01-01

    Fuel cycle, design, and irradiation performance requirements impose restraints on the fabrication processes. Both kernel and coating fabrication processes are flexible enough to adapt to the needs of the various existing and proposed high-temperature gas-cooled reactors. Extensive experience has demonstrated that fuel kernels with excellent sphericity and uniformity can be produced by wet chemical processes. Similarly experience has shown that the various multilayer coatings can be produced to fully meet design and specification requirements. Quality reliability of coated fuel particles is ensured by quality control and quality assurance programs operated by an aduiting system that includes licensing officials and the customer

  12. 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)

  13. Transmutation Fuel Fabrication-Fiscal Year 2016

    Energy Technology Data Exchange (ETDEWEB)

    Fielding, Randall Sidney [Idaho National Lab. (INL), Idaho Falls, ID (United States); Grover, Blair Kenneth [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-12-01

    ABSTRACT Nearly all of the metallic fuel that has been irradiated and characterized by the Advanced Fuel Campaign, and its earlier predecessors, has been arc cast. Arc casting is a very flexible method of casting lab scale quantities of materials. Although the method offers flexibility, it is an operator dependent process. Small changes in parameter space or alloy composition may affect how the material is cast. This report provides a historical insight in how the casting process has been modified over the history of the advanced fuels campaign as well as the physical parameters of the fuels cast in fiscal year 2016.

  14. Fabrication of DUPIC fuel pellets using high burn-up spent PWR fuel

    International Nuclear Information System (INIS)

    Lee, Jung-Won; Park, Geun-Il; Choi, Yong

    2012-01-01

    Technology for the direct usage of a spent PWR fuel in CANDU reactors (DUPIC) was developed in KAERI to reduce the amount of spent fuel. DUPIC fuel pellets were fabricated using a dry processing method to re-fabricate CANDU fuel from spent PWR fuel without any intentional separation of fissile materials or fission products. The DUPIC fuel element fabrication process satisfied a quality assurance program in accordance with the Canadian standard. For the DUPIC fuels with various fuel burn-ups between 27,300 and 65,000 MWd/tU, the sintered pellet density decreased with increasing fuel burn-ups. Fission gas releases and powder properties of the spent fuel also influenced the DUPIC fuel characteristics. Measurement of cesium content released from green pellets revealed that their sintered density significantly depended on sintering temperature history. It was useful to establish a DUPIC fuel fabrication technology in which a high-burn-up fuel with 65,000 MWd/tU was treated. (author)

  15. Financing Strategies for Nuclear Fuel Cycle Facility

    International Nuclear Information System (INIS)

    David Shropshire; Sharon Chandler

    2005-01-01

    To help meet our nation's energy needs, reprocessing of spent nuclear fuel is being considered more and more as a necessary step in a future nuclear fuel cycle, but incorporating this step into the fuel cycle will require considerable investment. This report presents an evaluation of financing scenarios for reprocessing facilities integrated into the nuclear fuel cycle. A range of options, from fully government owned to fully private owned, was evaluated using a DPL (Dynamic Programming Language) 6.0 model, which can systematically optimize outcomes based on user-defined criteria (e.g., lowest life-cycle cost, lowest unit cost). Though all business decisions follow similar logic with regard to financing, reprocessing facilities are an exception due to the range of financing options available. The evaluation concludes that lowest unit costs and lifetime costs follow a fully government-owned financing strategy, due to government forgiveness of debt as sunk costs. Other financing arrangements, however, including regulated utility ownership and a hybrid ownership scheme, led to acceptable costs, below the Nuclear Energy Agency published estimates. Overwhelmingly, uncertainty in annual capacity led to the greatest fluctuations in unit costs necessary for recovery of operating and capital expenditures; the ability to determine annual capacity will be a driving factor in setting unit costs. For private ventures, the costs of capital, especially equity interest rates, dominate the balance sheet; the annual operating costs dominate the government case. It is concluded that to finance the construction and operation of such a facility without government ownership could be feasible with measures taken to mitigate risk, and that factors besides unit costs should be considered (e.g., legal issues, social effects, proliferation concerns) before making a decision on financing strategy

  16. Development of Fabrication Technology for Ceramic Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, H. S.; Lee, Y. W.; Na, S. H.; Kim, Y. G.; Jung, C. Y.; Kim, S. H.; Lee, S. C.; Son, D. S

    2006-04-15

    Purpose and Necessity Research purposes for the 3rd stage were to reaffirm the MOX fabrication processes and to establish the process database, based on the fabrication technology developed during the previous stage. This project was also proceeded to improve the fuel performance and to accomplish the inherent MOX technology for PWR. The fabrication processes should be proceeded in the glove boxes because the raw powders of MOX fuel is very toxic. Therefore, some special technology were needed to develop besides the fuel fabrication technology. Both core technology and steadiness of fabrication process are important to obtain homogeneity and thermo-physical properties of MOX fuel pellet. By developing these technology in fashion unique to ourselves, we can take the initiative in the nuclear fuel for next generation. The uranium price has been increasing along with the oil price recently. We have to secure the MOX fabrication technology which serves the effective use of uranium resource. Improvement of pellet characteristics along with the MOX irradiation analysis: Collection and monitoring of the MOX irradiation data, Establishment of the improvement methods of pellet characteristics Establishment of the MOX pellet fabrication process by the unique technology, Establishment of database with the MOX fabrication parameters and characteristics, Analysis of co-relation and re-appearance of the pellet characteristics affected by each process parameter, Construction of feedback system between database and process, Application of the unique fabrication technology to the industrial spot. Applicability of the unique fabrication processes to the glove box technology, Installment of process equipment in the glove box and development of operation skill, Methods for modifying, handling, maintaining and fixing of glove box and subsidiary, Construction of transport channel for the connection between glove boxes - MOX fabrication by the unique technology in the glove box. Research

  17. Design, Fabrication, and Initial Operation of a Reusable Irradiation Facility

    Energy Technology Data Exchange (ETDEWEB)

    Heatherly, D.W.; Thoms, K.R.; Siman-Tov, I.I.; Hurst, M.T.

    1999-06-01

    A Heavy-Section Steel Irradiation (HSSI) Program project, funded by the US Nuclear Regulatory Commission, was initiated at Oak Ridge National Laboratory to develop reusable materials irradiation facilities in which metallurgical specimens of reactor pressure vessel steels could be irradiated. As a consequence, two new, identical, reusable materials irradiation facilities have been designed, fabricated, installed, and are now operating at the Ford Nuclear Reactor at the University of Michigan. The facilities are referred to as the HSSI-IAR facilities with the individual facilities being designated as IAR-1 and IAR-2. This new and unique facility design requires no cutting or grinding operations to retrieve irradiated specimens, all capsule hardware is totally reusable, and materials transported from site to site are limited to specimens only. At the time of this letter report, the facilities have operated successfully for approximately 2500 effective full-power hours.

  18. 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

  19. Permeation of Military Fuels Through Nitrile-Coated Fabrics Used for Collapsible Fuel Storage Containers

    Science.gov (United States)

    2014-03-01

    materials when challenged with three separate military fuels. ..................................................................8 Figure 6. Effect of...three separate military fuels. The permeation rates for the ASTM Ref Fuel B were significantly larger for six of the nine fabrics. The Bell Avon...the Compatibility of Nitrile Rubber with Brazilian Biodiesel , Energy 2013, 49, 102. 6. Seil, D.; Wolf, F. Nitrile and Polyacrylic Rubbers, Rubber

  20. Gel-sphere-pac reactor fuel fabrication and its application to a variety of fuels

    International Nuclear Information System (INIS)

    Olsen, A.R.; Judkins, R.R.

    1979-12-01

    The gel-sphere-pac fuel fabrication option was evaluated for its possible application to commercial scale fuel fabrication for 19 fuel element designs that use oxide fuel in metal clad rods. The dry gel spheres are prepared at the reprocessing plant and are then calcined, sintered, inspected, and loaded into fuel rods and packed by low-energy vibration. A fuel smear density of 83 to 88% theoretical can be obtained. All fuel fabrication process steps were defined and evaluated from fuel receiving to finished fuel element shipping. The evaluation also covers the feasibility of the process, the current status of technology, estimates of the required time and cost to develop the technology to commercial status, and the safety and licensability of commercial scale plants. The primary evaluation was for a Light-Water Reactor fuel element containing (U,Pu)O 2 fuel. The other 18 fuel element types - 3 for Light-Water Reactors, 1 for a Heavy-Water Reactor, 1 for a Gas-Cooled Fast Reactor, 7 for Liquid-Metal-Cooled Fast Breeder Reactors, and 3 pairs for Light-Water Prebreeder and Breeder Reactors - were compared with the Light-Water Reactor. The gel-sphere-pac option was found applicable to 17 of the 19 element types; the characteristics of a commercial scale plant were defined for these for making cost estimates for such plants. The evaluation clearly shows the gel-sphere-pac process to be a viable fuel fabrication option. Estimates indicate a significant potential fabrication cost advantage for the gel-sphere-pac process if a remotely operated and remotely maintained fuel fabrication plant is required

  1. Fabric-based alkaline direct formate microfluidic fuel cells.

    Science.gov (United States)

    Domalaon, Kryls; Tang, Catherine; Mendez, Alex; Bernal, Franky; Purohit, Krutarth; Pham, Linda; Haan, John; Gomez, Frank A

    2017-04-01

    Fabric-based microfluidic fuel cells (MFCs) serve as a novel, cost-efficient alternative to traditional FCs and batteries, since fluids naturally travel across fabric via capillary action, eliminating the need for an external pump and lowering production and operation costs. Building on previous research with Y-shaped paper-based MFCs, fabric-based MFCs mitigate fragility and durability issues caused by long periods of fuel immersion. In this study, we describe a microfluidic fabric-based direct formate fuel cell, with 5 M potassium formate and 30% hydrogen peroxide as the anode fuel and cathode oxidant, respectively. Using a two-strip, stacked design, the optimized parameters include the type of encasement, the barrier, and the fabric type. Surface contact of the fabric and laminate sheet expedited flow and respective chemical reactions. The maximum current (22.83 mA/cm 2 ) and power (4.40 mW/cm 2 ) densities achieved with a 65% cotton/35% polyester blend material are a respective 8.7% and 32% higher than previous studies with Y-shaped paper-based MFCs. In series configuration, the MFCs generate sufficient energy to power a handheld calculator, a thermometer, and a spectrum of light-emitting diodes. © 2017 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  2. Remotely replaceable fuel and feed nozzles for the new waste calcining facility calciner vessel

    International Nuclear Information System (INIS)

    Fletcher, R.D.; Carter, J.A.; May, K.W.

    1978-01-01

    The development and testing of remotely replaceable fuel and feed nozzles for calcination of liquid radioactive wastes in the calciner vessel of the New Waste Calcining Facility being built at the Idaho National Engineering Laboratory is described. A complete fuel nozzle assembly was fabricated and tested at the Remote Maintenance Development Facility to evolve design refinements, identify required support equipment, and develop handling techniques. The design also provided for remote replacement of the nozzle support carriage and adjacent feed and fuel pipe loops using two pairs of master-slave manipulators

  3. Redundancy of Supply in the International Nuclear Fuel Fabrication Market: Are Fabrication Services Assured?

    International Nuclear Information System (INIS)

    Seward, Amy M.; Toomey, Christopher; Ford, Benjamin E.; Wood, Thomas W.; Perkins, Casey J.

    2011-01-01

    For several years, Pacific Northwest National Laboratory (PNNL) has been assessing the reliability of nuclear fuel supply in support of the U.S. Department of Energy/National Nuclear Security Administration. Three international low enriched uranium reserves, which are intended back up the existing and well-functioning nuclear fuel market, are currently moving toward implementation. These backup reserves are intended to provide countries credible assurance that of the uninterrupted supply of nuclear fuel to operate their nuclear power reactors in the event that their primary fuel supply is disrupted, whether for political or other reasons. The efficacy of these backup reserves, however, may be constrained without redundant fabrication services. This report presents the findings of a recent PNNL study that simulated outages of varying durations at specific nuclear fuel fabrication plants. The modeling specifically enabled prediction and visualization of the reactors affected and the degree of fuel delivery delay. The results thus provide insight on the extent of vulnerability to nuclear fuel supply disruption at the level of individual fabrication plants, reactors, and countries. The simulation studies demonstrate that, when a reasonable set of qualification criteria are applied, existing fabrication plants are technically qualified to provide backup fabrication services to the majority of the world's power reactors. The report concludes with an assessment of the redundancy of fuel supply in the nuclear fuel market, and a description of potential extra-market mechanisms to enhance the security of fuel supply in cases where it may be warranted. This report is an assessment of the ability of the existing market to respond to supply disruptions that occur for technical reasons. A forthcoming report will address political disruption scenarios.

  4. Redundancy of Supply in the International Nuclear Fuel Fabrication Market: Are Fabrication Services Assured?

    Energy Technology Data Exchange (ETDEWEB)

    Seward, Amy M.; Toomey, Christopher; Ford, Benjamin E.; Wood, Thomas W.; Perkins, Casey J.

    2011-11-14

    For several years, Pacific Northwest National Laboratory (PNNL) has been assessing the reliability of nuclear fuel supply in support of the U.S. Department of Energy/National Nuclear Security Administration. Three international low enriched uranium reserves, which are intended back up the existing and well-functioning nuclear fuel market, are currently moving toward implementation. These backup reserves are intended to provide countries credible assurance that of the uninterrupted supply of nuclear fuel to operate their nuclear power reactors in the event that their primary fuel supply is disrupted, whether for political or other reasons. The efficacy of these backup reserves, however, may be constrained without redundant fabrication services. This report presents the findings of a recent PNNL study that simulated outages of varying durations at specific nuclear fuel fabrication plants. The modeling specifically enabled prediction and visualization of the reactors affected and the degree of fuel delivery delay. The results thus provide insight on the extent of vulnerability to nuclear fuel supply disruption at the level of individual fabrication plants, reactors, and countries. The simulation studies demonstrate that, when a reasonable set of qualification criteria are applied, existing fabrication plants are technically qualified to provide backup fabrication services to the majority of the world's power reactors. The report concludes with an assessment of the redundancy of fuel supply in the nuclear fuel market, and a description of potential extra-market mechanisms to enhance the security of fuel supply in cases where it may be warranted. This report is an assessment of the ability of the existing market to respond to supply disruptions that occur for technical reasons. A forthcoming report will address political disruption scenarios.

  5. Outline of a fuel treatment facility in NUCEF

    International Nuclear Information System (INIS)

    Sugikawa, Susumu; Umeda, Miki; Kokusen, Junya

    1997-03-01

    This report presents outline of the nuclear fuel treatment facility for the purpose of preparing solution fuel used in Static Experiment Critical Facility (STACY) and Transient Experiment Critical Facility (TRACY) in Nuclear Fuel Cycle Safety Engineering Research Facility (NUCEF), including descriptions of process conditions and dimensions of major process equipments on dissolution system of oxide fuel, chemical adjustment system, purification system, acid recovery system, solution fuel storage system, and descriptions of safety design philosophy such as safety considerations of criticality, solvent fire, explosion of hydrogen and red-oil and so on. (author)

  6. Chemical process safety at fuel cycle facilities

    International Nuclear Information System (INIS)

    Ayres, D.A.

    1997-08-01

    This NUREG provides broad guidance on chemical safety issues relevant to fuel cycle facilities. It describes an approach acceptable to the NRC staff, with examples that are not exhaustive, for addressing chemical process safety in the safe storage, handling, and processing of licensed nuclear material. It expounds to license holders and applicants a general philosophy of the role of chemical process safety with respect to NRC-licensed materials; sets forth the basic information needed to properly evaluate chemical process safety; and describes plausible methods of identifying and evaluating chemical hazards and assessing the adequacy of the chemical safety of the proposed equipment and facilities. Examples of equipment and methods commonly used to prevent and/or mitigate the consequences of chemical incidents are discussed in this document

  7. Sintering furnace for remote fuel fabrication

    International Nuclear Information System (INIS)

    Bowen, W.W.

    1978-10-01

    Component testing and evaluation of a chemical vapor deposition Re/W muffle has been initiated. Hydrogen permeation testing and thermal cycling behavior will be evaluated. Fabrication of prototype 10-12 Kg furnace is scheduled for completion late in 1979, at which time testing of the system will be initiated

  8. 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

  9. Description of a reference mixed oxide fuel fabrication plant (MOFFP)

    International Nuclear Information System (INIS)

    1978-01-01

    In order to evaluate the environment impact, due to the Mixed Oxide Fuel Fabrication Plants, work has been initiated to describe the general design and operating conditions of a reference Mixed Oxide Fuel Fabrication Plant (MOFFP) for the 1990 time frame. The various reference data and basic assumptions for the reference MOFFP plant have been defined after discussion with experts. The data reported in this document are only made available to allow an evaluation of the environmental impact due to a reference MOFFP plant. These data have therefore not to be used as recommandation, standards, regulatory guides or requirements

  10. Develpment of quality assurance manual for fabrication of DUPIC fuel

    International Nuclear Information System (INIS)

    Lee, Young Gun; Lee, J. W.; Kim, S. S. and others

    2001-09-01

    The Quality Assurance Manual for the fabrication of DUPIC fuel with high quality was developed. The Quality Assurance Policy established by this manual is to assure that the DUPIC fuel element supplied to customer conform to the specified requirements of customer, applicable codes and standards. The management of KAERI is committed to implementation and maintenance of the program described by this manual. This manual describes the quality assurance program for DUPIC fuel fabrication to comply with CAN3-Z299.2-85 to the extent as needed and appropriate. This manual describes the methods which DUPIC Fuel Development Team(DFDT) personnel must follow to achieve and assure high quality of our product. This manual also describes the quality management system applicable to the activities performed at DFDT

  11. Fabrication and control of fuel cans

    International Nuclear Information System (INIS)

    Henckes, J.

    The two requirements stipulated for the cladding tube are (1) to form an initial barrier and this implies a chemical, mechanical, thermal and nuclear resistance, and (2) to be as transparent as possible to neutrons and thermal exchanges. It is shown how these two contradictory conditions led to the choice of Zircaloy 4. The fabrication, which includes the rolling, the thermal treatment and the finishing are described, as is the testing and inspection at each stage of production [fr

  12. Study on the nitride fuel fabrication for FBR cycle (1)

    International Nuclear Information System (INIS)

    Shinkai, Yasuo; Ono, Kiyoshi; Tanaka, Kenya

    2002-07-01

    In the phase-II of JNC's 'Feasibility Study on Commercialized Fuel Reactor Cycle System (the F/S)', the nitride fuels are selected as candidate for fuels for heavy metal cooled reactor, gas cooled reactor, and small scale reactor. In particular, the coated fuel particles are a promising concept for gas cooled reactor. In addition, it is necessary to study in detail the application possibility of pellet nitride fuel and vibration compaction nitride fuel for heavy metal cooled reactor and small scale reactor in the phase-II. In 2001, we studied more about additional equipments for the nitride fuel fabrication in processes from gelation to carbothermic reduction in the vibration compaction method. The result of reevaluation of off-gas mass flow around carbothermic reduction equipment in the palletizing method, showed that quantity of off-gas flow reduced and its reduction led the operation cost to decrease. We studied the possibility of fabrication of large size particles in the coated fuel particles for helium gas cooled reactor and we made basic technical issues clear. (author)

  13. Experiences in transferring of AFA 3G fuel assembly fabrication

    International Nuclear Information System (INIS)

    Yang Xiaodong; Wu Zhiming; Luo Jiankang

    2002-01-01

    Implementation program is developed for the transferring of AFA 3G technology, together with the project management experts designated by Framatome Company, to facilitate the technology import under the guidance of strict program. Technical documents and quality insurance management documents are developed based on the full understanding of the information provided by Framatome to guide the fabrication of AFA 3G fuel elements. Technical requirement suggested by Framatome is adopted as much as possible, considering the practical process capability of YFP. The focus is the technology about fabrication difficulties in the AFA 3G technology, to insure the successful transfer of the AFA 3G fabrication technology

  14. Literature search on Light Water Reactor (LWR) fuel and absorber rod fabrication, 1960--1976

    Energy Technology Data Exchange (ETDEWEB)

    Sample, C R [comp.

    1977-02-01

    A literature search was conducted to provide information supporting the design of a conceptual Light Water Reactor (LWR) Fuel Fabrication plant. Emphasis was placed on fuel processing and pin bundle fabrication, effects of fuel impurities and microstructure on performance and densification, quality assurance, absorber and poison rod fabrication, and fuel pin welding. All data have been taken from publicly available documents, journals, and books. This work was sponsored by the Finishing Processes-Mixed Oxide (MOX) Fuel Fabrication Studies program at HEDL.

  15. Literature search on Light Water Reactor (LWR) fuel and absorber rod fabrication, 1960--1976

    International Nuclear Information System (INIS)

    Sample, C.R.

    1977-02-01

    A literature search was conducted to provide information supporting the design of a conceptual Light Water Reactor (LWR) Fuel Fabrication plant. Emphasis was placed on fuel processing and pin bundle fabrication, effects of fuel impurities and microstructure on performance and densification, quality assurance, absorber and poison rod fabrication, and fuel pin welding. All data have been taken from publicly available documents, journals, and books. This work was sponsored by the Finishing Processes-Mixed Oxide (MOX) Fuel Fabrication Studies program at HEDL

  16. OECD/NEA WGFCS Workshop: Safety Assessment of Fuel Cycle Facilities - Regulatory Approaches and Industry Perspectives

    International Nuclear Information System (INIS)

    2013-01-01

    Nuclear fuel is produced, processed, and stored mainly in industrial-scale facilities. Uranium ores are processed and refined to produce a pure uranium salt stream, Uranium is converted and enriched, nuclear fuel is fabricated (U fuel and U/Pu fuel for the closed cycle option); and spent fuel is stored and reprocessed in some countries (close cycle option). Facilities dedicated to the research and development of new fuel or new processes are also considered as Fuel Cycle Facilities. The safety assessment of nuclear facilities has often been led by the methodology and techniques initially developed for Nuclear Power Plants. As FCFs cover a wide diversity of installations the various approaches of national regulators, and their technical support organizations, for the Safety Assessment of Fuel Cycle Facilities are also diverse, as are the approaches by their industries in providing safety justifications for their facilities. The objective of the Working Group on Fuel Cycle Safety is to advance the understanding for both regulators and operators of relevant aspects of nuclear fuel cycle safety in member countries. A large amount of experience is available in safety assessment of FCFs, which should be shared to develop ideas in this field. To contribute to this task, the Workshop on 'Safety Assessment of Fuel Cycle Facilities - Regulatory Approaches and Industry Perspectives' was held in Toronto, on 27 - 29 September 2011. The workshop was hosted by Canadian Nuclear Safety Commission. The current proceedings provide summary of the results of the workshop with the text of the papers given and presentations made

  17. Brazing process in nuclear fuel element fabrication

    International Nuclear Information System (INIS)

    Katam, K.; Sudarsono

    1982-01-01

    The purpose of the brazing process is to join the spacers and pads of fuel pins, so that the process is meant as a soldering technique and not only as a hardening or reinforcing process such as in common brazing purposes. There are some preliminary processes before executing the brazing process such as: materials preparation, sand blasting, brazing metal coating tack welding the spacers and pads on the fuel cladding. The metal brazing used is beryllium in strip form which will be evaporated in vacuum condition to coat the spacers and pads. The beryllium vapor and dust is very hazardous to the workers, so all the line process of brazing needs specials safety protection and equipment to protect the workers and the processing area. Coating process temperature is 2470 deg C with a vacuum pressure of 10 -5 mmHg. Brazing process temperature process is 1060 deg C with a vacuum pressure of 10 -6 mmHg. The brazing process with beryllium coating probably will give metallurgical structural change in the fuel cladding metal at the locations of spacers and pads. The quality of brazing is highly influenced by and is depending on the chemical composition of the metal and the brazing metal, materials preparations, temperature, vacuum pressure, time of coating and brazing process. The quality control of brazing could be performed with methods of visuality geometry, radiography and metallography. (author)

  18. 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)

  19. Fission product behaviors in spent fuel materials during DUPIC fuel fabrication process

    International Nuclear Information System (INIS)

    Kim, J. H.; Na, S. H.; Lee, J. H.; Yang, M. S.

    2002-01-01

    In order to obtain the fundamental data for the analysis of fission product behaviors during DUPIC fuel fabrication process, which is to convert spent PWR fuel into CANDU reactor fuel, the measurement system of radioactivity in spent fuel materials by gamma spectrometry technique was installed in IMEF M6 hot cell,and the preliminary analysis on the release behaviors of fission gas during the DUPIC fuel fabrication process were conducted. Based on the radioactivity measurement for the spent oxidized powder, green pellet and the sintered pellet produced from DUPIC fabrication process, it was found that little Cs-137 was released during OREOX process, but almost 99% of Cs-137 was released during sintering process. The release rate of both Zr-95 and Ru-103 was not so high during sintering process

  20. Transfer tunnel transporter system for the Fuels and Materials Examination Facility

    International Nuclear Information System (INIS)

    Petty, J.A.; Miller, S.C.; Richards, J.T.

    1981-01-01

    The detail design is complete and fabrication is approximately 75% complete on the Transfer Tunnel Transporter System. This system provides material handling capability for large, bulky equipment between two hot cells in a new Breeder Reactor Program support facility, the Fuels and Materials Examination Facility. One hot cell has an air atmosphere, the other a high purity inert gas atmosphere which must be maintained during transfer operations. System design features, operational capabilities and remote recovery provisions are described

  1. Fabrication and irradiation performance of French HTGR fuel

    International Nuclear Information System (INIS)

    Moreau, C.; Ballagny, A.; Bujas, R.; Holder, J.; Ranc, R.; Soulhier, R.

    1977-01-01

    The fabrication of fuel for high-temperature reactors has been carried out in France either by using the techniques of the General Atomic Company or by adapting European techniques to the characteristics of the American fuel. Currently, the principal efforts are directed towards fuel containing enriched uranium and thorium. Prior to this, a large body of experience had been accumulated on the low enriched fuel of the type used in the Dragon reactor. Installations now in service permit fabrication, at realistic levels, of all the different parts comprising the fuel block. The kernels of fissile particles are made by loading of ion-exchange resin beads and conversion by thermal treatment. The fertile kernels are obtained by dry agglomeration of thorine powder. The coatings, of either BISO or TRISO design (pyrocarbon or pyrocarbon and silicon carbide), are deposited in a fluidized bed in a 127-mm-diameter industrial furnace. The fuel rods (or compacts) are obtained by injecting a pitch loaded with graphite powder into a bed of particles. Another technique, which is now under development, consists of hot-pressing the particles which have been mixed with resin-impregnated graphite powder and then polymerizing them. In both cases, a final thermal treatment is carried out at 1800 0 C. Research and development in the machining of the prismatic graphite blocks has been performed by the Atelier ''Graphite'' at the Marcoule Research Centre. A large programme of irradiations has permitted testing of the individual fuel components followed by tests on complete fuel assemblies. Particles, fuel rods and graphite have been tested in capsules placed in experimental French reactors. Fuel elements have been proof-tested either in the irradiation loops of the Osiris and Pegasus reactors or in the Dragon reactor. The results of these tests have shown that the French fuel can withstand the operating conditions in a power reactor (temperature, fast fluence, burnup) without exceeding the

  2. Design and fabrication procedures of Super-Phenix fuel elements

    International Nuclear Information System (INIS)

    Leclere, J.; Vialard, J.-L.; Delpeyroux, P.

    1975-01-01

    For Super-Phenix fuel assemblies, Phenix technological arrangements will be used again, but they will be simplified as far as possible. The maximum fuel can temperature has been lowered in order to obtain a good behavior of hexagonal tubes and cans at high irradiation levels. An important experimental programme and the experience gained from Phenix operation will confirm the merits of the options retained. The fuel element fabrication is envisaged to take place in the plutonium workshop at Cadarache. Usual procedures will be employed and both reliability and automation will be increased [fr

  3. Fabrication of oxide dispersion strengthened ferritic clad fuel pins

    International Nuclear Information System (INIS)

    Zirker, L.R.; Bottcher, J.H.; Shikakura, S.; Tsai, C.L.

    1991-01-01

    A resistance butt welding procedure was developed and qualified for joining ferritic fuel pin cladding to end caps. The cladding are INCO MA957 and PNC ODS lots 63DSA and 1DK1, ferritic stainless steels strengthened by oxide dispersion, while the end caps are HT9 a martensitic stainless steel. With adequate parameter control the weld is formed without a residual melt phase and its strength approaches that of the cladding. This welding process required a new design for fuel pin end cap and weld joint. Summaries of the development, characterization, and fabrication processes are given for these fuel pins. 13 refs., 6 figs., 1 tab

  4. Nuclear Cryogenic Propulsion Stage Fuel Design and Fabrication

    Science.gov (United States)

    Hickman, Robert; Broadway, Jeramie; Mireles, Omar; Webb, Jon; Qualls, Lou

    2012-01-01

    Nuclear Cryogenic Propulsion Stage (NCPS) is a game changing technology for space exploration. Goal of assessing the affordability and viability of an NCPS includes these overall tasks: (1) Pre-conceptual design of the NCPS and architecture integration (2) NCPS Fuel Design and Testing (3) Nuclear Thermal Rocket Element Environmental Simulator (NTREES) (4) Affordable NCPS Development and Qualification Strategy (5) Second Generation NCPS Concepts. There is a critical need for fuels development. Fuel task objectives are to demonstrate capabilities and critical technologies using full scale element fabrication and testing.

