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.

76 FR 65544 - Standard Format and Content of License Applications for Mixed Oxide Fuel Fabrication Facilities

Science.gov (United States)

2011-10-21

... NUCLEAR REGULATORY COMMISSION [NRC-2009-0323] Standard Format and Content of License Applications... revision to regulatory guide (RG) 3.39, ``Standard Format and Content of License Applications for Mixed Oxide Fuel Fabrication Facilities.'' This guide endorses the standard format and content for license...

Design of the MOX fuel fabrication facility

International Nuclear Information System (INIS)

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

2001-01-01

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

Design of the MOX fuel fabrication facility

Energy Technology Data Exchange (ETDEWEB)

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

2001-07-01

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

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)

Waste management state-of-the-art review for mixed-oxide fuel fabrication facilities

International Nuclear Information System (INIS)

Woodsum, H.C.; Goodman, J.

1977-11-01

This report provides a state-of-the-art review of the waste management for mixed-oxide (MOX) fuel fabrication facilities. The intent of this report is to focus on those processes and regulatory issues which have a direct bearing on existing and anticipated future management of transuranic (TRU) wastes from a commercial MOX fuel fabrication faciity. Recent government agency actions are reviewed with regard to their impact on existing and projected waste management regulations; and it is concluded that acceleration in the development of regulations, standards, and criteria is one of the most important factors in the implementation of improved MOX plant waste management techniques. ERDA development programs pertaining to the management of TRU wastes have been reviewed and many promising methods for volume reduction of both solid and liquid wastes are discussed. For solid wastes, these methods include compaction, shredding and baling, combustion, acid digestion, and decontamination by electropolishing or by electrolytic treatment. For liquid wastes, treatment options include evaporation, drying, calcination, flocculation, ion exchange, filtration, reverse osmosis, combustion (of combustible organics), and bioprocessing. Based on this review, it is recommended that ERDA continue with its combustible solid waste volume reduction program and complete these development activities by 1979. Following this, a critical evaluation of solid waste volume reduction techniques should be made to select the most promising systems for a commercial MOX fuel facility

Case study application of the IAEA safeguards assessment methodology to a mixed oxide fuel fabrication facility

International Nuclear Information System (INIS)

Swartz, J.; McDaniel, T.

1981-01-01

Science Applications, Inc. has prepared a case study illustrating the application of an assessment methodology to an international system for safeguarding mixed oxide (MOX) fuel fabrication facilities. This study is the second in a series of case studies which support an effort by the International Atomic Energy Agency (IAEA) and an international Consultant Group to develop a methodology for assessing the effectiveness of IAEA safeguards. 3 refs

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

Critical experiments with mixed oxide fuel

International Nuclear Information System (INIS)

Harris, D.R.

1997-01-01

This paper very briefly outlines technical considerations in performing critical experiments on weapons-grade plutonium mixed oxide fuel assemblies. The experiments proposed would use weapons-grade plutonium and Er 2 O 3 at various dissolved boron levels, and for specific fuel assemblies such as the ABBCE fuel assembly with five large water holes. Technical considerations described include the core, the measurements, safety, security, radiological matters, and licensing. It is concluded that the experiments are feasible at the Rensselaer Polytechnic Institute Reactor Critical Facility. 9 refs

Solid oxide fuel cells fueled with reducible oxides

Science.gov (United States)

Chuang, Steven S.; Fan, Liang Shih

2018-01-09

A direct-electrochemical-oxidation fuel cell for generating electrical energy includes a cathode provided with an electrochemical-reduction catalyst that promotes formation of oxygen ions from an oxygen-containing source at the cathode, a solid-state reduced metal, a solid-state anode provided with an electrochemical-oxidation catalyst that promotes direct electrochemical oxidation of the solid-state reduced metal in the presence of the oxygen ions to produce electrical energy, and an electrolyte disposed to transmit the oxygen ions from the cathode to the solid-state anode. A method of operating a solid oxide fuel cell includes providing a direct-electrochemical-oxidation fuel cell comprising a solid-state reduced metal, oxidizing the solid-state reduced metal in the presence of oxygen ions through direct-electrochemical-oxidation to obtain a solid-state reducible metal oxide, and reducing the solid-state reducible metal oxide to obtain the solid-state reduced metal.

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

Mixed U/Pu oxide fabrication facility for gel-sphere-pac fuel

International Nuclear Information System (INIS)

1978-09-01

This paper describes a conceptual plant which uses the gel-sphere-pac process to fabricate mixed oxide (MOX) fuel and covers (1) fabrication of co-processed MOX fuel and (2) fabrication of co-processed spiked MOX fuel, using 60 Co. The report describes: the fuel fabrication process and plant layout, including scrap and waste processing; and maintenance safety and ventilation measures. A description of the conversion of U and Pu nitrate using a gel sphere process is given in Appendix A

Fast Flux Test Facility fuel and test management: The first 10 years

International Nuclear Information System (INIS)

Bennett, R.A.; Bennett, C.L.; Campbell, L.R.; Dobbin, K.D.; Tang, E.L.

1991-07-01

Core design and fuel and test management have been performed efficiently at the Fast Flux Test Facility. No outages have been extended to adjust core loadings. Development of mixed oxide fuels for advanced liquid metal breeder reactors has been carried out successfully. In fact, the fuel performance is extraordinary. Failures have been so infrequent that further development and refinement of fuel requirements seem appropriate and could lead to a significant reduction in projected electrical busbar costs. The Fast Flux Test Facility is also involved in early metal fuel development tests and appears to be an ideal test bed for any further fuel development or refinement testing. 3 refs., 4 figs., 2 tabs

Safety study of fire protection for nuclear fuel cycle facility

International Nuclear Information System (INIS)

2013-01-01

Insufficiencies in the fire protection system of the nuclear reactor facilities were pointed out when the fire occurred due to the Niigata prefecture-Chuetsu-oki Earthquake in July, 2007. This prompted the revision of the fire protection safety examination guideline for nuclear reactors as well as commercial guidelines. The commercial guidelines have been endorsed by the regulatory body. Now commercial fire protection standards for nuclear facilities such as the design guideline and the management guideline for protecting fire in the Light Water Reactors (LWRs) are available, however, those to apply to the nuclear fuel cycle facilities such as mixed oxide fuel fabrication facility (MFFF) have not been established. For the improvement of fire protection system of the nuclear fuel cycle facilities, the development of a standard for the fire protection, corresponding to the commercial standard for LWRs were required. Thus, Japan Nuclear Energy Safety Organization (JNES) formulated a fire protection guidelines for nuclear fuel cycle facilities as a standard relevant to the fire protection of the nuclear fuel cycle facilities considering functions specific to the nuclear fuel cycle facilities. In formulating the guidelines, investigation has been conduced on the commercial guidelines for nuclear reactors in Japan and the standards relevant to the fire protection of nuclear facilities in USA and other countries as well as non-nuclear industrial fire protection standards. The guideline consists of two parts; Equipments and Management, as the commercial guidances of the nuclear reactor. In addition, the acquisition of fire evaluation data for a components (an electric cabinet, cable, oil etc.) targeted for spread of fire and the evaluation model of fire source were continued for the fire hazard analysis (FHA). (author)

Seismic technology of nuclear fuel cycle facilities: A view of BNFL's approach and methods

International Nuclear Information System (INIS)

Morris, I.R.

2001-01-01

The approach BNFL employs in the seismic qualification of its nuclear fuel cycle facilities is described in this paper. The overall seismic qualification process from design to installation and commissioning is considered. The approach for new facilities, such as the Sellafield Mixed Oxide Fuel Plant and Windscale Vitrification Plant Line 3 currently under construction, is examined. (author)

Dry air oxidation kinetics of K-Basin spent nuclear fuel

International Nuclear Information System (INIS)

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

1998-06-01

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

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

Spent fuel storage facility, Kalpakkam

International Nuclear Information System (INIS)

Shreekumar, B.; Anthony, S.

2017-01-01

Spent Fuel Storage Facility (SFSF), Kalpakkam is designed to store spent fuel arising from PHWRs. Spent fuel is transported in AERB qualified/authorized shipping cask by NPCIL to SFSF by road or rail route. The spent fuel storage facility at Kalpakkam was hot commissioned in December 2006. All systems, structures and components (SSCs) related to safety are designed to meet the operational requirements

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.

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

Development of cables for nuclear fuel processing facilities

International Nuclear Information System (INIS)

Kamimura, Seiji; Seki, Ikuo; Yamamoto, Yasuaki; Matsuyama, Shigeki; Endo, Shigeru; Yagi, Toshiaki; Kawakami, Waichiro.

1988-01-01

Accompanying the development of nuclear power stations, the expansion and repletion of the facilities for nuclear fuel cycle such as fuel reprocessing facilities and waste treatment facilities are requested. In these facilities, there is the environment which is exposed to very high level radiation, and in this case, the cables withstanding 10 MGy radiation dose are required. As the cables meeting such requirement, the new cables having excellent flexibility and radiation resistance were developed. In this paper, the points of material development and the characteristics of cables are reported. Considering the radiation resistance and others, ethylene propylene rubber was selected as the base polymer of the insulator, and polyethylene chlorosulfonate was selected as the sheath material. In order to give excellent radiation resistance, as the anti-rad, energy transfer type aromatic oil that absorbs and dissipates radiation energy and radical trap type anti-oxidant of amine group that catches and stabilizes the radicals generated in the polymer were added. The bromine group burning retarding agent having excellent radiation resistance was applied. In this way, the cables withstanding high radiation dose up to 10 MGy were able to be developed. (K.I.)

46 CFR 108.489 - Helicopter fueling facilities.

Science.gov (United States)

2010-10-01

... 46 Shipping 4 2010-10-01 2010-10-01 false Helicopter fueling facilities. 108.489 Section 108.489... AND EQUIPMENT Fire Extinguishing Systems Fire Protection for Helicopter Facilities § 108.489 Helicopter fueling facilities. (a) Each helicopter fueling facility must have a fire protection system that...

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

Dissolving method for nuclear fuel oxide

International Nuclear Information System (INIS)

Tomiyasu, Hiroshi; Kataoka, Makoto; Asano, Yuichiro; Hasegawa, Shin-ichi; Takashima, Yoichi; Ikeda, Yasuhisa.

1996-01-01

In a method of dissolving oxides of nuclear fuels in an aqueous acid solution, the oxides of the nuclear fuels are dissolved in a state where an oxidizing agent other than the acid is present together in the aqueous acid solution. If chlorate ions (ClO 3 - ) are present together in the aqueous acid solution, the chlorate ions act as a strong oxidizing agent and dissolve nuclear fuels such as UO 2 by oxidation. In addition, a Ce compound which generates Ce(IV) by oxidation is added to the aqueous acid solution, and an ozone (O 3 ) gas is blown thereto to dissolve the oxides of nuclear fuels. Further, the oxides of nuclear fuels are oxidized in a state where ClO 2 is present together in the aqueous acid solution to dissolve the oxides of nuclear fuels. Since oxides of the nuclear fuels are dissolved in a state where the oxidizing agent is present together as described above, the oxides of nuclear fuels can be dissolved even at a room temperature, thereby enabling to use a material such as polytetrafluoroethylene and to dissolve the oxides of nuclear fuels at a reduced cost for dissolution. (T.M.)

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...

Regulation of fuel cycle facilities in the UK

International Nuclear Information System (INIS)

Ascroft-Hutton, H.W.

2001-01-01

The UK has facilities for the production of uranium hexafluoride, its enrichment, conversion into fuel and for the subsequent reprocessing of irradiated fuel and closure of the fuel cycle. All of these facilities must be licensed under UK legislation. HM Nuclear Installations Inspectorate has delegated powers to issue the licence and to attach any conditions it considers necessary in the interests of safety. The fuel cycle facilities in the UK have been licensed since 1971. This paper describes briefly the UK nuclear regulatory framework and the fuel cycle facilities involved. It considers the regulatory practices adopted together with similarities and differences between regulation of fuel cycle facilities and power reactors. The safety issues associated with the fuel cycle are discussed and NII's regulatory strategy for these facilities is set out. (author)

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

International Nuclear Information System (INIS)

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

2012-01-01

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

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

Accountability control system in plutonium fuel facility

International Nuclear Information System (INIS)

Naruki, Kaoru; Aoki, Minoru; Mizuno, Ohichi; Mishima, Tsuyoshi

1979-01-01

More than 30 tons of plutonium-uranium mixed-oxide fuel have been manufactured at the Plutonium Facility in PNC for JOYO, FUGEN and DCA (Deuterium Critical Assembly) and for the purpose of irradiation tests. This report reviews the nuclear material accountability control system adopted in the Plutonium Facility. Initially, the main objective of the system was the criticality control of fissible materials at various stages of fuel manufacturing. The first part of this report describes the functions and the structure of the control system. A flow chart is provided to show the various stages of material flow and their associated computer files. The system is composed of the following three sub-systems: procedures of nuclear material transfer; PIT (Physical Inventory Taking); data retrieval, report preparation and file maintenance. OMR (Optical Mark Reader) sheets are used to record the nuclear material transfer. The MUF (Materials Unaccounted For) are evaluated by PIT every three months through computer processing based on the OMR sheets. The MUF ratio of Pu handled in the facility every year from 1966 to 1977 are presented by a curve, indicating that the MUF ratio was kept well under 0.5% for every project (JOYO, FUGEN, and DCA). As for the Pu safeguards, the MBA (Material Balance Area) and the KMP (Key Measurement Point) in the facility of PNC are illustrated. The general idea of the projected PINC (Plutonium Inventory Control) system in PNC is also shortly explained. (Aoki, K.)

Licensed fuel facility status report

International Nuclear Information System (INIS)

1990-04-01

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 NRC 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

Licensed fuel facility status report

International Nuclear Information System (INIS)

Joy, D.; Brown, C.

1993-04-01

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 NRC 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

Simulation with GOTHIC of experiments Oxidation of fuel in Air

International Nuclear Information System (INIS)

Martinez-Murillo Mendez, J. C.

2012-01-01

In the present work has been addressed for the first time la simulation with the GOTHIC code, experiments oxidation and ignition of SFP in phase 1. This work represents a solid starting point for analysis of specific degradation of fuel in the pools of our facilities.

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

Long term integrity of spent fuel and construction materials for dry storage facilities

Energy Technology Data Exchange (ETDEWEB)

Saegusa, T [CRIEPI (Japan)

2012-07-01

In Japan, two dry storage facilities at reactor sites have already been operating since 1995 and 2002, respectively. Additionally, a large scale dry storage facility away from reactor sites is under safety examination for license near the coast and desired to start its operation in 2010. Its final storage capacity is 5,000tU. It is therefore necessary to obtain and evaluate the related data on integrity of spent fuels loaded into and construction materials of casks during long term dry storage. The objectives are: - Spent fuel rod: To evaluate hydrogen migration along axial fuel direction on irradiated claddings stored for twenty years in air; To evaluate pellet oxidation behaviour for high burn-up UO{sub 2} fuels; - Construction materials for dry storage facilities: To evaluate long term reliability of welded stainless steel canister under stress corrosion cracking (SCC) environment; To evaluate long term integrity of concrete cask under carbonation and salt attack environment; To evaluate integrity of sealability of metal gasket under long term storage and short term accidental impact force.