  5. Nuclear Fuel Test Rod Fabrication for Data Acquisition Test

    International Nuclear Information System (INIS)

    Joung, Chang-Young; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho

    2014-01-01

    A nuclear fuel test rod must be fabricated with precise welding and assembly technologies, and confirmed for their soundness. Recently, we have developed various kinds of processing systems such as an orbital TIG welding system, a fiber laser welding system, an automated drilling system and a helium leak analyzer, which are able to fabricate the nuclear fuel test rods and rigs, and keep inspection systems to confirm the soundness of the nuclear fuel test rods and rids. The orbital TIG welding system can be used with two kinds of welding methods. One can perform the round welding for end-caps of a nuclear fuel test rod by an orbital head mounted in a low-pressure chamber. The other can do spot welding for a pin-hole of a nuclear fuel test rod in a high-pressure chamber to fill up helium gas of high pressure. The fiber laser welding system can weld cylindrical and 3 axis samples such as parts of a nuclear fuel test rod and instrumentation sensors which is moved by an index chuck and a 3 axis (X, Y, Z) servo stage controlled by the CNC program. To measure the real-time temperature change at the center of the nuclear fuel during the irradiation test, a thermocouple should be instrumented at that position. Therefore, a hole needs to be made at the center of fuel pellet to instrument the thermocouple. An automated drilling system can drill a fine hole into a fuel pellet without changing tools or breaking the work-piece. The helium leak analyzer (ASM-380 model of DEIXEN Co.) can check the leak of the nuclear fuel test rod filled with helium gas. This paper describes not only the assembly and fabrication methods used by the process systems, but also the results of the data acquisition test for the nuclear fuel test rod. A nuclear fuel test rod for the data acquisition test was fabricated using the welding and assembling echnologies acquired from previous tests

  6. Fabrication details for wire wrapped fuel assembly components

    International Nuclear Information System (INIS)

    Bosy, B.J.

    1978-09-01

    Extensive hydraulic testing of simulated LMFBR blanket and fuel assemblies is being carried out under this MIT program. The fabrication of these test assemblies has involved development of manufacturing procedures involving the wire wrapped pins and the flow housing. The procedures are described in detail in the report

  7. Method to fabricate block fuel elements for high temperature reactors

    International Nuclear Information System (INIS)

    Hrovat, M.; Rachor, L.

    1977-01-01

    The fabrication of block fuel elements for gas-cooled high temperature reactors can be improved upon by adding 0.2 to 2 wt.% of a hydrocarbon compound to the lubricating mixture prior to pressing. Hexanol or octanol are named as substances. The dimensional accuracy of the block is thus improved. 2 examples illustrate the method. (RW) [de

  8. Standard model for safety analysis report of fuel fabrication plants

    International Nuclear Information System (INIS)

    1980-09-01

    A standard model for a safety analysis report of fuel fabrication plants is established. This model shows the presentation format, the origin, and the details of the minimal information required by CNEN (Comissao Nacional de Energia Nuclear) aiming to evaluate the requests of construction permits and operation licenses made according to the legislation in force. (E.G.) [pt

  9. Fabrication of nano-structured UO2 fuel pellets

    International Nuclear Information System (INIS)

    Yang, Jae Ho; Kang, Ki Won; Rhee, Young Woo; Kim, Dong Joo; Kim, Jong Heon; Kim, Keon Sik; Song, Kun Woo

    2007-01-01

    Nano-structured materials have received much attention for their possibility for various functional materials. Ceramics with a nano-structured grain have some special properties such as super plasticity and a low sintering temperature. To reduce the fuel cycle costs and the total mass of spent LWR fuels, it is necessary to extend the fuel discharged burn-up. In order to increase the fuel burn-up, it is important to understand the fuel property of a highly irradiated fuel pellet. Especially, research has focused on the formation of a porous and small grained microstructure in the rim area of the fuel, called High Burn-up Structure (HBS). The average grain size of HBS is about 300nm. This paper deals with the feasibility study on the fabrication of nano-structured UO 2 pellets. The nano sized UO 2 particles are prepared by a combined process of a oxidation-reducing and a mechanical milling of UO 2 powder. Nano-structured UO 2 pellets (∼300nm) with a density of ∼93%TD can be obtained by sintering nano-sized UO 2 compacts. The SEM study reveals that the microstructure of the fabricated nano-structure UO 2 pellet is similar to that of HBS. Therefore, this bulk nano-structured UO 2 pellet can be used as a reference pellet for a measurement of the physical properties of HBS

  10. 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.)

  11. Training development in Juzbado's Fuel Cycle Facility

    International Nuclear Information System (INIS)

    Perez, A.; Cunado, E.; Ortiz, D.

    2003-01-01

    In Juzbado's fuel cycle facility, because of the special activities developed, training is a very important issues. Training has been evolved, due to changes on the standards applicable each moment, and also due to the technological resources available. Both aspects have resulted in an evolution of the documents referred to training, such as training programs procedures, Radiation Protection Manual as well as the teaching methods. In the report we are going to present, we will show more precisely the changes that take place, referring to the different training methods used, present training sanitations, and the improvements already planned in training subjects as well as tools used, accomplishing with the legislation and improving in our effort of a better assimilation of the necessary knowledge. (Author)

  12. Product Conversion: The Link between Separations and Fuel Fabrication

    International Nuclear Information System (INIS)

    Felker, L.K.; Vedder, R.J.; Walker, E.A.; Collins, E.D.

    2008-01-01

    Several chemical processing flowsheets are under development for the separation and isolation of the actinide, lanthanide, and fission product streams in spent nuclear fuel. The conversion of these product streams to solid forms, typically oxides, is desired for waste disposition and recycle of product fractions back into transmutation fuels or targets. The modified direct denitration (MDD) process developed at Oak Ridge National Laboratory (ORNL) in the 1980's offers significant advantages for the conversion of the spent fuel products to powder form suitable for direct fabrication into recycle fuels. A glove-box-contained MDD system and a fume-hood-contained system have been assembled at ORNL for the purposes of testing the co-conversion of uranium and mixed-actinide products. The current activities are focused on the conversion of the first products from the processing of spent nuclear fuel in the Coupled End-to-End Demonstration currently being conducted at ORNL. (authors)

  13. Establishing a LEU MTR fuel manufacturing facility in South Africa

    International Nuclear Information System (INIS)

    Jamie, R.W.; Kocher, A.

    2010-01-01

    The South African MTR Fuel Manufacturing Facility was established in the 1970's to supply SAFARI-1 with Fuel Elements and Control Rods. South African capability was developed in parallel with the uranium enrichment program to meet the needs of the Reactor. Further to the July 2005 decision by the South African Governmnent to convert both SAFARI-1 and the Fuel Plant to LEU, the SAFARI-1 phase has been successfully completed and Necsa has commenced with the conversion of the MTR Fuel Manufacturing Facility. In order to establish, validate and qualify the facility, Necsa has entered into a co-operation and technology transfer agreement with AREVA CERCA, the French manufacturer of Research Reactor fuel elements. Past experiences, conversion challenges and the status of the MTR Fuel Facility Project are discussed. On-going co-operation with AREVA CERCA to implement the local manufacture of LEU fuel is explained and elaborated on. (author)

  14. Advanced methods for fabrication of PHWR and LMFBR fuels

    International Nuclear Information System (INIS)

    Ganguly, C.

    1988-01-01

    For self-reliance in nuclear power, the Department of Atomic Energy (DAE), India is pursuing two specific reactor systems, namely the pressurised heavy water reactors (PHWR) and the liquid metal cooled fast breeder reactors (LMFBR). The reference fuel for PHWR is zircaloy-4 clad high density (≤ 96 per cent T.D.) natural UO 2 pellet-pins. The advanced PHWR fuels are UO 2 -PuO 2 (≤ 2 per cent), ThO 2 -PuO 2 (≤ 4 per cent) and ThO 2 -U 233 O 2 (≤ 2 per cent). Similarly, low density (≤ 85 per cent T.D.) (UPu)O 2 pellets clad in SS 316 or D9 is the reference fuel for the first generation of prototype and commercial LMFBRs all over the world. However, (UPu)C and (UPu)N are considered as advanced fuels for LMFBRs mainly because of their shorter doubling time. The conventional method of fabrication of both high and low density oxide, carbide and nitride fuel pellets starting from UO 2 , PuO 2 and ThO 2 powders is 'powder metallurgy (P/M)'. The P/M route has, however, the disadvantage of generation and handling of fine powder particles of the fuel and the associated problem of 'radiotoxic dust hazard'. The present paper summarises the state-of-the-art of advanced methods of fabrication of oxide, carbide and nitride fuels and highlights the author's experience on sol-gel-microsphere-pelletisation (SGMP) route for preparation of these materials. The SGMP process uses sol gel derived, dust-free and free-flowing microspheres of oxides, carbide or nitride for direct pelletisation and sintering. Fuel pellets of both low and high density, excellent microhomogeneity and controlled 'open' or 'closed' porosity could be fabricated via the SGMP route. (author). 5 tables, 14 figs., 15 refs

  15. Nuclear fuel cycle facilities and RP: the case of Brazil

    International Nuclear Information System (INIS)

    Tranjan Filho, Alfredo; Costa, Cesar Gustavo S.

    2008-01-01

    Full text: The renewed nuclear energy scenario, national and worldwide, calls for the strengthening of all activities involving the nuclear fuel production, from uranium extraction at the mines to fuel assemblies delivery at the nuclear power plants, which in Brazil is the mission of the Industrias Nucleares do Brasil (INB). With only a third of its territory prospected, Brazil currently has the sixth largest uranium reserve in the world. Brazil's three main deposits are: the Caldas mine (in the state of Minas Gerais) the first mineral-industrial complex that processed uranium, developed in 1982, and presently being decommissioned; Caetite mine and processing facility (located in the state of Bahia), nowadays operational and with a current production capacity of 400 tonnes per year of uranium concentrates, being in trend of doubling its annual capacity; and the Itataia/Santa Quiteria deposit (in Ceara State), the largest geological uranium reserve in Brazil, although its feasible future production depends on the exploration of the phosphate associated to it. Concerning the nuclear fuel fabrication, INB plant at Resende (in the state of Rio de Janeiro) is responsible for the conversion of Uf 6 to UO 2 the production of fuel pellets and the assembly of the fuel elements, in order to supply the demands of Brazil's two operating PWR (Angra 1 and Angra 2). In addition, in May 2006, INB-Resende inaugurated the uranium enrichment facility, employing the ultra-centrifugation technology. Today still in its first phase of operation, when completed the enrichment facility is intended to provide 100 percent of the domestic requirements, eventually by the year 2015. Detailing present status and future perspectives of INB, in face of the global and national renaissance of nuclear energy, this paper addresses the Radiation Protection (RP) aspects related to INB's achievements and performance, as well as the pressing future challenges to be dealt with, in order to guarantee

  16. 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

  17. Fabrication of TREAT Fuel with Increased Graphite Loading

    Energy Technology Data Exchange (ETDEWEB)

    Luther, Erik Paul [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Leckie, Rafael M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dombrowski, David E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Papin, Pallas A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-02-05

    As part of the feasibility study exploring the replacement of the HEU fuel core of the TREAT reactor at Idaho National Laboratory with LEU fuel, this study demonstrates that it is possible to increase the graphite content of extruded fuel by reformulation. The extrusion process was use to fabricate the “upgrade” core1 for the TREAT reactor. The graphite content achieved is determined by calculation and has not been measured by any analytical method. In conjunction, a technique, Raman Spectroscopy, has been investigated for measuring the graphite content. This method shows some promise in differentiating between carbon and graphite; however, standards that would allow the technique to be calibrated to quantify the graphite concentration have yet to be fabricated. Continued research into Raman Spectroscopy is on going. As part of this study, cracking of graphite extrusions due to volatile evolution during heat treatment has been largely eliminated. Continued research to optimize this extrusion method is required.

  18. Development of challengeable reprocessing and fuel fabrication technologies for advanced fast reactor fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Nomura, S.; Aoshima, T.; Myochin, M. [Japan Nuclear Cycle Development Institute, Tokai Works (Japan)

    2001-07-01

    R and D in the next five years in Feasibility Study Phase-2 are focused on selected key technologies for the advanced fuel cycle. These are the reference technology of simplified aqueous extraction and fuel pellet short process based on the oxide fuel and the innovative technology of oxide-electrowinning and metal- electrorefining process and their direct particle/metal fuel fabrication methods in a hot cell. Automatic and remote handling system operation in both reprocessing and fuel manufacturing can handle MA and LLFP concurrently with Pu and U attaining the highest recovery and an accurate accountability of these materials. (author)

  19. Development of challengeable reprocessing and fuel fabrication technologies for advanced fast reactor fuel cycle

    International Nuclear Information System (INIS)

    Nomura, S.; Aoshima, T.; Myochin, M.

    2001-01-01

    R and D in the next five years in Feasibility Study Phase-2 are focused on selected key technologies for the advanced fuel cycle. These are the reference technology of simplified aqueous extraction and fuel pellet short process based on the oxide fuel and the innovative technology of oxide-electrowinning and metal- electrorefining process and their direct particle/metal fuel fabrication methods in a hot cell. Automatic and remote handling system operation in both reprocessing and fuel manufacturing can handle MA and LLFP concurrently with Pu and U attaining the highest recovery and an accurate accountability of these materials. (author)

  20. Comparison of concepts for independent spent fuel storage facilities

    International Nuclear Information System (INIS)

    Held, Ch.; Hintermayer, H.P.

    1978-01-01

    The design and the construction costs of independent spent fuel storage facilities show significant differences, reflecting the fuel receiving rate (during the lifetime of the power plant or within a very short period), the individual national policies and the design requirements in those countries. Major incremental construction expenditures for storage facilities originate from the capacity and the type of the facilities (casks or buildings), the method of fuel cooling (water or air), from the different design of buildings, the redundancy of equipment, an elaborate quality assurance program, and a single or multipurpose design (i.e. interim or long-term storage of spent fuel, interim storage of high level waste after fuel storage). The specific costs of different designs vary by a factor of 30 to 60 which might in the high case increase the nuclear generating costs remarkably. The paper also discusses the effect of spent fuel storage on fuel cycle alternatives with reprocessing or disposal of spent fuel. (author)

  1. Nuclear fuel cycle facility accident analysis handbook

    International Nuclear Information System (INIS)

    1998-03-01

    The purpose of this Handbook is to provide guidance on how to calculate the characteristics of releases of radioactive materials and/or hazardous chemicals from nonreactor nuclear facilities. In addition, the Handbook provides guidance on how to calculate the consequences of those releases. There are four major chapters: Hazard Evaluation and Scenario Development; Source Term Determination; Transport Within Containment/Confinement; and Atmospheric Dispersion and Consequences Modeling. These chapters are supported by Appendices, including: a summary of chemical and nuclear information that contains descriptions of various fuel cycle facilities; details on how to calculate the characteristics of source terms for releases of hazardous chemicals; a comparison of NRC, EPA, and OSHA programs that address chemical safety; a summary of the performance of HEPA and other filters; and a discussion of uncertainties. Several sample problems are presented: a free-fall spill of powder, an explosion with radioactive release; a fire with radioactive release; filter failure; hydrogen fluoride release from a tankcar; a uranium hexafluoride cylinder rupture; a liquid spill in a vitrification plant; and a criticality incident. Finally, this Handbook includes a computer model, LPF No.1B, that is intended for use in calculating Leak Path Factors. A list of contributors to the Handbook is presented in Chapter 6. 39 figs., 35 tabs

  2. Nuclear fuel cycle facility accident analysis handbook

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-01

    The purpose of this Handbook is to provide guidance on how to calculate the characteristics of releases of radioactive materials and/or hazardous chemicals from nonreactor nuclear facilities. In addition, the Handbook provides guidance on how to calculate the consequences of those releases. There are four major chapters: Hazard Evaluation and Scenario Development; Source Term Determination; Transport Within Containment/Confinement; and Atmospheric Dispersion and Consequences Modeling. These chapters are supported by Appendices, including: a summary of chemical and nuclear information that contains descriptions of various fuel cycle facilities; details on how to calculate the characteristics of source terms for releases of hazardous chemicals; a comparison of NRC, EPA, and OSHA programs that address chemical safety; a summary of the performance of HEPA and other filters; and a discussion of uncertainties. Several sample problems are presented: a free-fall spill of powder, an explosion with radioactive release; a fire with radioactive release; filter failure; hydrogen fluoride release from a tankcar; a uranium hexafluoride cylinder rupture; a liquid spill in a vitrification plant; and a criticality incident. Finally, this Handbook includes a computer model, LPF No.1B, that is intended for use in calculating Leak Path Factors. A list of contributors to the Handbook is presented in Chapter 6. 39 figs., 35 tabs.

  3. Engineered safeguards system activities at Sandia Laboratories for back-end fuel cycle facilities

    International Nuclear Information System (INIS)

    Sellers, T.A.; Fienning, W.C.; Winblad, A.E.

    1978-01-01

    Sandia Laboratories have been developing concepts for safeguards systems to protect facilities in the back-end of the nuclear fuel cycle against potential threats of sabotage and theft of special nuclear material (SNM). Conceptual designs for Engineered Safeguards Systems (ESSs) have been developed for a Fuel Reprocessing Facility (including chemical separations, plutonium conversion, and waste solidification), a Mixed-Oxide Fuel Fabrication Facility, and a Plutonium Transport Vehicle. Performance criteria for the various elements of these systems and a candidate systematic design approach have been defined. In addition, a conceptual layout for a large-scale Fuel-Cycle Plutonium Storage Facility has been completed. Work is continuing to develop safeguards systems for spent fuel facilities, light-water reactors, alternative fuel cycles, and improved transportation systems. Additional emphasis will be placed on the problems associated with national diversion of special nuclear material. The impact on safeguards element performance criteria for surveillance and containment to protect against national diversion in various alternative fuel cycle complexes is also being investigated

  4. Fabrication of metallic channel-containing UO2 fuels

    International Nuclear Information System (INIS)

    Yang, Jae Ho; Song, Kun Woo; Kim, Keon Sik; Jung, Youn Ho

    2004-01-01

    The uranium dioxide is widely used as a fuel material in the nuclear industry, owing to many advantages. But it has a disadvantage of having the lowest thermal conductivity of all kinds of nuclear fuels; metal, carbide, nitride. It is well known that the thermal conductivity of UO 2 fuel is enhanced by making, so called, the CERMET (ceramic-metal) composite which consists of both continuous body of highly thermal-conducting metal and UO 2 islands. The CERMET fuel fabrication technique needs metal phase of at least 30%, mostly more than 50%, of the volume of the pellet in order to keep the metal phase interconnected. This high volume fraction of metal requires such a high enrichment of U that the parasitic effect of metal should be compensated. Therefore, it is attractive to develop an innovative composite fuel that can form continuous metal phase with a small amount of metal. In this investigation, a feasibility study was made on how to make such an innovative fuel. Candidate metals (W, Mo, Cr) were selected, and fabrication process was conceptually designed from thermodynamic calculations. We have experimentally found that a metal phase envelops perfectly UO 2 grains, forming continuous channel throughout the pellet, and improving the thermal conductivity of pellet

  5. U-10Mo Baseline Fuel Fabrication Process Description

    Energy Technology Data Exchange (ETDEWEB)

    Hubbard, Lance R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Arendt, Christina L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Dye, Daniel F. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Clayton, Christopher K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lerchen, Megan E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lombardo, Nicholas J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lavender, Curt A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Zacher, Alan H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2017-09-27

    This document provides a description of the U.S. High Power Research Reactor (USHPRR) low-enriched uranium (LEU) fuel fabrication process. This document is intended to be used in conjunction with the baseline process flow diagram (PFD) presented in Appendix A. The baseline PFD is used to document the fabrication process, communicate gaps in technology or manufacturing capabilities, convey alternatives under consideration, and as the basis for a dynamic simulation model of the fabrication process. The simulation model allows for the assessment of production rates, costs, and manufacturing requirements (manpower, fabrication space, numbers and types of equipment, etc.) throughout the lifecycle of the USHPRR program. This document, along with the accompanying PFD, is updated regularly

  6. 76 FR 44049 - Guidance for Fuel Cycle Facility Change Processes

    Science.gov (United States)

    2011-07-22

    ... COMMISSION Guidance for Fuel Cycle Facility Change Processes AGENCY: Nuclear Regulatory Commission. ACTION... for Fuel Cycle Facility Change Processes'' in the Federal Register for a 30 day public comment period... Change Processes.'' By e-mail ] dated July 7, 2011, the Nuclear Energy Institute (ADAMS Accession No...

  7. 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)

  8. Remote handling technology for nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    Sakai, Akira; Maekawa, Hiromichi; Ohmura, Yutaka

    1997-01-01

    Design and R and D on nuclear fuel cycle facilities has intended development of remote handling and maintenance technology since 1977. IHI has completed the design and construction of several facilities with remote handling systems for Power Reactor and Nuclear Fuel Development Corporation (PNC), Japan Atomic Energy Research Institute (JAERI), and Japan Nuclear Fuel Ltd. (JNFL). Based on the above experiences, IHI is now undertaking integration of specific technology and remote handling technology for application to new fields such as fusion reactor facilities, decommissioning of nuclear reactors, accelerator testing facilities, and robot simulator-aided remote operation systems in the future. (author)

  9. Final safety analysis report for the irradiated fuels storage facility

    International Nuclear Information System (INIS)

    Bingham, G.E.; Evans, T.K.

    1976-01-01

    A fuel storage facility has been constructed at the Idaho Chemical Processing Plant to provide safe storage for spent fuel from two commercial HTGR's, Fort St. Vrain and Peach Bottom, and from the Rover nuclear rocket program. The new facility was built as an addition to the existing fuel storage basin building to make maximum use of existing facilities and equipment. The completed facility provides dry storage for one core of Peach Bottom fuel (804 elements), 1 1 / 2 cores of Fort St. Vrain fuel (2200 elements), and the irradiated fuel from the 20 reactors in the Rover program. The facility is designed to permit future expansion at a minimum cost should additional storage space for graphite-type fuels be required. A thorough study of the potential hazards associated with the Irradiated Fuels Storage Facility has been completed, indicating that the facility is capable of withstanding all credible combinations of internal accidents and pertinent natural forces, including design basis natural phenomena of a 10,000 year flood, a 175-mph tornado, or an earthquake having a bedrock acceleration of 0.33 g and an amplification factor of 1.3, without a loss of integrity or a significant release of radioactive materials. The design basis accident (DBA) postulated for the facility is a complete loss of cooling air, even though the occurrence of this situation is extremely remote, considering the availability of backup and spare fans and emergency power. The occurrence of the DBA presents neither a radiation nor an activity release hazard. A loss of coolant has no effect upon the fuel or the facility other than resulting in a gradual and constant temperature increase of the stored fuel. The temperature increase is gradual enough that ample time (28 hours minimum) is available for corrective action before an arbitrarily imposed maximum fuel centerline temperature of 1100 0 F is reached

  10. Technology of Fabrication for Sodium-cooled Fast Reactor Metallic Fuel

    International Nuclear Information System (INIS)

    Oh, S. J.; Kim, K. H.; Lee, C. T.; Ryu, H. J.; Ko, Y. M.; Woo, W. M.; Jang, S. J.; Lee, Y. S.; Lee, C. B.

    2008-02-01

    The fabrication process of metallic fuel for SFR(sodium fast reactor) of Generation-IV candidate reactors is composed of the fabrication of fuel pin, fuel rod, and fuel assembly. The key technology of the fabrication process for SFR can be referred to the fabrication technology of fuel pin. As SFR fuel contains MA(minor actinide) elements proceeding the recycling of actinide elements, it is so important to extinguish MA during irradiation in SFR, included in nuclear fuel through collection of volatile MA elements during fabrication of fuel pin. Hence, it is an imminent circumstance to develop the fabrication process of fuel pin. This report is an state-of art report related to the characteristics of irradiation performance for U-Zr- Pu metallic fuel, and the apparatus and the technology of conventional injection casting process. In addition, to overcome the drawbacks of the conventional injection casting and the U-Zr-Pu fuel, new fabrication technologies such as the gravity casting process, the casting of fuel pin to metal-barrier mold, the fabrication of particulate metallic fuel utilizing centrifugal atomization is surveyed and summarized. The development of new U-10Mo-X metallic fuel as nuclear fuel having a single phase in the temperature range between 550 and 950 .deg. C, reducing the re-distribution of the fuel elements and improving the compatibility between fuel and cladding, is also surveyed and summarized

  11. Fabrication and use of zircaloy/tantalum-sheathed cladding thermocouples and molybdenum/rhenium-sheathed fuel centerline thermocouples

    International Nuclear Information System (INIS)

    Wilkins, S.C.; Sepold, L.K.

    1985-01-01

    The thermocouples described in this report are zircaloy/tantalum-sheathed and molybdenum/rhenium alloy-sheathed instruments intended for fuel rod cladding and fuel centerline temperature measurements, respectively. Both types incorporate beryllium oxide insulation and tungsten/rhenium alloy thermoelements. These thermocouples, operated at temperatures of 2000 0 C and above, were developed for use in the internationally sponsored Severe Fuel Damage test series in the Power Burst Facility. The fabrication steps for both thermocouple types are described in detail. A laser-welding attachment technique for the cladding-type thermocouple is presented, and experience with alternate materials for cladding and fuel therocouples is discussed

  12. Radiological health aspects of commercial uranium conversion, enrichment, and fuel fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Stoetzel, G.A.; Hoenes, G.R.; Cummings, F.M.; McCormack, W.D.

    1982-11-01

    Detailed information concerning occupational exposures, health physics practices, and regulatory procedures at commercial conversion, enrichment and fuel fabrication facilities is given. Sites visits were the primary source of information, which is divided into four sections. The first section discusses health physics practices that are common to the conversion, enrichment, and fuel fabrication phases of the commercial uranium industry. The next three sections review process descriptions, radiological health practices, and regulatory procedures for the three phases. Nonradiological exposures are considered only as they influence the interpretation of the health effects of radiological exposures. The review of regulatory procedures indicates the types of exposure evaluation records being kept on uranium workers and the responsibility for maintaining the records.

  13. 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)

  14. Uranium accountability for ATR fuel fabrication. Part I. A description of the existing system

    International Nuclear Information System (INIS)

    Dolan, C.A.; Nieschmidt, E.B.; Vegors, S.H. Jr.; Wagner, E.P. Jr.

    1977-06-01

    An evaluation of the materials accountability program at the Atomics International fuel fabrication facility in Canoga Park, California, with regard to the fabrication of highly enriched uranium fuel for the Advanced Test Reactor is presented. An analysis is given of the existing standards program, the existing measurements program and the existing statistical analysis procedures. In addition a short discussion is given of our evaluation of the safeguards procedures at Atomics International together with suggestions for possible modifications and improvements. Appendices of this report contain a rather complete description of the Atomics International plant and the flow of highly enriched uranium through the plant as well as the principal documents used for material accountability records

  15. Spent fuel receipt and lag storage facility for the spent fuel handling and packaging program

    International Nuclear Information System (INIS)

    Black, J.E.; King, F.D.

    1979-01-01

    Savannah River Laboratory (SRL) is participating in the Spent Fuel Handling and Packaging Program for retrievable, near-surface storage of spent light water reactor (LWR) fuel. One of SRL's responsibilities is to provide a technical description of the wet fuel receipt and lag storage part of the Spent Fuel Handling and Packaging (SFHP) facility. This document is the required technical description

  16. Regulations concerning the fabricating business of nuclear fuel materials

    International Nuclear Information System (INIS)

    1987-01-01

    Regulations specified here cover application for such matters as permission for an undertaking of processing, alteration (of location, structure, arrangements, processing method, etc.), approval of design and construction plan, approval of alteration (of design and construction plan of processing facilities), and inspection of the facilities. The regulations also cover execution of facilities inspection, certificate of facilities inspection, processing facilities subject to welding inspection, application for welding inspection, execution of welding inspection, facilities not subject to welding inspection, approval of welding method, welding inspection for imported equipment, certificate of welding inspection, application for approval of joint management, notice of alteration, etc., cancellation of permission, record keeping, restriction on access to areas under management measures concerning exposure to radioactive rays, patrol and checking in processing facilities, operation of processing equipment, transportation within plant or operation premises, storage, waste disposal within plant or operation premises, safety rules, public notification concerning examination and successful applicants, procedure for application for examination, reissue of certificate for nuclear fuel handling expert, return of certificate for nuclear fuel handling expert, submission of report, measures for emergency, notice of abolition of business, measures concerning cancellation of permission, identification card, etc. (Nogami, K.)

  17. 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.)

  18. Fuel conditioning facility electrorefiner start-up results

    International Nuclear Information System (INIS)

    Goff, K.M.; Mariani, R.D.; Vaden, D.; Bonomo, N.L.; Cunningham, S.S.

    1996-01-01

    At ANL-West, there are several thousand kilograms of metallic spent nuclear fuel containing bond sodium. This fuel will be treated in the Fuel Conditioning Facility (FCF) at ANL-West to produce stable waste forms for storage and disposal. The treatment operations will make use of an electrometallurgical process employing molten salts and liquid metals. The treatment equipment is presently undergoing testing with depleted uranium. Operations with irradiated fuel will commence when the environmental evaluation for FCF is complete

  19. Method to fabricate high performance tubular solid oxide fuel cells

    Science.gov (United States)

    Chen, Fanglin; Yang, Chenghao; Jin, Chao

    2013-06-18

    In accordance with the present disclosure, a method for fabricating a solid oxide fuel cell is described. The method includes forming an asymmetric porous ceramic tube by using a phase inversion process. The method further includes forming an asymmetric porous ceramic layer on a surface of the asymmetric porous ceramic tube by using a phase inversion process. The tube is co-sintered to form a structure having a first porous layer, a second porous layer, and a dense layer positioned therebetween.

  20. Fabrication of small-orifice fuel injectors for diesel engines.

    Energy Technology Data Exchange (ETDEWEB)

    Woodford, J. B.; Fenske, G. R.

    2005-04-08

    Diesel fuel injector nozzles with spray hole diameters of 50-75 {micro}m have been fabricated via electroless nickel plating of conventionally made nozzles. Thick layers of nickel are deposited onto the orifice interior surfaces, reducing the diameter from {approx}200 {micro}m to the target diameter. The nickel plate is hard, smooth, and adherent, and covers the orifice interior surfaces uniformly.

  1. 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

  2. Achievement report (edition B) for fiscal 1999 on development of technology to manufacture coal gas for fuel cells. Studies by using pilot test facilities (Volumes for equipment fabrication and constructions, and trial run design); 1999 nendo seika hokokusho (B ban). Nenryo denchi you sekitan gas seizo gijutsu kaihatsu - Pilot setsubi ni yoru kenkyu (kisokoji kiki seisaku hen shiunten sekkei hen)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-03-01

    With an objective to develop a coal gas manufacturing system for fuel cells, research and development has been performed on a oxygen-blown coal gasifier and researches on a technology to purify gases for fuel cells. This paper summarizes the achievements in fiscal 1999. The current fiscal year has performed, among other the pilot plant construction works, execution of the above-the-ground constructions for the operation center and compressor room building, construction of the cooling water tanks, and partial improvement of roads in the plant site. In the gasifier facilities, items of equipment were fabricated, some of the outsourced articles were procured, and the installations thereof were carried out. For the gas purifying equipment, installation of the gas analyzer room was executed. In the trial run design, discussions were given on the systematic improvements in the test items, the gas sampling procedures, the unit protecting interlock, and the facility protecting logic. For the trial run design, establishment has been implemented on the efficient and functional test plans by establishing priority on the tests to be executed, so that the development items demanded in the pilot test and research can be achieved within the limited test processes. (NEDO)

  3. Nitration by a simulated fuel technique for nitride fuel re-fabrication

    International Nuclear Information System (INIS)

    Lee, Young-Woo; Ryu, Ho Jin; Lee, Jae Won; Lee, Jung Won; Park, Geun Il

    2009-01-01

    Nitration reaction of a spent nuclear oxide fuel through a carbothermic reduction and the change in thermal conductivity of the resultant nitride specimens were investigated by a simulated fuel technique for use in nitride fuel re-fabrication from spent oxide fuel. The simulated spent oxide fuel was formed by compacting and sintering a powder mixture of UO 2 and stable oxide fission product impurities. It was pulverized by a 3-cycle successive oxidation-reduction treatment and converted into nitride pellet specimens through the carbothermic reduction. The rate of the nitration reaction of the simulated spent oxide fuel was decreased due to the fission product impurities when compared with pure uranium dioxide. The amount of Ba and Sr in the simulated spent oxide fuel was considerably reduced after the nitride fuel re-fabrication. The thermal conductivity of the nitride pellet specimen in the range 295-373 K was lower than that of the pure uranium nitride but higher than the simulated spent oxide fuel containing fission product impurities.

  4. A Facile Way to Fabricate Transparent Superhydrophobic Surfaces.

    Science.gov (United States)

    Shi, Wentao; He, Ran; Yunus, Doruk E; Yang, Jie; Liu, Yaling

    2018-07-01

    A fast, easy, and low-cost way to fabricate transparent superhydrophobic (SHP) surfaces is developed. By simply mixing silica nanoparticles (SiNPs), polydimethylsiloxane (PDMS) and heptane to form a suspension, dip- or drop-coating the suspension onto different surfaces, transparent SHP surfaces can be obtained. By tuning the ratio of the three components above, transparency of the coating can reach more than 90% transmittance in the visible region, while static water contact angle of the coating can reach as high as 162°. Dynamic contact angle study shows the advancing contact angle and receding contact angle of water can be as high as 168° and 161°, and the resulting contact angle hysteresis can be as low as 7°. The reported facile way of fabricating transparent superhydrophobic (SHP) surfaces is potential for applications which need both optical transparency and self-cleaning capability, such as solar cells, optical equipment, and visible microfluidic chips.

  5. 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)

  6. Conceptual development of a test facility for spent fuel management

    International Nuclear Information System (INIS)

    Park, S.W.; Lee, H.H.; Lee, J.Y.; Lee, J.S.; Ro, S.G.

    1997-01-01

    Spent fuel management is an important issue for nuclear power program, requiring careful planning and implementation. With the wait-and-see policy on spent fuel management in Korea, research efforts are directed at KAERI to develop advanced technologies for safer and more efficient management of the accumulating spent fuels. In support of these research perspectives, a test facility of pilot scale is being developed with provisions for integral demonstration of a multitude of technical functions required for spent fuel management. The facility, baptized SMART (Spent fuel MAnagement technology Research and Test facility), is to be capable of handling full size assembly of spent PWR fuel (as well as CANDU fuel) with a maximum capacity of 10 MTU/y (about 24 assemblies of PWR type). Major functions of the facility are consolidation of spent PWR fuel assembly into a half-volume package and optionally transformation of the fuel rod into a fuel of CANDU type (called DUPIC). Objectives of these functions are to demonstrate volume reduction of spent fuel (for either longer-term dry storage or direct disposal ) in the former case and direct refabrication of the spent PWR fuel into CANDU-type DUPIC fuel for reuse in CANDU reactors in the latter case, respectively. In addition to these major functions, there are other associated technologies to be demonstrated : such as waste treatment, remote maintenance, safeguards, etc. As the facility is to demonstrate not only the functional processes but also the safety and efficiency of the test operations, engineering criteria equivalent to industrial standards are incorporated in the design concept. The hot cell structure enclosing the radioactive materials is configured in such way to maximize costs within the given functional and operational requirements. (author). 3 tabs., 4 figs

  7. Nuclear-fuel-cycle risk assessment: descriptions of representative non-reactor facilities. Sections 1-14

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, K.J.

    1982-09-01

    The Fuel Cycle Risk Assessment Program was initiated to provide risk assessment methods for assistance in the regulatory process for nuclear fuel cycle facilities other than reactors. This report, the first from the program, defines and describes fuel cycle elements that are being considered in the program. One type of facility (and in some cases two) is described that is representative of each element of the fuel cycle. The descriptions are based on real industrial-scale facilities that are current state-of-the-art, or on conceptual facilities where none now exist. Each representative fuel cycle facility is assumed to be located on the appropriate one of four hypothetical but representative sites described. The fuel cycles considered are for Light Water Reactors with once-through flow of spent fuel, and with plutonium and uranium recycle. Representative facilities for the following fuel cycle elements are described for uranium (or uranium plus plutonium where appropriate): mining, milling, conversion, enrichment, fuel fabrication, mixed-oxide fuel refabrication, fuel reprocessing, spent fuel storage, high-level waste storage, transuranic waste storage, spent fuel and high-level and transuranic waste disposal, low-level and intermediate-level waste disposal, and transportation. For each representative facility the description includes: mainline process, effluent processing and waste management, facility and hardware description, safety-related information and potential alternative concepts for that fuel cycle element. The emphasis of the descriptive material is on safety-related information. This includes: operating and maintenance requirements, input/output of major materials, identification and inventories of hazardous materials (particularly radioactive materials), unit operations involved, potential accident driving forces, containment and shielding, and degree of hands-on operation.

  8. Nuclear-fuel-cycle risk assessment: descriptions of representative non-reactor facilities. Sections 1-14

    International Nuclear Information System (INIS)

    Schneider, K.J.

    1982-09-01

    The Fuel Cycle Risk Assessment Program was initiated to provide risk assessment methods for assistance in the regulatory process for nuclear fuel cycle facilities other than reactors. This report, the first from the program, defines and describes fuel cycle elements that are being considered in the program. One type of facility (and in some cases two) is described that is representative of each element of the fuel cycle. The descriptions are based on real industrial-scale facilities that are current state-of-the-art, or on conceptual facilities where none now exist. Each representative fuel cycle facility is assumed to be located on the appropriate one of four hypothetical but representative sites described. The fuel cycles considered are for Light Water Reactors with once-through flow of spent fuel, and with plutonium and uranium recycle. Representative facilities for the following fuel cycle elements are described for uranium (or uranium plus plutonium where appropriate): mining, milling, conversion, enrichment, fuel fabrication, mixed-oxide fuel refabrication, fuel reprocessing, spent fuel storage, high-level waste storage, transuranic waste storage, spent fuel and high-level and transuranic waste disposal, low-level and intermediate-level waste disposal, and transportation. For each representative facility the description includes: mainline process, effluent processing and waste management, facility and hardware description, safety-related information and potential alternative concepts for that fuel cycle element. The emphasis of the descriptive material is on safety-related information. This includes: operating and maintenance requirements, input/output of major materials, identification and inventories of hazardous materials (particularly radioactive materials), unit operations involved, potential accident driving forces, containment and shielding, and degree of hands-on operation

  9. 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

  10. Facile Fabrication of Ultrafine Copper Nanoparticles in Organic Solvent

    Directory of Open Access Journals (Sweden)

    Siegert Uwe

    2009-01-01

    Full Text Available Abstract A facile chemical reduction method has been developed to fabricate ultrafine copper nanoparticles whose sizes can be controlled down to ca. 1 nm by using poly(N-vinylpyrrolidone (PVP as the stabilizer and sodium borohyrdride as the reducing agent in an alkaline ethylene glycol (EG solvent. Transmission electron microscopy (TEM results and UV–vis absorption spectra demonstrated that the as-prepared particles were well monodispersed, mostly composed of pure metallic Cu nanocrystals and extremely stable over extended period of simply sealed storage.

  11. Research on plant of metal fuel fabrication using casting process (2)

    International Nuclear Information System (INIS)

    Senda, Yasuhide; Yamada, Seiya

    2005-02-01

    In this research work for the metal fuel fabrication system (38 tHM/y), the studies of the concept of the main process equipments were performed based on the previous studies on the process design and the quality control system design. In this study the handling equipment of the products were also designed, according to these designs the handling periods were evaluated. Consequently the numbers of the equipments were assessed taking into account for the method of the blending the fuel composition. (1) Structural concept design of the major equipments, the fuel handling machine and the gravimetries in the main fabrication process. The structural concept were designed for the fuel composition blending equipment, the fuel pin assembling equipment, the sodium bonding equipment, the handling equipment for fuel slug palettes, the handling equipment for fuel pins and the gravimetries. (2) Re-assessment of the numbers of the equipments taking account of the handling periods. Based on the results of item (1) the periods were evaluated for the fuel slug and pin handling. Processing time of demolder is short, then the number of it is increased to two. Three vehicles are also added to transfer the slugs and a heel smoothly. (3) Design of the buffer storages. The buffer storages among the equipments were designed through the comparison of the process speed between the equipments taking into account for the handling periods. The required amount of the structural parts (for example cladding materials) was assessed for the buffer in the same manner and the amount of the buffer facilities were optimized. (author)

  12. Needs of Advanced Safeguards Technologies for Future Nuclear Fuel Cycle (FNFC) Facilities and a Trial Application of SBD Concept to Facility Design of a Hypothetical FNFC Facility

    International Nuclear Information System (INIS)

    Seya, M.; Hajima, R.; Nishimori, N.; Hayakawa, T.; Kikuzawa, N.; Shizuma, T.; Fujiwara, M.

    2010-01-01

    Some of future nuclear fuel cycle (FNFC) facilities are supposed to have the characteristic features of very large throughput of plutonium, low decontamination reprocessing (no purification process; existence of certain amount of fission products (FP) in all process material), full minor actinides (MA) recycle, and treatment of MOX with FP and MA in fuel fabrication. In addition, the following international safeguards requirements have to be taken into account for safeguards approaches of the FNFC facilities. -Application of integrated safeguards (IS) approach; -Remote (unattended) verification; - 'Safeguards by Design' (SBD) concept. These features and requirements compel us to develop advanced technologies, which are not emerged yet. In order to realize the SBD, facility designers have to know important parts of design information on advanced safeguards systems before starting the facility design. The SBD concept requires not only early start of R and D of advanced safeguards technologies (before starting preliminary design of the facility) but also interaction steps between researchers working on safeguards systems and nuclear facility designers. The interaction steps are follows. Step-1; researchers show images of advanced safeguards systems to facility designers based on their research. Step-2; facility designers take important design information on safeguards systems into process systems of demonstration (or test) facility. Step-3; demonstration and improvement of both systems based on the conceptual design. Step-4; Construction of a FNFC facility with the advanced safeguards systems We present a trial application of the SBD concept to a hypothetical FNFC facility with an advanced hybrid K-edge densitometer and a Pu NDA system for spent nuclear fuel assembly using laser Compton scattering (LCS) X-rays and γ-rays and other advanced safeguards systems. (author)

  13. Summary of Off-Normal Events in US Fuel Cycle Facilities for AFCI Applications

    Energy Technology Data Exchange (ETDEWEB)

    L. C. Cadwallader; S. J. Piet; S. O. Sheetz; D. H. McGuire; W. B. Boore

    2005-09-01

    This report is a collection and review of system operation and failure experiences for facilities comprising the fission reactor fuel cycle, with the exception of reactor operations. This report includes mines, mills, conversion plants, enrichment plants, fuel fabrication plants, transportation of fuel materials between these centers, and waste storage facilities. Some of the facilities discussed are no longer operating; others continue to produce fuel for the commercial fission power plant industry. Some of the facilities discussed have been part of the military’s nuclear effort; these are included when the processes used are similar to those used for commercial nuclear power. When reading compilations of incidents and accidents, after repeated entries it is natural to form an opinion that there exists nothing but accidents. For this reason, production or throughput values are described when available. These adverse operating experiences are compiled to support the design and decisions needed for the Advanced Fuel Cycle Initiative (AFCI). The AFCI is to weigh options for a new fission reactor fuel cycle that is efficient, safe, and productive for US energy security.

  14. LEMA facility and equipments for minor actinides compounds fabrication and characterisation

    International Nuclear Information System (INIS)

    Donnet, L.

    2008-01-01

    The LEMA (Actinide based materials study laboratory) is mainly involved in minor actinides materials development and fabrication, from raw materials choice and synthesis to finished products including pin assembly. The aim of the technological analyses is to establish choices of raw materials and manufacturing techniques. The LEMA is located in the ATALANTE facility in Marcoule. It consists in two shielded chains (one specific for neutrons) and three hot laboratories. The laboratory has various apparatuses in hot cells such as: ball mills, press, dilatometer, TGA (thermo-gravimetry analyser), calcination and sintering furnaces (2000 deg. C). The laboratory has also characterisation apparatuses such as XRD and SEM (scanning electron microscopy) dedicated to structural and microstructural studies. Thanks to the diversity of its equipment, the LEMA has well established worldwide collaborations and takes part in international fuels/target fabrication and irradiation experiments. (author)

  15. 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

  16. Fabrication procedures for manufacturing high uranium concentration dispersion fuel elements

    International Nuclear Information System (INIS)

    Souza, Jose Antonio Batista de

    2011-01-01

    IPEN-CNEN/SP developed the technology to produce the dispersion type fuel elements for research reactors and made it available for routine production. Today, the fuel produced in IPEN-CNEN/SP is limited to the uranium concentration of 3.0 gU/cm 3 for U 3 Si 2 -Al dispersion-based and 2.3 gU/cm 3 for U 3 O 8 -Al dispersion. The increase of uranium concentration in fuel plates enables the reactivity of the reactor core reactivity to be higher and extends the fuel life. Concerning technology, it is possible to increase the uranium concentration in the fuel meat up to the limit of 4.8 gU/cm 3 in U 3 Si 2 -Al dispersion and 3.2 gU/cm 3 U 3 O 8 -Al dispersion. These dispersions are well qualified worldwide. This work aims to develop the manufacturing process of both fuel meats with high uranium concentrations, by redefining the manufacturing procedures currently adopted in the Nuclear Fuel Center of IPEN-CNEN/SP. Based on the results, it was concluded that to achieve the desired concentration, it is necessary to make some changes in the established procedures, such as in the particle size of the fuel powder and in the feeding process inside the matrix, before briquette pressing. These studies have also shown that the fuel plates, with a high concentration of U 3 Si 2 -Al, met the used specifications. On the other hand, the appearance of the microstructure obtained from U 3 O 8 -Al dispersion fuel plates with 3.2 gU/cm 3 showed to be unsatisfactory, due to the considerably significant porosity observed. The developed fabrication procedure was applied to U 3 Si 2 production at 4.8 gU/cm 3 , with enriched uranium. The produced plates were used to assemble the fuel element IEA-228, which was irradiated in order to check its performance in the IEA-R1 reactor at IPEN-CNEN/SP. These new fuels have potential to be used in the new Brazilian Multipurpose Reactor - RMB. (author)

  17. Fabrication of fully ceramic microencapsulated fuel by hot pressing

    International Nuclear Information System (INIS)

    Lee, H. G.; Kim, D. J; Park, J. Y.; Kim, W. J.; Lee, S. J

    2014-01-01

    Fully ceramic microencapsulated(FCM) nuclear fuel is one of the recently suggested concept to enhance stability nuclear fuel itself. The requirements to increase the accident tolerance of nuclear fuel are mainly two parts: First, the performance has to be maintained compared to the existing UO 2 nuclear fuel and zircaloy cladding system under the normal operation condition. Second, under the severe accident condition, the high temperature structural integrity has to be kept and the generation rate of hydrogen has to be decrease largely. FCM nuclear fuel consists of tristructural isotropic(TRISO) fuel particle and SiC matrix. The relative thermal conductivity of the SiC matrix as compared to UO 2 is quite good, yielding as-irradiated fuel centerline temperature compared to high temperature for the existing fuel leading to reduced stored energy in the core and reduced operational release of fission products from the fuel. Generally SiC ceramics are fabricated via liquid phase sintering due to strong covalent bonding property and low self-diffusivity coefficient. Hot pressing is very effective method to conduct sintering of SiC powder including different second phase. In this study, SiC-matrix composite including TRISO particles were sintered by hot pressing with Al 2 O 3 -Y 2 O 3 additive system. Various sintering condition were investigated to obtain high relative density above 95%. The internal distribution of TRISO particles within SiC-matrix composite was observed by x-ray radiograph. From the analysis of the cross-section of SiC-matrix composite, the fracture of TRISO particles was investigated. In order to uniform distribution of TRISO particle embedded in the SiC matrix, SiC powder overcoating is considered. SiC matrix composite including TRISO was fabricated by hot pressing. FCM pallets with full density were obtained with Al 2 O 3 -Y 2 O 3 additive system. From the microstructure image, the effect of the sintering additive contents and sintering mechanism

  18. A Swedish nuclear fuel facility and public acceptance

    International Nuclear Information System (INIS)

    Andersson, Bengt A.

    1989-01-01

    For more than ten years the ABB Atom Nuclear Fuel Facility has gained a lot of public attention in Sweden. When the nuclear power debate was coming up in the middle of the seventies, the Nuclear Fuel Facility very soon became a spectacular object. It provided a possibility to bring factual information about nuclear power to the public. Today that public interest still exists. For ABB Atom the Facility works as a tool of information activities in several ways, as a solid base for ABB Atom company presentations. but also as a very practical demonstration of the nuclear power technology to the public. This is valid especially to satisfy the local school demand for a real life object complementary to the theoretical nuclear technology education. Beyond the fact that the Nuclear Fuel Facility is a very effective fuel production plant, it is not too wrong to see it as an important resource for education as well as a tool for improved public relations

  19. 300 Area fuel supply facilities deactivation function analysis report

    International Nuclear Information System (INIS)

    Lund, D.P.

    1995-09-01

    The document contains the functions, function definitions, function interfaces, function interface definitions, Input Computer Automated Manufacturing Definition (IDEFO) diagrams, and a function hierarchy chart that describe what needs to be performed to deactivate the 300 Area Fuel Supply Facilities

  20. Development of CANDU high-burnup fuel fabrication technology

    International Nuclear Information System (INIS)

    Sim, Ki Seob; Suk, H. C.; Kwon, H. I.; Ji, C. G.; Cho, M. S.; Chang, H. I.

    1997-07-01

    This study is focused on the achievement of the fabrication process improvement of CANFLEX-NU and for this purpose, following two areas of basic research were executed this year. 1) development of amorphous alloy for use in brazing of nuclear materials. 2) development of ECT techniques for the end-cap weld inspection. Also, preliminary feasibility analyses on the characteristics and handling techniques of CANFLEX-RU fuel were executed this year. - Selection of optimum conversion process of RU power -Characterization of the composition of RU power - Radiological characterization of RU power and sintered pellets - Compaction and sintering characteristics of RU power - Required special process for the production of CANFLEX-RU fuel - Development of technical specification for RU powder and pellets. In addition, technical support activities were performed for in-pile and out-pile fuel performance tests such as precision measurement of out-pile test fuel dimensions, establishment of quality control technique on fuel bundle by providing bundle kits to AECL for use in-pile irradiation tests in the NRU research reactor. (author). 57 refs., 16 tabs.,40 figs

  1. Analytical methodology and facility description spent fuel policy

    International Nuclear Information System (INIS)

    1978-08-01

    Three generic environmental impact statements (GEISs) on domestic fuels, foreign fuels, and storage charges are being prepared to provide environmental input into decisions on whether, and if so how the 1977 Presidential policy on spent fuel storage should be implmented. This report provides background information for two of these environmental impact statements: Storage of U.S. Spent Power Reactor Fuel and Storage of Foreign Spent Power Reactor Fuel. It includes the analytical methodology used in GEISs to assess the environmental effects and a description of the facilities used in the two GEISs

  2. Analytical methodology and facility description spent fuel policy

    Energy Technology Data Exchange (ETDEWEB)

    1978-08-01

    Three generic environmental impact statements (GEISs) on domestic fuels, foreign fuels, and storage charges are being prepared to provide environmental input into decisions on whether, and if so how the 1977 Presidential policy on spent fuel storage should be implmented. This report provides background information for two of these environmental impact statements: Storage of U.S. Spent Power Reactor Fuel and Storage of Foreign Spent Power Reactor Fuel. It includes the analytical methodology used in GEISs to assess the environmental effects and a description of the facilities used in the two GEISs.

  3. Coordinated safeguards for materials management in a mixed-oxide fuel facility

    Energy Technology Data Exchange (ETDEWEB)

    Shipley, J.P.; Cobb, D.D.; Dietz, R.J.; Evans, M.L.; Schelonka, E.P.; Smith, D.B.; Walton, R.B.

    1977-02-01

    A coordinated safeguards system is described for safeguarding strategic quantities of special nuclear materials in mixed-oxide recycle fuel fabrication facilities. The safeguards system is compatible with industrial process requirements and combines maximum effectiveness consistent with modest cost and minimal process interference. It is based on unit process accounting using a combination of conventional and state-of-the-art NDA measurement techniques. The effectiveness of the system against single and multiple thefts is evaluated using computer modeling and simulation techniques.

  4. Coordinated safeguards for materials management in a mixed-oxide fuel facility

    International Nuclear Information System (INIS)

    Shipley, J.P.; Cobb, D.D.; Dietz, R.J.; Evans, M.L.; Schelonka, E.P.; Smith, D.B.; Walton, R.B.

    1977-02-01

    A coordinated safeguards system is described for safeguarding strategic quantities of special nuclear materials in mixed-oxide recycle fuel fabrication facilities. The safeguards system is compatible with industrial process requirements and combines maximum effectiveness consistent with modest cost and minimal process interference. It is based on unit process accounting using a combination of conventional and state-of-the-art NDA measurement techniques. The effectiveness of the system against single and multiple thefts is evaluated using computer modeling and simulation techniques

  5. Structural analysis of hatch cover plates on Fuels and Materials Examination Facility high bay mezzanine

    International Nuclear Information System (INIS)

    Dixson, G.E.

    1997-01-01

    In order to move the Idaho National Engineering Laboratory (INEL) Light Duty Utility Arm (LDUA) trailer into position for testing on the Fuels and Materials Examination Facility (FMEF) 42 ft level mezzanine one of the trailer's wheels will have to sit on a circular hatch cover fabricated from one-inch thick steel plate. The attached calculations verify that the hatch cover plate is strong enough to support the weight of the INEL LDUA trailer's wheel

  6. Impact of certain safeguards considerations on fuel-cycle facility design

    International Nuclear Information System (INIS)

    Darby, J.L.; de Montmollin, J.M.

    1979-01-01

    Both physical protection and containment/surveillance systems impact plant design and operations. Effective physical protection systems can be systematically designed; work on designing containment/surveillance systems is in progress. Fuel fabrication facility designers need to be cognizant of these safeguards system developments to enable effective implementation of them with as little effect on plant functions as possible. This brief overview provides a general indication of what the impacts of the systems might be, and current thinking on their structure

  7. 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.

  8. Improvements in the fabrication of HTR fuel elements

    International Nuclear Information System (INIS)

    Brähler, Georg; Hartung, Markus; Fachinger, Johannes; Grosse, Karl-Heinz; Seemann, Richard

    2012-01-01

    The application of High Temperature Reactor (HTR) Technology in the course of the continuously increasing world wide demand on energy is taken more and more under serious consideration in the power supply strategy of various countries. Especially for the emerging nations the HTR Technology has become of special interest because of its inherent safety feature and due to the alternative possibilities of applications, e.g. in the production of liquid hydrocarbons or the alternative application in H 2 generation. The HTR fuel in its various forms (spheres or prismatic fuel blocks) is based on small fuel kernels of about 500 μm in diameter. Each of these uranium oxide or carbide kernels are coated with several layers of pyrocarbon (PyC) as well as an additional silicon carbide (SiC) layer. While the inner pyrocarbon layer is porous and capable to absorb gaseous fission products, the dense outer PyC layer forms the barrier against fission product release. The SiC layer improves the mechanical strengths of this barrier and considerably increases the retention capacity for solid fission products that tent to diffuse at these temperatures. Especially the high quality German LEU TRISO spherical fuel based on the NUKEM design, has demonstrated the best fission product release rate, particular at high temperatures. The ∼10% enriched uranium triple-coated particles are embedded in a moulded graphite sphere. A fuel sphere consists of approximately 9 g of uranium (some 15,000 particles) and has a diameter of 60 mm. As the unique safety features, especially the inherent safety of the HTR is based on the fuel design, this paper shall reflect the complexity but also developments and economical aspects of the fabrication processes for HTR fuel elements.

  9. Thermal fuel research and development facilities in BNFL

    International Nuclear Information System (INIS)

    Roberts, V.A.; Vickers, J.

    1996-01-01

    BNFL is committed to providing high quality, cost effective nuclear fuel cycle services to customers on a National and International level. BNFL's services, products and expertise span the complete fuel cycle; from fuel manufacture through to fuel reprocessing, transport, waste management and decommissioning and the Company maintains its technical and commercial lead by investment in continued research and development (R and D). This paper discusses BNFL's involvement in R and D and gives an account of the current facilities available together with a description of the advanced R and D facilities constructed or planned at Springfields and Sellafield. It outlines the work being carried out to support the company fuel technology business, to (1) develop more cost effective routes to existing fuel products; (2) maximize the use of recycled uranium, plutonium and tails uranium and (3) support a successful MOX business

  10. 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

  11. The Atalante facility at CEA/Marcoule: towards Gen IV systems fuel cycle

    International Nuclear Information System (INIS)

    Bordier, Gilles; Warin, Dominique; Masson, Michel

    2008-01-01

    The Atalante facility is a complete set of 18 hot labs and 9 shielded cells devoted to the research and development on fuel cycle. The activities correspond to 4 major sectors of nuclear research: -) to support the operation of actual reprocessing plants with the aim of adapting the head of the process to the increase of the spend fuel burn-up and to different types of new burnt fuels to be reprocessed (including MOX, USi or UMo fuels); -) to develop the COEX TM process that jointly manages uranium and plutonium from the dissolution of spent fuel to the production of UPuO 2 powder and the fabrication of MOX fuel pellets; -) to prepare the recycling of minor actinides (MA) by partitioning or by grouped actinide extraction, and by MA bearing fuel fabrication; -) to study the long term behavior of high level waste conditioning matrices and especially self irradiation and leaching of vitrified waste. The first hot lab of Atalante was operated in 1992, the process shielded cell (CBP) in 2003 and the last LN1 lab in 2005, while at the same time a large scale demonstration test on the DIAMEX-SANEX MA partitioning process was performed. Now some new challenges involve further necessary evolutions of the facility. Some are related to safety assessment and operating flexibility; the major evolutions will come from new scientific goals and research programs. Furthermore, minor actinides materials irradiation tests in fast reactors will be prepared in the framework of a large international cooperation (GACID program) and need the production of significant amounts of MA bearing mixed U-Pu oxide compounds in new hot labs and shielded cells equipment. The major new research tools are presented and we highlight how Atalante is a unique facility which brings a real opportunity to reinforce the European and international scientific cooperation in order to prepare the next Gen IV fuel cycle. (authors)

  12. Fuel cell hybrid drive train test facility

    NARCIS (Netherlands)

    J. Bruinsma; I. Zafina; H. Bosma; Edwin Tazelaar; Bram Veenhuizen

    2009-01-01

    Fuel cells are expected to play an important role in the near future as prime energy source on board of road-going vehicles. In order to be able to test all important functional aspects of a fuel cell hybrid drive train, the Automotive Institute of the HAN University has decided to realize a

  13. Nuclear Fuel Cycle Information System. A directory of nuclear fuel cycle facilities. 2009 ed

    International Nuclear Information System (INIS)

    2009-04-01

    The Nuclear Fuel Cycle Information System (NFCIS) is an international directory of civilian nuclear fuel cycle facilities, published online as part of the Integrated Nuclear Fuel Cycle Information System (iNFCIS: http://www-nfcis.iaea.org/). This is the fourth hardcopy publication in almost 30 years and it represents a snapshot of the NFCIS database as of the end of 2008. Together with the attached CD-ROM, it provides information on 650 civilian nuclear fuel cycle facilities in 53 countries, thus helping to improve the transparency of global nuclear fuel cycle activities

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

    International Nuclear Information System (INIS)

    Hoover, M.D.; Newton, G.J.; Yeh, H.C.; Edison, A.F.

    1983-01-01

    Information on the characteristics of aerosols in three industrial facilities during fabrication of uranium and plutonium mixed-oxide reactor fuels is summarized. Aerosol characterization included measurements of air concentrations particle size distributions, particle morphology, aerosol electrostatic charge characteristics and in vitro dissolution. These were the physicochemical properties considered most relevant to evaluating the consequences of potential accidents wherein aerosols might be released from their normal containment and inhaled by workers. Alpha radioactivity concentrations in air ranged about 1-15,000 nCi/l, depending on the fabrication step and facility. Particle size distributions were approximately log-normal, and activity median aerodynamic diameter (AMAD) ranged 1.6-3.5 μm, with a geometric standard deviation of about 1.6. Electron microscopy indicated that particles were irregularly shaped agglomerates. In vitro dissolution half-times were on the order of years, with some dependence on aerosol size, temperature history and fabrication step. No unique potential exposure conditions were found in any step of the fabrication processes

  15. Reassembling technique for irradiation vehicle at Fuel Monitoring Facility (FMF)

    International Nuclear Information System (INIS)

    Maeda, Koji; Nagamine, Tsuyoshi; Nakamura, Yasuo; Mitsugi, Takeshi; Matsumoto, Shinichiro

    1999-01-01

    The remote handling technique has been developed and demonstrated by Fuel Monitoring Facility (FMF) operated by Japan Nuclear Cycle Development Institute (JNC). In particular, the reassembling of irradiated fuels has been successfully performed, and reassembled irradiation vehicles were reinserted to Japanese experimental fast reactor 'JOYO'. This paper describes following four items; (A) Irradiation vehicle, (B) Disassembling and interim examination, (C) Decontamination of fuel pin or capsule, (D) Reassembling machine, which are necessary for the reinsertion. (J.P.N.)