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

International Nuclear Information System (INIS)

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

1977-01-01

ozon in the cryogenic system to remove krypton 85. Therefore, the off-gas processing and cleanup systems receive particular attention with respect to safety of nuclear fuel cycle facilities. The consequences of credible accidents in mixed oxide fuel fabrication plants, i.e. for instance criticality, explosions, leaks in glove boxes or Pu-nitrate tanks are considered. It is shown that credible accidents can be controlled and release of radioactive contaminants can be minimized by the combination of multiple physical barriers with graduated underpressure, corresponding off-gas systems and process safety systems

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

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

Direct Coal Oxidation in Modified Solid Oxide Fuel Cells

DEFF Research Database (Denmark)

Deleebeeck, Lisa; Gil, Vanesa; Ippolito, Davide

2017-01-01

Hybrid direct carbon fuel cells employ a classical solid oxide fuel cell together with carbon dispersed in a carbonate melt on the anode side. In a European project, the utilization of various coals has been investigated with and without addition of an oxidation catalyst to the carbon-carbonate s......Hybrid direct carbon fuel cells employ a classical solid oxide fuel cell together with carbon dispersed in a carbonate melt on the anode side. In a European project, the utilization of various coals has been investigated with and without addition of an oxidation catalyst to the carbon......-carbonate slurry or anode layer. The nature of the coal affects both open circuit voltage and power output. Highest OCV and power densities were observed for bituminous coal and by adding manganese oxide or praseodymium-doped ceria to the carbon/carbonate mixture. Comparing the carbon black fueled performance...... bituminous coal (73 mW/cm2)....

Remote mixed oxide fabrication facility development. Volume 2. State-of-the-art review of remote maintenance system technology

International Nuclear Information System (INIS)

Horgos, R.M.; Masch, M.L.

1979-06-01

This report provides a state-of-the-art review of remote systems technology, which includes manipulators, process connectors, vision systems and specialized process systems. A proposed mixed oxide fuel fabrication facility was reviewed and evaluated for identification of major remote maintenance and repair tasks. The technological areas were evaluated on the basis of their suitability or applicability for remote maintenance and repair of a proposed fully remote operating mixed oxide fuel fabrication facility. A technological base exists from which the design criteria for a reliable, remote operating facility can be established. Commercially available systems and components, along with those remote technologies now in development, will require modifications to adapt them to specific plant designs and requirements

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

Direct Coal Oxidation in Modified Solid Oxide Fuel Cells

DEFF Research Database (Denmark)

Deleebeeck, Lisa; Gil, Vanesa; Ippolito, Davide

2015-01-01

Hybrid direct carbon fuel cells employ a classical solid oxide fuel cell together with carbon dispersed in a carbonate melt on the anode side. In a European project, the utilization of various coals has been investigated with and without addition of an oxidation catalyst to the carbon-carbonate s......Hybrid direct carbon fuel cells employ a classical solid oxide fuel cell together with carbon dispersed in a carbonate melt on the anode side. In a European project, the utilization of various coals has been investigated with and without addition of an oxidation catalyst to the carbon......-carbonate slurry or anode layer. The nature of the coal affects both open circuit voltage and power output. Highest OCV and power densities were observed for bituminous coal and by adding manganese oxide or praseodymium-doped ceria to the carbon/carbonate mixture. Comparing the carbon black fueled performance...... bituminous coal (73 mW/cm2). © 2015 ECS - The Electrochemical Society...

Design impacts of safeguards and security requirements for a US MOX fuel fabrication facility

International Nuclear Information System (INIS)

Erkkila, B.H.; Rinard, P.M.; Thomas, K.E.; Zack, N.R.; Jaeger, C.D.

1998-01-01

The disposition of plutonium that is no longer required for the nation's defense is being structured to mitigate risks associated with the material's availability. In the 1997 Record of Decision, the US Government endorsed a dual-track approach that could employ domestic commercial reactors to effect the disposition of a portion of the plutonium in the form of mixed oxide (MOX) reactor fuels. To support this decision, the Office of Materials Disposition requested preparation of a document that would review US requirements for safeguards and security and describe their impact on the design of a MOX fuel fabrication facility. The intended users are potential bidders for the construction and operation of the facility. The document emphasizes the relevant DOE Orders but also considers the Nuclear Regulatory Commission (NRC) requirements. Where they are significantly different, the authors have highlighted this difference and provided guidance on the impact to the facility design. Finally, the impacts of International Atomic Energy Agency (IAEA) safeguards on facility design are discussed. Security and materials control and accountability issues that influence facility design are emphasized in each area of discussion. This paper will discuss the prepared report and the issues associated with facility design for implementing practical, modern safeguards and security systems into a new MOX fuel fabrication facility

Heating subsurface formations by oxidizing fuel on a fuel carrier

Science.gov (United States)

Costello, Michael; Vinegar, Harold J.

2012-10-02

A method of heating a portion of a subsurface formation includes drawing fuel on a fuel carrier through an opening formed in the formation. Oxidant is supplied to the fuel at one or more locations in the opening. The fuel is combusted with the oxidant to provide heat to the formation.

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)

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 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. 5 refs., 4 figs

Fuel-cycle cost comparisons with oxide and silicide fuels

International Nuclear Information System (INIS)

Matos, J.E.; Freese, K.E.

1982-01-01

This paper addresses fuel cycle cost comparisons for a generic 10 MW reactor with HEU aluminide fuel and with LEU oxide and silicide fuels in several fuel element geometries. The intention of this study is to provide a consistent assessment of various design options from a cost point of view. Fuel cycle cost benefits could result if a number of reactors were to utilize fuel elements with the same number or different numbers of the same standard fuel plate. Data are presented to quantify these potential cost benefits. This analysis shows that there are a number of fuel element designs using LEU oxide or silicide fuels that have either the same or lower total fuel cycle costs than the HEU design. Use of these fuels with the uranium densities considered requires that they are successfully demonstrated and licensed

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

Energy Technology Data Exchange (ETDEWEB)

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. (DLC)

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

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)

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

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)

The cascad spent fuel dry storage facility

International Nuclear Information System (INIS)

Guay, P.; Bonnet, C.

1991-01-01

France has a wide variety of experimental spent fuels different from LWR spent fuel discharged from commercial reactors. Reprocessing such fuels would thus require the development and construction of special facilities. The French Atomic Energy Commission (CEA) has consequently opted for long-term interim storage of these spent fuels over a period of 50 years. Comparative studies of different storage concepts have been conducted on the basis of safety (mainly containment barriers and cooling), economic, modular design and operating flexibility criteria. These studies have shown that dry storage in a concrete vault cooled by natural convection is the best solution. A research and development program including theoretical investigations and mock-up tests confirmed the feasibility of cooling by natural convection and the validity of design rules applied for fuel storage. A facility called CASCAD was built at the CEA's Cadarache Nuclear Research Center, where it has been operational since mid-1990. This paper describes the CASCAD facility and indicates how its concept can be applied to storage of LWR fuel assemblies

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.)

Jet Fuel Based High Pressure Solid Oxide Fuel Cell System

Science.gov (United States)

Gummalla, Mallika (Inventor); Yamanis, Jean (Inventor); Olsommer, Benoit (Inventor); Dardas, Zissis (Inventor); Bayt, Robert (Inventor); Srinivasan, Hari (Inventor); Dasgupta, Arindam (Inventor); Hardin, Larry (Inventor)

2015-01-01

A power system for an aircraft includes a solid oxide fuel cell system which generates electric power for the aircraft and an exhaust stream; and a heat exchanger for transferring heat from the exhaust stream of the solid oxide fuel cell to a heat requiring system or component of the aircraft. The heat can be transferred to fuel for the primary engine of the aircraft. Further, the same fuel can be used to power both the primary engine and the SOFC. A heat exchanger is positioned to cool reformate before feeding to the fuel cell. SOFC exhaust is treated and used as inerting gas. Finally, oxidant to the SOFC can be obtained from the aircraft cabin, or exterior, or both.

A conceptual model for the fuel oxidation of defective fuel

International Nuclear Information System (INIS)

Higgs, J.D.; Lewis, B.J.; Thompson, W.T.; He, Z.

2007-01-01

A mechanistic conceptual model has been developed to predict the fuel oxidation behaviour in operating defective fuel elements for water-cooled nuclear reactors. This theoretical work accounts for gas-phase transport and sheath reactions in the fuel-to-sheath gap to determine the local oxygen potential. An improved thermodynamic analysis has also been incorporated into the model to describe the equilibrium state of the oxidized fuel. The fuel oxidation kinetics treatment accounts for multi-phase transport including normal diffusion and thermodiffusion for interstitial oxygen migration in the solid, as well as gas-phase transport in the fuel pellet cracks. The fuel oxidation treatment is further coupled to a heat conduction equation. A numerical solution of the coupled transport equations is obtained by a finite-element technique with the FEMLAB 3.1 software package. The model is able to provide radial-axial profiles of the oxygen-to-uranium ratio and the fuel temperatures as a function of time in the defective element for a wide range of element powers and defect sizes. The model results are assessed against coulometric titration measurements of the oxygen-to-metal profile for pellet samples taken from ten spent defective elements discharged from the National Research Universal Reactor at the Chalk River Laboratories and commercial reactors

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)

Solid Oxide Fuel Cell Experimental Laboratory

Data.gov (United States)

Federal Laboratory Consortium — NETL’s Solid Oxide Fuel Cell Experimental Laboratory in Morgantown, WV, gives researchers access to models and simulations that predict how solid oxide fuel cells...

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

International Nuclear Information System (INIS)

Hilton, B.A.

2000-01-01

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

Proceedings of the Topical Meeting on the safety of nuclear fuel cycle intermediate storage facilities

International Nuclear Information System (INIS)

1998-01-01

The CSNI Working Group on Fuel Cycle Safety held an International Topical Meeting on safety aspects of Intermediate Storage Facilities in Newby Bridge, England, from 28 to 30 October 1997. The main purpose of the meeting was to provide a forum for the exchange of information on the technical issues on the safety of nuclear fuel cycle facilities (intermediate storage). Titles of the papers are: An international view on the safety challenges to interim storage of spent fuel. Interim storage of intermediate and high-level waste in Belgium: a description and safety aspects. Encapsulated intermediate level waste product stores at Sellafield. Safety of interim storage facilities of spent fuel: the international dimension and the IAEA's activities. Reprocessing of irradiated fuel and radwaste conditioning at Belgoprocess site: an overview. Retrieval of wastes from interim storage silos at Sellafield. Outline of the fire and explosion of the bituminization facility and the activities of the investigation committee (STAIJAERI). The fire and explosion incident of the bituminization facility and the lessons learned from the incident. Study on the scenario of the fire incident and related analysis. Study on the scenario of the explosion incident and related analysis. Accident investigation board report on the May 14, 1997 chemical explosion at the plutonium reclamation facility, Hanford site, Richland, Washington. Dry interim storage of spent nuclear fuel elements in Germany. Safe and effective system for the bulk receipt and storage of light water reactor fuel prior to reprocessing. Receiving and storage of glass canisters at vitrified waste storage center of Japan Nuclear Fuel Ltd. Design and operational experience of dry cask storage systems. Sellafield MOX plant; Plant safety design (BNFL). The assessment of fault studies for intermediate term waste storage facilities within the UK nuclear regulatory regime. Non-active and active commissioning of the thermal oxide

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

Emergency planning for fuel cycle facilities

International Nuclear Information System (INIS)

Lacey, L.R.

1991-01-01

In April 1989, NRC published new emergency planning regulations which apply to certain by-product, source, and special nuclear materials licensees including most fuel cycle facilities. In addition to these NRC regulations, other regulatory agencies such as EPA, OSHA, and DOT have regulations concerning emergency planning or notification that may apply to fuel cycle facilities. Emergency planning requirements address such areas as emergency classification, organization, notification and activation, assessment, corrective and protective measures, emergency facilities and equipment, maintaining preparedness, records and reports, and recovery. This article reviews applicable regulatory requirements and guidance, then concentrates on implementation strategies to produce an effective emergency response capability

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

Electrolytes for solid oxide fuel cells

Science.gov (United States)

Fergus, Jeffrey W.

The high operating temperature of solid oxide fuel cells (SOFCs), as compared to polymer electrolyte membrane fuel cells (PEMFCs), improves tolerance to impurities in the fuel, but also creates challenges in the development of suitable materials for the various fuel cell components. In response to these challenges, intermediate temperature solid oxide fuel cells (IT-SOFCs) are being developed to reduce high-temperature material requirements, which will extend useful lifetime, improve durability and reduce cost, while maintaining good fuel flexibility. A major challenge in reducing the operating temperature of SOFCs is the development of solid electrolyte materials with sufficient conductivity to maintain acceptably low ohmic losses during operation. In this paper, solid electrolytes being developed for solid oxide fuel cells, including zirconia-, ceria- and lanthanum gallate-based materials, are reviewed and compared. The focus is on the conductivity, but other issues, such as compatibility with electrode materials, are also discussed.

Electrolytes for solid oxide fuel cells

Energy Technology Data Exchange (ETDEWEB)

Fergus, Jeffrey W. [Auburn University, Materials Research and Education Center, 275 Wilmore Laboratories, Auburn, AL 36849 (United States)

2006-11-08

The high operating temperature of solid oxide fuel cells (SOFCs), as compared to polymer electrolyte membrane fuel cells (PEMFCs), improves tolerance to impurities in the fuel, but also creates challenges in the development of suitable materials for the various fuel cell components. In response to these challenges, intermediate temperature solid oxide fuel cells (IT-SOFCs) are being developed to reduce high-temperature material requirements, which will extend useful lifetime, improve durability and reduce cost, while maintaining good fuel flexibility. A major challenge in reducing the operating temperature of SOFCs is the development of solid electrolyte materials with sufficient conductivity to maintain acceptably low ohmic losses during operation. In this paper, solid electrolytes being developed for solid oxide fuel cells, including zirconia-, ceria- and lanthanum gallate-based materials, are reviewed and compared. The focus is on the conductivity, but other issues, such as compatibility with electrode materials, are also discussed. (author)

Behaviour of defective CANDU fuel: fuel oxidation kinetic and thermodynamic modelling

International Nuclear Information System (INIS)

Higgs, J.

2005-01-01

The thermal performance of operating CANDU fuel under defect conditions is affected by the ingress of heavy water into the fuel element. A mechanistic model has been developed to predict the extent of fuel oxidation in defective fuel and its affect on fuel thermal performance. A thermodynamic treatment of such oxidized fuel has been performed as a basis for the boundary conditions in the kinetic model. Both the kinetic and thermodynamic models have been benchmarked against recent experimental work. (author)

Review of the literature for dry reprocessing oxide, metal, and carbide fuel: The AIROX, RAHYD, and CARBOX pyrochemical processes

Energy Technology Data Exchange (ETDEWEB)

Hoyt, R.C.; Rhee, B.W. [Rockwell International Corp., Canoga Park, CA (United States). Energy Systems Group

1979-09-30

The state of the art of dry processing oxide, carbide, and metal fuel has been determined through an extensive literature review. Dry processing in one of the most proliferation resistant fuel reprocessing technologies available to date, and is one of the few which can be exported to other countries. Feasibility has been established for oxide, carbide, and metal fuel on a laboratory scale, and large-scale experiments on oxide and carbide fuel have shown viability of the dry processing concept. A complete dry processing cycle has been demonstrated by multicycle processing-refabrication-reirradiation experiments on oxide fuel. Additional experimental work is necessary to: (1) demonstrate the complete fuel cycle for carbide and metal fuel, (2) optimize dry processing conditions, and (3) establish fission product behavior. Dry process waste management is easier than for an aqueous processing facility since wastes are primarily solids and gases. Waste treatment can be accomplished by techniques which have been, or are being, developed for aqueous plants.

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

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

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

Engineering study: Fast Flux Test Facility fuel reprocessing

International Nuclear Information System (INIS)

Beary, M.M.; Raab, G.J.; Reynolds, W.R. Jr.; Yoder, R.A.

1974-01-01

Several alternatives were studied for reprocessing FFTF fuels at Hanford. Alternative I would be to decontaminate and trim the fuel at T Plant and electrolytically dissolve the fuel at Purex. Alternative II would be to decontaminate and shear leach the fuels in a new facility near Purex. Alternative III would be to decontaminate and store fuel elements indefinitely at T Plant for subsequent offsite shipment. Alternative I, 8 to 10 M$ and 13 quarter-years; for Alternative II, 24 to 28 M$ and 20 quarter-years; for Alternative III, 3 to 4 M$ and 8 quarter-years. Unless there is considerable slippage in the FFTF shipping schedule, it would not be possible to build a new facility as described in Alternative II in time without building temporary storage facilities at T Plant, as described in Alternative III

Thermochemical aspects of fuel-cladding and fuel-coolant interactions in LMFBR oxide fuel pins

International Nuclear Information System (INIS)

Adamson, M.G.; Aitken, E.A.; Caputi, R.W.; Potter, P.E.; Mignanelli, M.A.

1979-01-01

This paper examines several thermochemical aspects of the fuel-cladding, fuel-coolant and fuel-fission product interactions that occur in LMFBR austenitic stainless steel-clad mixed (U,Pu)-oxide fuel pins during irradiation under normal operating conditions. Results are reported from a variety of high temperature EMF cell experiments in which continuous oxygen activity measurements on reacting and equilibrium mixtures of metal oxides and (excess) liquid alkali metal (Na, K, Cs) were performed. Oxygen potential and 0:M thresholds for Na-fuel reactions are re-evaluated in the light of new measurements and newly-assessed thermochemical data, and the influence on oxygen potential of possible U-Pu segregation between oxide and urano-plutonate (equilibrium) phases has been analyzed. (orig./RW) [de

Dissolution of mixed oxide spent fuel from FBR

International Nuclear Information System (INIS)

Sanyoshi, H.; Nishina, H.; Toyota, O.; Yamamoto, R.; Nemoto, S.; Okamoto, F.; Togashi, A.; Kawata, T.; Hayashi, S.

1991-01-01

At the Tokai Works of the Power Reactor and Nuclear Fuel Development Corporation (PNC), the Chemical Processing Facility (CPF) has been continuing operation since 1982 for laboratory scale hot experiments on reprocessing of FBR mixed oxide fuel. As a part of these experiments, dissolution experiments have been performed to define the key parameters affecting dissolution rates such as concentration of nitric acid, temperature and burnup and also to confirm the amount of insoluble residue. The dissolution rate of the irradiated fuel was determined to be in proportion to the 1.7 power of the nitric acid concentration. The activation energy determined from the experiments varied from 6 to 11 kcal/mol depending on the method of dissolution. The dissolution rate decreased as the fuel burnup increased in low nitric acid media below 5 mol/l. However, it was found that the effect of the burnup became negligible in a high concentration of nitric acid media. The amount of insoluble residue and its constituents were evaluated by changing the dissolution condition. (author)

DoD Fuel Facilities Criteria

Science.gov (United States)

2015-04-27

Pantograph Feb-2010 UFGS 33 58 00 Leak Detection for Fueling Systems Apr-2008 UFGS 33 52 43.13 Aviation Fuel Piping Feb-2010 UFGS 33 59 00 Tightness of... Pipeline Pressure Testing Guidelines  Specifications  Questions 2 7/12/2017 3 7/12/2017 DoD Fuels Facilities Documents  Unified...UFGS)  Most in the 33 nn nn series  Associated with Standard Designs  Available on WBDG site  Coating Systems 4 7/12/2017 Pipeline

33 CFR 149.655 - What are the requirements for helicopter fueling facilities?

Science.gov (United States)

2010-07-01

... helicopter fueling facilities? 149.655 Section 149.655 Navigation and Navigable Waters COAST GUARD... EQUIPMENT Design and Equipment Helicopter Fueling Facilities § 149.655 What are the requirements for helicopter fueling facilities? Helicopter fueling facilities must comply with 46 CFR 108.489 or an equivalent...

Over view of nuclear fuel cycle examination facility at KAERI

Energy Technology Data Exchange (ETDEWEB)

Lee, Key-Soon; Kim, Eun-Ga; Joe, Kih-Soo; Kim, Kil-Jeong; Kim, Ki-Hong; Min, Duk-Ki [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-09-01

Nuclear fuel cycle examination facilities at the Korea Atomic Energy Research Institute (KAERI) consist of two post-irradiation examination facilities (IMEF and PIEF), one chemistry research facility (CRF), one radiowaste treatment facility (RWTF) and one radioactive waste form examination facility (RWEF). This paper presents the outline of the nuclear fuel cycle examination facilities in KAERI. (author)

Transient survivability of LMR oxide fuel pins

International Nuclear Information System (INIS)

Weber, E.T.; Pitner, A.L.; Bard, F.E.; Culley, G.E.; Hunter, C.W.

1986-01-01

Fuel pin integrity during transient events must be assessed for both the core design and safety analysis phases of a reactor project. A significant increase in the experience related to limits of integrity for oxide fuel pins in transient overpower events has been realized from testing of fuel pins irradiated in FFTF and PFR. Fourteen FFTF irradiated fuel pins were tested in TREAT, representing a range of burnups, overpower ramp rates and maximum overpower conditions. Results of these tests along with similar testing in the PFR/TREAT program, provide a demonstration of significant safety margins for oxide fuel pins. Useful information applied in analytical extrapolation of fuel pin test data have been developed from laboratory transient tests on irradiated fuel cladding (FCTT) and on unirradiated fuel pellet deformation. These refinements in oxide fuel transient performance are being applied in assessment of transient capabilities of long lifetime fuel designs using ferritic cladding

A Swedish nuclear fuel facility and public acceptance

Energy Technology Data Exchange (ETDEWEB)

Andersson, Bengt A [ABB Atom (Sweden)

1989-07-01

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.

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

Significant incidents in nuclear fuel cycle facilities

Energy Technology Data Exchange (ETDEWEB)

NONE

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.

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

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)

Studies on the dissolution of mixed oxide spent fuel from FBR

International Nuclear Information System (INIS)

Nemoto, Shin-ichi; Shibata, Atsuhiro; Shioura, Takao; Okamoto, Fumitoshi; Tanaka, Yasumasa

1995-01-01

At the Chemical Processing Facility(CPF) in the Tokai Works of the Power Reactor and Nuclear Fuel Development Corporation(PNC), since 1982 Laboratory scale hot experiments have been carried out on the development of reprocessing technology for FBR mixed oxide fuel. The spent fuel pins which have been used in out experiments were irradiated in Experimental Fast Reactor 'Joyo' Phenix (France) and DFR(UK). Burn-up of the fuel pins were 4,400-100,000 MWd/t. This paper Summarizes a dissolution study that have been performed to define the Key parameters affecting dissolution rate such as concentration of nitric acid, burn-up, and temperature. And this paper also discusses about the character of releasing 85 Kr in chopping and dissolution process, and about the amount of insoluble residue. (author)

Fuel cycle cost comparisons with oxide and silicide fuels

Energy Technology Data Exchange (ETDEWEB)

Matos, J E; Freese, K E [RERTR Program, Argonne National Laboratory (United States)

1983-09-01

This paper addresses fuel cycle cost comparisons for a generic 10 MW reactor with HEU aluminide fuel and with LEU oxide and silicide fuels in several fuel element geometries. The intention of this study is to provide a consistent assessment of various design options from a cost point of view. The status of the development and demonstration of the oxide and silicide fuels are presented in several papers in these proceedings. Routine utilization of these fuels with the uranium densities considered here requires that they are successfully demonstrated and licensed. Thermal-hydraulic safety margins, shutdown margins, mixed cores, and transient analyses are not addressed here, but analyses of these safety issues are in progress for a limited number of the most promising design options. Fuel cycle cost benefits could result if a number of reactors were to utilize fuel elements with the same number or different numbers of the same standard fuel plate. Data is presented to quantify these potential cost benefits. This analysis shows that there are a number of fuel element designs using LEU oxide or silicide fuels that have either the same or lower total fuel cycle costs than the HEU design. Use of these fuels with the uranium densities considered requires that they are successfully demonstrated and licensed. All safety criteria for the reactor with these fuel element designs need to be satisfied as well. With LEU oxide fuel, 31 g U/cm{sup 3} 1 and 0.76 mm--thick fuel meat, elements with 18-22 plates 320-391 g {sup 235}U) result in the same or lower total costs than with the HEU element 23 plates, 280 g {sup 235}U). Higher LEU loadings (more plates per element) are needed for larger excess reactivity requirements. However, there is little cost advantage to using more than 20 of these plates per element. Increasing the fuel meat thickness from 0.76 mm to 1.0 mm with 3.1 g U/cm{sup 3} in the design with 20 plates per element could result in significant cost reductions if the

Safeguards approach for conditioning facility for spent fuel

International Nuclear Information System (INIS)

Younkin, J.M.; Barham, M.; Moran, B.W.

1999-01-01

A safeguards approach has been developed for conditioning facilities associated with the final disposal of spent fuel in geologic repositories. The proposed approach is based on a generic conditioning facility incorporating common features of conditioning facility designs currently proposed. The generic facility includes a hot cell for consolidation of spent fuel pins and repackaging of spent fuel items such as assemblies and cans of pins. The consolidation process introduces safeguards concerns which have not previously been addressed in traditional safeguards approaches. In developing the safeguards approach, diversion of spent fuel was assessed in terms of potential target items, operational activities performed on the items, containment of the items, and concealment activities performed on the items. The combination of these factors defines the potential diversion pathways. Diversion pathways were identified for spent fuel pellets, pins, assemblies, canisters, and casks. Diversion activities provide for opportunities of detection along the diversion paths. Potential detection methods were identified at several levels of diversion activities. Detection methods can be implemented through safeguards measures. Safeguards measures were proposed for each of the primary safeguards techniques of design information verification (DIV), containment and surveillance (C/S), and material accountancy. Potential safeguards approaches were developed by selection of appropriate combinations of safeguards measures. For all candidate safeguards approaches, DIV is a fundamental component. Variations in the approaches are mainly in the degree of C/S measures and in the types and numbers of material accountancy verification measures. The candidate safeguards approaches were evaluated toward the goal of determining a model safeguards approach. This model approach is based on the integrated application of selected safeguards measures to use International Atomic Energy Agency resources

Experience with oxide fuel for advanced reactors

International Nuclear Information System (INIS)

Leggett, R.D.

1984-01-01

This paper focuses on the use and potential of oxide fuel systems for the LMFBR. The flawless performance of mixed oxide (UO 2 -PuO 2 ) fuel in FFTF to 100,000 MWd/MTM is reviewed and means for achieving 200,000 MWd/MTM are presented. This includes using non-swelling alloys for cladding and ducts to overcome the limitations caused by swelling of the current alloys. Examples are provided of the inherently safe characteristics of oxide fuel including a large negative Doppler coefficient, its dispersive nature under hypothetical accident scenarios, and the low energy molten fuel-coolant interaction. Developments in fuel fabrication and reprocessing that stress safety and reduced personnel exposure are presented. Lastly, the flexibility to design for maximum fuel supply (high breeding gain) or minimum fuel cost (long lifetime) is shown

Experience with oxide fuel for advanced reactors

International Nuclear Information System (INIS)

Leggett, R.D.

1984-04-01

This paper focuses on the use and potential of oxide fuel system for the LMFBR. The flawless performance of mixed oxide (UO 2 -PuO 2 ) fuel in FFTF to 100,000 MWd/MTM is reviewed and means for achieving 200,000 MWd/MTM are presented. This includes using non-swelling alloys for cladding and ducts to overcome the limitations caused by swelling of the current alloys. Exampled are provided of the inherently safe characteristics of oxide fuel including a large negative Doppler coefficient, its dispersive nature under hypothetical accident scenarios, and the low energy molten fuel-coolant interaction. Developments in fuel fabrication and reprocessing that stress safety and reduced personnel exposure are presented. Lastly, the flexibility to design for maximum fuel supply (high breeding gain) or minimum fuel cost (long lifetime) is shown

Whole-Pin Furnace system: An experimental facility for studying irradiated fuel pin behavior under potential reactor accident conditions

International Nuclear Information System (INIS)

Liu, Y.Y.; Tsai, H.C.; Donahue, D.A.; Pushis, D.O.; Savoie, F.E.; Holland, J.W.; Wright, A.E.; August, C.; Bailey, J.L.; Patterson, D.R.

1990-05-01

The whole-pin furnace system is a new in-cell experimental facility constructed to investigate how irradiated fuel pins may fail under potential reactor accident conditions. Extensive checkouts have demonstrated excellent performance in remote operation, temperature control, pin breach detection, and fission gas handling. The system is currently being used in testing of EBIR-II-irradiated Integral Fast Reactor (IFR) metal fuel pins; future testing will include EBR-II-irradiated mixed-oxide fuel pins. 7 refs., 4 figs

Internal fuel pin oxidizer

International Nuclear Information System (INIS)

Andrews, M.G.

1978-01-01

A nuclear fuel pin has positioned within it material which will decompose to release an oxidizing agent which will react with the cladding of the pin and form a protective oxide film on the internal surface of the cladding

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

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

Energy Technology Data Exchange (ETDEWEB)

Bordier, Gilles; Warin, Dominique; Masson, Michel [CEA/Marcoule Direction, BP 17171 - 30207 - Bagnols-sur-Ceze Cedex (France)

2008-07-01

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{sup TM} process that jointly manages uranium and plutonium from the dissolution of spent fuel to the production of UPuO{sub 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)

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)

Nuclear fuel treatment facility for 'Mutsu'

International Nuclear Information System (INIS)

Kanazawa, Toshio; Fujimura, Kazuo; Horiguchi, Eiji; Kobayashi, Tetsuji; Tamekiyo, Yoshizou

1989-01-01

A new fixed mooring harbor in Sekinehama and surrounding land facilities to accommodate a test voyage for the nuclear-powered ship 'Mutsu' in 1990 were constructed by the Japan Atomic Energy Research Institute. Kobe Steel took part in the construction of the nuclear fuel treatment process in various facilities, beginning in October, 1988. This report describes the outline of the facility. (author)

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.)

Status of spent fuel storage facilities in Switzerland

International Nuclear Information System (INIS)

Beyeler, P.C.; Lutz, H.R.; Heesen, W. von

1999-01-01

Planning of a dry spent fuel storage facility in Switzerland started already 15 years ago. The first site considered for a central interim storage facility was the cavern of the decommissioned pilot nuclear plant at Lucens in the French-speaking part of Switzerland. This project was terminated in the late eighties because of lack of public acceptance. The necessary acceptance was found in the small town of Wuerenlingen which has hosted for many years the Swiss Reactor Research Centre. The new project consists of centralised interim storage facilities for all types of radioactive waste plus a hot cell and a conditioning and incinerating facility. It represents a so-called integrated storage solution. In 1990, the new company 'ZWILAG Zwischenlager Wuerenlingen AG' (ZWILAG) was founded and the licensing procedures according to the Swiss Atomic law were initiated. On August 26, 1996 ZWILAG got the permit for construction of the whole facility including the operating permit for the storage facilities. End of construction and commissioning are scheduled for autumn 1999. The nuclear power station Beznau started planning a low level waste and spent fuel storage facility on its own, because in 1990 its management thought that by 1997 the first high active waste from the reprocessing facilities in France would have to be taken back. This facility at the Beznau site, called ZWIBEZ, was licensed according to a shorter procedure so its construction was finished by 1997. The two facilities for high level waste and spent fuel provide space for a total of 278 casks, which is sufficient for the waste and spent fuel of the four Swiss nuclear power stations including their life extension programme. (author)

Comparison of neutron dose measured by Albedo TLD and etched tracks detector at PNC plutonium fuel facilities

International Nuclear Information System (INIS)

Tsujimura, N.; Momose, T.; Shinohara, K.; Ishiguro, H.