  16. Health and environmental aspects of nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    1996-11-01

    The purpose of the present publication is to give a generic description of health and environmental aspects of nuclear fuel cycle facilities. Primarily the report is meant to stand alone; however, because of the content of the publication and in the context of the DECADES project, it may serve as a means of introducing specialists in other fuel cycles to the nuclear fuel cycle. Refs, figs, tabs

  17. Part 6. Internationalization and collocation of FBR fuel cycle facilities

    International Nuclear Information System (INIS)

    Stevenson, M.G.; Abramson, P.B.; LeSage, L.G.

    1980-01-01

    This report examines some of the non-proliferation, technical, and institutional aspects of internationalization and/or collocation of major facilities of the Fast Breeder Reactor (FBR) fuel cycle. The national incentives and disincentives for establishment of FBR Fuel Cycle Centers are enumerated. The technical, legal, and administrative considerations in determining the feasibility of FBR Fuel Cycle Centers are addressed by making comparisons with Light Water Reactor (LWR) centers which have been studied in detail by the IAEA and UNSRC

  18. 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

  19. Release characteristics of cesium from green pellet fabricated with spent fuel under different sintering conditions

    International Nuclear Information System (INIS)

    Park, Geun Il; Lee, Dou Youn; Lee, Young Soon; Kim, Woong Ki; Lee, Jae Won; Lee, Jung Won; Yang, Myung Seung

    2005-01-01

    The dry process, known as DUPIC(Direct Use of spent PWR fuel in CANDU reactor), for fabricating fuel pellets from spent fuel as recycling technology has been well demonstrated by establishing an optimization process for fuel fabrication through a number of batch processes using typical PWR spent fuel. As considering a strategy for extending the burn-up in LWR fuel, experimental verification for analyzing the effect of spent fuel burn-up on fuel fabrication is necessary in some respects that one of key parameters influencing the fuel fabrication characteristic would amount of fission products contained as impurity elements in spent fuel. A high burn-up spent fuel has a higher amount of fission products compared with typical spent fuel irradiated in about 27,000 MWd/tU. A preliminary study showed that the sintered pellet density fabricated with a high burn-up fuel has a lower value than that of common fuel burn-ups of about 30,000 MWd/tU. To provide better understanding a remote fuel fabrication characteristic in an aspect of wide ranges of spent fuel generated from PWR reactor, the influence of fission products release on fabrication characteristics of the dry processed fuel with a high burn-up fuel of 65,000 MWd/tU were experimentally evaluated. It is expected that key fission product affecting fabrication characteristics in dry process is cesium isotope due to the boiling point of 670 .deg. C and the low dissociation temperature of its oxides(<700 .deg. C). This study focus to analyze the release characteristics of cesium from green pellets fabricated with a variation of compaction pressure under different sintering conditions using tubular furnace in IMEF M6 hot cell

  20. Implications of plutonium and americium recycling on MOX fuel fabrication

    International Nuclear Information System (INIS)

    Renard, A.; Pilate, S.; Maldague, Th.; La Fuente, A.; Evrard, G.

    1995-01-01

    The impact of the multiple recycling of plutonium in power reactors on the radiation dose rates is analyzed for the most critical stage in a MOX fuel fabrication plant. The limitation of the number of Pu recycling in light water reactors would rather stem from reactor core physics features. The case of recovering americium with plutonium is also considered and the necessary additions of shielding are evaluated. A comparison between the recycling of Pu in fast reactors and in light water reactors is presented. (author)

  1. 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

  2. Activities related to fuel fabrication at the Chilean Commission for Nuclear Energy (CCHEN)

    International Nuclear Information System (INIS)

    Testart Tobar, E.; Castillo V, F.J.; Cortes N, D.; Chavez P, J.C.; Hechenleitner S, H.; Marin E, J.

    1987-01-01

    A brief description is herein presented of the activities related to fuel elements fabrication for research reactors, starting with U 3 O 8 powder preparation, U 3 O 8 - Al miniplates fabrication, dummy fuel elements assembling and disassembling and the reassembling of fuel elements. (Author)

  3. Establishment of technological basis for fabrication of U-Pu-Zr ternary alloy fuel pins for irradiation tests in Japan

    International Nuclear Information System (INIS)

    Kikuchi, Hironobu; Iwai, Takashi; Nakajima, Kunihisa; Arai, Yasuo; Nakamura, Kinya; Ogata, Takanari

    2011-01-01

    A high-purity Ar gas atmosphere glove box accommodating injection casting and sodium-bonding apparatuses was newly installed in the Plutonium Fuel Research Facility of Oarai Research and Development Center, Japan Atomic Energy Agency, in which several nitride and carbide fuel pins were fabricated for irradiation tests. The experiences led to the establishment of the technological basis of the fabrication of U-Pu-Zr alloy fuel pins for the first time in Japan. After the injection casting of the U-Pu-Zr alloy, the metallic fuel pins were fabricated by welding upper and lower end plugs with cladding tubes of ferritic-martensitic steel. Subsequent to the sodium bonding for filling the annular gap region between the U-Pu-Zr alloy and the cladding tube with the melted sodium, the fuel pins for irradiation tests are inspected. This paper shows the apparatuses and the technological basis for the fabrication of U-Pu-Zr alloy fuel pins for the irradiation test planned at the experimental fast test reactor Joyo. (author)

  4. Facility for electrochemical dissolution of rejected fuel elements

    International Nuclear Information System (INIS)

    Deniskin, V.P.; Filatov, O.N.; Konovalov, E.A.; Kolesnikov, B.P.; Bukharin, A.D.

    2003-01-01

    A facility for electrochemical dissolution of rejected fuel elements with the stainless steel can and uranium of 90% enrichment is described. The start-adjustment works and trial-commercial tests of the facility are carried out. A s a result its technological parameters are determined [ru

  5. 300 Area fuel supply shutdown facility hazards assessment

    International Nuclear Information System (INIS)

    Campbell, L.R.

    1998-01-01

    This document establishes the technical basis in support of Emergency Planning activities for the 300 Area Fuel Supply Shutdown Facilities on the Hanford Site. Through this document, the technical basis for the development of facility specific Emergency Action Levels and Emergency Planning Zone, is demonstrated

  6. DoD Fuel Facilities Criteria

    Science.gov (United States)

    2015-04-27

    078-24-33  Published in July 2010  Cut and Cover storage tanks are steel-lined reinforced concrete with leak monitoring capability.  They are...System Submittal Requirements Feb-2010 UFGS 33 52 43.28 Filter Separator, Aviation Fueling System Nov-2010 UFGS 32 13 15.20 Concrete Pavement for...Interstitial Space May-2012 UFGS 33 52 10 Service Piping, Fuel Systems Apr-2008 UFGS 33 56 63 Fuel Impermeable Liner System Apr-2006 UFGS 33 52 43.11

  7. Fuel Storage Facility Final Safety Analysis Report. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Linderoth, C.E.

    1984-03-01

    The Fuel Storage Facility (FSF) is an integral part of the Fast Flux Test Facility. Its purpose is to provide long-term storage (20-year design life) for spent fuel core elements used to provide the fast flux environment in FFTF, and for test fuel pins, components and subassemblies that have been irradiated in the fast flux environment. This Final Safety Analysis Report (FSAR) and its supporting documentation provides a complete description and safety evaluation of the site, the plant design, operations, and potential accidents.

  8. 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)

  9. Reducing Proliferation Rick Through Multinational Fuel Cycle Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Amanda Rynes

    2010-11-01

    With the prospect of rapid expansion of the nuclear energy industry and the ongoing concern over weapons proliferation, there is a growing need for a viable alternative to traditional nation-based fuel production facilities. While some in the international community remain apprehensive, the advantages of multinational fuel cycle facilities are becoming increasingly apparent, with states on both sides of the supply chain able to garner the security and financial benefits of such facilities. Proliferation risk is minimized by eliminating the need of states to establish indigenous fuel production capabilities and the concept's structure provides an additional internationally monitored barrier against the misuse or diversion of nuclear materials. This article gives a brief description of the arguments for and against the implementation of a complete multinational fuel cycle.

  10. Fuel cell power generation facility; Nenryo denchi hatsuden sochi

    Energy Technology Data Exchange (ETDEWEB)

    Hara, T.; Kato, K. [Aishin AW Co. Ltd., Aichi (Japan); Tanizaki, K.; Ishiko, C. [Equos Research Co. Ltd., Tokyo (Japan)

    1996-07-12

    In a fuel cell power generation facility with a carbon monoxide removal equipment to remove carbon monoxide in a reformed gas by oxidation, a reversed flow of the residual reformed gas in the reformed gas feeding line to the fuel reforming equipment cannot be prevented by a previous suspension technique. This invention aims to present a method to prevent a reversed flow of the residual reformed gas and to protect effectively the fuel cell power generation facility against the danger of catching fire or explosion. This invention relates to a fuel cell power generation facility installed with a fuel reforming equipment and a carbon monoxide removal equipment to remove carbon monoxide in the reformed gas by oxidation, which is equipped with reversed flow prevention devices on the oxidizer gas feeding duct to the carbon monoxide removal equipment, on the reformed gas feeding duct from the fuel reforming equipment to the carbon monoxide removal equipment and on the source fuel feeding duct to the fuel reforming equipment. The reversed flow prevention devices are made of direction control valves or pressure control valves. 2 figs.

  11. Deactivating a major nuclear fuels reprocessing facility

    International Nuclear Information System (INIS)

    LeBaron, G.J.

    1997-01-01

    This paper describes three key processes used in deactivating the Plutonium Uranium Extraction (PUREX) Facility, a large, complex nuclear reprocessing facility, 15 months ahead of schedule and $77 million under budget. The organization was reengineered to refine its business processes and more effectively organize around the deactivation work scope. Multi-disciplined work teams were formed to be self-sufficient and empowered to make decisions and perform work. A number of benefits were realized by reengineering. A comprehensive process to develop end points which clearly identified specific results and the post-project facility configuration was developed so all areas of a facility were addressed. Clear and specific end points allowed teams to focus on completing deactivation activities and helped ensure there were no unfulfilled end-of-project expectations. The RCRA regulations require closure of permitted facilities within 180 days after cessation of operations which may essentially necessitate decommissioning. A more cost effective approach was adopted which significantly reduced risk to human health and the environment by taking the facility to a passive, safe, inexpensive-to-maintain surveillance and maintenance condition (deactivation) prior to disposition. PUREX thus became the first large reprocessing facility with active TSD [treatment, storage, and disposal] units to be deactivated under the RCRA regulations

  12. Direct fabrication of 238PuO2 fuel forms

    International Nuclear Information System (INIS)

    Burney, G.A.; Congdon, J.W.

    1982-07-01

    The current process for the fabrication of 238 PuO 2 heat sources includes precipitation of small particle plutonium oxalate crystals (4 to 6 μm diameter), a calcination to PuO 2 , ball milling, cold pressing, granulation (60 to 125 μm), and granule sintering prior to hot pressing the fuel pellet. A new two-step direct-strike Pu(III) oxalate precipitation method which yields mainly large well-developed rosettes (50 to 100 μm diameter) has been demonstrated in the laboratory and in the plant. These large rosettes are formed by agglomeration of small (2 to 4 μm) crystals, and after calcining and sintering, were directly hot pressed into fuel forms, thus eliminating several of the powder conditioning steps. Conditions for direct hot pressing of the large heat-treated rosettes were determined and a full-scale General Purpose Heat Source pellet was fabricated. The pellet had the desired granule-type microstructure to provide dimensional stability at high temperature. 27 figures

  13. Safety analysis of IFR fuel processing in the Argonne National Laboratory Fuel Cycle Facility

    International Nuclear Information System (INIS)

    Charak, I; Pedersen, D.R.; Forrester, R.J.; Phipps, R.D.

    1993-01-01

    The Integral Fast Reactor (IFR) concept developed by Argonne National Laboratory (ANL) includes on-site processing and recycling of discharged core and blanket fuel materials. The process is being demonstrated in the Fuel Cycle Facility (FCF) at ANL's Idaho site. This paper describes the safety analyses that were performed in support of the FCF program; the resulting safety analysis report was the vehicle used to secure authorization to operate the facility and carry out the program, which is now under way. This work also provided some insights into safety-related issues of a commercial IFR fuel processing facility. These are also discussed

  14. Direction of Improvement for Licensing Advanced Fuel Cycle Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Il; Yook, Dae Sik; Jeong, Seung Yeong; Jeong, Chan Woo [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-05-15

    The final spent fuel management method has yet to be established, and the South Korean government is expected to decide on the final spent fuel management method under a national consensus. In particular, two methods of spent fuel management are under consideration: Direct disposal in base rock several hundred meters underground and recycling. The present study reviewed the direction of improvement of the regulatory system that can be applied when an advanced fuel cycle for recycling spent fuel is adopted as the final management method. For recycling to be adopted as the domestic final spent fuel management method, there remains the task of having to overcome the stumbling blocks of a national consensus and the Agreement for Cooperation between the Government of the Republic of Korea and the Government of the United States of America concerning the Civil Use of Atomic Energy (Korea-US Atomic Energy Agreement). To resolve this and to construct and operate advanced fuel cycle facilities, it is necessary to establish an applicable legal system, which the present study reviewed. The results of the present study are expected to be used as the basic data in improving the legal system after the realization of advanced fuel cycles in the future. In addition, research on the development of technical standards and safety requirements for advanced fuel cycle facilities will continue to be necessary.

  15. Study of internal exposure to uranium compounds in fuel fabrication plants in Brazil

    International Nuclear Information System (INIS)

    Santos, Maristela Souza

    2006-01-01

    The International Commission on Radiological Protection (ICRP) Publication 66 and Supporting Guidance 3) strongly recommends that specific information on lung retention parameters should be used in preference to default values wherever appropriate, for the derivation of effective doses and for bioassay interpretation of monitoring data. A group of 81 workers exposed to UO 2 at the fuel fabrication facility in Brazil was selected to evaluate the committed effective dose. The workers were monitored for determination of uranium content in the urinary and faecal excretion. The contribution of intakes by ingestion and inhalation were assessed on the basis of the ratios of urinary to fecal excretion. For the selected workers it was concluded that inhalation dominated intake. According to ICRP 66, uranium oxide is classified as insoluble Type S compound. The ICRP Supporting Guidance 3 and some recent studies have recommended specific lung retention parameters to UO 2 . The solubility parameters of the uranium oxide compound handled by the workers at the fuel fabrication facility in Brazil was evaluated on the basis of the ratios of urinary to fecal excretion. Excretion data were corrected for dietary intakes. This paper will discuss the application of lung retention parameters recommended by the ICRP models to these data and also the dependence of the effective committed dose on the lung retention parameters. It will also discuss the problems in the interpretation of monitoring results, when the worker is exposed to several uranium compounds of different solubilities. (author)

  16. IAEA physical inventory verification procedures implemented at US and Canadian fuel fabrication plants

    International Nuclear Information System (INIS)

    Gough, J.; Wredberg, L.; Zobor, E.; Zuccaro-Labellarte, G.

    1988-01-01

    IAEA has implemented safeguards at three Low Enriched Uranium (LEU) fuel fabrication plants in the USA during the period 1982 to 1987, and it is in the process of safeguarding a fourth plant from 01 January 1988. In Canada IAEA safeguards inspections were implemented at all Natural Uranium (NU) fuel fabrication plants form 1972 onwards, and there are, at present, three plants under safeguards. The direct responsibility for the implementation of safeguards inspections in the USA and Canada lies with the Division of Operations B (SGOB) within the IAEA Department of Safeguards. The senior staff that is at present directly engaged in the implementation activities has accumulated supervising inspection experience at about 50 Physical Inventory Verification (PIV) inspections at the Canadian and US fabrication plants during the period 1978 to 1987. This experience has been gained in close cooperation with the facility operators and with the support of the state authorities. The paper describes the latest PIV inspections at the Westinghouse Columbia plant and the Zircatec Precision Industries Inc. Port Hope plant. Furthermore, the paper describes the initial activities for the 1988 PIV inspection at the General Electric Wilmington plant including computerized book audit activities

  17. Fabrication and control of fuels made of mixed carbides (U, Pu)C

    International Nuclear Information System (INIS)

    Lorenzelli, R.; Delaroche, P.

    1980-01-01

    Fabrication of this type of advanced fuel is described. The fuel is prepared by reduction of oxides with carbon and natural sintering. Density, thermal stability and thermal conductibility are more particularly studied [fr

  18. 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)

  19. Use of probabilistic risk assessment in fuel cycle facilities

    International Nuclear Information System (INIS)

    Gonzalez, Felix; Gonzalez, Michelle; Wagner, Brian

    2013-01-01

    As expressed in its Policy Statement on the Use of Probabilistic Risk Assessment (PRA) Methods in Nuclear Regulatory Activities, the U.S Nuclear Regulatory Commission has been working for decades to increase the use of PRA technology in its regulatory activities. Since the policy statement was issued in 1995, PRA has become a core component of the nuclear power plant (NPP) licensing and oversight processes. In the last several years, interest has increased in PRA technologies and their possible application to other areas including, but not limited to, spent fuel handling, fuel cycle facilities, reprocessing facilities, and advanced reactors. This paper describes the application of PRA technology currently used in NPPs and its application in other areas such as fuel cycle facilities and advanced reactors. It describes major challenges that are being faced in the application of PRA into new technical areas and possible ways to resolve them. (authors)

  20. Fuels Selection Alternatives for Army Facilities

    Science.gov (United States)

    1986-12-01

    feeding and a less expensive boiler; however, both sulfur capture and carbon burnup may be less with this design than with an underfeed system...Ash-handling equipment costs were calculated by scaling according to ash mass flowrate. About nine employees were included in estimating direct... calculate the modified present-worth factors needed to rank fuel options. In developing an actual project, current policy must be determined, as explained

  1. Pacific Northwest Laboratory (PNL) spent fuel transportation and handling facility models

    International Nuclear Information System (INIS)

    Andrews, W.B.; Bower, J.C.; Burnett, R.A.; Engel, R.L.; Rolland, C.W.

    1979-09-01

    A spent fuel logistics study was conducted in support of the US DOE program to develop facilities for preparing spent unreprocessed fuel from commercial LWRs for geological storage. Two computerized logistics models were developed. The first one was the site evaluation model. Two studies of spent fuel handling facility and spent fuel disposal facility siting were completed; the first postulates a single spent fuel handling facility located at any of six DOE laboratory sites, while the second study examined siting strategies with the spent fuel repository relative to the spent fuel handling facility. A second model to conduct storage/handling facility simulations was developed

  2. PHOEBUS/UHTREX: a preliminary study of a low-cost facility for transient tests of LMFBR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kirk, W.L. (comp.)

    1976-08-01

    The results of a brief preliminary design study of a facility for transient nuclear tests of fast breeder reactor fuel are described. The study is based on the use of a reactor building originally built for the UHTREX reactor, and the use of some reactor hardware and reactor design and fabrication technology remaining from the Phoebus-2 reactor of the Rover nulcear rocket propulsion program. The facility is therefore currently identified as the PHOEBUS/UHTREX facility. This facility is believed capable of providing early information regarding fast reactor core accident energetics issues which will be very valuable to the overall LMFBR safety program. Facility performance in conjunction with a reference 127-fuel pin experiment is described. Low cost and early availability of the facility were emphasized in the selection of design features and parameters.

  3. Fabrication of electroslag welded Magnox fuel transport flasks

    International Nuclear Information System (INIS)

    Tuliani, S.S.

    1979-01-01

    The high weld metal deposition rate of the electroslag welding process offers an attractive method of fabricating nuclear fuel transport flasks from 370 mm (14.5in) thick steel plates. The paper describes pre-production trials carried out on full scale corner-section joints to establish that the weld metal meets the exacting mechanical property requirements for the Nuclear Industry. The paper presents results obtained on welds produced using two base metal compositions and two wires, one recommended for submerged-arc and the other for electroslag welding processes. Details of mechanical tests and metallographic examinations are given which led to the selection of the latter type of wire. It was found that while the weld metal deposited by this process may be sensitive to cracking, this can be avoided by careful selection of welding consumables and sound joints can be obtained under production conditions. (author)

  4. Nuclear criticality safety program at the Fuel Cycle Facility

    International Nuclear Information System (INIS)

    Lell, R.M.; Fujita, E.K.; Tracy, D.B.; Klann, R.T.; Imel, G.R.; Benedict, R.W.; Rigg, R.H.

    1994-01-01

    The Fuel Cycle Facility (FCF) is designed to demonstrate the feasibility of a novel commercial-scale remote pyrometallurgical process for metallic fuels from liquid metal-cooled reactors and to show closure of the Integral Fast Reactor (IFR) fuel cycle. Requirements for nuclear criticality safety impose the most restrictive of the various constraints on the operation of FCF. The upper limits on batch sizes and other important process parameters are determined principally by criticality safety considerations. To maintain an efficient operation within appropriate safety limits, it is necessary to formulate a nuclear criticality safety program that integrates equipment design, process development, process modeling, conduct of operations, a measurement program, adequate material control procedures, and nuclear criticality analysis. The nuclear criticality safety program for FCF reflects this integration, ensuring that the facility can be operated efficiently without compromising safety. The experience gained from the conduct of this program in the Fuel cycle Facility will be used to design and safely operate IFR facilities on a commercial scale. The key features of the nuclear criticality safety program are described. The relationship of these features to normal facility operation is also described

  5. Aerial infrared monitoring for nuclear fuel cycle facilities in Ukraine

    International Nuclear Information System (INIS)

    Stankevich, S.A.; Dudar, T.V.; Kovalenko, G.D.; Kartashov, V.V.

    2015-01-01

    The scientific research overall objective is rapid express detection and preliminary identification of pre-accidental conditions at nuclear fuel cycle facilities. We consider development of a miniature unmanned aerial vehicle equipped with high-precision infrared spectroradiometer able to detect remotely internal warming up of hazardous facilities by its thermal infrared radiation. The possibility of remote monitoring using unmanned aerial vehicle is considered at the example of the dry spent fuel storage facility of the Zaporizhzhya Nuclear Power Plant. Infrared remote monitoring is supposed to present additional information on the monitored facilities based on different physical principles rather than those currently in use. Models and specifications towards up-to-date samples of infrared surveying equipment and its small-sized unmanned vehicles are presented in the paper.

  6. 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

  7. Understanding and Managing Aging of Spent Nuclear Fuel and Facility Components in Wet Storage

    International Nuclear Information System (INIS)

    Johnson, A. B.

    2007-01-01

    Storage of nuclear fuel after it has been discharged from reactors has become the leading spent fuel management option. Many storage facilities are being required to operate longer than originally anticipated. Aging is a term that has emerged to focus attention on the consequences of extended operation on systems, structures, and components that comprise the storage facilities. The key to mitigation of age-related degradation in storage facilities is to implement effective strategies to understand and manage aging of the facility materials. A systematic approach to preclude serious effects of age-related degradation is addressed in this paper, directed principally to smaller facilities (test and research reactors). The first need is to assess the materials that comprise the facility and the environments that they are subject to. Access to historical data on facility design, fabrication, and operation can facilitate assessment of expected materials performance. Methods to assess the current condition of facility materials are summarized in the paper. Each facility needs an aging management plan to define the scope of the management program, involving identification of the materials that need specific actions to manage age-related degradation. For each material identified, one or more aging management programs are developed and become part of the plan Several national and international organizations have invested in development of comprehensive and systematic approaches to aging management. A method developed by the US Nuclear Regulatory Commission is recommended as a concise template to organize measures to effectively manage age-related degradation of storage facility materials, including the scope of inspection, surveillance, and maintenance that is needed to assure successful operation of the facility over its required life. Important to effective aging management is a staff that is alert for evidence of materials degradation and committed to carry out the aging

  8. The role of spent fuel test facilities in the fuel cycle strategy

    International Nuclear Information System (INIS)

    Huang, S. T.; Gross, D. L.; Snyder, N. W.; Woods, W. D.

    1988-01-01

    Disposal of commercial spent nuclear fuels in the major industrialized countries may be categorized into two broad approaches: a once-through policy which will dispose of spent fuels and recycle fissile materials. Within reprocess spent fuels and recycle fissile materials. Within each policy, various technical, licensing, institutional and public issues exist. These issues tend to complicate the formulation of an effective and acceptable fuel cycle strategy which will meet various cost, schedule, and legislative constraints. This paper examines overall fuel cycle strategies from the viewpoint of these underlying technical issues and assesses the roles of spent fuel test facilities in the overall fuel cycles steps. Basic functions of such test facilities are also discussed. The main emphasis is placed on the once-through policy although the reprocessing / recycle policy is also discussed. Benefits of utilizing test facilities in the fuel cycle strategies are explored. The results indicate that substantial benefits may be obtained in terms of minimizing programmatic risks, increasing public confidence, and more effective utilization of overall budgetary resources by structuring and highlighting the test facilities as an important element in the overall strategy

  9. Constrained Sintering in Fabrication of Solid Oxide Fuel Cells.

    Science.gov (United States)

    Lee, Hae-Weon; Park, Mansoo; Hong, Jongsup; Kim, Hyoungchul; Yoon, Kyung Joong; Son, Ji-Won; Lee, Jong-Ho; Kim, Byung-Kook

    2016-08-09

    Solid oxide fuel cells (SOFCs) are inevitably affected by the tensile stress field imposed by the rigid substrate during constrained sintering, which strongly affects microstructural evolution and flaw generation in the fabrication process and subsequent operation. In the case of sintering a composite cathode, one component acts as a continuous matrix phase while the other acts as a dispersed phase depending upon the initial composition and packing structure. The clustering of dispersed particles in the matrix has significant effects on the final microstructure, and strong rigidity of the clusters covering the entire cathode volume is desirable to obtain stable pore structure. The local constraints developed around the dispersed particles and their clusters effectively suppress generation of major process flaws, and microstructural features such as triple phase boundary and porosity could be readily controlled by adjusting the content and size of the dispersed particles. However, in the fabrication of the dense electrolyte layer via the chemical solution deposition route using slow-sintering nanoparticles dispersed in a sol matrix, the rigidity of the cluster should be minimized for the fine matrix to continuously densify, and special care should be taken in selecting the size of the dispersed particles to optimize the thermodynamic stability criteria of the grain size and film thickness. The principles of constrained sintering presented in this paper could be used as basic guidelines for realizing the ideal microstructure of SOFCs.

  10. Quality control of CANDU6 fuel element in fabrication process

    International Nuclear Information System (INIS)

    Li Yinxie; Zhang Jie

    2012-01-01

    To enhance the fine control over all aspects of the production process, improve product quality, fuel element fabrication process for CANDU6 quality process control activities carried out by professional technical and management technology combined mode, the quality of the fuel elements formed around CANDU6 weak links - - end plug , and brazing processes and procedures associated with this aspect of strict control, in improving staff quality consciousness, strengthening equipment maintenance, improved tooling, fixtures, optimization process test, strengthen supervision, fine inspection operations, timely delivery carry out aspects of the quality of information and concerns the production environment, etc., to find the problem from the improvement of product quality and factors affecting the source, and resolved to form the active control, comprehensive and systematic analysis of the problem of the quality management concepts, effectively reducing the end plug weld microstructure after the failure times and number of defects zirconium alloys brazed, improved product quality, and created economic benefits expressly provided, while staff quality consciousness and attention to detail, collaboration department, communication has been greatly improved and achieved very good management effectiveness. (authors)

  11. Description of fuel element brush assembly's fabrication for 105-K west

    International Nuclear Information System (INIS)

    Maassen, D.P.

    1997-01-01

    This report is a description of the process to redesign and fabricate, as well as, describe the features of the Fuel Element Brush Assembly used in the 105-K West Basin. This narrative description will identify problems that occurred during the redesigning and fabrication of the 105-K West Basin Fuel Element Brush Assembly and specifically address their solutions

  12. Plutonium Consumption Program, CANDU Reactor Project: Feasibility of BNFP Site as MOX Fuel Supply Facility. Final report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-06-30

    An evaluation was made of the technical feasibility, cost, and schedule for converting the existing unused Barnwell Nuclear Fuel Facility (BNFP) into a Mixed Oxide (MOX) CANDU fuel fabrication plant for disposition of excess weapons plutonium. This MOX fuel would be transported to Ontario where it would generate electricity in the Bruce CANDU reactors. Because CANDU MOX fuel operates at lower thermal load than natural uranium fuel, the MOX program can be licensed by AECB within 4.5 years, and actual Pu disposition in the Bruce reactors can begin in 2001. Ontario Hydro will have to be involved in the entire program. Cost is compared between BNFP and FMEF at Hanford for converting to a CANDU MOX facility.

  13. 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

  14. Fabrication experience and integrity confirmation tests of the first-loading-fuel of the HTTR

    International Nuclear Information System (INIS)

    Sawa, Kazuhiro; Tobita, Tsutomu; Sumita, Junya; Mogi, Haruyoshi; Yoshimuta, Shigeharu; Suzuki, Shuuichi; Deushi, Kouzaburou; Kato, Shigeru

    1999-01-01

    The first-loading core of the High Temperature Engineering Test Reactor (HTTR) consists of 150 fuel assemblies. An HTTR fuel assembly is so-called a pin-in-block type of hexagonal graphite block. A fuel rod consists of a graphite sleeve and of 14 fuel compacts. In a fuel compact, about 13,000 TRISO coated fuel panicles are dispersed densely. The coated fuel panicle is TRISO (Tri-isotropic) type with four coating layers. The fuel kernel is low-enriched (average 6wt%) UO 2 . The fabrication of the first-loading fuel started from June 1995. A total of 4,770 fuel rods were successfully produced and transferred to the reactor building of the HTTR. Finally, in the reactor building, the fuel rods were inserted to the graphite blocks to form fuel assemblies. On December 1997, 150 fuel assemblies were completely formed and were stored in new fuel storage cells. The cells were filled with helium gas to keep the fuel blocks in dry condition. Fabrication technology of the HTTR fuel was established through a lot of R and D activities and fabrication experiences of irradiation examination samples spread over about 30 years. 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 optimization of the compaction conditions. In the safety design of HTGR fuel, it is important to retain fission products within the coated fuel panicles so that their release to the primary coolant may not exceed an acceptable level. From this point of view, as-fabricated failure fraction is important. In the specification, SiC-failure and exposed uranium fractions were determined to be less than 1.5x10 -3 and l.5x10 -4 , respectively. The quality of the first loading fuel fully satisfied the design specifications for the fuel. The fuel compacts contained almost no through-coatings failed particles and few SiC-defective particles. Average through-coatings and Si

  15. Summary engineering description of underwater fuel storage facility for foreign research reactor spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Dahlke, H.J.; Johnson, D.A.; Rawlins, J.K.; Searle, D.K.; Wachs, G.W.