1996-01-01

Power Reactor and Nuclear Fuel Development Corporation (PNC) has fabricated Plutonium and Uranium Mixed OXide (MOX) fuel for FBR MONJU at Tokai works. In this site, PNC/Panasonic albedo TLDs/1/ are used for personnel neutron monitoring. And a part of workers wore Etched Tracks Detector (ETD) combined with TLD in order to check the accuracy of the neutron dose estimated by albedo TLD. In this paper, the neutron dose measured by TLD and ETD in the routine monitoring is compared at PNC plutonium fuel facilities. (author)

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...

Fuel cracking in relation to fuel oxidation in support of an out-reactor instrumented defected fuel experiment

Energy Technology Data Exchange (ETDEWEB)

Quastel, A.; Thiriet, C. [Atomic Energy of Canada Limited, Chalk River, ON (Canada); Lewis, B., E-mail: brent.lewis@uoit.ca [Univ. of Ontario Inst. of Tech., Oshawa, ON (Canada); Corcoran, E., E-mail: emily.corcoran@rmc.ca [Royal Military College of Canada, Kingston, ON (Canada)

2014-07-01

An experimental program funded by the CANDU Owners Group (COG) is studying an out-reactor instrumented defected fuel experiment in Stern Laboratories (Hamilton, Ontario) with guidance from Atomic Energy of Canada Limited (AECL). The objective of this test is to provide experimental data for validation of a mechanistic fuel oxidation model. In this experiment a defected fuel element with UO{sub 2} pellets will be internally heated with an electrical heater element, causing the fuel to crack. By defecting the sheath in-situ the fuel will be exposed to light water coolant near normal reactor operating conditions (pressure 10 MPa and temperature 265-310{sup o}C) causing fuel oxidation, especially near the hotter regions of the fuel in the cracks. The fuel thermal conductivity will change, resulting in a change in the temperature distribution of the fuel element. This paper provides 2D r-θ plane strain solid mechanics models to simulate fuel thermal expansion, where conditions for fuel crack propagation are investigated with the thermal J integral to predict fuel crack stress intensity factors. Finally since fuel crack geometry can affect fuel oxidation this paper shows that the solid mechanics model with pre-set radial cracks can be coupled to a 2D r-θ fuel oxidation model. (author)

UK experience on fuel and cladding interaction in oxide fuels

Energy Technology Data Exchange (ETDEWEB)

Batey, W [Dounreay Experimental Reactor Establishment, Thurso, Caithness (United Kingdom); Findlay, J R [AERE, Harwell, Didcot, Oxon (United Kingdom)

1977-04-01

The occurrence of fuel cladding interactions in fast reactor fuels has been observed in UK irradiations over a period of years. Chemical incompatibility between fuel and clad represents a potential source of failure and has, on this account, been studied using a variety of techniques. The principal fuel of interest to the UK for fast reactor application is mixed uranium plutonium oxide clad in stainless steel and it is in this field that the majority of work has been concentrated. Some consideration has been given to carbide fuels, because of their application as an advanced fuel. This experience is described in the accompanying paper. Several complementary initiatives have been followed to investigate the interactions in oxide fuel. The principal source of experimental information is from the experimental fuel irradiation programme in the Dounreay Fast Reactor (DFR). Supporting information has been obtained from irradiation programmes in Materials Testing Reactors (MTR). Conditions approaching those in a fast reactor are obtained and the effects of specific variables have been examined in specifically designed experiments. Out-of-reactor experiments have been used to determine the limits of fuel and cladding compatibility and also to give indications of corrosion The observations from all experiments have been examined in the light of thermo-dynamic predictions of fuel behaviour to assess the relative significance of various observations and operating conditions. An experimental programme to control and limit the interactions in oxide fuel is being followed.

UK experience on fuel and cladding interaction in oxide fuels

International Nuclear Information System (INIS)

Batey, W.; Findlay, J.R.

1977-01-01

The occurrence of fuel cladding interactions in fast reactor fuels has been observed in UK irradiations over a period of years. Chemical incompatibility between fuel and clad represents a potential source of failure and has, on this account, been studied using a variety of techniques. The principal fuel of interest to the UK for fast reactor application is mixed uranium plutonium oxide clad in stainless steel and it is in this field that the majority of work has been concentrated. Some consideration has been given to carbide fuels, because of their application as an advanced fuel. This experience is described in the accompanying paper. Several complementary initiatives have been followed to investigate the interactions in oxide fuel. The principal source of experimental information is from the experimental fuel irradiation programme in the Dounreay Fast Reactor (DFR). Supporting information has been obtained from irradiation programmes in Materials Testing Reactors (MTR). Conditions approaching those in a fast reactor are obtained and the effects of specific variables have been examined in specifically designed experiments. Out-of-reactor experiments have been used to determine the limits of fuel and cladding compatibility and also to give indications of corrosion The observations from all experiments have been examined in the light of thermo-dynamic predictions of fuel behaviour to assess the relative significance of various observations and operating conditions. An experimental programme to control and limit the interactions in oxide fuel is being followed

Features and safety aspects of spent fuel storage facility, Tarapur

International Nuclear Information System (INIS)

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

2017-01-01

Spent Fuel Storage Facility (SFSF), Tarapur is designed to store spent fuel arising from PHWRs in different parts of the country. Spent fuel is transported in AERB qualified/authorized shipping cask by NPCIL to SFSF by road or rail route. The spent fuel storage facility at Tarapur was hot commissioned after regulatory clearances

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)

Gas transport in solid oxide fuel cells

CERN Document Server

He, Weidong; Dickerson, James

2014-01-01

This book provides a comprehensive overview of contemporary research and emerging measurement technologies associated with gas transport in solid oxide fuel cells. Within these pages, an introduction to the concept of gas diffusion in solid oxide fuel cells is presented. This book also discusses the history and underlying fundamental mechanisms of gas diffusion in solid oxide fuel cells, general theoretical mathematical models for gas diffusion, and traditional and advanced techniques for gas diffusivity measurement.

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

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

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

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

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

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.

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)

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

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.

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

Electrometallurgical treatment of oxide spent fuels

International Nuclear Information System (INIS)

Karell, E. J.

1999-01-01

The Department of Energy (DOE) inventory of spent nuclear fuel contains a wide variety of oxide fuel types that may be unsuitable for direct repository disposal in their current form. The molten-salt electrometallurgical treatment technique developed by Argonne National Laboratory (ANL) has the potential to simplify preparing and qualifying these fuels for disposal by converting them into three uniform product streams: uranium metal, a metal waste form, and a ceramic waste form. This paper describes the major steps in the electrometallurgical treatment process for oxide fuels and provides the results of recent experiments performed to develop and scale up the process

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

International Nuclear Information System (INIS)

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

1978-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1978-01-01

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

Oxidation behavior of fuel cladding tube in spent fuel pool accident condition

International Nuclear Information System (INIS)

Nemoto, Yoshiyuki; Kaji, Yoshiyuki; Ogawa, Chihiro; Nakashima, Kazuo; Tojo, Masayuki

2017-01-01

In spent fuel pool (SFP) under loss-of-cooling or loss-of-coolant severe accident condition, the spent fuels will be exposed to air and heated by their own residual decay heat. Integrity of fuel cladding is crucial for SFP safety therefore study on cladding oxidation in air at high temperature is important. Zircaloy-2 (Zry2) and zircaloy-4 (Zry4) were applied for thermogravimetric analyses (TGA) in different temperatures in air at different flow rates to evaluate oxidation behavior. Oxidation rate increased with testing temperature. In a range of flow rate of air which is predictable in spent fuel lack during a hypothetical SFP accident, influence of flow rate was not clearly observed below 950degC for the Zry2, or below 1050degC for Zry4. In higher temperature, oxidation rate was higher in high rate condition, and this trend was seen clearer when temperature increased. Oxide layers were carefully examined after the TGA analyses and compared with mass gain data to investigate detail of oxidation process in air. It was revealed that the mass gain data in pre-breakaway regime reflects growth of dense oxide film on specimen surface, meanwhile in post-breakaway regime, it reflects growth of porous oxide layer beneath fracture of the dense oxide film. (author)

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)

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

International Nuclear Information System (INIS)

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

1997-01-01

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

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

300 Area fuel supply facilities deactivation mission analysis report

International Nuclear Information System (INIS)

Lund, D.P.

1995-01-01

This report presents the results of the 300 Area fuel supply facilities (formerly call ''N reactor fuel fabrication facilities'') Deactivation Project mission analysis. Hanford systems engineering (SE) procedures call for a mission analysis. The mission analysis is an important first step in the SE process

Conceptual development of a test facility for spent fuel management

Energy Technology Data Exchange (ETDEWEB)

Park, S.W.; Lee, H.H.; Lee, J.Y.; Lee, J.S.; Ro, S.G. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-12-31

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.

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

Decommissioning of nuclear fuel cycle facilities. Safety guide

International Nuclear Information System (INIS)

2001-01-01

The objective of this Safety Guide is to provide guidance to regulatory bodies and operating organizations on planning and provision for the safe management of the decommissioning of non-reactor nuclear fuel cycle facilities. While the basic safety considerations for the decommissioning of nuclear fuel cycle facilities are similar to those for nuclear power plants, there are important differences, notably in the design and operating parameters for the facilities, the type of radioactive material and the support systems available. It is the objective of this Safety Guide to provide guidance for the shutdown and eventual decommissioning of such facilities, their individual characteristics being taken into account

Fuel performance of DOE fuels in water storage

International Nuclear Information System (INIS)

Hoskins, A.P.; Scott, J.G.; Shelton-Davis, C.V.; McDannel, G.E.

1993-01-01

Westinghouse Idaho Nuclear Company operates the Idaho Chemical Processing Plant (ICPP) at the Idaho National Engineering Laboratory. In April of 1992, the U.S. Department of Energy (DOE) decided to end the fuel reprocessing mission at ICPP. Fuel performance in storage received increased emphasis as the fuel now needs to be stored until final dispositioning is defined and implemented. Fuels are stored in four main areas: an original underwater storage facility, a modern underwater storage facility, and two dry fuel storage facilities. As a result of the reactor research mission of the DOE and predecessor agencies, the Energy Research and Development Administration and the Atomic Energy Commission, many types of nuclear fuel have been developed, used, and assigned to storage at the ICPP. Fuel clad with stainless steel, zirconium, aluminum, and graphite are represented. Fuel matrices include uranium oxide, hydride, carbide, metal, and alloy fuels, resulting in 55 different fuel types in storage. Also included in the fuel storage inventory is canned scrap material

Safeguards System for the Advanced Spent Fuel Conditioning Process Facility

International Nuclear Information System (INIS)

Kim, Ho-dong; Lee, T.H.; Yoon, J.S.; Park, S.W; Lee, S.Y.; Li, T.K.; Menlove, H.; Miller, M.C.; Tolba, A.; Zarucki, R.; Shawky, S.; Kamya, S.

2007-01-01

The advanced spent fuel conditioning process (ACP) which is a part of a pyro-processing has been under development at Korean Atomic Energy Research Institute (KAERI) since 1997 to tackle the problem of an accumulation of spent fuel. The concept is to convert spent oxide fuel into a metallic form in a high temperature molten salt in order to reduce the heat energy, volume, and radioactivity of a spent fuel. Since the inactive tests of the ACP have been successfully implemented to confirm the validity of the electrolytic reduction technology, a lab-scale hot test will be undertaken in a couple of years to validate the concept. For this purpose, the KAERI has built the ACP Facility (ACPF) at the basement of the Irradiated Material Examination Facility (IMEF) of KAERI, which already has a reserved hot-cell area. Through the bilateral arrangement between US Department of Energy (DOE) and Korean Ministry of Science and Technology (MOST) for safeguards R and D, the KAERI has developed elements of safeguards system for the ACPF in cooperation with the Los Alamos National Laboratory (LANL). The reference safeguards design conditions and equipment were established for the ACPF. The ACPF safeguards system has many unique design specifications because of the particular characteristics of the pyro-process materials and the restrictions during a facility operation. For the material accounting system, a set of remote operation and maintenance concepts has been introduced for a non-destructive assay (NDA) system. The IAEA has proposed a safeguards approach to the ACPF for the different operational phases. Safeguards measures at the ACPF will be implemented during all operational phases which include a 'Cold Test', a 'Hot Test' and at the end of a 'Hot test'. Optimization of the IAEA's inspection efforts was addressed by designing an effective safeguards approach that relies on, inter alia, remote monitoring using cameras, installed NDA instrumentation, gate monitors and seals

Operation of post-irradiation examination facility

Energy Technology Data Exchange (ETDEWEB)

Kim, Eun Ka; Park, Kwang Joon; Jeon, Yong Bum [and others; Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-02-01

In 1995, the post-irradiation examination (PIE) of nuclear fuels was performed as follows. The relation between burnup and top nozzle spring force of fuel assembly was obtained by measuring the holddown spring force on the Kori-1 reactor fuel assemblies. The resonance ultrasonic test for inspection of defect and moisture in fuel rod was carried out on fuel rods of C15 and J14 assemblies, and the change of fuel rod condition by storing in pool has been analyzed on the intentionally defected fuel rods (ID-C and ID-L) as well as intact fuel rod (1-2) by NDT in ht cell. The oxide layer thickness on cladding surface of J44-L12 fuel rod was measured by NDT method and metallography to reveal the oxidation as a function of temperature in the fuel rod, and the burnup of J44 fuel assembly was measured by chemical analysis. HVAC system and pool water treatment system of the PIE facility were continuously operated for air filtration and water purification. The monitoring of radiation and pool water in PIE facility has been carried out to maintain the facility safety, and electric power supply system was checked and maintained to supply the electric power to the facility normally. The developed measurement techniques of oxide layer thickness on fuel rod cladding and holddown spring force of top nozzle in fuel assembly were applied to examine the nuclear fuels. Besides, a radiation shielding glove box was designed and a hot cell compressor for volume reduction of radioactive materials was fabricated. 19 tabs., 38 figs., 7 refs. (Author) .new.

Design information questionnaire for a model mixed oxide fuel fabrication facility

International Nuclear Information System (INIS)

Glancy, J.E.

1976-05-01

The model fuel plant is based on the proposed Westinghouse Anderson, S.C., plant and is typical of plants that will be constructed and operated in 1980 to 1990. A number of plant systems and procedures are uncertain, and in these cases judgment was used in describing relevant parameters in order to provide a complete model on which to design an inspection plan. The model plant does not, therefore, strictly represent any planned facility nor does it strictly represent the ideas of Westinghouse on plant design and material accountability. This report is divided into two sections. The first section is the IAEA Design Information Questionnaire form that contains an outline of all information requested. The second section is a complete listing of design information

Radiation protection at nuclear fuel cycle facilities

International Nuclear Information System (INIS)

Endo, K.; Momose, T.; Furuta, S.

2011-01-01

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. (authors)

The burnup dependence of light water reactor spent fuel oxidation

International Nuclear Information System (INIS)

Hanson, B.D.

1998-07-01

Over the temperature range of interest for dry storage or for placement of spent fuel in a permanent repository under the conditions now being considered, UO 2 is thermodynamically unstable with respect to oxidation to higher oxides. The multiple valence states of uranium allow for the accommodation of interstitial oxygen atoms in the fuel matrix. A variety of stoichiometric and nonstoichiometric phases is therefore possible as the fuel oxidizers from UO 2 to higher oxides. The oxidation of UO 2 has been studied extensively for over 40 years. It has been shown that spent fuel and unirradiated UO 2 oxidize via different mechanisms and at different rates. The oxidation of LWR spent fuel from UO 2 to UO 2.4 was studied previously and is reasonably well understood. The study presented here was initiated to determine the mechanism and rate of oxidation from UO 2.4 to higher oxides. During the early stages of this work, a large variability in the oxidation behavior of samples oxidized under nearly identical conditions was found. Based on previous work on the effect of dopants on UO 2 oxidation and this initial variability, it was hypothesized that the substitution of fission product and actinide impurities for uranium atoms in the spent fuel matrix was the cause of the variable oxidation behavior. Since the impurity concentration is roughly proportional to the burnup of a specimen, the oxidation behavior of spent fuel was expected to be a function of both temperature and burnup. This report (1) summarizes the previous oxidation work for both unirradiated UO 2 and spent fuel (Section 2.2) and presents the theoretical basis for the burnup (i.e., impurity concentration) dependence of the rate of oxidation (Sections 2.3, 2.4, and 2.5), (2) describes the experimental approach (Section 3) and results (Section 4) for the current oxidation tests on spent fuel, and (3) establishes a simple model to determine the activation energies associated with spent fuel oxidation (Section 5)

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

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

Development of Experimental Facilities for Advanced Spent Fuel Management Technology

Energy Technology Data Exchange (ETDEWEB)

You, G. S.; Jung, W. M.; Ku, J. H. [and others

2004-07-01

The advanced spent fuel management process(ACP), proposed to reduce the overall volume of the PWR spent fuel and improve safety and economy of the long-term storage of spent fuel, is under research and development. This technology convert spent fuels into pure metal-base uranium with removing the highly heat generating materials(Cs, Sr) efficiently and reducing of the decay heat, volume, and radioactivity from spent fuel by 1/4. In the next phase(2004{approx}2006), the demonstration of this technology will be carried out for verification of the ACP in a laboratory scale. For this demonstration, the hot cell facilities of {alpha}-{gamma} type and auxiliary facilities are required essentially for safe handling of high radioactive materials. As the hot cell facilities for demonstration of the ACP, a existing hot cell of {beta}-{gamma} type will be refurbished to minimize construction expenditures of hot cell facility. In this study, the design requirements are established, and the process detail work flow was analysed for the optimum arrangement to ensure effective process operation in hot cell. And also, the basic and detail design of hot cell facility and process, and safety analysis was performed to secure conservative safety of hot cell facility and process.

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

Solid Oxide Fuel Cells Operating on Alternative and Renewable Fuels

Energy Technology Data Exchange (ETDEWEB)

Wang, Xiaoxing; Quan, Wenying; Xiao, Jing; Peduzzi, Emanuela; Fujii, Mamoru; Sun, Funxia; Shalaby, Cigdem; Li, Yan; Xie, Chao; Ma, Xiaoliang; Johnson, David; Lee, Jeong; Fedkin, Mark; LaBarbera, Mark; Das, Debanjan; Thompson, David; Lvov, Serguei; Song, Chunshan

2014-09-30

This DOE project at the Pennsylvania State University (Penn State) initially involved Siemens Energy, Inc. to (1) develop new fuel processing approaches for using selected alternative and renewable fuels – anaerobic digester gas (ADG) and commercial diesel fuel (with 15 ppm sulfur) – in solid oxide fuel cell (SOFC) power generation systems; and (2) conduct integrated fuel processor – SOFC system tests to evaluate the performance of the fuel processors and overall systems. Siemens Energy Inc. was to provide SOFC system to Penn State for testing. The Siemens work was carried out at Siemens Energy Inc. in Pittsburgh, PA. The unexpected restructuring in Siemens organization, however, led to the elimination of the Siemens Stationary Fuel Cell Division within the company. Unfortunately, this led to the Siemens subcontract with Penn State ending on September 23rd, 2010. SOFC system was never delivered to Penn State. With the assistance of NETL project manager, the Penn State team has since developed a collaborative research with Delphi as the new subcontractor and this work involved the testing of a stack of planar solid oxide fuel cells from Delphi.

Construction and engineering report for advanced nuclear fuel development facility

International Nuclear Information System (INIS)

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 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

Storage facilities of spent nuclear fuel in dry for Mexican nuclear facilities

International Nuclear Information System (INIS)

Salmeron V, J. A.; Camargo C, R.; Nunez C, A.; Mendoza F, J. E.; Sanchez J, J.

2013-10-01

In this article the relevant aspects of the spent fuel storage and the questions that should be taken in consideration for the possible future facilities of this type in the country are approached. A brief description is proposed about the characteristics of the storage systems in dry, the incorporate regulations to the present Nuclear Regulator Standard, the planning process of an installation, besides the approaches considered once resolved the use of these systems; as the modifications to the system, the authorization periods for the storage, the type of materials to store and the consequent environmental impact to their installation. At the present time the Comision Nacional de Seguridad Nuclear y Salvaguardias (CNSNS) considers the possible generation of two authorization types for these facilities: Specific, directed to establish a new nuclear installation with the authorization of receiving, to transfer and to possess spent fuel and other materials for their storage; and General, focused to those holders that have an operation license of a reactor that allows them the storage of the nuclear fuel and other materials that they possess. Both authorizations should be valued according to the necessities that are presented. In general, this installation type represents a viable solution for the administration of the spent fuel and other materials that require of a temporary solution previous to its final disposal. Its use in the nuclear industry has been increased in the last years demonstrating to be appropriate and feasible without having a significant impact to the health, public safety and the environment. Mexico has two main nuclear facilities, the nuclear power plant of Laguna Verde of the Comision Federal de Electricidad (CFE) and the facilities of the TRIGA Reactor of the Instituto Nacional de Investigaciones Nucleares (ININ) that will require in a future to use this type of disposition installation of the spent fuel and generated wastes. (Author)

Fuel Supply Shutdown Facility Interim Operational Safety Requirements

International Nuclear Information System (INIS)

BENECKE, M.W.

2000-01-01

The Interim Operational Safety Requirements for the Fuel Supply Shutdown (FSS) Facility define acceptable conditions, safe boundaries, bases thereof, and management of administrative controls to ensure safe operation of the facility

Use of Pilot Plants for Developing Used Nuclear Fuel Recycling Facilities

Energy Technology Data Exchange (ETDEWEB)

Phillips, Chris; Arm, Stuart [EnergySolutions LLC (United States); Banfield, Zara; Jeapes, Andrew; Taylor, Richard [National Nuclear Laboratory (United Kingdom)

2009-06-15

EnergySolutions and its teaming partners are working with the US Department of Energy (DOE) to develop processes, equipment and facilities for recycling used nuclear fuel (UNF). Recycling significantly reduces the volume of wastes that ultimately will be consigned to the National Geologic Repository, enables the re-use in new fuel of the valuable uranium and plutonium in the UNF, and allows the long-lived minor actinides to be treated separately so they do not become long term heat emitters in the Repository. A major requirement of any new UNF recycling facility is that pure plutonium is not separated anywhere in the process, so as to reduce the nuclear proliferation attractiveness of the facility. EnergySolutions and its team partner the UK National Nuclear Laboratory (NNL) have developed the NUEX process to achieve this and to handle appropriately the treatment of other species such as krypton, tritium, neptunium and technetium. NUEX is based on existing successful commercial UNF recycling processes deployed in the UK, France and imminently in Japan, but with a range of modifications to the flowsheet to keep some uranium with the plutonium at all times and to minimize aerial and liquid radioactive discharges. NNL's long-term experience in developing the recycling and associated facilities at the Sellafield site in the UK, and its current duties to support technically the operation of the Thermal Oxide Reprocessing Plant (THORP) at Sellafield provides essential input to the design of the US NUEX-based facility. Development work for THORP and other first-of-kind nuclear plants employed miniature scale fully radioactive through large scale inactive pilot plants. The sequence of development work that we have found most successful is to (i) perform initial process development at small (typically 1/5000) scale in gloveboxes using trace active materials, (ii) demonstrate the processes at the same small scale with actual irradiated fuel in hot cells and (iii

Spent fuels conditioning and irradiated nuclear fuel elements examination: the STAR facility and its abilities

Energy Technology Data Exchange (ETDEWEB)

Boussard, F.; Huillery, R. [CEA Centre d`Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d`Etudes des Combustibles; Averseng, J.L.; Serpantie, J.P. [Novatome Industries, 92 - Le Plessis-Robinson (France)

1994-12-31

This paper is a presentation of the STAR facility, a high activity laboratory located in Cadarache Nuclear Research Center (France). The purpose of the STAR facility and of the associated processes, is the treatment, cleaning and conditioning of spent fuels from Gas Cooled Reactors (GCR) and in particular of about 2300 spent GCR fuel cartridges irradiated more than 20 years ago in Electricite de France (EDF) or CEA Uranium Graphite GCR. The processes are: to separate the nuclear fuel from the clad remains, to chemically stabilize the nuclear material and to condition it in sealed canisters. An additional objective of STAR consists in non-destructive or destructive examinations and tests on PWR rods or FBR pins in the frame of fuel development programs. The paper describes the STAR facility conceptual design (safety design rules, hot cells..) and the different options corresponding to the GCR reconditioning process and to further research and development works on various fuel types. (J.S.). 3 figs.

Spent fuels conditioning and irradiated nuclear fuel elements examination: the STAR facility and its abilities

International Nuclear Information System (INIS)

Boussard, F.; Huillery, R.

1994-01-01

This paper is a presentation of the STAR facility, a high activity laboratory located in Cadarache Nuclear Research Center (France). The purpose of the STAR facility and of the associated processes, is the treatment, cleaning and conditioning of spent fuels from Gas Cooled Reactors (GCR) and in particular of about 2300 spent GCR fuel cartridges irradiated more than 20 years ago in Electricite de France (EDF) or CEA Uranium Graphite GCR. The processes are: to separate the nuclear fuel from the clad remains, to chemically stabilize the nuclear material and to condition it in sealed canisters. An additional objective of STAR consists in non-destructive or destructive examinations and tests on PWR rods or FBR pins in the frame of fuel development programs. The paper describes the STAR facility conceptual design (safety design rules, hot cells..) and the different options corresponding to the GCR reconditioning process and to further research and development works on various fuel types. (J.S.). 3 figs

High-Level Functional and Operational Requirements for the Advanced Fuel Cycle Facility

International Nuclear Information System (INIS)

Charles Park

2006-01-01

This document describes the principal functional and operational requirements for the proposed Advanced Fuel Cycle Facility (AFCF). The AFCF is intended to be the world's foremost facility for nuclear fuel cycle research, technology development, and demonstration. The facility will also support the near-term mission to develop and demonstrate technology in support of fuel cycle needs identified by industry, and the long-term mission to retain and retain U.S. leadership in fuel cycle operations. The AFCF is essential to demonstrate a more proliferation-resistant fuel cycle and make long-term improvements in fuel cycle effectiveness, performance and economy

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

Fuel Assemblies Thermal Analysis in the New Spent Fuel Storage Facility at Inshass Site

International Nuclear Information System (INIS)

Khattab, M.; Mariy, Ahmed

1999-01-01

New Wet Storage Facility (NSF) is constructed at Inshass site to solve the problem of spent fuel storage capacity of ETRR-1 reactor . The Engineering Safety Heat Transfer Features t hat characterize the new facility are presented. Thermal analysis including different scenarios of pool heat load and safety limits are discussed . Cladding temperature limit during handling and storage process are specified for safe transfer of fuel

Spent Fuel Handling and Packaging Program: a survey of hot cell facilities

International Nuclear Information System (INIS)

Menon, M.N.

1978-07-01

Hot cell facilities in the United States were surveyed to determine their capabilities for conducting integral fuel assembly and individual fuel rod examinations that are required in support of the Spent Fuel Handling and Packaging Program. The ability to receive, handle, disassemble and reconstitute full-length light water reactor spent fuel assemblies, and the ability to conduct nondestructive and destructive examinations on full-length fuel rods were of particular interest. Three DOE-supported facilities and three commercial facilities were included in the survey. This report provides a summary of the findings

Advanced methods of solid oxide fuel cell modeling

CERN Document Server

Milewski, Jaroslaw; Santarelli, Massimo; Leone, Pierluigi

2011-01-01

Fuel cells are widely regarded as the future of the power and transportation industries. Intensive research in this area now requires new methods of fuel cell operation modeling and cell design. Typical mathematical models are based on the physical process description of fuel cells and require a detailed knowledge of the microscopic properties that govern both chemical and electrochemical reactions. ""Advanced Methods of Solid Oxide Fuel Cell Modeling"" proposes the alternative methodology of generalized artificial neural networks (ANN) solid oxide fuel cell (SOFC) modeling. ""Advanced Methods

Behavior of molybdenum in mixed-oxide fuel

International Nuclear Information System (INIS)

Giacchetti, G.; Sari, C.

1976-01-01

Metallic molybdenum, Mo--Ru--Rh--Pd alloys, barium, zirconium, and tungsten were added to uranium and uranium--plutonium oxides by coprecipitation and mechanical mixture techniques. This material was treated in a thermal gradient similar to that existing in fuel during irradiation to study the behavior of molybdenum in an oxide matrix as a function of the O/(U + Pu) ratio and some added elements. Result of ceramographic and microprobe analysis shows that when the overall O/(U + Pu) ratio is less than 2, molybdenum and Mo--Ru--Rh--Pd alloy inclusions are present in the uranium--plutonium oxide matrix. If the O/(U + Pu) ratio is greater than 2, molybdenum oxidizes to MoO 2 , which is gaseous at a temperature approximately 1000 0 C. Molybdenum oxide vapor reacts with barium oxide and forms a compound that exists as a liquid phase in the columnar grain region. Molybdenum oxide also reacts with tungsten oxide (tungsten is often present as an impurity in the fuel) and forms a compound that contains approximately 40 wt percent of actinide metals. The apparent solubility of molybdenum in uranium and uranium--plutonium oxides, determined by electron microprobe, was found to be less than 250 ppM both for hypo- and hyperstoichiometric fuels

Gas detection for alternate-fuel vehicle facilities.

Science.gov (United States)

Ferree, Steve

2003-05-01

Alternative fuel vehicles' safety is driven by local, state, and federal regulations in which fleet owners in key metropolitan [table: see text] areas convert much of their fleet to cleaner-burning fuels. Various alternative fuels are available to meet this requirement, each with its own advantages and requirements. This conversion to alternative fuels leads to special requirements for safety monitoring in the maintenance facilities and refueling stations. A comprehensive gas and flame monitoring system needs to meet the needs of both the user and the local fire marshal.