    1994-10-01

    This document is a summary description for an Underwater Fuel Storage Facility (UFSF) for foreign research reactor (FRR) spent nuclear fuel (SNF). A FRR SNF environmental Impact Statement (EIS) is being prepared and will include both wet and dry storage facilities as storage alternatives. For the UFSF presented in this document, a specific site is not chosen. This facility can be sited at any one of the five locations under consideration in the EIS. These locations are the Idaho National Engineering Laboratory, Savannah River Site, Hanford, Oak Ridge National Laboratory, and Nevada Test Site. Generic facility environmental impacts and emissions are provided in this report. A baseline fuel element is defined in Section 2.2, and the results of a fission product analysis are presented. Requirements for a storage facility have been researched and are summarized in Section 3. Section 4 describes three facility options: (1) the Centralized-UFSF, which would store the entire fuel element quantity in a single facility at a single location, (2) the Regionalized Large-UFSF, which would store 75% of the fuel element quantity in some region of the country, and (3) the Regionalized Small-UFSF, which would store 25% of the fuel element quantity, with the possibility of a number of these facilities in various regions throughout the country. The operational philosophy is presented in Section 5, and Section 6 contains a description of the equipment. Section 7 defines the utilities required for the facility. Cost estimates are discussed in Section 8, and detailed cost estimates are included. Impacts to worker safety, public safety, and the environment are discussed in Section 9. Accidental releases are presented in Section 10. Standard Environmental Impact Forms are included in Section 11.

  16. CEA and AREVA R and D on V/HTR fuel fabrication with the CAPRI experimental manufacturing line

    International Nuclear Information System (INIS)

    Charollais, Francois; Fonquernie, Sophie; Perrais, Christophe; Perez, Marc; Cellier, Francois; Vitali, Marie-Pierre

    2006-01-01

    In the framework of the French V/HTR fuel development and qualification program, the Commissariat a l'Energie Atomique (CEA) and AREVA through its program called ANTARES (Areva New Technology for Advanced Reactor Energy Supply) conduct R and D projects covering the mastering of UO 2 coated particle and fuel compact fabrication technology. To fulfill this task, a review of past knowledge, of existing technologies and a preliminary laboratory scale work program have been conducted with the aim of retrieving the know-how on HTR coated particle and compact manufacture: - The different stages of UO 2 kernel fabrication GSP Sol-Gel process have been reviewed, reproduced and improved; - The experimental conditions for the chemical vapour deposition (CVD) of coatings have been defined on dummy kernels and development of innovative characterization methods has been carried out; - Former CERCA compacting process has been reviewed and updated. In parallel, an experimental manufacturing line for coated particles, named GAIA, and a compacting line based on former CERCA compacting experience have been designed, constructed and are in operation since early 2005 at CEA Cadarache and CERCA Romans, respectively. These two facilities constitute the CAPRI line (CEA and AREVA PRoduction Integrated line). The major objectives of the CAPRI line are: - to recover and validate past knowledge; - to permit the optimisation of reference fabrication processes for kernels and coatings and the investigation of alternative and innovative fuel design (UCO kernel, ZrC coating); - to test alternative compact process options; - to fabricate and characterize fuel required for irradiation and qualification purpose; - to specify needs for the fabrication of representative V/HTR TRISO fuel meeting industrial standards. This paper presents the progress status of the R and D conducted on V/HTR fuel particle and compact manufacture by mid 2005. (authors)

  17. Fabrication and characterization of MX-type fuels and fuel pins

    International Nuclear Information System (INIS)

    Richter, K.; Bartscher, W.; Benedict, U.; Gueugnon, J.F.; Kutter, H.; Sari, C.; Schmidt, H.E.

    1978-01-01

    This paper summarizes the most important fabrication parameters and characterization of fuel and fuel pins obtained during the investigation of uranium-plutonium carbides, oxicarbides, carbonitrides and nitrides in the past years at the European Institute for Transuranium Elements at Karlsruhe. All preparation methods discussed are based on carbothermic reduction of a mechanical blend of uranium-plutonium oxide and carbon powder. General data for carbothermic reduction processes are discussed (influence of starting material, homogeneity, control of degree of reaction, etc). A survey of different preparation methods investigated is given. Limitations with respect to temperature and atmosphere for both carbothermic reduction processes and sintering conditions for the different compounds are summarized. A special preparation process for mixed carbonitrides with low nitrogen content (U,Pu)sub(1-x)Nsub(x) in the range 0.1 0 C to 1400 0 C by means of a modulated electron beam technique. A scheme is proposed, which allows to predict the thermal properties of MX fuels on the basis of their chemical composition and porosity. Preparation, preirradiation characterization and final controls of fuel test pins for pellet and vibrocompacted type of pins are described and the most important data summarized for all advanced fuels irradiated at Dounreay (DN1) and Rapsodie Fast Reactor (DN2) within the TU irradiation programme

  18. Regulatory cross-cutting topics for fuel cycle facilities.

    Energy Technology Data Exchange (ETDEWEB)

    Denman, Matthew R.; Brown, Jason; Goldmann, Andrew Scott; Louie, David

    2013-10-01

    This report overviews crosscutting regulatory topics for nuclear fuel cycle facilities for use in the Fuel Cycle Research & Development Nuclear Fuel Cycle Evaluation and Screening study. In particular, the regulatory infrastructure and analysis capability is assessed for the following topical areas: Fire Regulations (i.e., how applicable are current Nuclear Regulatory Commission (NRC) and/or International Atomic Energy Agency (IAEA) fire regulations to advance fuel cycle facilities) Consequence Assessment (i.e., how applicable are current radionuclide transportation tools to support risk-informed regulations and Level 2 and/or 3 PRA) While not addressed in detail, the following regulatory topic is also discussed: Integrated Security, Safeguard and Safety Requirement (i.e., how applicable are current Nuclear Regulatory Commission (NRC) regulations to future fuel cycle facilities which will likely be required to balance the sometimes conflicting Material Accountability, Security, and Safety requirements.)

  19. Risk analysis for nuclear spent fuel storage facility

    International Nuclear Information System (INIS)

    Dina, Dumitru; Andrei, Veronica; Ghita, Sorin; Glodeanu, Florin

    2004-01-01

    In June 2003, the first capacity of the Intermediate Dry Spent Fuel Storage Facility (DICA) was commissioned at Cernavoda Nuclear Power Plant (Cernavoda NPP). The facility is a dry system type facility; its designed lifetime is for a minimum of 50 years and capacity for two nuclear power units' lifetime. The storage structures are monolith reinforced concrete modules offering a very good isolation of the spent fuel from the environment. The spent fuel is confined by a system of double barriers that prevents radioactive emissions and ensures protection of the population and environment. The security functions of the facility are operational through passive means. In Romania, the National Commission for Nuclear Activities Control, CNCAN, is the authority that licenses the nuclear activities. CNCAN issued the commissioning and operating licenses for DICA following a complex process. The Final Nuclear Safety Report represents basic documentation for licensing and one of its main chapters presents the risk analysis results. The risk analysis performed for DICA covers normal operational regimes and accident cases considered as design basis events (DBE). The results of risk analysis for Cernavoda NNP DICA demonstrates that risks for the population and environment are much lower than the authorization limits established by CNCAN and in agreement with values for proven safe spent fuel storage technologies from European Union and worldwide. (authors)

  20. Seismic design considerations of nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    2001-10-01

    An Advisory Group Meeting (AGM) on Seismic Technologies of Nuclear Fuel Cycle Facilities was convened in Vienna from 12 to 14 November 1997. The main objective of the meeting was the investigation of the present status of seismic technologies in nuclear fuel cycle facilities in Member States as a starting point for understanding of the most important directions and trends of national initiatives, including research and development, in the area of seismic safety. The AGM gave priority to the establishment of a consistent programme for seismic assessment of nuclear fuel cycle facilities worldwide. A consultants meeting subsequently met in Vienna from 16 to 19 March 1999. At this meeting the necessity of a dedicated programme was further supported and a technical background to the initiative was provided. This publication provides recommendations both for the seismic design of new plants and for re-evaluation projects of nuclear fuel cycle facilities. After a short introduction of the general IAEA approach, some key contributions from Member State participants are presented. Each of them was indexed separately

  1. Safety Analysis of Spent Nuclear Fuel and Radwaste Facilities

    International Nuclear Information System (INIS)

    Poskas, P.; Ragaisis, V.

    2001-01-01

    The overview of the activities in the Laboratory of Heat Transfer in Nuclear Reactors related with the assessment of thermal, neutronic and radiation characteristics in spent nuclear fuel and radwaste facilities are performed. Activities related with decommissioning of Ignalina NPP are also reviewed. (author)

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  3. Remote maintenance in TOR fast reactor fuel reprocessing facility

    International Nuclear Information System (INIS)

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

    1986-11-01

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

  4. Costs of fuel cycle industrial facilities: an international review

    International Nuclear Information System (INIS)

    Macias, R.M.

    2004-01-01

    This document presents, comments, and compares economic and financial data for industrial facilities concerning different aspects of the nuclear fuel cycle. It first comments the present situation and the short term trends for the natural uranium market, the conversion market, the enrichment market, the reprocessing market, the storage market. It gives an assessment of the elementary costs of the existing facilities for the different stages and processes: reprocessing, spent fuel warehousing (example of the CLAB in Sweden and comparison with other available data), warehousing of all types of wastes (examples of Habog in Netherlands, Zwilag in Switzerland), spent fuel storage (example of Yucca Mountain in the USA, Onkalo in Finland, projects and studies in Sweden), storage of vitrified wastes in Belgium, storing of transuranic wastes in the USA, storage of low and intermediate level and short life wastes in Sweden

  5. Operational experience of the fuel cleaning facility of Joyo

    International Nuclear Information System (INIS)

    Mukaibo, R.; Matsuno, Y.; Sato, I.; Yoneda, Y.; Ito, H.

    1978-01-01

    Spent fuel assemblies in 'Joyo', after they are taken out of the core, are taken to the Fuel Cleaning Facility in the reactor service building and sodium removal is done. The cleaning process is done by cooling the assembly with argon gas, steam charging and rinsing by demineralized water. Deposited sodium was 50 ∼ 60 g per assembly. The sodium and steam reaction takes about 15 minutes to end and the total time the fuel is placed in the pot is about an hour. The total number of assemblies cleaned in the facility was 95 as of November 1977. In this report the operational experience together with discussions of future improvements are given. (author)

  6. The OSU Hydro-Mechanical Fuel Test Facility: Standard Fuel Element Testing

    Energy Technology Data Exchange (ETDEWEB)

    Wade R. Marcum; Brian G. Woods; Ann Marie Phillips; Richard G. Ambrosek; James D. Wiest; Daniel M. Wachs

    2001-10-01

    Oregon State University (OSU) and the Idaho National Laboratory (INL) are currently collaborating on a test program which entails hydro-mechanical testing of a generic plate type fuel element, or standard fuel element (SFE), for the purpose of qualitatively demonstrating mechanical integrity of uranium-molybdenum monolithic plates as compared to that of uranium aluminum dispersion, and aluminum fuel plates due to hydraulic forces. This test program supports ongoing work conducted for/by the fuel development program and will take place at OSU in the Hydro-Mechanical Fuel Test Facility (HMFTF). Discussion of a preliminary test matrix, SFE design, measurement and instrumentation techniques, and facility description are detailed in this paper.

  7. Facile fabrication of a superhydrophobic fabric with mechanical stability and easy-repairability.

    Science.gov (United States)

    Zhu, Xiaotao; Zhang, Zhaozhu; Yang, Jin; Xu, Xianghui; Men, Xuehu; Zhou, Xiaoyan

    2012-08-15

    The poor mechanical stability of superhydrophobic fabrics severely hindered their use in practical applications. Herein, to address this problem, we fabricated a superhydrophobic fabric with both mechanical stability and easy-repairability by a simple method. The mechanical durability of the obtained superhydrophobic fabric was evaluated by finger touching and abrasion with sandpaper. The results show that rough surface textures of the fabric were retained, and the fabric surface still exhibited superhydrophobicity after tests. More importantly, when the fabric lost its superhydrophobicity after a long-time abrasion, it can be easily rendered with superhydrophobicity once more by a regeneration process. Copyright © 2012 Elsevier Inc. All rights reserved.

  8. Modified Fabrication Method of Metal Fuel Slug for Preventing Evaporation of Volatile Elements

    International Nuclear Information System (INIS)

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

    2014-01-01

    In order to develop innovative fabrication method of metal fuel slugs 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, U-10wt.%Zr-Mn fuel slugs for SFR have been fabricated by modified casting method and characterized to evaluate the feasibility of the alternative fabrication method. Alternative casting such as modified casting has been applied to develop fabrication method of fuel slugs for preventing the evaporation of volatile elements such as Am. U-10wt.%Zr-Mn containing a volatile surrogate Mn fuel slug was soundly fabricated under inert atmosphere with dimensions of L250mm. Mass fraction of fuel loss was so low, upto 0.2%. Mn element was most recovered with prevention in evaporation of Mn. It was seen that the losses of volatile Am can be effectively controlled to below detectable levels using modest pressure

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  10. Characterization of fuel miniplates fabricated with U(Mo) particles dispersed in Al-Si matrices

    International Nuclear Information System (INIS)

    Arico, S F; Mirandou, M I; Balart, S N; Fabro, J O

    2012-01-01

    In 2011 ECRI facility (Depto. ECRI, GCCN, CNEA) restarted the development for the fabrication of dispersion miniplates fuel elements in Al-Si matrix. This miniplates are fabricated with atomized U-7wt%Mo particles dispersed in a matrix formed by a mixture of pure Al and pure Si powders. The first results for an Al-4wt%Si matrix were presented at the AATN 2011 Annual Meeting. In this work, new results from the microstructural characterization of the meat in Al- 2wt%Si and pure Al miniplates are presented and compared with the previous ones. It is the intention to study the influence of the fabrication parameters as well as different Si concentration in the matrix, on the formation and characteristics of the interaction layer formed between the particles and the matrix at the end of the fabrication process. According to the results presented in this work an improvement can be observed on miniplates with Al-Si matrix respect to the one with pure Al. On the miniplates with Al- Si matrix, almost 100 % of the U(Mo) particles presented, at least in some fraction of its surface, an interaction layer composed by phases that contain Si. Moreover its morphological characteristics are independent of the crystallographic state of the U(Mo) particles. However, the oxide layer formed on the U(Mo) during the hot rolling acts as a barrier to the formation of the interaction layer. As a consequence, it is then mandatory to introduce some changes on the fabrication parameters to avoid, or at least minimize, this oxide layer (author)

  11. Fabrication procedures for manufacturing High Flux Isotope Reactor fuel elements - 2

    International Nuclear Information System (INIS)

    Knight, R.W.; Morin, R.A.

    1999-01-01

    The original fabrication procedures written in 1968 delineated the manufacturing procedures at that time. Since 1968, there have been a number of procedural changes. This rewrite of the fabrication procedures incorporates these changes. The entire fuel core of this reactor is made up of two fuel elements. Each element consists of one annular array of fuel plates. These annuli are identified as the inner and outer fuel elements, since one fits inside the other. The inner element consists of 171 identical fuel plates, and the outer element contains 369 identical fuel plates differing slightly from those in the inner element. Both sets of fuel plates contain U 3 O 8 powder as the fuel, dispersed in an aluminum powder matrix and clad with aluminum. Procedures for manufacturing and inspection of the fuel elements are described and illustrated

  12. Fabrication procedures for manufacturing High Flux Isotope Reactor fuel elements - 2

    Energy Technology Data Exchange (ETDEWEB)

    Knight, R.W.; Morin, R.A.

    1999-12-01

    The original fabrication procedures written in 1968 delineated the manufacturing procedures at that time. Since 1968, there have been a number of procedural changes. This rewrite of the fabrication procedures incorporates these changes. The entire fuel core of this reactor is made up of two fuel elements. Each element consists of one annular array of fuel plates. These annuli are identified as the inner and outer fuel elements, since one fits inside the other. The inner element consists of 171 identical fuel plates, and the outer element contains 369 identical fuel plates differing slightly from those in the inner element. Both sets of fuel plates contain U{sub 3}O{sub 8} powder as the fuel, dispersed in an aluminum powder matrix and clad with aluminum. Procedures for manufacturing and inspection of the fuel elements are described and illustrated.

  13. Power Burst Facility Severe Fuel Damage test series

    International Nuclear Information System (INIS)

    Buescher, B.J.; Osetek, D.J.; Ploger, S.A.

    1982-01-01

    The Severe Fuel Damage (SFD) tests planned for the Power Burst Facility (PBF) are described. Bundles containing 32 zircaloy-clad, PWR-type fuel rods will be subjected to severe overheating transients in a high-pressure, superheated-steam environment. Cladding temperatures are expected to reach 2400 0 K, resulting in cladding ballooning and rupture, severe cladding oxidation, cladding melting, fuel dissolution, fuel rod fragmentation, and possibly, rubble bed formation. An experiment effluent collection system is being installed and the PBF fission product monitoring system is being upgraded to meet the special requirements of the SFD tests. Scoping calculations were performed to evaluate performance of the SFD test design and to establish operational requirements for the PBF loop

  14. Storage for the Fast Flux Test Facility unirradiated fuel in the Plutonium Finishing Plant Complex, Hanford Site, Richland, Washington

    International Nuclear Information System (INIS)

    1992-01-01

    This Environmental Assessment evaluates the proposed action to relocate and store unirradiated Fast Flux Test Facility fuel in the Plutonium Finishing Plant Complex on the Hanford Site, Richland, Washington. The US Department of Energy has decided to cease fuel fabrication activities in the 308 Building in the 300 Area. This decision was based on a safety concern over the ability of the fuel fabrication portion of the 308 Building to withstand a seismic event. The proposed action to relocate and store the fuel is based on the savings that could be realized by consolidating security costs associated with storage of the fuel. While the 308 Building belowgrade fuel storage areas are not at jeopardy by a seismic event, the US Department of Energy is proposing to cease storage operations along with the related fabrication operations. The US Department of Energy proposes to remove the unirradiated fuel pins and fuel assemblies from the 308 Building and store them in Room 192A, within the 234-5Z Building, a part of the Plutonium Finishing Plant Complex, located in the 200 West Area. Minor modifications to Room 192A would be required to accommodate placement of the fuel. The US Department of Energy estimates that removing all of the fuel from the 308 Building would save $6.5 million annually in security expenditures for the Fast Flux Test Facility. Environmental impacts of construction, relocation, and operation of the proposed action and alternatives were evaluated. This evaluation concluded that the proposed action would have no significant impacts on the human environment

  15. Improving of spent fuel monitoring in condition of Slovak wet interim spent fuel storage facility

    International Nuclear Information System (INIS)

    Miklos, M.; Krsjak, V.; Bozik, M.; Vasina, D.

    2008-01-01

    Monitoring of WWER fuel assemblies condition in Slovakia is presented in the paper. The leak tightness results of fuel assemblies used in Slovak WWER units in last 20 years are analyzed. Good experiences with the 'Sipping system' are described. The Slovak wet interim spent fuel storage facility in NPP Jaslovske Bohunice was build and put in operation in 1986. Since 1999, leak tests of WWER-440 fuel assemblies are provided by special leak tightness detection system 'Sipping in Pool' delivered by Framatome-ANP facility with external heating for the precise detection of active specimens. Another system for monitoring of fuel assemblies condition was implemented in December 2006 under the name 'SVYPP-440'. First non-active tests started at February 2007 and are described in the paper. Although those systems seems to be very effective, the detection time of all fuel assemblies in one storage pool is too long (several months). Therefore, a new 'on-line' detection system, based on new sorbent KNiFC-PAN for effective 134 Cs and 137 Cs activity was developed. This sorbent was compared with another type of sorbent NIFSIL and results are presented. The design of this detection system and its possible application in the Slovak wet spent fuel storage facility is discussed. For completeness, the initial results of the new system are also presented. (authors)

  16. Experiences with the first prototype MOX fuel rods fabricated at Argentina

    International Nuclear Information System (INIS)

    Marino, Armando Carlos; Perez, Edmundo; Adelfang, Pablo

    1996-01-01

    The irradiation of the first Argentine prototypes of pressurized heavy water reactor (PHWR) (U,Pu)O sub 2 MOX fuels began in 1986. These experiments were carried out in the High Flux Reactor (HFR)-Petten, Holland. The rods were prepared and controlled in the C NEA's alpha Facility. The first rod has been used for destructive pre-irradiation analysis. The second one as a pathfinder to adjust systems in the HFR. Two additional rods including iodine doped pellets were intended to simulate 15000 MWd/T(M) burnup. The remaining two rods were irradiated until 15000 MWd/T(M). One of them underwent a final ramp with the aim of verifying fabrication processes and studying the behaviour under power transients. BACO (BArra COmbustible) code was used to define the power histories and to analyse the experiments. This paper presents a description of the different experiments and a comparison between the results of the postirradiation examinations and the BACO outputs

  17. Spent Nuclear Fuel Project Cold Vacuum Drying Facility Operations Manual

    International Nuclear Information System (INIS)

    IRWIN, J.J.

    1999-01-01

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

  18. Main trends and content of works on fabrication of fuel rods with MOX fuel for the WWER-1000 reactor

    International Nuclear Information System (INIS)

    Tsykanov, V.A.; Golovanov, V.N.; Mayorshin, A.A.; Yurchenko, A.D.; Ilyenko, S.A.; Syuzev, V.N.

    2000-01-01

    The main trends of production of pellet MOX-fuel for the WWER reactors using the trial-experimental equipment at SSC RF RIAR are set forth. The main realized parameters of fabrication of MOX-fuel pellets are presented. The content of the reactor tests program is considered with allowance for their licensing requirements for the WWER reactors. (author)

  19. Fuel conditioning facility electrorefiner cadmium vapor trap operation

    International Nuclear Information System (INIS)

    Vaden, D. E.

    1998-01-01

    Processing sodium-bonded spent nuclear fuel at the Fuel Conditioning Facility at Argonne National Laboratory-West involves an electrometallurgical process employing a molten LiCl-KCl salt covering a pool of molten cadmium. Previous research has shown that the cadmium dissolves in the salt as a gas, diffuses through the salt layer and vaporizes at the salt surface. This cadmium vapor condenses on cool surfaces, causing equipment operation and handling problems. Using a cadmium vapor trap to condense the cadmium vapors and reflux them back to the electrorefiner has mitigated equipment problems and improved electrorefiner operations

  20. Impact of receipt of coprocessed uranium/plutonium on advanced accountability concepts and fabrication facilities. Addendum 1 to application of advanced accountability concepts in mixed oxide fabrication

    International Nuclear Information System (INIS)

    Bastin, J.J.; Jump, M.J.; Lange, R.A.; Randall, C.C.

    1977-11-01

    The Phase I study of the application of advanced accountability methods (DYMAC) in a uranium/plutonium mixed oxide facility was extended to assess the effect of coprocessed UO 2 --PuO 2 feed on the observations made in the original Phase I effort and on the proposed Phase II program. The retention of plutonium mixed with uranium throughout the process was also considered. This addendum reports that coprocessed feed would have minimal effect on the DYMAC program, except in the areas of material specifications, starting material delivery schedule, and labor requirements. Each of these areas is addressed, as are the impact of coprocessed feed at a large fuel fabrication facility and the changes needed in the dirty scrap recovery process to maintain the lower plutonium levels which may be required by future nonproliferation philosophy. An amended schedule for Phase II is included

  1. 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

  2. Criticality safety evaluation of the fuel cycle facility electrorefiner

    International Nuclear Information System (INIS)

    Lell, R.M.; Mariani, R.D.; Fujita, E.K.; Benedict, R.W.; Turski, R.B.

    1993-01-01

    The integral Fast Reactor (IFR) being developed by Argonne National Laboratory (ANL) combines the advantages of metal-fueled, liquid-metal cooled reactors and a closed-loop fuel cycle. Some of the primary advantages are passive safety for the reactor and resistance to diversion for the heavy metal in the fuel cycle. in addition, the IFR pyroprocess recycles all the long-lived actinide activation products for casting into new fuel pins so that they may be burned in the reactor. A key component in the Fuel Cycle Facility (FCF) recycling process is the electrorefiner (ER) in which the actinides are separated from the fission products. In the process, the metal fuel is electrochemically dissolved into a high-temperature molten salt, and electrorefined uranium or uranium/plutonium products are deposited at cathodes. This report addresses the new and innovative aspects of the criticality analysis ensuing from processing metallic fuel, rather than metal oxide fuel, and from processing the spent fuel in batch operations. in particular, the criticality analysis employed a mechanistic approach as opposed to a probabilistic one. A probabilistic approach was unsuitable because of a lack of operational experience with some of the processes, rendering the estimation of accident event risk factors difficult. The criticality analysis also incorporated the uncertainties in heavy metal content attending the process items by defining normal operations envelopes (NOES) for key process parameters. The goal was to show that reasonable process uncertainties would be demonstrably safe toward criticality for continuous batch operations provided the key process parameters stayed within their NOES. Consequently the NOEs became the point of departure for accident events in the criticality analysis

  3. Operation results of WWER fuel fabricated by OAO MSZ

    International Nuclear Information System (INIS)

    Asatiani, A.; Balabanov, S.; Komarova, T.

    2013-01-01

    This report presents a statistical analysis concerning the operation results of WWER fuel manufactured by OAO MSZ. For different fuel modifications we noted individual trends demonstrating the change in the achievable fuel assembly (FA) burnup figures; we also gave statistical evaluations of fuel rod (FR) leakage levels occurring in different WWER power units. (authors)

  4. 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

  5. In pile testing program of mixed oxide fuels fabricated under the new GSP method

    International Nuclear Information System (INIS)

    Marinucci, G.; Nobili, A.; Lanchi, M.; Caracchini, R.; Dupont, G.; Galtier, J.

    1983-01-01

    ENEA, AGN and CEA have collaborated in developing and effectuating the new fuel fabrication technique GSP. In that framework, three irradiation experiments in the SILOE reactor are projected for studying certain characteristics of GSP fuel in comparison with standard fuels (type FBR). The three experiments facilitate the analysis of the thermal behaviour and the fuel pin deformations up to a high burnup. In this paper, the experimental devices are described (5 designs) and the irradiation conditions are briefly discussed. (G.J.P.)

  6. Experience of RepU fuel fabrication and operation in WWER reactors

    International Nuclear Information System (INIS)

    Kolosovsky, V.; Asatiani, I.; Sannikov, E.; Novikov, V.; Kuleshov, A.; Mikheev, E.; Proselkov, V.; Plyashkevich, V.; Semchenkov, Y.; Spirkin, E.; Ionov, V.; Pimenov, Y.

    2008-01-01

    Russia has a long-lasting experience in successful utilization of the reprocessed uranium fuel with different types of reactors. Stages of implementing the RepU fuel for WWER reactors are presented.The nuclear design assays, radiation and nuclear safety analysis during fabrication and handling of the fuel are made. The operational experience of commercial batches is summarized. It is shown that the RepU fuel characteristics meet the design limitations, approved for WWER reactors. (authors)

  7. Air quality impacts due to construction of LWR waste management facilities

    International Nuclear Information System (INIS)

    1977-06-01

    Air quality impacts of construction activities and induced housing growth as a result of construction activities were evaluated for four possible facilities in the LWR fuel cycle: a fuel reprocessing facility, fuel storage facility, fuel fabrication plant, and a nuclear power plant. Since the fuel reprocessing facility would require the largest labor force, the impacts of construction of that facility were evaluated in detail

  8. Fuel Flexible Gas Turbine Combustor Flametube Facility Upgraded

    Science.gov (United States)

    Little, James E.; Nemets, Steve A.; Tornabene, Robert T.; Smith, Timothy D.; Frankenfeld, Bruce J.

    2004-01-01

    In fiscal year 2003, test cell 23 of the Research Combustion Laboratory (RCL 23) at the NASA Glenn Research Center was upgraded with the addition of gaseous hydrogen as a working propellant and the addition of a 450-psig air-supply system. Test flexibility was further enhanced by upgrades to the facility control systems. RCL 23 can now test with gaseous hydrogen flow rates up to 0.05 lbm/sec and jet fuel flow rates up to 0.62 lbm/sec. Research airflow rates up to 3 lbm/sec are possible with the 450-psig supply system over a range of inlet temperatures. Nonvitiated, heated air is supplied from a shell and tube heat exchanger. The maximum nonvitiated facility air temperature is 1100 F at 1.5 lbm/sec. Research-section exhaust temperatures are limited to 3200 F because of material and cooling capacity limits. A variety of support systems are available depending on the research hardware configuration. Test section ignition can be provided via either a hydrogen air torch system or an electronic spark system. Emissions measurements are obtained with either pneumatically or electromechanically actuated gas sample probes, and the electromechanical system allows for radial measurements at a user-specified axial location for measurement of emissions profiles. Gas analysis data can be obtained for a variety of species, including carbon monoxide (CO), carbon dioxide (CO2), nitrogen oxides (NO and NOx), oxygen (O2), unburnt hydrocarbons, and unburnt hydrogen. Facility control is accomplished with a programmable logic control system. Facility operations have been upgraded to a system based on graphical user interface control screens. A data system is available for real-time acquisition and monitoring of both measurements in engineering units and performance calculations. The upgrades have made RCL 23 a highly flexible facility for research into low emissions gas turbine combustor concepts, and the flame tube configuration inherently allows for a variety of fuel nozzle

  9. Fabrication of MOX fuel element clusters for irradiation in PWL, CIRUS

    International Nuclear Information System (INIS)

    Roy, P.R.; Purushotham, D.S.C.; Majumdar, S.