Analysis of fuel management in the KIPT neutron source facility

Energy Technology Data Exchange (ETDEWEB)

Zhong Zhaopeng, E-mail: zzhong@anl.gov [Nuclear Engineering Division, Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Gohar, Yousry; Talamo, Alberto [Nuclear Engineering Division, Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

2011-05-15

Research highlights: > Fuel management of KIPT ADS was analyzed. > Core arrangement was shuffled in stage wise. > New fuel assemblies was added into core periodically. > Beryllium reflector could also be utilized to increase the fuel life. - Abstract: Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of an experimental neutron source facility consisting of an electron accelerator driven sub-critical assembly. The neutron source driving the sub-critical assembly is generated from the interaction of 100 KW electron beam with a natural uranium target. The sub-critical assembly surrounding the target is fueled with low enriched WWR-M2 type hexagonal fuel assemblies. The U-235 enrichment of the fuel material is <20%. The facility will be utilized for basic and applied research, producing medical isotopes, and training young specialists. With the 100 KW electron beam power, the total thermal power of the facility is {approx}360 kW including the fission power of {approx}260 kW. The burnup of the fissile materials and the buildup of fission products continuously reduce the system reactivity during the operation, decrease the neutron flux level, and consequently impact the facility performance. To preserve the neutron flux level during the operation, the fuel assemblies should be added and shuffled for compensating the lost reactivity caused by burnup. Beryllium reflector could also be utilized to increase the fuel life time in the sub-critical core. This paper studies the fuel cycles and shuffling schemes of the fuel assemblies of the sub-critical assembly to preserve the system reactivity and the neutron flux level during the operation.

Capabilities for processing shipping casks at spent fuel storage facilities

International Nuclear Information System (INIS)

Baker, W.H.; Arnett, L.M.

1978-01-01

Spent fuel is received at a storage facility in heavily shielded casks transported either by rail or truck. The casks are inspected, cooled, emptied, decontaminated, and reshipped. The spent fuel is transferred to storage. The number of locations or space inside the building provided to perform each function in cask processing will determine the rate at which the facility can process shipping casks and transfer spent fuel to storage. Because of the high cost of construction of licensed spent fuel handling and storage facilities and the difficulty in retrofitting, it is desirable to correctly specify the space required. In this paper, the size of the cask handling facilities is specified as a function of rate at which spent fuel is received for storage. The minimum number of handling locations to achieve a given throughput of shipping casks has been determined by computer simulation of the process. The simulation program uses a Monte Carlo technique in which a large number of casks are received at a facility with a fixed number of handling locations in each process area. As a cask enters a handling location, the time to process the cask at that location is selected at random from the distribution of process time. Shipping cask handling times are based on experience at the General Electric Storage Facility, Morris, Illinois. Shipping cask capacity is based on the most recent survey available of the expected capability of reactors to handle existing rail or truck casks

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

Safety of Nuclear Fuel Cycle Facilities. Safety Requirements (Arabic Edition)

International Nuclear Information System (INIS)

2015-01-01

This publication covers the broad scope of requirements for fuel cycle facilities that, in light of the experience and present state of technology, must be satisfied to ensure safety for the lifetime of the facility. Topics of specific relevance include aspects of nuclear fuel generation, storage, reprocessing and disposal

Large-scale fuel cycle centres

International Nuclear Information System (INIS)

Smiley, S.H.; Black, K.M.

1977-01-01

The US Nuclear Regulatory Commission (NRC) has considered the nuclear energy centre concept for fuel cycle plants in the Nuclear Energy Centre Site Survey 1975 (NECSS-75) Rep. No. NUREG-0001, an important study mandated by the US Congress in the Energy Reorganization Act of 1974 which created the NRC. For this study, the NRC defined fuel cycle centres as consisting of fuel reprocessing and mixed-oxide fuel fabrication plants, and optional high-level waste and transuranic waste management facilities. A range of fuel cycle centre sizes corresponded to the fuel throughput of power plants with a total capacity of 50,000-300,000MW(e). The types of fuel cycle facilities located at the fuel cycle centre permit the assessment of the role of fuel cycle centres in enhancing the safeguard of strategic special nuclear materials - plutonium and mixed oxides. Siting fuel cycle centres presents a smaller problem than siting reactors. A single reprocessing plant of the scale projected for use in the USA (1500-2000t/a) can reprocess fuel from reactors producing 50,000-65,000MW(e). Only two or three fuel cycle centres of the upper limit size considered in the NECSS-75 would be required in the USA by the year 2000. The NECSS-75 fuel cycle centre evaluation showed that large-scale fuel cycle centres present no real technical siting difficulties from a radiological effluent and safety standpoint. Some construction economies may be achievable with fuel cycle centres, which offer opportunities to improve waste-management systems. Combined centres consisting of reactors and fuel reprocessing and mixed-oxide fuel fabrication plants were also studied in the NECSS. Such centres can eliminate shipment not only of Pu but also mixed-oxide fuel. Increased fuel cycle costs result from implementation of combined centres unless the fuel reprocessing plants are commercial-sized. Development of Pu-burning reactors could reduce any economic penalties of combined centres. The need for effective fissile

Decommissioning nuclear facilities

International Nuclear Information System (INIS)

Harmon, K.M.; Jenkins, C.E.; Waite, D.A.; Brooksbank, R.E.; Lunis, B.C.; Nemec, J.F.

1976-01-01

This paper describes the currently accepted alternatives for decommissioning retired light water reactor fuel cycle facilities and the current state of decommissioning technology. Three alternatives are recognized: Protective Storage; Entombment; and Dismantling. Application of these alternatives to the following types of facilities is briefly described: light water reactors; fuel reprocessing plants, and mixed oxide fuel fabrication plants. Brief descriptions are given of decommissioning operations and results at a number of sites, and recent studies of the future decommissioning of prototype fuel cycle facilities are reviewed. An overview is provided of the types of operations performed and tools used in common decontamination and decommissioning techniques and needs for improved technology are suggested. Planning for decommissioning a nuclear facility is dependent upon the maximum permitted levels of residual radioactive contamination. Proposed guides and recently developed methodology for development of site release criteria are reviewed. 21 fig, 32 references

Thermal gradient effects on the oxidation of Zircaloy fuel cladding

International Nuclear Information System (INIS)

Klein, A.C.; Reyes, J.N. Jr.; Maguire, M.A.

1990-01-01

A Thermal Gradient Test Facility (TGTF) has been designed and constructed to measure the thermal gradient effect on pressurized water reactor (PWR) fuel rod cladding. The TGTF includes a heat flux simulator assembly capable of producing a wide range of PWR operating conditions including water flow velocities and temperatures, water chemistry conditions, cladding temperatures, and heat fluxes ranging to 160 W/cm 2 . It is fully instrumented including a large number of thermocouples both inside the water flow channel and inside the cladding. Two test programs are in progress. First, cladding specimens are pre-oxidized in air at 500 deg. C and in 400 deg. C steam for various lengths of time to develop a range of uniform oxide thicknesses from 1 to 60 micrometers. The pre-oxidized specimens are placed in the TGTF to characterize the oxide thermal conductivity under a variety of water flow and heat flux conditions. Second, to overcome the long exposure times required under typical PWR conditions a series of tests with the addition of high concentrations of lithium hydroxide to the water are being considered. Static autoclave tests have been conducted with lithium hydroxide concentrations ranging from 0 to 2 moles per liter at 300, 330, and 360 deg. C for up to 36 hours. Results for zircaloy-4 show a considerable increase in the weight gain for the exposed samples with oxidation rate enhancement factors as high as 70 times that of pure water. Operation of the TGTF with elevated lithium hydroxide levels will yield real-time information concerning the effects of a heat flux on the oxidation kinetics of zircaloy fuel rod cladding. (author). 5 refs, 5 figs, 2 tabs

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.

Physicochemical analysis of interaction of oxide fuel with pyrocarbon coatings of fuel particles

International Nuclear Information System (INIS)

Lyutikov, R.A.; Khromov, Yu.F.; Chernikov, A.S.

1990-01-01

Equilibrium pressure of (CO+Kr,Xe) gases inside fuel particle with oxide kern depending on design features of fuel particle, on temperature. on (O/U) initial composition and fuel burnup is calculated using the suggested model. Analysis of possibility for gas pressure reduction by means of uranium carbide alloying of kern and degree increase of solid fission product retention (Cs for example) during alumosilicate alloying of uranium oxide is conducted

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

International Nuclear Information System (INIS)

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

1978-12-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1978-12-01

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

PND fuel handling decontamination: facilities and techniques

Energy Technology Data Exchange (ETDEWEB)

Pan, R Y [Ontario Hydro, Toronto, ON (Canada)

1997-12-31

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.

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

Scope and dissolution studies and characterization of irradiated nuclear fuel in Atalante Hot Cell Facilities (abstract and presentation slides)

Energy Technology Data Exchange (ETDEWEB)

Dancausse, Jean-Philippe; Reynier Tronche, Nathalie; Ferlay, Gilles; Herlet, Nathalie; Eysseric, Cathrine; Esbelin, Eric

2005-01-01

Since 1999, several studies on nuclear fuels were realised in C11/C12 Atalante Hot Cell. This paper presents firstly an overview of the apparatus used for fuel dissolution and characterisation like reactor design, gas trapping flask and solid/liquid separation. Then, the general methodology is described as a function of fuel, temperature, reagents, showing for each step, the reachable experimental data: Dissolution rate, chemical and radiochemical fuel composition including volatile LLRN, insoluble mass, composition, morphology, cladding chemical, radiochemical and physical characterisation using SIMS (made in Cadarache/LECA facilities), MEB. To conclude, some of the obtained results on 129I and 14C composition of oxide fuels, rate of dissolution and first results on dissolution studies of RERTR UMo fuel will be detailed. (Author)

Development of examination technique for oxide layer thickness measurement of irradiated fuel rods

International Nuclear Information System (INIS)

Koo, D. S.; Park, S. W.; Kim, J. H.; Seo, H. S.; Min, D. K.; Kim, E. K.; Chun, Y. B.; Bang, K. S.

1999-06-01

Technique for oxide layer thickness measurement of irradiated fuel rods was developed to measure oxide layer thickness and study characteristic of fuel rods. Oxide layer thickness of irradiated fuels were measured, analyzed. Outer oxide layer thickness of 3 cycle-irradiated fuel rods were 20 - 30 μm, inner oxide layer thickness 0 - 10 μm and inner oxide layer thickness on cracked cladding about 30 μm. Oxide layer thickness of 4 cycle-irradiated fuel rods were about 2 times as thick as those of 1 cycle-irradiated fuel rods. Oxide layer on lower region of irradiated fuel rods was thin and oxide layer from lower region to upper region indicated gradual increase in thickness. Oxide layer thickness from 2500 to 3000 mm showed maximum and oxide layer thickness from 3000 to top region of irradiated fuel rods showed decreasing trend. Inner oxide layer thicknesses of 4 cycle-irradiated fuel rod were about 8 μm at 750 - 3500 mm from the bottom end of fuel rod. Outer oxide layer thickness were about 8 μm at 750 - 1000 mm from the bottom end of fuel rod. These indicated gradual increase up to upper region from the bottom end of fuel rod. These indicated gradual increase up to upper region from the bottom end of fuel. Oxide layer thickness technique will apply safety evaluation and study of reactor fuels. (author). 6 refs., 14 figs

Spent fuel. Dissolution and oxidation

International Nuclear Information System (INIS)

Grambow, B.

1989-03-01

Data from studies of the low temperature air oxidation of spent fuel were retrieved in order to provide a basis for comparison between the mechanism of oxidation in air and corrosion in water. U 3 O 7 is formed by diffusion of oxygen into the UO 2 lattice. A diffusion coefficient of oxygen in the fuel matric was calculated for 25 degree C to be in the range of 10 -23 to 10 -25 m 2 /s. The initial rates of U release from spent fuel and from UO 2 appear to be similar. The lowest rates (at 25 degree c >10 -4 g/(m 2 d)) were observed under reducing conditions. Under oxidizing conditions the rates depend mainly of the nature and concentraion of the oxidant and/or on corbonate. In contact with air, typical initial rates at room temperature were in the range between 0.001 and 0.1 g/(m 2 d). A study of apparent U solubility under oxidizing conditions was performed and it was suggested that the controlling factor is the redox potential at the UO 2 surface rather than the E h of the bulk solution. Electrochemical arguments were used to predict that at saturation, the surface potential will eventually reach a value given by the boundaries at either the U 3 O 7 /U 3 O 8 or the U 3 O 7 /schoepite stability field, and a comparison with spent fuel leach data showed that the solution concentration of uranium is close to the calculated U solubility at the U 3 O 7 /U 3 O 8 boundary. The difference in the cumulative Sr and U release was calculated from data from Studsvik laboratory. The results reveal that the rate of Sr release decreases with the square root of time under U-saturated conditions. This time dependence may be rationalized either by grain boundary diffusion or by diffusion into the fuel matrix. Hence, there seems to be a possibility of an agreement between the Sr release data, structural information and data for oxygen diffusion in UO 2 . (G.B.)

Alternative anode materials for solid oxide fuel cells

Energy Technology Data Exchange (ETDEWEB)

Goodenough, John B.; Huang, Yun-Hui [Texas Materials Institute, ETC 9.102, 1 University Station, C2200, The University of Texas at Austin, Austin, TX 78712 (United States)

2007-11-08

The electrolyte of a solid oxide fuel cell (SOFC) is an O{sup 2-}-ion conductor. The anode must oxidize the fuel with O{sup 2-} ions received from the electrolyte and it must deliver electrons of the fuel chemisorption reaction to a current collector. Cells operating on H{sub 2} and CO generally use a porous Ni/electrolyte cermet that supports a thin, dense electrolyte. Ni acts as both the electronic conductor and the catalyst for splitting the H{sub 2} bond; the oxidation of H{sub 2} to H{sub 2}O occurs at the Ni/electrolyte/H{sub 2} triple-phase boundary (TPB). The CO is oxidized at the oxide component of the cermet, which may be the electrolyte, yttria-stabilized zirconia, or a mixed oxide-ion/electron conductor (MIEC). The MIEC is commonly a Gd-doped ceria. The design and fabrication of these anodes are evaluated. Use of natural gas as the fuel requires another strategy, and MIECs are being explored for this application. The several constraints on these MIECs are outlined, and preliminary results of this on-going investigation are reviewed. (author)

Interim results from UO2 fuel oxidation tests in air

International Nuclear Information System (INIS)

Campbell, T.K.; Gilbert, E.R.; Thornhill, C.K.; White, G.D.; Piepel, G.F.; Griffin, C.W.j.

1987-08-01

An experimental program is being conducted at Pacific Northwest Laboratory (PNL) to extend the characterization of spent fuel oxidation in air. To characterize oxidation behavior of irradiated UO 2 , fuel oxidation tests were performed on declad light-water reactor spent fuel and nonirradited UO 2 pellets in the temperature range of 135 to 250 0 C. These tests were designed to determine the important independent variables that might affect spent fuel oxidation behavior. The data from this program, when combined with the test results from other programs, will be used to develop recommended spent fuel dry-storage temperature limits in air. This report describes interim test results. The initial PNL investigations of nonirradiated and spent fuels identified the important testing variables as temperature, fuel burnup, radiolysis of the air, fuel microstructure, and moisture in the air. Based on these initial results, a more extensive statistically designed test matrix was developed to study the effects of temperature, burnup, and moisture on the oxidation behavior of spent fuel. Oxidation tests were initiated using both boiling-water reactor and pressurized-water reactor fuels from several different reactors with burnups from 8 to 34 GWd/MTU. A 10 5 R/h gamma field was applied to the test ovens to simulate dry storage cask conditions. Nonirradiated fuel was included as a control. This report describes experimental results from the initial tests on both the spent and nonirradiated fuels and results to date on the tests in a 10 5 R/h gamma field. 33 refs., 51 figs., 6 tabs

Aluminum cladding oxidation of prefilmed in-pile fueled experiments

Energy Technology Data Exchange (ETDEWEB)

Marcum, W.R., E-mail: marcumw@engr.orst.edu [Oregon State University, School of Nuclear Science and Engineering, 116 Radiation Center, Corvallis, OR 97331 (United States); Wachs, D.M.; Robinson, A.B.; Lillo, M.A. [Idaho National Laboratory, Nuclear Fuels & Materials Department, 2525 Fremont Ave., Idaho Falls, ID 83415 (United States)

2016-04-01

A series of fueled irradiation experiments were recently completed within the Advanced Test Reactor Full size plate In center flux trap Position (AFIP) and Gas Test Loop (GTL) campaigns. The conduct of the AFIP experiments supports ongoing efforts within the global threat reduction initiative (GTRI) to qualify a new ultra-high loading density low enriched uranium-molybdenum fuel. This study details the characterization of oxide growth on the fueled AFIP experiments and cross-correlates the empirically measured oxide thickness values to existing oxide growth correlations and convective heat transfer correlations that have traditionally been utilized for such an application. This study adds new and valuable empirical data to the scientific community with respect to oxide growth measurements of highly irradiated experiments, of which there is presently very limited data. Additionally, the predicted oxide thickness values are reconstructed to produce an oxide thickness distribution across the length of each fueled experiment (a new application and presentation of information that has not previously been obtainable in open literature); the predicted distributions are compared against experimental data and in general agree well with the exception of select outliers. - Highlights: • New experimental data is presented on oxide layer thickness of irradiated aluminum fuel. • Five oxide growth correlations and four convective heat transfer correlations are used to compute the oxide layer thickness. • The oxide layer thickness distribution is predicted via correlation for each respective experiment. • The measured experiment and predicted distributions correlate well, with few outliers.