    1983-01-01

    Three clusters, each containing 6 zircaloy-2 clad short length fuel elements of either MOX or UO 2 fuel pellets were fabricated for irradiation in pressurized water loop of CIRUS. The major objectives of the programme were: (a) to optimize the various fabrication parameters for developing a flow sheet for MOX fuel element fabrication; (b) to study the performance of the MOX fuel elements at a peak heat flux of 110 W/cm 2 ; and (c) to study the effect of various fuel pellet design changes on the behaviour of the fuel element under irradiation. Two clusters, one each of UO 2 and MOX, have been successfully irradiated to the required burn-up level and are now awaiting post irradiation examinations. The third MOX cluster is still undergoing irradiation. Fabrication of these fuel elements involved considerable amount of developing work related to the fabrication of the MOX fuel pellets and the element welding technique and is reported in detail in this report. (author)

  10. Radiation and physical protection challenges at advanced nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    Pickett, Susan E.

    2008-01-01

    Full text: The purpose of this study is to examine challenges and opportunities for radiation protection in advanced nuclear reactors and fuel facilities proposed under the Generation IV (GEN IV) initiative which is examining and pursuing the exploration and development of advanced nuclear science and technology; and the Global Nuclear Energy Partnership (GNEP), which seeks to develop worldwide consensus on enabling expanded use of economical, carbon-free nuclear energy to meet growing energy demand. The International Energy Agency projects nuclear power to increase at a rate of 1.3 to 1.5 percent a year over the next 20 years, depending on economic growth. Much of this growth will be in Asia, which, as a whole, currently has plans for 40 new nuclear power plants. Given this increase in demand for new nuclear power facilities, ranging from light water reactors to advanced fuel processing and fabrication facilities, it is necessary for radiation protection and physical protection technologies to keep pace to ensure both worker and public health. This paper is based on a review of current initiatives and the proposed reactors and facilities, primarily the nuclear fuel cycle facilities proposed under the GEN IV and GNEP initiatives. Drawing on the Technology Road map developed under GEN IV, this work examines the potential radiation detection and protection challenges and issues at advanced reactors, including thermal neutron spectrum systems, fast neutron spectrum systems and nuclear fuel recycle facilities. The thermal neutron systems look to improve the efficiency of production of hydrogen or electricity, while the fast neutron systems aim to enable more effective management of actinides through recycling of most components in the discharged fuel. While there are components of these advanced systems that can draw on the current and well-developed radiation protection practices, there will inevitably be opportunities to improve the overall quality of radiation

  11. Development of a Micro- and Nano-Fabrication Facility

    National Research Council Canada - National Science Library

    Couzis, Alexander; Gilchrist, Lane; Maldarelli, Charles; Rumschitzki, David; Vazquez, Maribel

    2004-01-01

    .... This facility has fostered several research grant awards and applications from the PIs, has encouraged the development of undergraduate and graduate courses involving micro and nanotechnology at CCNY...

  12. Fabrication and Testing of Full-Length Single-Cell Externally Fueled Converters for Thermionic Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Schock, Alfred

    1995-08-01

    Paper presented at the 29th IECEC in Monterey, CA in August 1994. The present paper describes the fabrication and testing of full-length prototypcial converters, both unfueled and fueled, and presents parametric results of electrically heated tests.

  13. Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors - I: DUPIC Fuel Fabrication Cost

    International Nuclear Information System (INIS)

    Choi, Hangbok; Ko, Won Il; Yang, Myung Seung

    2001-01-01

    A preliminary conceptual design of a Direct Use of spent Pressurized water reactor (PWR) fuel In Canada deuterium uranium (CANDU) reactors (DUPIC) fuel fabrication plant was studied, which annually converts spent PWR fuel of 400 tonnes heavy element (HE) into CANDU fuel. The capital and operating costs were estimated from the viewpoint of conceptual design. Assuming that the annual discount rate is 5% during the construction (5 yr) and operation period (40 yr) and contingency is 25% of the capital cost, the levelized unit cost (LUC) of DUPIC fuel fabrication was estimated to be 616 $/kg HE, which is mostly governed by annual operation and maintenance costs that correspond to 63% of LUC. Among the operation and maintenance cost components being considered, the waste disposal cost has the dominant effect on LUC (∼49%). From sensitivity analyses of production capacity, discount rate, and contingency, it was found that the production capacity of the plant is the major parameter that affects the LUC

  14. Global nuclear energy partnership fuels transient testing at the Sandia National Laboratories nuclear facilities : planning and facility infrastructure options

    International Nuclear Information System (INIS)

    Kelly, John E.; Wright, Steven Alan; Tikare, Veena; MacLean, Heather J.; Parma, Edward J.Jr; Peters, Curtis D.; Vernon, Milton E.; Pickard, Paul S.

    2007-01-01

    The Global Nuclear Energy Partnership fuels development program is currently developing metallic, oxide, and nitride fuel forms as candidate fuels for an Advanced Burner Reactor. The Advance Burner Reactor is being designed to fission actinides efficiently, thereby reducing the long-term storage requirements for spent fuel repositories. Small fuel samples are being fabricated and evaluated with different transuranic loadings and with extensive burnup using the Advanced Test Reactor. During the next several years, numerous fuel samples will be fabricated, evaluated, and tested, with the eventual goal of developing a transmuter fuel database that supports the down selection to the most suitable fuel type. To provide a comparative database of safety margins for the range of potential transmuter fuels, this report describes a plan to conduct a set of early transient tests in the Annular Core Research Reactor at Sandia National Laboratories. The Annular Core Research Reactor is uniquely qualified to perform these types of tests because of its wide range of operating capabilities and large dry central cavity which extents through the center of the core. The goal of the fuels testing program is to demonstrate that the design and fabrication processes are of sufficient quality that the fuel will not fail at its design limit--up to a specified burnup, power density, and operating temperature. Transient testing is required to determine the fuel pin failure thresholds and to demonstrate that adequate fuel failure margins exist during the postulated design basis accidents

  15. Global nuclear energy partnership fuels transient testing at the Sandia National Laboratories nuclear facilities : planning and facility infrastructure options.

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, John E.; Wright, Steven Alan; Tikare, Veena; MacLean, Heather J. (Idaho National Laboratory, Idaho Falls, ID); Parma, Edward J., Jr.; Peters, Curtis D.; Vernon, Milton E.; Pickard, Paul S.

    2007-10-01

    The Global Nuclear Energy Partnership fuels development program is currently developing metallic, oxide, and nitride fuel forms as candidate fuels for an Advanced Burner Reactor. The Advance Burner Reactor is being designed to fission actinides efficiently, thereby reducing the long-term storage requirements for spent fuel repositories. Small fuel samples are being fabricated and evaluated with different transuranic loadings and with extensive burnup using the Advanced Test Reactor. During the next several years, numerous fuel samples will be fabricated, evaluated, and tested, with the eventual goal of developing a transmuter fuel database that supports the down selection to the most suitable fuel type. To provide a comparative database of safety margins for the range of potential transmuter fuels, this report describes a plan to conduct a set of early transient tests in the Annular Core Research Reactor at Sandia National Laboratories. The Annular Core Research Reactor is uniquely qualified to perform these types of tests because of its wide range of operating capabilities and large dry central cavity which extents through the center of the core. The goal of the fuels testing program is to demonstrate that the design and fabrication processes are of sufficient quality that the fuel will not fail at its design limit--up to a specified burnup, power density, and operating temperature. Transient testing is required to determine the fuel pin failure thresholds and to demonstrate that adequate fuel failure margins exist during the postulated design basis accidents.

  16. Key differences in the fabrication of US and German TRISO-coated particle fuel, and their implications on fuel performance

    International Nuclear Information System (INIS)

    Petti, D.A.; Buongiorno, J.; Maki, J.T.; Miller, G.K.; Hobbins, R.R.

    2002-01-01

    Historically, the irradiation performance of TRISO-coated gas reactor particle fuel in Germany has been superior to that in the US. German fuel generally displayed in-pile gas release values that were three orders of magnitude lower than US fuel. Thus, we have critically examined the TRISO-coated fuel fabrication processes in the US and German and the associated irradiation database with a goal of understanding why the German fuel behaves acceptably, why the US fuel has not faired as well, and what process/production parameters impart the reliable performance to this fuel form. The postirradiation examination results are also reviewed to identify failure mechanisms that may be the cause of the poorer US irradiation performance. This comparison will help determine the roles that particle fuel process/product attributes and irradiation conditions (burnup, fast neutron fluence, temperature, degree of acceleration, power per particle) have on the behavior of the fuel during irradiation and provide a more quantitative linkage between acceptable processing parameters, as-fabricated fuel properties and subsequent in-reactor performance. (author)

  17. Plant overview of JNFL MOX fuel fabrication plant (J-MOX)

    International Nuclear Information System (INIS)

    Hiruta, Kazuhiko; Suzuki, Masataka; Shimizu, Junji; Suzuki, Kazumi; Yamamoto, Yutaka; Deguchi, Morimoto; Fujimaki, Kazunori

    2005-01-01

    In April 2005, JNFL submitted METI an application for the permission of MOX fuel fabrication business for JNFL MOX Fuel Fabrication Plant (J-MOX). Accordingly, safeguards formalities and discussion with the Agency have been also started for J-MOX as an official project. This report describes J-MOX plant overview and also presents outline of J-MOX by focusing on safeguards features and planned material accountancy method. (author)

  18. Improvements in the consistency of fabrication and the reliability of nuclear fuels through quality assurance

    International Nuclear Information System (INIS)

    Sifferlen, R.

    1976-01-01

    By establishing correlations between rejection level and fabrication parameters, quality assurance can guide corrective action for improving the consistency of fabrication and the reliability of fuel elements. The author cites examples relating to the quality of the uranium in metallic fuels, the influence of the parent metal on the welding of zirconium alloys, the behaviour of the springs inside the cladding during the welding of plugs and the behaviour of uranium oxide pellets. (author)

  19. Basis for Interim Operation for Fuel Supply Shutdown Facility

    International Nuclear Information System (INIS)

    BENECKE, M.W.

    2003-01-01

    This document establishes the Basis for Interim Operation (BIO) for the Fuel Supply Shutdown Facility (FSS) as managed by the 300 Area Deactivation Project (300 ADP) organization in accordance with the requirements of the Project Hanford Management Contract procedure (PHMC) HNF-PRO-700, ''Safety Analysis and Technical Safety Requirements''. A hazard classification (Benecke 2003a) has been prepared for the facility in accordance with DOE-STD-1027-92 resulting in the assignment of Hazard Category 3 for FSS Facility buildings that store N Reactor fuel materials (303-B, 3712, and 3716). All others are designated Industrial buildings. It is concluded that the risks associated with the current and planned operational mode of the FSS Facility (uranium storage, uranium repackaging and shipment, cleanup, and transition activities, etc.) are acceptable. The potential radiological dose and toxicological consequences for a range of credible uranium storage building have been analyzed using Hanford accepted methods. Risk Class designations are summarized for representative events in Table 1.6-1. Mitigation was not considered for any event except the random fire event that exceeds predicted consequences based on existing source and combustible loading because of an inadvertent increase in combustible loading. For that event, a housekeeping program to manage transient combustibles is credited to reduce the probability. An additional administrative control is established to protect assumptions regarding source term by limiting inventories of fuel and combustible materials. Another is established to maintain the criticality safety program. Additional defense-in-depth controls are established to perform fire protection system testing, inspection, and maintenance to ensure predicted availability of those systems, and to maintain the radiological control program. It is also concluded that because an accidental nuclear criticality is not credible based on the low uranium enrichment

  20. Refabrication of spent fuel pins provided with instrumentation by means of the FABRICE process

    International Nuclear Information System (INIS)

    Garcia, A.; Malard, J.P.; Mery, M.; Vouillot, M.

    1986-05-01

    The Fabrice ''hot cell refabrication'' process for small pins from very long irradiated fuel elements was developed at the CEA to allow parametic studies of the irradiation behavior of pins from nuclear power plants. The present note describes the adaptation of techniques to the fabrication of pins provided with important instrumentations [fr

  1. Facile fabrication of dendritic silver structures and their surface ...

    Indian Academy of Sciences (India)

    A simple and efficient approach was developed to fabricate silver dendrites by Cu reducing Ag+ in AgNO3 solution. The growth speed, morphologies and structures of the silver dendrites strongly depend on AgNO3 concentration and reaction time. The silver dendrites were formed from nanosheets and the crystal structure ...

  2. An advanced C/S system for fuel cycle facilities

    International Nuclear Information System (INIS)

    Mizuno, O.; Seya, M.; Ohtani, T.; Takahashi, S.

    1987-01-01

    The IAEA have given a top priority to the R and D theme of ''Speed up Review of Video Tape'' in the ''Safeguards Needs List (APRIL, 1986) by WG-R and D Needs Under Support Programmes.'' Because it always consumes work time for inspector to review conventional surveillance records (film or video tape). PNC is now developing Advanced Containment and Surveillance (C/S) system which enables quick reviewing and direct retrieval of surveillance image records. In recent years, a rapid progress has been made in the field of optical image processing technology. Combining this optical technology with appropriate trigger devices, all activities to be surveyed in a specified area (storage) can be recorded as a series of fine motion pictures. Electronic seal and γ-ray detector have been developed as the trigger signal generators for personal entry and SNM passage. Inspector can review all operators' recorded activities in a short time through easy-operating man-machine system. Also, a direct retrieval system based on frame index contributes to save much time for reviewing. This advanced C/S system could be applied to storages of nuclear fuel cycle facilities. A Proto-type system has been introduced at the fuel assembly storage in the PNC MOX (Mixed Oxide) plant PPFF and a demonstration system will be introduced in the new MOX plant PFPF (Plutonium Fuel Production Facility)

  3. Criticality control during conditioning of spent nuclear fuel in the Fuel Cycle Facility

    International Nuclear Information System (INIS)

    Lell, R.M.; Khalil, H.S.

    1994-01-01

    Spent nuclear fuel may be unacceptable for direct repository storage because of composition, enrichment, form, physical condition, or the presence of undesirable materials such as sodium. Fuel types which are not acceptable for direct storage must be processed or conditioned to produce physical forms which can safely be stored in a repository. One possible approach to conditioning is the pyroprocess implemented in the Fuel Cycle Facility (FCF) at Argonne National Laboratory-West. Conditioning of binary (U-Zr) and ternary (U-Pu-Zr) metallic fuels from the EBR-2 reactor is used to demonstrate the process. Criticality safety considerations limit batch sizes during the conditioning steps and provide one constraint on the final form of conditioned material. Criticality safety during conditioning is assured by the integration of criticality safety analysis, equipment design, process development, a measurement program, accountability procedures, and a computerized Mass Tracking System. Criticality issues related to storage and shipment of conditioned material have been examined

  4. Facile Fabrication of Multifunctional Hybrid Silk Fabrics with Controllable Surface Wettability and Laundering Durability.

    Science.gov (United States)

    Chen, Fengxiang; Yang, Huiyu; Liu, Xin; Chen, Dongzhi; Xiao, Xingfang; Liu, Keshuai; Li, Jing; Cheng, Fan; Dong, Binhai; Zhou, Yingshan; Guo, Zhiguang; Qin, Yong; Wang, Shimin; Xu, Weilin

    2016-03-02

    To obtain a hydrophobic surface, TiO2 coatings are deposited on the surface of silk fabric using atomic layer deposition (ALD) to realize a hierarchical roughness structure. The surface morphology and topography, structure, and wettability properties of bare silk fabric and TiO2-coated silk fabrics thus prepared are evaluated using scanning electron microscopy (SEM), field-emission scanning electron microscopy (FESEM), scanning probe microscope (SPM), X-ray diffraction (XRD), static water contact angles (WCAs), and roll-off angles, respectively. The surfaces of the silk fabrics with the TiO2 coatings exhibit higher surface roughnesses compared with those of the bare silk fabric. Importantly, the hydrophobic and laundering durability properties of the TiO2-coated silk fabrics are largely improved by increasing the thickness of the ALD TiO2 coating. Meanwhile, the ALD process has a litter effect on the service performance of silk fabric. Overall, TiO2 coating using an ALD process is recognized as a promising approach to produce hydrophobic surfaces for elastic materials.

  5. Present status of quality control on MOX fuel fabrication in JNC

    International Nuclear Information System (INIS)

    Kobayashi, H.; Suzuki, T.; Nakae, N.; Ohtani, T.

    2000-01-01

    Japan Nuclear Cycle Development Institute (JNC has developed MOX fuels for Advanced Thermal Reactor FUGEN and Fast Reactors JOYO and MONJU for these 30 years. The total amount of MOX fuels fabricated in JNC reached to be 157 tons as of the end of March 1999. No fuel pin failure has been found in the fuels loaded in these reactors. This result shows that the fabrication technology of MOX has been established and the quality control has also been done adequately in the fabrication process. This paper describes the basic concept and associated procedures of quality control related to MOX fuel fabrication in JNC high-lighting the key procedures of quality control in MOX fuel fabrication, measurements of plutonium enriched zone, determination of plutonium concentration, and the advanced method of visual inspection for the surface appearance of the products. The homogeneity of plutonium and uranium is one of key properties from the viewpoint of local heating in reactor operation and dissolubility in reprocessing. Thus plutonium enriched zone (Pu spot) should be detected with an adequate method. The alpha autoradiography has been used to identify the characteristics of the zone. The procedure of alpha autoradiography is established and is successfully applied to the MOX fabrication process. The determination of plutonium concentration is also one of key properties for quality control. JNC has gained the experience relating to Titration, Coulometry, and Isotope Dilution Mass Spectrometry (IDMS) which are known as determination method. IDMS has been used for the determination of plutonium concentration in MOX fabricated in JNC. The advanced method for the visual inspection of the surface appearance of the product. has been developed and demonstrated. The method utilizing the VTR is considered to be well harmonized with fully automated MOX fuel fabrication process. (author)

  6. Procedure for estimating facility decommissioning costs for non-fuel-cycle nuclear facilities

    International Nuclear Information System (INIS)

    Short, S.M.

    1988-01-01

    The Nuclear Regulatory Commission (NRC) staff has been reappraising its regulatory position relative to the decommissioning of nuclear facilities over the last several years. Approximately 30 reports covering the technology, safety, and costs of decommissioning reference nuclear facilities have been published during this period in support of this effort. One of these reports, Technology, Safety, and Costs of Decommissioning Reference Non-Fuel-Cycle Nuclear Facilities (NUREG/CR-1754), was published in 1981 and was felt by the NRC staff to be outdated. The Pacific Northwest Laboratory (PNL) was asked by the NRC staff to revise the information provided in this report to reflect the latest information on decommissioning technology and costs and publish the results as an addendum to the previous report. During the course of this study, the NRC staff also asked that PNL provide a simplified procedure for estimating decommissioning costs of non-fuel-cycle nuclear facilities. The purpose being to provide NRC staff with the means to easily generate their own estimate of decommissioning costs for a given facility for comparison against a licensee's submittal. This report presents the procedure developed for use by NRC staff

  7. Fabrication, fabrication control and in-core follow up of 4 LEU leader fuel elements based on U{sub 3}Si{sub 2} in RECH-1

    Energy Technology Data Exchange (ETDEWEB)

    Chavez, J.C.; Barrera, M.; Olivares, L.; Lisboa, J. [Department of Nuclear Materials, Chilean Nuclear Energy Commission, Santiago (Chile)

    1999-07-01

    The RECH-1 MTR reactor has been converted from HEU to MEU (45% enrichment) and the decision to a LEU (20% enrichment) conversion was taken some years ago. This LEU conversion decision involved a local fuel development and fabrication based on U{sub 3}Si{sub 2}-Al dispersion fuel, and a fabrication qualification stage that resulted in four fuel elements fully complying with established fabrication standards for this type of fuel. This report-presents relevant points of these four leaders fuel elements fabrication, in particular a fuel plate core homogeneity control development. A summary of the intended in core follow-up studies for the leaders fuel elements is also presented here. (author)

  8. Fuels and materials testing capabilities in Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Baker, R.B.; Chastain, S.A.; Culley, G.E.; Ethridge, J.L.; Lovell, A.J.; Newland, D.J.; Pember, L.A.; Puigh, R.J.; Waltar, A.E.

    1989-01-01

    The Fast Flux Test Facility (FFTF) reactor, which started operating in 1982, is a 400 MWt sodium-cooled fast neutron reactor located in Hanford, Washington State, and operated by Westinghouse Hanford Co. under contract with U.S. Department of Energy. The reactor has a wide variety of functions for irradiation tests and special tests, and its major purpose is the irradiation of fuel and material for liquid metal reactor, nuclear reactor and space reactor projects. The review first describes major technical specifications and current conditions of the FFTF reactor. Then the plan for irradiation testing is outlined focusing on general features, fuel pin/assembly irradiation tests, and absorber irradiation tests. Assemblies for special tests include the material open test assembly (MOTA), fuel open test assembly (FOTA), closed loop in-reactor assembly (CLIRA), and other special fuel assemblies. An interim examination and maintenance cell (FFTF/IEM cell) and other hot cells are used for nondestructive/destructive tests and physical/mechanical properties test of material after irradiation. (N.K.)

  9. The final disposal facility of spent nuclear fuel

    International Nuclear Information System (INIS)

    Prvakova, S.; Necas, V.

    2001-01-01

    Today the most serious problem in the area of nuclear power engineering is the management of spent nuclear fuel. Due to its very high radioactivity the nuclear waste must be isolated from the environment. The perspective solution of nuclear fuel cycle is the final disposal into geological formations. Today there is no disposal facility all over the world. There are only underground research laboratories in the well developed countries like the USA, France, Japan, Germany, Sweden, Switzerland and Belgium. From the economical point of view the most suitable appears to build a few international repositories. According to the political and social aspect each of the country prepare his own project of the deep repository. The status of those programmes in different countries is described. The development of methods for the long-term management of radioactive waste is necessity in all countries that have had nuclear programmes. (authors)

  10. Facile fabrication of superhydrophobic surfaces with hierarchical structures.

    Science.gov (United States)

    Lee, Eunyoung; Lee, Kun-Hong

    2018-03-06

    Hierarchical structures were fabricated on the surfaces of SUS304 plates using a one-step process of direct microwave irradiation under a carbon dioxide atmosphere. The surface nanostructures were composed of chrome-doped hematite single crystals. Superhydrophobic surfaces with a water contact angle up to 169° were obtained by chemical modification of the hierarchical structures. The samples maintained superhydrophobicity under NaCl solution up to 2 weeks.

  11. Fabrication of Tungsten-UO 2 Hexagonal-Celled Fuel-Element Configurations

    Energy Technology Data Exchange (ETDEWEB)

    Goetsch, R.R.; Cover, P.W.; Gripshover, P.J.; Wilson, W.J.

    1964-12-04

    The gas-pressure-bonding process is being evaluated as a means of fabricating tungsten-UO 2 hexagonal-celled fuel geometries. A two-part study was initiated to optimize the fuel materials and to develp the required fixturing and loading techniques. Production of fueled tungsten-coated UO 2 particles in in progress so that geometries embodying coated particles or coated particles plus fine tungsten powder can be evaluated. Tests to data have shown the rquirement for a pretreatment in which a gaseous oxide phase is removed. Initial loading and fixturing procedures were proven satisfactory by the fabrication of a 19-cylindrical-hole hexagonal-type composite.

  12. Financing Strategies For A Nuclear Fuel Cycle Facility

    International Nuclear Information System (INIS)

    David Shropshire; Sharon Chandler

    2006-01-01

    To help meet the nation's energy needs, recycling of partially used nuclear fuel is required to close the nuclear fuel cycle, but implementing this step will require considerable investment. This report evaluates financing scenarios for integrating recycling facilities into the nuclear fuel cycle. A range of options from fully government owned to fully private owned were evaluated using DPL (Decision Programming Language 6.0), which can systematically optimize outcomes based on user-defined criteria (e.g., lowest lifecycle cost, lowest unit cost). This evaluation concludes that the lowest unit costs and lifetime costs are found for a fully government-owned financing strategy, due to government forgiveness of debt as sunk costs. However, this does not mean that the facilities should necessarily be constructed and operated by the government. The costs for hybrid combinations of public and private (commercial) financed options can compete under some circumstances with the costs of the government option. This analysis shows that commercial operations have potential to be economical, but there is presently no incentive for private industry involvement. The Nuclear Waste Policy Act (NWPA) currently establishes government ownership of partially used commercial nuclear fuel. In addition, the recently announced Global Nuclear Energy Partnership (GNEP) suggests fuels from several countries will be recycled in the United States as part of an international governmental agreement; this also assumes government ownership. Overwhelmingly, uncertainty in annual facility capacity led to the greatest variations in unit costs necessary for recovery of operating and capital expenditures; the ability to determine annual capacity will be a driving factor in setting unit costs. For private ventures, the costs of capital, especially equity interest rates, dominate the balance sheet; and the annual operating costs, forgiveness of debt, and overnight costs dominate the costs computed for the

  13. Guide to Permitting Hydrogen Motor Fuel Dispensing Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Rivkin, Carl [National Renewable Energy Lab. (NREL), Golden, CO (United States); Buttner, William [National Renewable Energy Lab. (NREL), Golden, CO (United States); Burgess, Robert [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2016-03-28

    The purpose of this guide is to assist project developers, permitting officials, code enforcement officials, and other parties involved in developing permit applications and approving the implementation of hydrogen motor fuel dispensing facilities. The guide facilitates the identification of the elements to be addressed in the permitting of a project as it progresses through the approval process; the specific requirements associated with those elements; and the applicable (or potentially applicable) codes and standards by which to determine whether the specific requirements have been met. The guide attempts to identify all applicable codes and standards relevant to the permitting requirements.

  14. Fabrication experience of aluminium clad aluminium matrix dispersion fuels at BARC

    International Nuclear Information System (INIS)

    Ganguly, C.; Hedge, P.W.; Prasad, G.J.

    1995-01-01

    Aluminium clad, aluminium matrix plate type dispersion fuels have been fabricated in BARC in recent years as part of fuel development programme for small non-power research reactors. The present paper describes the flowsheet developed for fabrication of Al-UAl x , Al-U 3 Si 2 and Al-U 3 O 8 fuels at BARC. The Al-20% U alloy fuel for KAMINI neutron radiography reactor was prepared by 'melting and casting' route, followed by picture framing and roll-bonding. For higher 'U' density fuels namely, Al-UAl x , Al-U 3 O 8 and Al-U 3 Si 2 the 'powder metallurgy' route was followed for preparation of fuel meat. The novel features in fabrication route were: addition of Zr for stabilizing UAl 3 phase in Al-20% U alloy; x-ray radiography and microdensitometric scanning of radiographs for location of fuel outline inside fuel element and for confirming homogeneous distribution of fissile atoms; immersion ultrasonic testing for confirming good bonding between mating Al surface of the fuel plate. (author)

  15. Decontamination and decommissioning of the Argonne National Laboratory Building 350 Plutonium Fabrication Facility. Final report

    International Nuclear Information System (INIS)

    Kline, W.H.; Moe, H.J.; Lahey, T.J.

    1985-02-01

    In 1973, Argonne National Laboratory began consolidating and upgrading its plutonium-handling operations with the result that the research fuel-fabrication facility located in Building 350 was shut down and declared surplus. Sixteen of the twenty-three gloveboxes which comprised the system were disassembled and relocated for reuse or placed into controlled storage during 1974 but, due to funding constraints, full-scale decommissioning did not start until 1978. Since that time the fourteen remaining contaminated gloveboxes, including all internal and external equipment as well as the associated ventilation systems, have been assayed for radioactive content, dismantled, size reduced to fit acceptable packaging and sent to a US Department of Energy (DOE) transuranic retrievable-storage site or to a DOE low-level nuclear waste burial ground. The project which was completed in 1983, required 5 years to accomplish, 32 man years of effort, produced some 540 m 3 (19,000 ft 3 ) of radioactive waste of which 60% was TRU, and cost 2.4 million dollars

  16. Development of decommissioning technology for nuclear fuel facility

    International Nuclear Information System (INIS)

    Tanimoto, Ken-ichi

    1998-01-01

    There are many kinds of objects for decommissioning and their properties are greatly different in respects of morphology, constituent materials, contamination history, etc. Therefore, the techniques for decontamination and dismantlement are required to have a great applicability. In addition, most of contamination nuclides have long half-life and so, it is desirable to rapidly take measures to stop or close a contaminated facility. In consideration of these characteristics developments of elementary techniques for decontamination have been attempted. This report summarized the present states of decommissioning technology for nuclear fuel facility. The function and performance of each elementary technique were examined through test operation and simulation was made for the important techniques of them aiming at generalization and optimization. For remote handling technology, two operation tools; 'metal splitting saw cutting tool' and 'plasma cutting tool' were produced and utilizations of these tools in combination with a robot for conveyance are under investigation now. (M.N.)

  17. Implications of multinational arrangements for nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    Muench, E.; Richter, B.; Stein, G.

    1980-01-01

    In the recently concluded INFCE study a variety of possibilities to minimize the proliferation risk was discussed, and their applicability in the nuclear fuel cycle was investigated. It was found that safeguards still play a central part as an anti-proliferation measure. Aspect of institutional arrangements with the aim of placing nuclear material processing and storage facilities under multinational or international auspices is the basis and goal of this study, as in international discussions some degree of proliferation hindrance is attributed to such models. In the assessment of the internationalization of nuclear facilities as an anti-proliferation measure two aspects have to be emphasized: Firstly, internationalization may be understood as a political measure to hinder proliferation, and secondly, no additional control effort should be caused by the possible complementary character to safeguards. 5 refs

  18. Development of decommissioning technology for nuclear fuel facility

    Energy Technology Data Exchange (ETDEWEB)

    Tanimoto, Ken-ichi [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-01-01

    There are many kinds of objects for decommissioning and their properties are greatly different in respects of morphology, constituent materials, contamination history, etc. Therefore, the techniques for decontamination and dismantlement are required to have a great applicability. In addition, most of contamination nuclides have long half-life and so, it is desirable to rapidly take measures to stop or close a contaminated facility. In consideration of these characteristics developments of elementary techniques for decontamination have been attempted. This report summarized the present states of decommissioning technology for nuclear fuel facility. The function and performance of each elementary technique were examined through test operation and simulation was made for the important techniques of them aiming at generalization and optimization. For remote handling technology, two operation tools; `metal splitting saw cutting tool` and `plasma cutting tool` were produced and utilizations of these tools in combination with a robot for conveyance are under investigation now. (M.N.)

  19. 10 CFR Appendix O to Part 110 - Illustrative List of Fuel Element Fabrication Plant Equipment and Components Under NRC's Export...

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Illustrative List of Fuel Element Fabrication Plant... Appendix O to Part 110—Illustrative List of Fuel Element Fabrication Plant Equipment and Components Under NRC's Export Licensing Authority Note: Nuclear fuel elements are manufactured from source or special...

  20. Thermomechanical evaluation of the fuel assemblies fabricated in the ININ

    International Nuclear Information System (INIS)

    Hernandez L, H.; Ortiz V, J.

    2005-01-01

    The pilot plant of fuel production of the National Institute of Nuclear Research (ININ) provided to the Laguna Verde Nuclear Power Plant (CNLV) four fuel assemblies type GE9B. The fuel irradiation was carried out in the unit 1 of the CNLV during four operation cycles, highlighting the fact that in their third cycle the four assemblies were placed in the center of the reactor core. In the Nuclear Systems Department (DSN) of the ININ it has been carried out studies to evaluate their neutron performance and to be able to determine the exposure levels of this fuels. Its also outlines the necessity to carry out a study of the thermomechanical behavior of the fuel rods that compose the assemblies, through computational codes that simulate their performance so much thermal as mechanical. For such purpose has been developing in the DSN the FETMA code, together with the codes that compose the system Fuel Management System (FMS), which evaluates the thermomechanical performance of fuel elements. In this work were used the FETMA and FEMAXI codes (developed by JAERI) to study the thermomechanical performance of the fuel elements manufactured in the ININ. (Author)

  1. Facile and Gram-scale Synthesis of Metal-free Catalysts: Toward Realistic Applications for Fuel Cells

    Science.gov (United States)

    Kim, Ok-Hee; Cho, Yong-Hun; Chung, Dong Young; Kim, Min Jeong; Yoo, Ji Mun; Park, Ji Eun; Choe, Heeman; Sung, Yung-Eun

    2015-03-01

    Although numerous reports on nonprecious metal catalysts for replacing expensive Pt-based catalysts have been published, few of these studies have demonstrated their practical application in fuel cells. In this work, we report graphitic carbon nitride and carbon nanofiber hybrid materials synthesized by a facile and gram-scale method via liquid-based reactions, without the use of toxic materials or a high pressure-high temperature reactor, for use as fuel cell cathodes. The resulting materials exhibited remarkable methanol tolerance, selectivity, and stability even without a metal dopant. Furthermore, these completely metal-free catalysts exhibited outstanding performance as cathode materials in an actual fuel cell device: a membrane electrode assembly with both acidic and alkaline polymer electrolytes. The fabrication method and remarkable performance of the single cell produced in this study represent progressive steps toward the realistic application of metal-free cathode electrocatalysts in fuel cells.

  2. Gel-sphere-pac fuel for thermal reactors: assessment of fabrication technology and irradiation performance

    Energy Technology Data Exchange (ETDEWEB)

    Beatty, R.L. Norman, R.E.; Notz, K.J. (comps.)

    1979-11-01

    Recent interest in proliferation-resistant fuel cycles for light-water reactors has focused attention on spiked plutonium and /sup 233/U-Th fuels, requiring remote refabrication. The gel-sphere-pac process for fabricating metal-clad fuel elements has drawn special attention because it involves fewer steps. Gel-sphere-pac fabrication technology involves two major areas: the preparation of fuel spheres of high density and loading these spheres into rods in an efficiently packed geometry. Gel sphere preparation involves three major steps: preparation of a sol or of a special solution (broth), gelation of droplets of sol or broth to give semirigid spheres of controlled size, and drying and sintering these spheres to a high density. Gelation may be accomplished by water extraction (suitable only for sols) or ammonia gelation (suitable for both sols and broths but used almost exclusively with broths). Ammonia gelation can be accomplished either externally, via ammonia gas and ammonium hydroxide, or internally via an added ammonia generator such as hexamethylenetetramine. Sphere-pac fuel rod fabrication involves controlled blending and metering of three sizes of spheres into the rod and packing by low- to medium-energy vibration to achieve about 88% smear density; these sizes have diametral ratios of about 40:10:1 and are blended in size fraction amounts of about 60% coarse, 18% medium, and 22% fine. Irradiation test results indicate that sphere-pac fuel performs at least as well as pellet fuel, and may in fact offer an advantage in significantly reducing mechanical and chemical interaction between the fuel and cladding. The normal feed for gel sphere preparation, heavy metal nitrate solution, is the usual product of fuel reprocessing, so that fabrication of gel spheres performs all the functions performed by both conversion and pellet fabrication in the case of pellet technology.

  3. SEM and TEM Characterization of As-Fabricated U-7Mo Disperson Fuel Plates

    International Nuclear Information System (INIS)

    Keiser, D.D. Jr.; Yao, B.; Perez, E.; Sohn, Y.H.

    2009-01-01

    The starting microstructure of a dispersion fuel plate can have a dramatic impact on the overall performance of the plate during irradiation. To improve the understanding of the as-fabricated microstructures of dispersion fuel plates, SEM and TEM analysis have been performed on RERTR-9A archive fuel plates, which went through an additional hot isostatic procsssing (HIP) step during fabrication. The fuel plates had depleted U-7Mo fuel particles dispersed in either Al-2Si or 4043 Al alloy matrix. For the characterized samples, it was observed that a large fraction of the ?-phase U-7Mo alloy particles had decomposed during fabrication, and in areas near the fuel/matrix interface where the transformation products were present significant fuel/matrix interaction had occurred. Relatively thin Si-rich interaction layers were also observed around the U-7Mo particles. In the thick interaction layers, (U)(Al,Si)3 and U6Mo4Al43 were identified, and in the thin interaction layers U(Al,Si)3, U3Si3Al2, U3Si5, and USi1.88-type phases were observed. The U3Si3Al2 phase contained some Mo. Based on the results of this work, exposure of dispersion fuel plates to relatively high temperatures during fabrication impacts the overall microstructure, particularly the nature of the interaction layers around the fuel particles. The time and temperature of fabrication should be carefully controlled in order to produce the most uniform Si-rich layers around the U-7Mo particles.

  4. Facile fabrication of properties-controllable graphene sheet

    Science.gov (United States)

    Choi, Jin Sik; Choi, Hongkyw; Kim, Ki-Chul; Jeong, Hu Young; Yu, Young-Jun; Kim, Jin Tae; Kim, Jin-Soo; Shin, Jin-Wook; Cho, Hyunsu; Choi, Choon-Gi

    2016-04-01

    Graphene has been received a considerable amount of attention as a transparent conducting electrode (TCE) which may be able to replace indium tin oxide (ITO) to overcome the significant weakness of the poor flexibility of ITO. Given that graphene is the thinnest 2-dimensional (2D) material known, it shows extremely high flexibility, and its lateral periodic honeycomb structure of sp2-bonded carbon atoms enables ~2.3% of incident light absorption per layer. However, there is a trade-off between the electrical resistance and the optical transmittance, and the fixed absorption rate in graphene limits is use when fabricating devices. Therefore, a more efficient method which continuously controls the optical and electrical properties of graphene is needed. Here, we introduce a method which controls the optical transmittance and the electrical resistance of graphene through various thicknesses of the top Cu layers with a Cu/Ni metal catalyst structure used to fabricate a planar mesh pattern of single and multi-layer graphene. We exhibit a continuous transmittance change from 85% (MLG) to 97.6% (SLG) at an incident light wavelength of 550 nm on graphene samples simultaneously grown in a CVD quartz tube. We also investigate the relationships between the sheet resistances.

  5. Material accountancy in an electrometallurgical Fuel Conditioning Facility

    International Nuclear Information System (INIS)

    Vaden, D.; Benedict, R.W.; Goff, K.M.; Keyes, R.W.; Mariani, R.D.; Bucher, R.G.; Yacout, A.M.

    1996-01-01

    The Fuel Conditioning Facility (FCF) treats spent nuclear fuel using an electrometallurgical process that separates the uranium from the fission products, sodium thermal bond and cladding materials. Material accountancy is necessary at FCF for two reasons: first, it provides a mechanism for detecting a potential loss of nuclear material for safeguards and security; second, it provides a periodic check of inventories to ensure that processes and material are under control. By weighing material entering and leaving a process, and using sampling results to determine composition, an inventory difference (ID) results when the measured inventory is compared to the predicted inventory. The ID and its uncertainty, based on error propagation, determines the degree of assurance that an operation proceeded according to expectations. FCF uses the ID calculation in two ways: closeout, which is the ID and uncertainty for a particular operational step, and material accountancy, which determines an ID and its associated uncertainty for a material balance area through several operational steps. Material accountancy over the whole facility for a specified time period assists in detecting diversion of nuclear material. Data from depleted uranium operations are presented to illustrate the method used in FCF

  6. Development of the advanced CANDU technology -Development of CANDU advanced fuel fabrication technology-

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Chang Bum; Park, Choon Hoh; Park, Chul Joo; Kwon, Woo Joo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This project is carrying out jointly with AECL to develop CANFLEX fuel which can enhance reactor safety, fuel economy and can be used with various fuel cycles (natural U, slightly enriched U, other advanced fuel). The final goal of this research is to load the CANFLEX fuel in commercial CANDU reactor for demonstration irradiation. The annual portion of research activities performed during this year are followings ; The detail design of CANFLEX-NU fuel was determined. Based on this design, various fabrication drawings and process specifications were revised. The seventeen CANFLEX-NU fuel bundles for reactivity test in ZED-2 and out-pile test, two CANFLEX-SEU fuel bundles for demo-irradiation in NRU were fabricated. Advanced tack welding machine was designed and sequence control software of automatic assembly welder was developed. The basic researches related to fabrication processes, such as weld evaluation by ECT, effect of additives in UO{sub 2}, thermal stabilities of Zr based metallic glasses, were curried out. 51 figs, 22 tabs, 42 refs. (Author).

  7. Topical papers on heavy water, fuel fabrication and reactors

    International Nuclear Information System (INIS)

    1978-01-01

    A total of four papers is presented. The first contribution of the Federal Republic of Germany reviews the market situation for reactors and the relations between reactor producers and buyers as reflected in sales agreements. The second West German contribution gives a world-wide survey of fuel element production as well as of fuel and fuel element demand up to the year 2000. The Canadian paper discusses the future prospects of heavy-water production, while the Ecuador contribution deals with small and medium-sized nuclear power plants

  8. Visualization test facility of nuclear fuel rod emergency cooling system

    International Nuclear Information System (INIS)

    Candido, Marcos Antonio; Mesquita, Amir Zacarias; Rezende, Hugo Cesar; Santos, Andre Augusto Campagnole

    2013-01-01

    The nuclear reactors safety is determined according to their protection against the consequences that may result from postulated accidents. The Loss of Coolant Accident (LOCA) is one the most important design basis accidents (DBA). The failure may be due to rupture of the primary loop piping. Another accident postulated is due to lack of power in the pump motors in the primary circuit. In both cases the reactor shut down automatically due to the decrease of reactivity to maintain the fissions, and to the drop of control rods. In the event of an accident it is necessary to maintain the coolant flow to remove the fuel elements residual heat, which remains after shut down. This heat is a significant amount of the maximum thermal power generated in normal operation (about 7%). Recently this event has been quite prominent in the press due to the reactor accident in Fukushima nuclear power station. This paper presents the experimental facility under rebuilding at the Thermal Hydraulic Laboratory of the Nuclear Technology Development Center (CDTN) that has the objective of monitoring and visualization of the process of emergency cooling of a nuclear fuel rod simulator, heated by Joule effect. The system will help the comprehension of the heat transfer process during reflooding after a loss of coolant accident in the fuel of light water reactor core. (author)

  9. Low-Cost Facile Fabrication of Flexible Transparent Copper Electrodes by Nanosecond Laser Ablation

    KAUST Repository

    Paeng, Dongwoo

    2015-03-27

    © 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim. Low-cost Cu flexible transparent conducting electrodes (FTCEs) are fabricated by facile nanosecond laser ablation. The fabricated Cu FTCEs show excellent opto-electrical properties (transmittance: 83%, sheet resistance: 17.48 Ω sq-1) with outstanding mechanical durability. Successful demonstration of a touch-screen panel confirms the potential applicability of Cu FTCEs to the flexible optoelectronic devices.

  10. 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

  11. Investigation of small scale sphere-pac fuel fabrication plant with external gelation process

    International Nuclear Information System (INIS)

    Maekawa, Kazuhiko; Yoshimura, Tadahiro; Kikuchi, Toshiaki; Hoshino, Yasushi; Munekata, Hideki; Shimizu, Makoto

    2005-02-01

    In feasibility studies on commercialized FBR cycle system, comprehensive system investigation and properties evaluation for candidate FBR cycle systems have been implemented through view point of safety, economics, environmental burden reduction, non-proliferation resistivity, etc. As part of these studies, an investigation of small scale sphere-pac fuel fabrication plant with external gelation process was conducted. Until last fiscal year, equipment layout in cells and overall layout design of the 200t-HM/y scale fuel fabrication plant were conducted as well as schematical design studies on main equipments in gelation and reagent recovery processes of the plant. System property data concerning economics and environmental burden reduction of fuel fabrication plant was also acquired. In this fiscal year, the processes from vibropacking to fuel assemblies storage were added to the investigation range, and a conceptual design of whole fuel fabrication plant was studied as well as deepening the design study on main equipments. The conceptual design study was mainly conducted for small 50t-HM/y scale plant and a revising investigation was done for 200t-HM/y scale plant. Taking the planed comparative evaluation with pellet fuel fabrication system into account, design of equipments which should be equivalent with pellet system, especially in post-vibropacking processes, were standardized in each system. Based on these design studies, system properties data concerning economics and environmental burden reduction of the plant was also acquired. In comparison with existing design, the cell height was lowered on condition that plug type pneumatic system was adopted and fuel fabrication building was downsized by applying rationalized layout design of pellet system to post-vibropacking processes. Reduction of reagent usage at gelation process and rationalization of sintering and O/M controlling processes etc., are foremost tasks. (author)

  12. 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

  13. Evolution of verification technics in mixed oxide fuel fabrication plants

    International Nuclear Information System (INIS)

    Bonucci, A.; Fernandez, F.; Kaiser, S.; Dossogne, P.

    1991-01-01

    In the MOX fabrication plants the increase in throughput and the consequent changes in the conditioning of the nuclear material and in the fabrication process applied is requiring a constant tailoring of the verification methods and techniques to the operational constraints. This paper describes the evolution of the verification scheme as a response to the changed conditions and the experience gained on the technical means used (NDA, DA, C/S). It describes as well the foreseeable evolution and the means which will be necessary to develop or to adapt

  14. Waste to Watts and Water: Enabling Self-Contained Facilities Using Microbial Fuel Cells

    Science.gov (United States)

    2009-03-01

    biofuels and biomass, hydrogen 5 fuel cells, protein - or enzyme-based fuel cells, solar power, wind power, and desalination plants. A brief...Waste to Watts and Water Enabling Self-Contained Facilities Using Microbial Fuel Cells Amanda Sue Birch, P. E. Major, USAFR Air Command and...Water: Enabling Self-Contained Facilities Using Microbial Fuel Cells 5a. CONTRACT NUMBER 5b. GRANT NUMBER 5c. PROGRAM ELEMENT NUMBER 6. AUTHOR(S

  15. Material Control and Accountability Experience at the Fuel Conditioning Facility

    International Nuclear Information System (INIS)

    Vaden, D.; Fredrickson, G.L.

    2007-01-01

    The Fuel Conditioning Facility (FCF) at the Idaho National Laboratory (INL) treats spent nuclear fuel using an electrometallurgical process that separates the uranium from the fission products, sodium thermal bond, and cladding materials. Material accountancy is necessary at FCF for two reasons: 1) it provides a mechanism for detecting a potential loss of nuclear material for safeguards and security, and 2) it provides a periodic check of inventories to ensure that processes and materials are within control limits. Material Control and Accountability is also a Department of Energy (DOE) requirement (DOE Order 474.1). The FCF employs a computer based Mass Tracking (MTG) System to collect, store, retrieve, and process data on all operations that directly affect the flow of materials through the FCF. The MTG System is important for the operations of the FCF because it supports activities such as material control and accountability, criticality safety, and process modeling. To conduct material control and accountability checks and to monitor process performance, mass balances are routinely performed around the process equipment. The equipment used in FCF for pyro-processing consists of two mechanical choppers and two electro-refiners (the Mark-IV with the accompanying element chopper and Mark-V with the accompanying blanket chopper for processing driver fuel and blanket, respectively), and a cathode processor (used for processing both driver fuel and blanket) and casting furnace (mostly used for processing driver fuel). Performing mass balances requires the measurement of the masses and compositions of several process streams and equipment inventories. The masses of process streams are obtained via in-cell balances (i.e., load cells) that weigh containers entering and leaving the process equipment. Samples taken at key locations are analyzed to determine the composition of process streams and equipment inventories. In cases where equipment or containers cannot be

  16. Re-fabrication and re-instrumentation of irradiated LWR fuel rods for irradiation testing at the HFR Petten

    International Nuclear Information System (INIS)

    Fischer, B.; Markgraf, J.F.W.; Puschek, P.; Duijves, K.A.; Haan, K.W. de

    1996-01-01

    LWR fuel testing at the High Flux Reactor (HFR) and the Hot Cells at Petten has been successfully performed with pre-irradiated fuel rod segments. The testing methods have been extended with hot cell techniques for re-fabrication of test fuel rods from full length fuel rods from power reactors; re-instrumentation of pre-irradiated fuel rod segments with pressure sensors; and instrumentation of re-fabricated fuel rods or fuel rod segments with central thermocouple and/or pressure sensors. 5 refs, 9 figs, 5 tabs

  17. The improvement of the fire protections system for nuclear cycle facilities. Formulation of a fire protection guideline for nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    2012-04-01

    The private side Fire Protection Guideline was investigated with respect to the fire having taken place at the nuclear reactor site followed by the Chuetsu-Oki earthquake in Niigata Prefecture in 2007. To improve the fire protection system especially applicable to MOX fuel fabrication facilities, JNES (Japan Nuclear Energy Safety Organization) investigated private guidelines adopted in Japanese Light Water cooled Reactors, the standardized guidelines used in Nuclear Facilities in other countries including USA, and the standards in the chemical plants. The content of the guideline concerns the prevention of the fire breakout, the prevention of fire extension, the reduction of the fire effects, as well as the facility-characteristic protection countermeasures and the fire effect evaluations. (S. Ohno)

  18. Analytical methods used at model facility

    International Nuclear Information System (INIS)

    Wing, N.S.

    1984-01-01

    A description of analytical methods used at the model LEU Fuel Fabrication Facility is presented. The methods include gravimetric uranium analysis, isotopic analysis, fluorimetric analysis, and emission spectroscopy

  19. Design principles and fabrication method for a miniaturized fuel gas combustion reactor

    NARCIS (Netherlands)

    Zhao, Yiyuan; Veltkamp, Henk-Willem; de Boer, Meint J.; Zeng, Yaxiang; Groenesteijn, Jarno; Wiegerink, Remco J.; Lötters, Joost Conrad

    2017-01-01

    Wobbe Index meter is widely used to reflect the energy content and gas quality of a fuel gas mixture. It is highly in demand to downscale from the conventionally large Wobbe Index facility to a miniaturized Wobbe Index meter. Therefore spontaneously combustion of the fuel gas/air mixtures on a

  20. Spent fuel handling and storage facility for an LWR fuel reprocessing plant

    International Nuclear Information System (INIS)

    Baker, W.H.; King, F.D.

    1979-01-01

    The facility will have the capability to handle spent fuel assemblies containing 10 MTHM/day, with 30% if the fuel received in legal weight truck (LWT) casks and the remaining fuel received in rail casks. The storage capacity will be about 30% of the annual throughput of the reprocessing plant. This size will provide space for a working inventory of about 50 days plant throughput and empty storage space to receive any fuel that might be in transit of the reprocessing plant should have an outage. Spent LWR fuel assemblies outside the confines of the shipping cask will be handled and stored underwater. To permit drainage, each water pool will be designed so that it can be isolated from the remaining pools. Pool water quality will be controlled by a filter-deionizer system. Radioactivity in the water will be maintained at less than or equal to 2 x 10 -4 Ci/m 3 ; conductivity will be maintained at 1 to 2 μmho/cm. The temperature of the pool water will be maintained at less than or equal to 40 0 C to retard algae growth and reduce evaporation. Decay heat will be transferred to the environment via a heat exchanger-cooling tower system

  1. 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).

  2. Radiation protection training at uranium hexafluoride and fuel fabrication plants

    International Nuclear Information System (INIS)

    Brodsky, A.; Soong, A.L.; Bell, J.

    1985-05-01

    This report provides general information and references useful for establishing or operating radiation safety training programs in plants that manufacture nuclear fuels, or process uranium compounds that are used in the manufacture of nuclear fuels. In addition to a brief summary of the principles of effective management of radiation safety training, the report also contains an appendix that provides a comprehensive checklist of scientific, safety, and management topics, from which appropriate topics may be selected in preparing training outlines for various job categories or tasks pertaining to the uranium nuclear fuels industry. The report is designed for use by radiation safety training professionals who have the experience to utilize the report to not only select the appropriate topics, but also to tailor the specific details and depth of coverage of each training session to match both employee and management needs of a particular industrial operation. 26 refs., 3 tabs

  3. Fabrication experience with mixed-oxide LWR fuels at the BELGONUCLEAIRE plant

    International Nuclear Information System (INIS)

    Vanhellemont, G.

    1979-01-01

    For nearly 20 years BELGONUCLEAIRE has been involved in a steadily growing effort to increase its production of mixed oxides. This programme has ranged from basic research and process development through a pilot-scale unit to today's mixed-oxide fuel fabrication plant at Dessel, which has been in operation for just over 5 years. The reference fabrication flow sheet includes UO 2 , PuO 2 and a scraped powder preparation, sintered ground pellets as well as rod fabrication and assembling. With regard to quality, attention is especially paid to the process monitoring and quality controls at the qualification step and during the routine production. Entirely different types of thermal UO 2 -PuO 2 fuel pellets, rods and assemblies have been manufactured for PWR and BWR operation. For these fabrications, some diagrams of the results with regard to the required technical specifications are presented. Special emphasis is placed on the occasional deviations of some finished products from the specifications and on the solutions applied to avoid such problems. Concerning the actual capacity of the mixed-oxide fuel fabrication plant, several limiting factors due to the nature of plutonium itself are discussed. Taking into account all these ambient limitations, a reference PWR mixed-oxide fuel output of nominally 18 t/a is obtained. The industrial feasibility of UO 2 -PuO 2 fuel fabrication has been thoroughly demonstrated by the present BELGONUCLEAIRE plant. The experience obtained has led to progressive improvements of the fabrication process and adaptation of the product controls in order to ensure the requested quality levels. (author)

  4. MOX fuel fabrication and utilisation in LWRs worldwide

    International Nuclear Information System (INIS)

    Provost, J.-L.; Schrader, M.; Nomura, S.

    2000-01-01

    Early in the development of the nuclear programme, a large part of the countries using nuclear energy has studied the reprocessing and recycling option in order to develop a safe conditioning of fission products and to recycle fissile materials in reactors. In the sixties, the feasibility of recycling plutonium in LWRs has been successfully demonstrated by several experimentations of MOX rod irradiations in different countries. Based on the background of the MOX behaviour collected during the seventies and on the results of the important MOX experimentation program implemented during this period, a large part of the European utilities decided at the beginning of the eighties to use MOX fuel in LWRs on an industrial scale. The main goals of the utilities were to use as a fuel an available fissile material and to control the stockpile of separated plutonium. Today, the understanding of the behaviour of plutonium fuel has grown significantly since the launch of the first R and D programmes on LWR and FR MOX fuels. Plutonium oxide physical and neutron behaviour is well known, its modelling is now available as well as experimentally validated. Up to now, more than 750 tHM MOX fuel (more than 2000 FAs) have been loaded in 29 PWRs and in 2 BWRs in Europe, corresponding to the recycling of about 35 t of plutonium. Reprocessing/recycling technology has reached maturity in the main nuclear industry countries. Spent fuel reprocessing and recycling of the separated fissile materials remains the main option for the back-end cycle. Today, the operation of MOX-recycling LWRs is considered satisfactory. Experience feedback shows that, in global terms, MOX cores behaviour is equivalent to that of UO 2 cores in terms of operation and safety. (author)

  5. Microstructural Examination to Aid in Understanding Friction Bonding Fabrication Technique for Monolithic Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Karen L. Shropshire

    2008-04-01

    Monolithic nuclear fuel is currently being developed for use in research reactors, and friction bonding (FB) is a technique being developed to help in this fuel’s fabrication. Since both FB and monolithic fuel are new concepts, research is needed to understand the impact of varying FB fabrication parameters on fuel plate characteristics. This thesis research provides insight into the FB process and its application to the monolithic fuel design by recognizing and understanding the microstructural effects of varying fabrication parameters (a) FB tool load, and (b) FB tool face alloy. These two fabrication parameters help drive material temperature during fabrication, and thus the material properties, bond strength, and possible formation of interface reaction layers. This study analyzed temperatures and tool loads measured during those FB processes and examined microstructural characteristics of materials and bonds in samples taken from the resulting fuel plates. This study shows that higher tool load increases aluminum plasticization and forging during FB, and that the tool face alloy helps determine the tool’s heat extraction efficacy. The study concludes that successful aluminum bonds can be attained in fuel plates using a wide range of FB tool loads. The range of tool loads yielding successful uranium-aluminum bonding was not established, but it was demonstrated that such bonding can be attained with FB tool load of 48,900 N (11,000 lbf) when using a FB tool faced with a tungsten alloy. This tool successfully performed FB, and with better results than tools faced with other materials. Results of this study correlate well with results reported for similar aluminum bonding techniques. This study’s results also provide support and validation for other nuclear fuel development studies and conclusions. Recommendations are offered for further research.

  6. Analysis and consideration for the US criteria of nuclear fuel cycle facilities to resist natural disasters

    International Nuclear Information System (INIS)

    Shen Hong

    2013-01-01

    Natural disasters pose a threat to the safety of nuclear facilities. Fukushima nuclear accident tells us that nuclear safety in siting, design and construction shall be strengthened in case of external events caused by natural disasters. This paper first analyzes the DOE criteria of nuclear fuel cycle facilities to resist natural disasters. Then to develop our national criteria for natural disaster resistance of nuclear fuel cycle facilities is suggested, so as to ensure the safety of these facilities. (authors)

  7. Assessment of radioactive material released from a fuel fabrication plant under accidental conditions

    International Nuclear Information System (INIS)

    1981-01-01

    This report evaluates the amounts of fissile material released both inside and outside a mixed oxide fuel fabrication plant (MOFFP) for light water reactors. The first section begins with a descriptive study of fissile material containment systems, and the methods available for quantifying accident occurrence probabilities. In addition to accidents common to all industrial facilities, other much rarer accidents were considered, such as aircraft crashes. The minimum occurrence probability limit for consideration in this study was set at 10 -6 per annum. The second part of this report attempts to assess the consequences of the accidents considered (i.e. with occurrence probabilities exceeding 10 -6 per annum) by determining maximum values for such accidents. Acts of sabotage and other accidents of this type are beyond the scope of this study and were not taken into consideration. The most serious potential accident would be a fire involving all of the glove boxes in the PuO 2 powder calcination and preparation cell, which could release 76.5 mg of PuO 2 powder into the atmosphere; the occurrence probability of such an accident, however, is slight (less than 10 -5 per annum). The second possibility, is a specially nuclear hazard that would release fission products into the atmosphere. The occurrence probability of such an accident is currently evaluated at 10 -3 per annum

  8. 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)

  9. Development of MOX facilities and the impact on the nuclear fuel markets

    International Nuclear Information System (INIS)

    Patterson, J.

    1990-01-01

    Mixed-oxide (MOX) fuel is nearing maturity as a fuel supply option. This paper briefly reviews the history and current status of the MOX fuel market, including the projected increase in demand for MOX fuel as more plutonium becomes available from the operation of commercial irradiated fuel reprocessing plants in Europe. The uncertainties of such projected demand are discussed, together with the anticipated requirements from the next generation of MOX fabrication plants. The impact of the growing demand for MOX fuel is assessed in the traditional sectors of the uranium fuel cycle. Finally, the author turns to a generalized treatment of the economic aspects of MOX fuel utilization, showing the financially attractive regimes of MOX use which will benefit nuclear power utilities and continue to ensure that MOX fuel can consolidate its position as a mature fuel supply option in those countries that have opted to recycle their spent fuel

  10. Revised Analyses of Decommissioning Reference Non-Fuel-Cycle Facilities

    International Nuclear Information System (INIS)

    Bierschbach, M.C.; Haffner, D.R.; Schneider, K.J.; Short, S.M.

    2002-01-01

    Cost information is developed for the conceptual decommissioning of non-fuel-cycle nuclear facilities that represent a significant decommissioning task in terms of decontamination and disposal activities. This study is a re-evaluation of the original study (NUREG/CR-1754 and NUREG/CR-1754, Addendum 1). The reference facilities examined in this study are the same as in the original study and include: a laboratory for the manufacture of 3 H-labeled compounds; a laboratory for the manufacture of 14 C-labeled compounds; a laboratory for the manufacture of 123 I-labeled compounds; a laboratory for the manufacture of 137 Cs sealed sources; a laboratory for the manufacture of 241 Am sealed sources; and an institutional user laboratory. In addition to the laboratories, three reference sites that require some decommissioning effort were also examined. These sites are: (1) a site with a contaminated drain line and hold-up tank; (2) a site with a contaminated ground surface; and (3) a tailings pile containing uranium and thorium residues. Decommissioning of these reference facilities and sites can be accomplished using techniques and equipment that are in common industrial use. Essentially the same technology assumed in the original study is used in this study. For the reference laboratory-type facilities, the study approach is to first evaluate the decommissioning of individual components (e.g., fume hoods, glove boxes, and building surfaces) that are common to many laboratory facilities. The information obtained from analyzing the individual components of each facility are then used to determine the cost, manpower requirements and dose information for the decommissioning of the entire facility. DECON, the objective of the 1988 Rulemaking for materials facilities, is the decommissioning alternative evaluated for the reference laboratories because it results in the release of the facility for restricted or unrestricted use as soon as possible. For a facility, DECON requires

  11. Revised Analyses of Decommissioning Reference Non-Fuel-Cycle Facilities

    Energy Technology Data Exchange (ETDEWEB)

    MC Bierschbach; DR Haffner; KJ Schneider; SM Short

    2002-12-01

    Cost information is developed for the conceptual decommissioning of non-fuel-cycle nuclear facilities that represent a significant decommissioning task in terms of decontamination and disposal activities. This study is a re-evaluation of the original study (NUREG/CR-1754 and NUREG/CR-1754, Addendum 1). The reference facilities examined in this study are the same as in the original study and include: a laboratory for the manufacture of {sup 3}H-labeled compounds; a laboratory for the manufacture of {sup 14}C-labeled compounds; a laboratory for the manufacture of {sup 123}I-labeled compounds; a laboratory for the manufacture of {sup 137}Cs sealed sources; a laboratory for the manufacture of {sup 241}Am sealed sources; and an institutional user laboratory. In addition to the laboratories, three reference sites that require some decommissioning effort were also examined. These sites are: (1) a site with a contaminated drain line and hold-up tank; (2) a site with a contaminated ground surface; and (3) a tailings pile containing uranium and thorium residues. Decommissioning of these reference facilities and sites can be accomplished using techniques and equipment that are in common industrial use. Essentially the same technology assumed in the original study is used in this study. For the reference laboratory-type facilities, the study approach is to first evaluate the decommissioning of individual components (e.g., fume hoods, glove boxes, and building surfaces) that are common to many laboratory facilities. The information obtained from analyzing the individual components of each facility are then used to determine the cost, manpower requirements and dose information for the decommissioning of the entire facility. DECON, the objective of the 1988 Rulemaking for materials facilities, is the decommissioning alternative evaluated for the reference laboratories because it results in the release of the facility for restricted or unrestricted use as soon as possible. For a

  12. Argonne Fuel Cycle Facility ventilation system -- modeling and results

    International Nuclear Information System (INIS)

    Mohr, D.; Feldman, E.E.; Danielson, W.F.