Analysis of fuel management in the KIPT neutron source facility

International Nuclear Information System (INIS)

Zhong Zhaopeng; Gohar, Yousry; Talamo, Alberto

2011-01-01

Research highlights: → Fuel management of KIPT ADS was analyzed. → Core arrangement was shuffled in stage wise. → New fuel assemblies was added into core periodically. → Beryllium reflector could also be utilized to increase the fuel life. - Abstract: Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of an experimental neutron source facility consisting of an electron accelerator driven sub-critical assembly. The neutron source driving the sub-critical assembly is generated from the interaction of 100 KW electron beam with a natural uranium target. The sub-critical assembly surrounding the target is fueled with low enriched WWR-M2 type hexagonal fuel assemblies. The U-235 enrichment of the fuel material is <20%. The facility will be utilized for basic and applied research, producing medical isotopes, and training young specialists. With the 100 KW electron beam power, the total thermal power of the facility is ∼360 kW including the fission power of ∼260 kW. The burnup of the fissile materials and the buildup of fission products continuously reduce the system reactivity during the operation, decrease the neutron flux level, and consequently impact the facility performance. To preserve the neutron flux level during the operation, the fuel assemblies should be added and shuffled for compensating the lost reactivity caused by burnup. Beryllium reflector could also be utilized to increase the fuel life time in the sub-critical core. This paper studies the fuel cycles and shuffling schemes of the fuel assemblies of the sub-critical assembly to preserve the system reactivity and the neutron flux level during the operation.

Fuel supply shutdown facility interim operational safety requirements

International Nuclear Information System (INIS)

Besser, R.L.; Brehm, J.R.; Benecke, M.W.; Remaize, J.A.

1995-01-01

These Interim Operational Safety Requirements (IOSR) for the Fuel Supply Shutdown (FSS) facility define acceptable conditions, safe boundaries, bases thereof, and management or administrative controls to ensure safe operation. The IOSRs apply to the fuel material storage buildings in various modes (operation, storage, surveillance)

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.

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

Planar solid oxide fuel cells: the Australian experience and outlook

Science.gov (United States)

Godfrey, Bruce; Föger, Karl; Gillespie, Rohan; Bolden, Roger; Badwal, S. P. S.

Since 1992, Ceramic Fuel Cells (CFCL) has grown to what is now the largest focussed program globally for development of planar ceramic (solid oxide) fuel cell, SOFC, technology. A significant intellectual property position in know-how and patents has been developed, with over 80 people involved in the venture. Over $A60 million in funding for the activities of the company has been raised from private companies, government-owned corporations and government business-support programs, including from energy — particularly electricity — industry shareholders that can facilitate access to local markets for our products. CFCL has established state-of-the-art facilities for planar SOFC R&D, with their expansion and scaling-up to pilot manufacturing capability underway. We expect to achieve commercial introduction of our market-entry products in 2002, with prototype systems expected to be available from early 2001.

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)

Support of Construction and Verification of Out-of-Pile Fuel Assembly Test Facilities

International Nuclear Information System (INIS)

Park, Nam Gyu; Kim, K. T.; Park, J. K.

2006-12-01

Fuel assembly and components should be verified by the out-of-pile test facilities in order to load the developed fuel in reactor. Even though most of the component-wise tests have been performed using the facilities in land, the assembly-wise tests has been depended on the oversees' facility due to the lack of the facilities. KAERI started to construct the assembly-wise mechanical/hydraulic test facilities and KNF, as an end user, is supporting the mechanical/hydraulic test facility construction by using the technologies studied through the fuel development programs. The works performed are as follows: - Test assembly shipping container design and manufacturing support - Fuel handling tool design : Gripper, Upper and lower core simulators for assembly mechanical test facility, Internals for assembly hydraulic test facility - Manufacture of test specimens : skeleton and assembly for preliminary functional verification of assembly mechanical/hydraulic test facilities, two assemblies for the verification of assembly mechanical/hydraulic test facilities, Instrumented rod design and integrity evaluation - Verification of assembly mechanical/hydraulic test facilities : test data evaluation

Support of Construction and Verification of Out-of-Pile Fuel Assembly Test Facilities

Energy Technology Data Exchange (ETDEWEB)

Park, Nam Gyu; Kim, K. T.; Park, J. K. [KNF, Daejeon (Korea, Republic of)] (and others)

2006-12-15

Fuel assembly and components should be verified by the out-of-pile test facilities in order to load the developed fuel in reactor. Even though most of the component-wise tests have been performed using the facilities in land, the assembly-wise tests has been depended on the oversees' facility due to the lack of the facilities. KAERI started to construct the assembly-wise mechanical/hydraulic test facilities and KNF, as an end user, is supporting the mechanical/hydraulic test facility construction by using the technologies studied through the fuel development programs. The works performed are as follows: - Test assembly shipping container design and manufacturing support - Fuel handling tool design : Gripper, Upper and lower core simulators for assembly mechanical test facility, Internals for assembly hydraulic test facility - Manufacture of test specimens : skeleton and assembly for preliminary functional verification of assembly mechanical/hydraulic test facilities, two assemblies for the verification of assembly mechanical/hydraulic test facilities, Instrumented rod design and integrity evaluation - Verification of assembly mechanical/hydraulic test facilities : test data evaluation.

A novel direct carbon fuel cell by approach of tubular solid oxide fuel cells

Energy Technology Data Exchange (ETDEWEB)

Liu, Renzhu; Zhao, Chunhua; Li, Junliang; Zeng, Fanrong; Wang, Shaorong; Wen, Tinglian; Wen, Zhaoyin [CAS Key Laboratory of Materials for Energy Conversion, Shanghai Inorganic Energy Materials and Power Source Engineering Center, Shanghai Institute of Ceramics, Chinese Academy of Sciences (SICCAS), 1295 Dingxi Road, Shanghai 200050 (China)

2010-01-15

A direct carbon fuel cell based on a conventional anode-supported tubular solid oxide fuel cell, which consisted of a NiO-YSZ anode support tube, a NiO-ScSZ anode functional layer, a ScSZ electrolyte film, and a LSM-ScSZ cathode, has been successfully achieved. It used the carbon black as fuel and oxygen as the oxidant, and a preliminary examination of the DCFC has been carried out. The cell generated an acceptable performance with the maximum power densities of 104, 75, and 47 mW cm{sup -2} at 850, 800, and 750 C, respectively. These results demonstrate the feasibility for carbon directly converting to electricity in tubular solid oxide fuel cells. (author)

Analytical throughput-estimating methods for the Hot Fuel Examination Facility

International Nuclear Information System (INIS)

Keyes, R.W.; Phipps, R.D.

1983-01-01

The Hot Fuel Examination Facility (HFEF) supports the operation and experimental programs of the major Liquid Metal Fast Breeder Reactor (LMFBR) test facilities; specifically, the Fast Flux Test Facility (FFTF), the Experimental Breeder Reactor II (EBR-II), and the Transient Reactor Test (TREAT) Facility. Successful management of HFEF and of LMFBR safety and fuels and materials programs, therefore, requires reliable information regarding HFEF's capability to handle expected or proposed program work loads. This paper describes the 10-step method that has been developed to consider all variables which significantly affect the HFEF examination throughput and quickly provide the necessary planning information

Uranium plutonium oxide fuels

International Nuclear Information System (INIS)

Cox, C.M.; Leggett, R.D.; Weber, E.T.

1981-01-01

Uranium plutonium oxide is the principal fuel material for liquid metal fast breeder reactors (LMFBR's) throughout the world. Development of this material has been a reasonably straightforward evolution from the UO 2 used routinely in the light water reactor (LWR's); but, because of the lower neutron capture cross sections and much lower coolant pressures in the sodium cooled LMFBR's, the fuel is operated to much higher discharge exposures than that of a LWR. A typical LMFBR fuel assembly is shown. Depending on the required power output and the configuration of the reactor, some 70 to 400 such fuel assemblies are clustered to form the core. There is a wide variation in cross section and length of the assemblies where the increasing size reflects a chronological increase in plant size and power output as well as considerations of decreasing the net fuel cycle cost. Design and performance characteristics are described

Desulfurization of Diesel Fuel by Oxidation and Solvent Extraction

Directory of Open Access Journals (Sweden)

Wadood Taher Mohammed

2015-02-01

Full Text Available This research presents a study in ultra-desulfurization of diesel fuel produced from conventional hydro desulfurization process, using oxidation and solvent extraction techniques. Dibenzothiophene (DBT was the organosulfur compound that had been detected in sulfur removal. The oxidation process used hydrogen peroxide as an oxidant and acetic acid as homogeneous catalyst . The solvent extraction process used acetonitrile (ACN and N-methyl – 2 - pyrrolidone (NMP as extractants . Also the effect of five parameters (stirring speed :150 , 250 , 350 , and 450 rpm, temperature (30 , 40 , 45 , and 50 oC, oxidant/simulated diesel fuel ratio (0.5 , 0.75 , 1 , and 1.5 , catalyst/oxidant ratio(0.125,0.25,0.5,and0.75 , and solvent/simulated diesel fuel ratio(0.5,0.6,0.75,and1 were examined as well as solvent type. The results exhibit that the highest removal of sulfur is 98.5% using NMP solvent while it is 95.8% for ACN solvent. The set of conditions that show the highest sulfur removal is: stirring speed of 350 rpm , temperature 50oC , oxidant/simulated diesel fuel ratio 1 , catalyst/oxidant ratio 0.5 , solvent/simulated diesel fuel ratio 1. These best conditions were applied upon real diesel fuel (produced from Al-Dora refinerywith 1000 ppm sulfur content . It was found that sulfur removal was 64.4% using ACN solvent and 75% using NMP solvent.

Safety analysis of DUPIC fuel development facility

International Nuclear Information System (INIS)

Lee, H. H.; Park, J. J.; Shin, J. M.; Yang, M. S.; Baek, S. Y.; Ahn, J. Y.

2001-01-01

Various experimental facilities are necessary in order to perform experimental verification for development of DUPIC fuel fabrication technology. In special, since highly radioactive material such as spent PWR fuel is used for this experiment, DUPIC fuel fabrication has to be performed in hot cell by remote handling. Therefore, it should be provided with proper engineering requirement and safety. M6 hot cell of IMEF which is to used for DUPIC fuel fabrication experiment was constructed as an α-γ hot cell for material examination of small amount of high-burnup fuel. The characteristics and amount of spent fuel for DUPIC fuel fabrication experiment will be different from the original design criteria. Therefore, the increased amount of spent fuel and different characteristics of experiment result in not only change of shielding and enviornmental evaluation results but new requirement of nuclear criticality evaluation. Therefore, this study includes evaluation of shielding, environmental effect and nuclear criticality in case that IMEF M6 hot cell is used for DUPIC fuel fabrication

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

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

Assessment of bio-fuel options for solid oxide fuel cell applications

Science.gov (United States)

Lin, Jiefeng

Rising concerns of inadequate petroleum supply, volatile crude oil price, and adverse environmental impacts from using fossil fuels have spurred the United States to promote bio-fuel domestic production and develop advanced energy systems such as fuel cells. The present dissertation analyzed the bio-fuel applications in a solid oxide fuel cell-based auxiliary power unit from environmental, economic, and technological perspectives. Life cycle assessment integrated with thermodynamics was applied to evaluate the environmental impacts (e.g., greenhouse gas emission, fossil energy consumption) of producing bio-fuels from waste biomass. Landfill gas from municipal solid wastes and biodiesel from waste cooking oil are both suggested as the promising bio-fuel options. A nonlinear optimization model was developed with a multi-objective optimization technique to analyze the economic aspect of biodiesel-ethanol-diesel ternary blends used in transportation sectors and capture the dynamic variables affecting bio-fuel productions and applications (e.g., market disturbances, bio-fuel tax credit, policy changes, fuel specification, and technological innovation). A single-tube catalytic reformer with rhodium/ceria-zirconia catalyst was used for autothermal reformation of various heavy hydrocarbon fuels (e.g., diesel, biodiesel, biodiesel-diesel, and biodiesel-ethanol-diesel) to produce a hydrogen-rich stream reformates suitable for use in solid oxide fuel cell systems. A customized mixing chamber was designed and integrated with the reformer to overcome the technical challenges of heavy hydrocarbon reformation. A thermodynamic analysis, based on total Gibbs free energy minimization, was implemented to optimize the operating environment for the reformations of various fuels. This was complimented by experimental investigations of fuel autothermal reformation. 25% biodiesel blended with 10% ethanol and 65% diesel was determined to be viable fuel for use on a truck travelling with

Thermodynamic and kinetic modelling of fuel oxidation behaviour in operating defective fuel

International Nuclear Information System (INIS)

Lewis, B.J.; Thompson, W.T.; Akbari, F.; Thompson, D.M.; Thurgood, C.; Higgs, J.

2004-01-01

A theoretical treatment has been developed to predict the fuel oxidation behaviour in operating defective nuclear fuel elements. The equilibrium stoichiometry deviation in the hyper-stoichiometric fuel has been derived from thermodynamic considerations using a self-consistent set of thermodynamic properties for the U-O system, which emphasizes replication of solubilities and three-phase invariant conditions displayed in the U-O binary phase diagram. The kinetics model accounts for multi-phase transport including interstitial oxygen diffusion in the solid and gas-phase transport of hydrogen and steam in the fuel cracks. The fuel oxidation model is further coupled to a heat conduction model to account for the feedback effect of a reduced thermal conductivity in the hyper-stoichiometric fuel. A numerical solution has been developed using a finite-element technique with the FEMLAB software package. The model has been compared to available data from several in-reactor X-2 loop experiments with defective fuel conducted at the Chalk River Laboratories. The model has also been benchmarked against an O/U profile measurement for a spent defective fuel element discharged from a commercial reactor

The TMI regenerable solid oxide fuel cell

Science.gov (United States)

Cable, Thomas L.