    1995-01-01

    This paper describes an integrated study of the Argonne-West Fuel Cycle Facility (FCF) interconnected ventilation systems during various operations. Analyses and test results include first a nominal condition reflecting balanced pressures and flows followed by several infrequent and off-normal scenarios. This effort is the first study of the FCF ventilation systems as an integrated network wherein the hydraulic effects of all major air systems have been analyzed and tested. The FCF building consists of many interconnected regions in which nuclear fuel is handled, transported and reprocessed. The ventilation systems comprise a large number of ducts, fans, dampers, and filters which together must provide clean, properly conditioned air to the worker occupied spaces of the facility while preventing the spread of airborne radioactive materials to clean am-as or the atmosphere. This objective is achieved by keeping the FCF building at a partial vacuum in which the contaminated areas are kept at lower pressures than the other worker occupied spaces. The ventilation systems of FCF and the EBR-II reactor are analyzed as an integrated totality, as demonstrated. We then developed the network model shown in Fig. 2 for the TORAC code. The scope of this study was to assess the measured results from the acceptance/flow balancing testing and to predict the effects of power failures, hatch and door openings, single-failure faulted conditions, EBR-II isolation, and other infrequent operations. The studies show that the FCF ventilation systems am very controllable and remain stable following off-normal events. In addition, the FCF ventilation system complex is essentially immune to reverse flows and spread of contamination to clean areas during normal and off-normal operation

  13. Advanced fuel for fast breeder reactors: Fabrication and properties and their optimization

    International Nuclear Information System (INIS)

    1988-06-01

    The present design for FBR fuel rods includes usually MOX fuel pellets cladded into stainless steel tubes, together with UO 2 axial blanket and stainless steel hexagonal wrappers. Mixed carbide, nitride and metallic fuels have been tested as alternative fuels in test reactors. Among others, the objectives to develop these alternative fuels are to gain a high breeding ratio, short doubling time and high linear ratings. Fuel rod and assembly designers are now concentrating on finding the combination of optimized fuel, cladding and wrapper materials which could result in improvement of fuel operational reliability under high burnups and load-follow mode of operation. The purpose of the meeting was to review the experience of advanced FBR fuel fabrication technology, its properties before, under and after irradiation, peculiarities of the back-end of the nuclear fuel cycle, and to outline future trends. As a result of the panel discussion, the recommendations on future Agency activities in the area of advanced FBR fuels were developed. A separate abstract was prepared for each of the 10 presentations of this meeting. Refs, figs and tabs

  14. ROSA-IV Large Scale Test Facility (LSTF) system description for second simulated fuel assembly

    International Nuclear Information System (INIS)

    1990-10-01

    The ROSA-IV Program's Large Scale Test Facility (LSTF) is a test facility for integral simulation of thermal-hydraulic response of a pressurized water reactor (PWR) during small break loss-of-coolant accidents (LOCAs) and transients. In this facility, the PWR core nuclear fuel rods are simulated using electric heater rods. The simulated fuel assembly which was installed during the facility construction was replaced with a new one in 1988. The first test with this second simulated fuel assembly was conducted in December 1988. This report describes the facility configuration and characteristics as of this date (December 1988) including the new simulated fuel assembly design and the facility changes which were made during the testing with the first assembly as well as during the renewal of the simulated fuel assembly. (author)

  15. Continuous process of powder production for MOX fuel fabrication according to ''granat'' technology

    International Nuclear Information System (INIS)

    Morkovnikov, V.E.; Raginskiy, L.S.; Pavlinov, A.P.; Chernov, V.A.; Revyakin, V.V.; Varykhanov, V.S.; Revnov, V.N.

    2000-01-01

    During last years the problem of commercial MOX fuel fabrication for nuclear reactors in Russia was solved in a number of directions. The paper deals with the solution of the problem of creating a continuous pilot plant for the production of MOX fuel powders on the basis of the home technology 'Granat', that was tested before on a small-scale pilot-commercial batch-operated plant of the same name and confirmed good results. (authors)

  16. 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

  17. Summary report on fuel development and miniplate fabrication for the RERTR Program, 1978 to 1990

    Energy Technology Data Exchange (ETDEWEB)

    Wiencek, T.C. [Argonne National Lab., IL (United States). Energy Technology Div.

    1995-08-01

    This report summarizes the efforts of the Fabrication Technology Section at Argonne National Laboratory in the program of Reduced Enrichment Research and Test Reactors (RERTR). The main objective of this program was to reduce the amount of high enriched ({approx}93% {sup 235}U) uranium (HEU) used in nonpower reactors. Conversion from low-density (0.8--1.6 g U/cm{sup 3}) HEU fuel elements to highly loaded (up to 7 g U/cm{sup 3}) low-enrichment (<20% {sup 235}U) uranium (LEU) fuel elements allows the same reactor power levels, core designs and sizes to be retained while greatly reducing the possibility of illicit diversion of HEU nuclear fuel. This document is intended as an overview of the period 1978--1990, during which the Section supported this project by fabricating mainly powder metallurgy uranium-silicide dispersion fuel plates. Most of the subjects covered in detail are fabrication-related studies of uranium silicide fuels and fuel plate properties. Some data are included for out-of-pile experiments such as corrosion and compatibility tests. Also briefly covered are most other aspects of the RERTR program such as irradiation tests, full-core demonstrations, and technology transfer. References included are for further information on most aspects of the entire program. A significant portion of the report is devoted to data that were never published in their entirety. The appendices contain a list of previous RERTR reports, ANL fabrication procedures, calculations for phases present in two-phase fuels, chemical analysis of fuels, miniplate characteristics, and a summary of bonding runs made by hot isostatic pressing.

  18. Procedure for the fabrication of ceramic fuel pellets with an adjustable structure

    International Nuclear Information System (INIS)

    Henke, M.; Klemm, U.; Sobek, D.

    1986-01-01

    The invention concerns a procedure for the fabrication of ceramic fuel pellets of UO 2 , PuO 2 , ThO 2 and their mixtures with an adjustable structure. Before or during the milling the particle shaped fuel pellets have been added polyethylenglycol in a 20 - 60 % aqueous solution with an amount of 0.5 - 2.0 % in weight. This additive has an effect on a controlled pore formation and grain growth advancement

  19. Summary report on fuel development and miniplate fabrication for the RERTR Program, 1978 to 1990

    International Nuclear Information System (INIS)

    Wiencek, T.C.

    1995-08-01

    This report summarizes the efforts of the Fabrication Technology Section at Argonne National Laboratory in the program of Reduced Enrichment Research and Test Reactors (RERTR). The main objective of this program was to reduce the amount of high enriched (∼93% 235 U) uranium (HEU) used in nonpower reactors. Conversion from low-density (0.8--1.6 g U/cm 3 ) HEU fuel elements to highly loaded (up to 7 g U/cm 3 ) low-enrichment ( 235 U) uranium (LEU) fuel elements allows the same reactor power levels, core designs and sizes to be retained while greatly reducing the possibility of illicit diversion of HEU nuclear fuel. This document is intended as an overview of the period 1978--1990, during which the Section supported this project by fabricating mainly powder metallurgy uranium-silicide dispersion fuel plates. Most of the subjects covered in detail are fabrication-related studies of uranium silicide fuels and fuel plate properties. Some data are included for out-of-pile experiments such as corrosion and compatibility tests. Also briefly covered are most other aspects of the RERTR program such as irradiation tests, full-core demonstrations, and technology transfer. References included are for further information on most aspects of the entire program. A significant portion of the report is devoted to data that were never published in their entirety. The appendices contain a list of previous RERTR reports, ANL fabrication procedures, calculations for phases present in two-phase fuels, chemical analysis of fuels, miniplate characteristics, and a summary of bonding runs made by hot isostatic pressing

  20. Cost estimate of Olkiluoto disposal facility for spent nuclear fuel

    International Nuclear Information System (INIS)

    Kukkola, T.; Saanio, T.

    2005-03-01

    The cost estimate covers the underground rock characterisation facility ONKALO, the investment and the operating costs of the above and underground facilities, the decommissioning of the encapsulation plant and the closure costs of the repository. The above ground facility is a once-investment; a re-investment takes place after 37 years operation. The repository is extended stepwise thus also the investment take place in stages. Annual operating costs are calculated with different operating efficiencies. The total investment costs of the disposal facility are estimated to be 503 M euro (Million Euros), the total operating costs are 1,923 M euro and the decommissioning and the closure costs are 116 M euro totaling 2,542 M euro. The investment costs of the above ground facility are 142 M euro, the operating costs are 1,678 M euro. The repository investment costs are 360 M euro and the operating costs are 245 M euro. The decommissioning costs are 7 M euro and the closure costs are 109 M euro. The costs are calculated by using the price level of December 2003. The cost estimate is based on a plan, where the spent fuel is encapsulated and the disposal canisters are disposed into the bedrock at a depth of about 420 meters in one storey. In the encapsulation process, the fuel assemblies are closed into composite canisters, in which the inner part of the canister is made of nodular cast iron and the outer wall of copper having a thickness of 50 mm. The inner canister is closed gas-tight by a bolted steel lid, and the electron beam welding method is used to close the outer copper lid. The encapsulation plant is independent and located above the deep repository spaces. The disposal canisters are transported to the repository by the lift. The disposal tunnels are constructed and closed in stages according the disposal canisters disposal. The operating time of the Loviisa nuclear power plant units is assumed to be 50 years and the operating time of the Olkiluoto nuclear power

  1. Final report on the decontamination of the Curium Source Fabrication Facility

    International Nuclear Information System (INIS)

    Schaich, R.W.

    1983-12-01

    The Curium Source Fabrication Facility (CSFF) at Oak Ridge National Laboratory (ORNL) was decontaminated to acceptable contamination levels for maintenance activities, using standard decontamination techniques. Solid and liquid waste volumes were controlled to minimize discharges to the ORNL waste systems. This program required two years of decontamination effort at a total cost of approximately $700K. 5 references, 7 figures, 2 tables

  2. Radioactive Iodine and Krypton Control for Nuclear Fuel Reprocessing Facilities

    Directory of Open Access Journals (Sweden)

    Nick R. Soelberg

    2013-01-01

    Full Text Available The removal of volatile radionuclides generated during used nuclear fuel reprocessing in the US is almost certain to be necessary for the licensing of a reprocessing facility in the US. Various control technologies have been developed, tested, or used over the past 50 years for control of volatile radionuclide emissions from used fuel reprocessing plants. The US DOE has sponsored, since 2009, an Off-gas Sigma Team to perform research and development focused on the most pressing volatile radionuclide control and immobilization problems. In this paper, we focus on the control requirements and methodologies for 85Kr and 129I. Numerous candidate technologies have been studied and developed at laboratory and pilot-plant scales in an effort to meet the need for high iodine control efficiency and to advance alternatives to cryogenic separations for krypton control. Several of these show promising results. Iodine decontamination factors as high as 105, iodine loading capacities, and other adsorption parameters including adsorption rates have been demonstrated under some conditions for both silver zeolite (AgZ and Ag-functionalized aerogel. Sorbents, including an engineered form of AgZ and selected metal organic framework materials (MOFs, have been successfully demonstrated to capture Kr and Xe without the need for separations at cryogenic temperatures.

  3. A study on domino effect in nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    Bozzolan, Jean-Claude

    2006-01-01

    Accidents caused by domino effect are among the most severe accidents in the chemical and process industry. Although the destructive potential of these accidental scenarios is widely known, little attention has been paid to this problem in the technical literature and a complete methodology for quantitative assessment of domino accidents contribution to industrial risk is still lacking. The present study proposed a systematic procedure for the quantitative assessment of the risk caused by domino effect in chemical plants that are part of nuclear fuel cycle plants. This work is based on recent advances in the modeling of fire and explosion damage to process equipment due to different escalation vectors (heat radiation, overpressure and fragment projection). Available data from literature and specific vulnerability models derived for several categories of process equipment had been used in the present work. The proposed procedure is applied to a typical storage area of a reconversion plant situated in a complex that shelters other nuclear fuel cycle facilities. The top-events and escalation vectors are identified, their consequences estimated and credible domino scenarios selected on the basis of their frequencies. (author)

  4. Air conditioning facilities in a fuel reprocessing plant

    International Nuclear Information System (INIS)

    Kawasaki, Michitaka; Oka, Tsutomu

    1987-01-01

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

  5. The implications of plant design on the life-time costs for nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    Macphee, D.S.; Hexter, B.C.; Young, M.P.; Wilson, B.J.

    1997-01-01

    Utilising the experience gained during many years of design and project management of nuclear plant, BNFL is now approaching the final stages of the construction and commissioning of the Sellafield MOX Plant (SMP) in the UK. The paper uses the SMP project to highlight the benefits of these experiences, in particular addressing the implications of the approach to plant design on life time costs. In addition to providing BNFL with a state of the art, commercial scale MOX fuel fabrication facility, the construction of this 120 tHM/yr facility, which is currently in the advanced stages of commissioning, represents a significant demonstration of the design and project management skills of BNFL Engineering Ltd. As well as meeting the main process requirements, the plant design incorporates the highest standards of safety, together with input from the future plant operators and potential customers. As befits a commercial scale plutonium handling facility, SMP also incorporates material accountancy and security provisions that will meet all international requirements. Design, construction and commissioning of this complex and highly automated plant, has benefited from a totally integrated approach to design and documentation that considers not only project implementation but also overall lifetime costs. In addition, project management techniques, developed over many years of major project construction at Sellafield, have been utilised in order to ensure successful project implementation against a background of significant technical challenge and 'fast track' timescales. (author)

  6. Survey of European LWR fuel irradiation test facilities

    International Nuclear Information System (INIS)

    Hardt, P. von der

    1983-01-01

    The first European commercial nuclear power plants (1956) featured gas-cooled thermal reactors. Although there is now a general orientation towards light water cooled plants (with a slight preference for the PWR) a large fraction of the 1982 nuclear generating capacity is still invested in gas-cooled reactors. R and D also continues for the HTGR with its long-term development potential. This paper, however, is limited to a general survey of experimental programmes and facilities for light water reactor fuel testing in Western Europe, particularly inside the European Communities. As it turns out, over a dozen major installations are available, all connected to research reactors in government-funded R and D centres. Their equipment is briefly reviewed. Some 50% of the experimental programmes are carried out in large international collaboration, involving up to 20 organizations per project. Techniques and results are rapidly communicated through frequent meetings and conferences. It is anticipated that a part of the present research reactor-based work will gradually shift to power reactor pool side inspection facilities. (author)

  7. Liquefied Gaseous Fuels Spill Test Facility: Overview of STF capabilities

    International Nuclear Information System (INIS)

    Gray, H.E.

    1993-01-01

    The Liquefied Gaseous Fuels Spill Test Facility (STF) constructed at the Department of Energy's Nevada Test Site is a basic research tool for studying the dynamics of accidental releases of various hazardous liquids. This Facility is designed to (1) discharge, at a controlled rate, a measured volume of hazardous test liquid on a prepared surface of a dry lake bed (Frenchman Lake); (2) monitor and record process operating data, close-in and downwind meteorological data, and downwind gaseous concentration levels; and (3) provide a means to control and monitor these functions from a remote location. The STF will accommodate large and small-scale testing of hazardous test fluid release rates up to 28,000 gallons per minute. Spill volumes up to 52,800 gallons are achievable. Generic categories of fluids that can be tested are cryogenics, isothermals, aerosol-forming materials, and chemically reactive. The phenomena that can be studied include source definition, dispersion, and pool fire/vapor burning. Other capabilities available at the STF include large-scale wind tunnel testing, a small test cell for exposing personnel protective clothing, and an area for developing mitigation techniques

  8. Recycling Of Uranium- And Plutonium-Contaminated Metals From Decommissioning Of The Hanau Fuel Fabrication Plant

    International Nuclear Information System (INIS)

    Kluth, T.; Quade, U.; Lederbrink, F. W.

    2003-01-01

    Decommissioning of a nuclear facility comprises not only actual dismantling but also, above all, management of the resulting residual materials and waste. Siemens Decommissioning Projects (DP) in Hanau has been involved in this task since 1995 when the decision was taken to decommission and dismantle the Hanau Fuel Fabrication Plant. Due to the decommissioning, large amounts of contaminated steel scrap have to be managed. The contamination of this metal scrap can be found almost exclusively in the form of surface contamination. Various decontamination technologies are involved, as there are blasting and wiping. Often these methods are not sufficient to meet the free release limits. In these cases, SIEMENS has decided to melt the scrap at Siempelkamp's melting plant. The plant is licensed according to the German Radiation Protection Ordinance Section 7 (issue of 20.07.2001). The furnace is a medium frequency induction type with a load capacity of 3.2 t and a throughput of 2 t/h for steel melting. For safety reasons, the furnace is widely operated by remote handling. A highly efficient filter system of cyclone, bag filter and HEPA-filter in two lines retains the dust and aerosol activity from the off-gas system. The slag is solidified at the surface of the melt and gripped before pouring the liquid iron into a chill. Since 1989, in total 15,000 t have been molten in the plant, 2,000 t of them having been contaminated steel scrap from the decommissioning of fuel fabrication plants. Decontamination factors could be achieved between 80 and 100 by the high affinity of the uranium to the slag former. The activity is transferred to the slag up to nearly 100 %. Samples taken from metal, slag and dust are analyzed by gamma measurements of the 186 keV line of U235 and the 1001 keV line of Pa234m for U238. All produced ingots showed a remaining activity less than 1 Bq/g and could be released for industrial reuse

  9. Design, fabrication and operating experience of Monju ex-vessel fuel storage tank

    International Nuclear Information System (INIS)

    Yokota, Yoshio; Yamagishi, Yoshiaki; Kuroha, Mitsuo; Inoue, Tatsuya

    1995-01-01

    In FBRs there are two methods of storing and cooling the spent fuel - the in-vessel storage and the ex-vessel storage. Because of the sodium leaks through the tank at the beginning of pre-operation, the utilization of the ex-vessel fuel storage tank (EVST) of some FBR plant has been changed from the ex-vessel fuel storage to the interim fuel transfer tank. This led to reactor designers focusing on the material, structure and fabrication of the carbon steel sodium storage tanks worldwide. The Monju EVST was at the final stage of the design, when the leaks occurred. The lesson learned from that experience and the domestic fabrication technology are reflected to the design and fabrication of the Monju EVST. This paper describes the design, fabrication and R and D results for the tank, and operating experience in functional test. The items to be examined are as follows: (1) Overall structure of the tank and design philosophy on the function, (2) Structure of the cover shielding plug and its design philosophy, (3) Structures of the rotating rack and its bearings, and their design philosophy, (4) Cooling method and its design philosophy, (5) Structure and fabrication of the cooling coil support inside EVST with comparison of leaked case, (6) R and D effort for items above. The fabrication of the Monju EVST started in August 1986 and it was shipped to the site in March 1990. Installation was completed in November 1990, and sodium fill after pre-heating started in 1991. The operation has been continued since September 1992. In 1996 when the first spent fuel is stored, its total functions will be examined. (author)

  10. Composite fuel-cladding tubes and its fabrication

    International Nuclear Information System (INIS)

    Higashinakagawa, Emiko; Kawashima, Junko; Sato, Kanemitsu; Kuwae, Ryosho.

    1985-01-01

    Purpose: To reduce stress-corrosion cracks in a fuel cladding tube. Method: By inserting the sleeve of pure zirconium forming a liner layer into the hollow billet of zirconium alloy as an outer tube, then the composed tube is completed through hot-extruding. Then, the composite tube is pressed, by cold working through several passes, into a tube with a predetermined smaller inner-diameter and thinner wall. Heat treatment is applied between each of the passes of the cold working to recrystallize the zirconium-alloy outer tube substantially, as well as to integrally join the outer tube and the liner layer metal-lurgically. It serves as a buffer for moderating the mechanical interaction with the fuel pellets, whereby the resistance to stress-corrosion cracks can be increased. (Moriyama, K.)

  11. Validation of cleaning method for various parts fabricated at a Beryllium facility

    Energy Technology Data Exchange (ETDEWEB)

    Davis, Cynthia M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-12-15

    This study evaluated and documented a cleaning process that is used to clean parts that are fabricated at a beryllium facility at Los Alamos National Laboratory. The purpose of evaluating this cleaning process was to validate and approve it for future use to assure beryllium surface levels are below the Department of Energy’s release limits without the need to sample all parts leaving the facility. Inhaling or coming in contact with beryllium can cause an immune response that can result in an individual becoming sensitized to beryllium, which can then lead to a disease of the lungs called chronic beryllium disease, and possibly lung cancer. Thirty aluminum and thirty stainless steel parts were fabricated on a lathe in the beryllium facility, as well as thirty-two beryllium parts, for the purpose of testing a parts cleaning method that involved the use of ultrasonic cleaners. A cleaning method was created, documented, validated, and approved, to reduce beryllium contamination.

  12. Evaluating Fuel Leak and Aging Infrastructure at Red Hill, Hawaii, the Largest Underground Fuel Storage Facility in the United States

    Science.gov (United States)

    Learn about how EPA Region 9, Hawaii’s Department of Health, U.S. Navy, and Defense Logistics Agency are working tprotect human health and the environment at the Red Hill Bulk Fuel Storage Facility in Hawaii.

  13. Nuclear fuel cycle facilities in the world (excluding the centrally planned economies)

    International Nuclear Information System (INIS)

    1979-01-01

    Information on the existing, under construction and planned fuel cycle facilities in the various countries is presented. Some thirty countries have activities related to different nuclear fuel cycle steps and the information covers the capacity, status, location, and the names of owners of the facilities

  14. Design, fabrication and operation of a biomass fermentation facility. First quarterly report, October 1-December 31, 1978

    Energy Technology Data Exchange (ETDEWEB)

    O' Neil, D.J.; Colcord, A.R.; Bery, M.K.; Day, S.W.; Roberts, R.S.; El-Barbary, I.A.; Havlicek, S.C.; Anders, M.E.; Sondhi, D.

    1978-01-01

    The objective of this project is to design, fabricate, and operate a fermentation facility which will demonstrate on a pilot-scale level (3 oven-dry tons (ODT) per day of feedstock) the economic and technical feasibility of producing anhydrous ethyl alcohol from lignocellulosic biomass residues (wood, corn stover, and wheat straw principally). The resultant process development unit (PDU) will be flexibly designed so as to evaluate current and projected unit operations, materials of construction, chemical and enzymatic systems which offer the potential of significant technological and economic breakthroughs in alcohol production from biomass. The principal focus of the project is to generate fuels from biomass. As such, in addition to alcohol which can be used as a transportation fuel, by-products are to be directed where possible to fuel applications. The project consists of two parts: (1) conceptual design, and (2) detailed engineering design. The first quarter's activities have focused on a critical review of several aspects of the conceptual design of the 3 ODT/day PDU, viz.: (1) biomass cost, availability, and characterization; (2) pretreatment processes for lignocellulosic residues; (3) hydrolytic processes (enzymatic and acidic); (4) fermentation processes; (5) alcohol recovery systems; (6) by-product streams utilization; and (7) process economics.

  15. Standard format and content of license applications for plutonium processing and fuel fabrication plants

    International Nuclear Information System (INIS)

    1976-01-01

    The standard format suggested for use in applications for licenses to possess and use special nuclear materials in Pu processing and fuel fabrication plants is presented. It covers general description of the plant, summary safety assessment, site characteristics, principal design criteria, plant design, process systems, waste confinement and management, radiation protection, accident safety analysis, conduct of operations, operating controls and limits, and quality assurance

  16. Fabrication experience of Al- sup 233 U and Al-Pu plate fuel for the Purnima III and Kamini research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ganguly, C.; Prasad, G.J.; Mahule, N.; Ghosh, J.K.; Assari, K.V.J.; Chandrasekharan, K.N.P.; Muralidhar, S.; Balan, T.S.; Roy, P.R. (Bhabha Atomic Research Centre, Bombay 400085 (IN))

    1991-10-01

    This paper reports on aluminum-clad Al-20 wt% {sup 233}U and Al-23 wt% Pu plate fuel subassemblies that have been fabricated for the Purnima III critical facility and the Kamini research reactor. The fabrication flow sheet consists of preparing the master alloy using aluminum and uranium or plutonium metals as feed materials, remelting and casting the fuel alloy ingots, rolling, picture framing and sandwiching the fuel alloy between aluminum sheets, roll bonding, locating the fuel alloy core outline by x-ray radiography, and trimming and machining to final dimensions. Metallic molds produce better ingots than graphite ones. The addition of zirconium during melting improves the microstructure of the Al-U and Al-Pu castings and facilitates hot rolling of the ingots. In the subassembly the fuel plates are finally locked in aluminum spacer grooves by a novel roll-swaging technique. High-resolution x-ray radiographs and microdensitometric scans are utilized to confirm the homogeneous distribution of the fissile material in the fuel plates. Nonbond areas are detected by blister testing and immersion ultrasonic testing of the roll-bonded fuel plates.

  17. Fabrication of catalytic electrodes for molten carbonate fuel cells

    Science.gov (United States)

    Smith, James L.

    1988-01-01

    A porous layer of catalyst material suitable for use as an electrode in a molten carbonate fuel cell includes elongated pores substantially extending across the layer thickness. The catalyst layer is prepared by depositing particulate catalyst material into polymeric flocking on a substrate surface by a procedure such as tape casting. The loaded substrate is heated in a series of steps with rising temperatures to set the tape, thermally decompose the substrate with flocking and sinter bond the catalyst particles into a porous catalytic layer with elongated pores across its thickness. Employed as an electrode, the elongated pores provide distribution of reactant gas into contact with catalyst particles wetted by molten electrolyte.

  18. 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...

  19. Special equipment for the fabrication and quality control of nuclear fuel elements

    International Nuclear Information System (INIS)

    Guse, K.; Herbert, W.; Jaeger, K.

    1989-01-01

    For the fabrication of LWR fuel elements, columns are packed of up to 4 m in length, consisting of fuel pellets with different uranium enrichment, their weight and total length to be measured prior to further processing to fuel rods. An automated column packing device has been developed for this purpose. The packing jobs and other tasks are computer-controlled, measured data are stored and are printed out for quality documentation. The forces in the springs of fuel spacers of LWR fuel elements are to be measured and compared with the standard requirements, deviations to be documented. For this task, another computer-controlled, automated device has been developed for measuring the spring forces at all required positions after positioning and fixation of the spacers. (orig./DG) [de

  20. Characterization and property evaluation of (Th–3.75U)O{sub 2+x} fuel pellets fabricated by impregnated agglomerate pelletization (IAP) process

    Energy Technology Data Exchange (ETDEWEB)

    Baghra, C.; Sathe, D.B.; Prakash, A.; Mishra, A.K.; Afzal, Mohd [Advanced Fuel Fabrication Facility (AFFF), Bhabha Atomic Research Center (BARC), Tarapur Complex 401502, Maharashtra (India); Panakkal, J.P., E-mail: panakkal@barctara.gov.in [Advanced Fuel Fabrication Facility (AFFF), Bhabha Atomic Research Center (BARC), Tarapur Complex 401502, Maharashtra (India); Kamath, H.S. [Advanced Fuel Fabrication Facility (AFFF), Bhabha Atomic Research Center (BARC), Tarapur Complex 401502, Maharashtra (India)

    2013-06-15

    Highlights: ► Characterization of (Th–3.75%U)O{sub 2+x} pellet fabricated by a new technique i.e., impregnated agglomerate pelletization (IAP). ► Comparison with pellets fabricated by powder to pellet (POP) and coated agglomerate pelletization processes. ► Solid solution was more complete in POP and IAP pellets. ► Uranium distribution was less uniform in the pellets fabricated by the CAP route. -- Abstract: Impregnated agglomerate pelletization (IAP) process has been developed by Advanced Fuel Fabrication Facility (AFFF), Tarapur for the fabrication of (Th–3.75U)O{sub 2+x} mixed oxide fuel for Indian advanced heavy water reactors (AHWR). In this process, ThO{sub 2} spheroids were impregnated with uranyl nitrate solution and the resultant mixture was compacted to form green pellets which were sintered in the oxidizing atmosphere to obtain high density (Th–3.75%U)O{sub 2+x} pellets. An attempt has been made in this paper to characterize the pellets fabricated by IAP route. The characterization of the sintered IAP pellets was made by X-ray diffraction, EPMA, alpha autoradiography, chemical analysis, O/M ratio, immersion density and optical microscopy. The characteristics of (Th–U)O{sub 2} pellets fabricated by impregnated agglomerate pelletization (IAP) were also compared with those fabricated by coated agglomerate pelletization (CAP), and conventional powder pellet (POP) processes. In case of pellets fabricated by IAP route, XRD data showed the presence of single fluorite phase. The uranium concentration and grain size distribution were found to be uniform throughout the pellet. These phase characteristics were also found to be more uniform in pellets made by both IAP and POP processes as compared to pellets fabricated using the CAP process.