1995-04-01

Energy storage and production in space requires rugged, reliable hardware which minimizes weight, volume, and maintenance while maximizing power output and usable energy storage. These systems generally consist of photovoltaic solar arrays which operate during sunlight cycles to provide system power and regenerate fuel (hydrogen) via water electrolysis; during dark cycles, hydrogen is converted by the fuel cell into system. The currently preferred configuration uses two separate systems (fuel cell and electrolyzer) in conjunction with photovoltaic cells. Fuel cell/electrolyzer system simplicity, reliability, and power-to-weight and power-to-volume ratios could be greatly improved if both power production (fuel cell) and power storage (electrolysis) functions can be integrated into a single unit. The Technology Management, Inc. (TMI), solid oxide fuel cell-based system offers the opportunity to both integrate fuel cell and electrolyzer functions into one unit and potentially simplify system requirements. Based an the TMI solid oxide fuel cell (SOPC) technology, the TMI integrated fuel cell/electrolyzer utilizes innovative gas storage and operational concepts and operates like a rechargeable 'hydrogen-oxygen battery'. Preliminary research has been completed on improved H2/H2O electrode (SOFC anode/electrolyzer cathode) materials for solid oxide, regenerative fuel cells. Improved H2/H2O electrode materials showed improved cell performance in both fuel cell and electrolysis modes in reversible cell tests. ln reversible fuel cell/electrolyzer mode, regenerative fuel cell efficiencies (ratio of power out (fuel cell mode) to power in (electrolyzer model)) improved from 50 percent (using conventional electrode materials) to over 80 percent. The new materials will allow the TMI SOFC system to operate as both the electrolyzer and fuel cell in a single unit. Preliminary system designs have also been developed which indicate the technical feasibility of using the TMI SOFC

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)

West Valley facility spent fuel handling, storage, and shipping experience

International Nuclear Information System (INIS)

Bailey, W.J.

1990-11-01

The result of a study on handling and shipping experience with spent fuel are described in this report. The study was performed by Pacific Northwest Laboratory (PNL) and was jointly sponsored by the US Department of Energy (DOE) and the Electric Power Research Institute (EPRI). The purpose of the study was to document the experience with handling and shipping of relatively old light-water reactor (LWR) fuel that has been in pool storage at the West Valley facility, which is at the Western New York Nuclear Service Center at West Valley, New York and operated by DOE. A subject of particular interest in the study was the behavior of corrosion product deposits (i.e., crud) deposits on spent LWR fuel after long-term pool storage; some evidence of crud loosening has been observed with fuel that was stored for extended periods at the West Valley facility and at other sites. Conclusions associated with the experience to date with old spent fuel that has been stored at the West Valley facility are presented. The conclusions are drawn from these subject areas: a general overview of the West Valley experience, handling of spent fuel, storing of spent fuel, rod consolidation, shipping of spent fuel, crud loosening, and visual inspection. A list of recommendations is provided. 61 refs., 4 figs., 5 tabs

Investigations of fuel cladding chemical interaction in irradiated LMFBR type oxide fuel pins

International Nuclear Information System (INIS)

Roake, W.E.; Adamson, M.G.; Hilbert, R.F.; Langer, S.

1977-01-01

Understanding and controlling the chemical attack of fuel pin cladding by fuel and fission products are major objectives of the U.S. LMFBR Mixed Oxide Irradiation Testing Program. Fuel-cladding chemical interaction (FCCI) has been recognized as an important factor in the ability to achieve goal peak burnups of 8% (80.MWd/kg) in FFTF and in excess of 10% (100.MWd/kg) in the LMFBR demonstration reactors while maintaining coolant bulk outlet temperatures up to ∼60 deg. C (1100 deg. F). In this paper we review pertinent parts of the irradiation program and describe recent observation of FCCI in the fuel pins of this program. One goal of the FCCI investigations is to obtain a sufficiently quantitative understanding of FCCI such that correlations can be developed relating loss of effective cladding thickness to irradiation and fuel pin fabrication parameters. Wastage correlations being developed using different approaches are discussed. Much of the early data on FCCI obtained in the U.S. Mixed Oxide Fuel Program came from capsule tests irradiated in both fast and thermal flux facilities. The fast flux irradiated encapsulated fuel pins continue to provide valuable data and insight into FCCI. Currently, however, bare pins with prototypic fuels and cladding irradiated in the fast flux Experimental Breeder Reactor-II (EBR-II) as multiple pin assemblies under prototypic powers, temperatures and thermal gradients are providing growing quantities of data on FCCI characteristics and cladding thickness losses from FCCI. A few special encapsulated fuel pin tests are being conducted in the General Electric Test Reactor (GETR) and EBR-II, but these are aimed at providing specific information under irradiation conditions not achievable in the fast flux bare pin assemblies or because EBR-II Operation or Safety requirements dictate that the pins be encapsulated. The discussion in this paper is limited to fast flux irradiation test results from encapsulated pins and multiple pin

Investigations of fuel cladding chemical interaction in irradiated LMFBR type oxide fuel pins

Energy Technology Data Exchange (ETDEWEB)

Roake, W E [Westinghouse-Hanford Co., Richland, WA (United States); Adamson, M G [General Electric Company, Vallecitos Nuclear Center, Pleasanton, CA (United States); Hilbert, R F; Langer, S

1977-04-01

Understanding and controlling the chemical attack of fuel pin cladding by fuel and fission products are major objectives of the U.S. LMFBR Mixed Oxide Irradiation Testing Program. Fuel-cladding chemical interaction (FCCI) has been recognized as an important factor in the ability to achieve goal peak burnups of 8% (80.MWd/kg) in FFTF and in excess of 10% (100.MWd/kg) in the LMFBR demonstration reactors while maintaining coolant bulk outlet temperatures up to {approx}60 deg. C (1100 deg. F). In this paper we review pertinent parts of the irradiation program and describe recent observation of FCCI in the fuel pins of this program. One goal of the FCCI investigations is to obtain a sufficiently quantitative understanding of FCCI such that correlations can be developed relating loss of effective cladding thickness to irradiation and fuel pin fabrication parameters. Wastage correlations being developed using different approaches are discussed. Much of the early data on FCCI obtained in the U.S. Mixed Oxide Fuel Program came from capsule tests irradiated in both fast and thermal flux facilities. The fast flux irradiated encapsulated fuel pins continue to provide valuable data and insight into FCCI. Currently, however, bare pins with prototypic fuels and cladding irradiated in the fast flux Experimental Breeder Reactor-II (EBR-II) as multiple pin assemblies under prototypic powers, temperatures and thermal gradients are providing growing quantities of data on FCCI characteristics and cladding thickness losses from FCCI. A few special encapsulated fuel pin tests are being conducted in the General Electric Test Reactor (GETR) and EBR-II, but these are aimed at providing specific information under irradiation conditions not achievable in the fast flux bare pin assemblies or because EBR-II Operation or Safety requirements dictate that the pins be encapsulated. The discussion in this paper is limited to fast flux irradiation test results from encapsulated pins and multiple pin

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)

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)

Minor Actinide Laboratory at JRC-ITU: Fuel fabrication facility

International Nuclear Information System (INIS)

Fernandez, A.; McGinley, J.; Somers, J.

2008-01-01

The Minor Actinide Laboratory (MA-lab) of the Institute for Transuranium Elements is a unique facility for the fabrication of fuels and targets containing minor actinides (MA). It is of key importance for research on Partitioning and Transmutation in Europe, as it is one of the only dedicated facilities for the fabrication of MA containing materials, either for property measurements or for the production of test pins for irradiation experiments. In this paper a detailed description of the MA-Lab facility and the fabrication processes developed to fabricate fuels and samples containing high content of minor actinides is given. In addition, experience gained and improvements are also outlined. (authors)

Large-scale fuel cycle centers

International Nuclear Information System (INIS)

Smiley, S.H.; Black, K.M.

1977-01-01

The United States Nuclear Regulatory Commission (NRC) has considered the nuclear energy center concept for fuel cycle plants in the Nuclear Energy Center Site Survey - 1975 (NECSS-75) -- an important study mandated by the U.S. Congress in the Energy Reorganization Act of 1974 which created the NRC. For the study, NRC defined fuel cycle centers to consist of fuel reprocessing and mixed oxide fuel fabrication plants, and optional high-level waste and transuranic waste management facilities. A range of fuel cycle center sizes corresponded to the fuel throughput of power plants with a total capacity of 50,000 - 300,000 MWe. The types of fuel cycle facilities located at the fuel cycle center permit the assessment of the role of fuel cycle centers in enhancing safeguarding of strategic special nuclear materials -- plutonium and mixed oxides. Siting of fuel cycle centers presents a considerably smaller problem than the siting of reactors. A single reprocessing plant of the scale projected for use in the United States (1500-2000 MT/yr) can reprocess the fuel from reactors producing 50,000-65,000 MWe. Only two or three fuel cycle centers of the upper limit size considered in the NECSS-75 would be required in the United States by the year 2000 . The NECSS-75 fuel cycle center evaluations showed that large scale fuel cycle centers present no real technical difficulties in siting from a radiological effluent and safety standpoint. Some construction economies may be attainable with fuel cycle centers; such centers offer opportunities for improved waste management systems. Combined centers consisting of reactors and fuel reprocessing and mixed oxide fuel fabrication plants were also studied in the NECSS. Such centers can eliminate not only shipment of plutonium, but also mixed oxide fuel. Increased fuel cycle costs result from implementation of combined centers unless the fuel reprocessing plants are commercial-sized. Development of plutonium-burning reactors could reduce any

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

Energy Technology Data Exchange (ETDEWEB)

Lochard, Jacques

1989-08-01

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.

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

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

International Nuclear Information System (INIS)

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

2013-01-01

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

JAEA key facilities for global advanced fuel cycle R and D

Energy Technology Data Exchange (ETDEWEB)

Nomura, Shigeo; Yamamoto, Ryuichi [Nuclear Fuel Cycle Engineering Labos, JAEA, 4-33 Tokai-mura, Ibaraki, 319-1194 (Japan)

2008-07-01

Advanced fuel cycle will be realized with the mid and long term R and D during the long-term transition period from LWR cycle to advanced reactor fuel cycle. Most of JAEA facilities have been utilized to establish the current LWR and FBR (Fast Breeder Reactor) fuel cycle by implementing evolutionary R and D. An assessment of today's state experimental facilities concerning the following research issues: reprocessing, Mox fuel fabrication, irradiation and post-irradiation examination, waste management and nuclear data measurement, is made. The revolutionary R and D requests new issues to be studied: the TRU multi-recycling, minor actinide recycling, the assessment of proliferation resistance and the assessment of cost reduction. To implement the revolutionary R and D for advanced fuel cycle, however, these facilities should be refurbished to install new machines and process equipment to provide more flexible testing parameters.

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

Material Accounting Issues at the U.S. MOX Fuel Fabrication Facility

International Nuclear Information System (INIS)

Lyman, E.

2015-01-01

The Mixed Oxide Fuel Fabrication Facility (MFFF) is under construction in the United States. The plant is being licenced by the U.S. Nuclear Regulatory Commission (NRC), which as the U.S. SSAC regulates both domestic MC and A and compliance with international safeguards (where applicable). Among the NRC's MC and A requirements for Category I fuel cycle facilities are programmes for item and process monitoring. The NRC also has requirements for timely resolution of alarms and assessment of the validity of alleged thefts. NRC's item monitoring requirement specifies that the operator must be able to verify the 'presence and integrity' of items, with the goal of detecting the loss of items containing 2 kilogrammes of plutonium within certain time periods. The requirements for resolution of alarms and assessment of alleged thefts also generally require some capability to locate and verify items on demand. However, to the extent these regulations mandate that individual items be physically located and verified by hand, they can be difficult (or impossible) to meet for facilities with large numbers of items. The MFFF design was based largely on French facilities that were not subject to similar requirements. Consequently, the applicant proposed a novel item monitoring approach that relies on the data within the plant's computerized inventory and process control systems. This proposal was challenged in July 2010 by intervenors, raising questions such as whether computer systems could be used as the sole means for verification, given the potential for data to be compromised. In February 2014, the NRC's Atomic Safety and Licencing Board issued a decision upholding the applicant's plan, but one of the three judges issued a dissent, citing concern about cyberterrorism. This paper will discuss the issues argued during the hearing and their broader relevance. (author)

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

Photoelectrochemical and electrocatalytic properties of thermally oxidized copper oxide for efficient solar fuel production

KAUST Repository

Garcia Esparza, Angel T.; Limkrailassiri, Kevin; Leroy, Fré dé ric; Rasul, Shahid; Yu, Weili; Lin, Liwei; Takanabe, Kazuhiro

2014-01-01

We report the use of a facile and highly scalable synthesis process to control growth products of earth-abundant Cu-based oxides and their application in relevant photoelectrochemical and electrochemical solar fuel generation systems. Characterization of the synthesized Cu(I)/Cu(II) oxides indicates that their surface morphology and chemical composition can be simply tuned by varying two synthesis parameters (time and temperature). UV-Vis spectroscopy and impedance spectroscopy studies are performed to estimate the band structures and electronic properties of these p-type semiconductor materials. Photoelectrodes made of Cu oxides possess favorable energy band structures for production of hydrogen from water; the position of their conduction band is ≈1 V more negative than the water-reduction potential. High acceptor concentrations on the order of 1018-1019 cm-3 are obtained, producing large electric fields at the semiconductor-electrolyte interface and thereby enhancing charge separation. The highly crystalline pristine samples used as photocathodes in photoelectrochemical cells exhibit high photocurrents under AM 1.5G simulated illumination. When the samples are electrochemically reduced under galvanostatic conditions, the co-existence of the oxide with metallic Cu on the surface seems to function as an effective catalyst for the selective electrochemical reduction of CO2. © the Partner Organisations 2014.

Safeguardability assessment on pilot-scale advanced spent fuel conditioning facility

International Nuclear Information System (INIS)

Lee, S.Y.; Li, T.K.; Pickett, S.E.; Miller, M.C.; Ko, W.I.; Kim, H.D.

2006-01-01

Full text: In South Korea, approximately 6,000 metric tons of spent nuclear fuel from commercial reactor operation has been accumulated with the expectation of more than 30,000 metric tons, three times the present storage capacity, by the end of 2040. To resolve these challenges in spent fuel management, the Korea Atomic Energy Research Institute (KAERI) has been developing a dry reprocessing technology called Advanced Spent Fuel Conditioning Process (ACP). This is an electrometallurgical treatment technique to convert oxide-type spent fuel into a metallic form, and the electrolytic reduction (ER) technology developed recently is known as a more efficient concept for spent fuel conditioning. The goal of the ACP study is to recover more than 99% of the actinide elements into a metallic form with minimizing the volume and heat load of spent fuel. The significant reduction of the volume and heat load of spent fuel is expected to lighten the burden of final disposal in terms of disposal size, safety, and economics. In the framework of R and D collaboration for the ACP safeguards, a joint study on the safeguardability of the ACP technology has been performed by the Los Alamos National Laboratory (LANL) and KAERI. The purpose of this study is to address the safeguardability of the ACP technology, through analysis of material flow and development of a proper safeguards system that meet IAEA's comprehensive safeguards objective. The sub-processes and material flow of the pilot-scale ACP facility were analyzed, and subsequently the relevant material balance area (MBA) and key measurement point (KMP) were designed for material accounting. The uncertainties in material accounting were also estimated with international target values, and design requirements for the material accounting systems were derived

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)

Operational experience of the fuel cleaning facility of Joyo

Energy Technology Data Exchange (ETDEWEB)

Mukaibo, R; Matsuno, Y; Sato, I; Yoneda, Y; Ito, H [O-arai Engineering Centre, PNC, Ibaraki-ken, Tokio (Japan)

1978-08-01

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 {approx} 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)

Selection of away-from-reactor facilities for spent fuel storage. A guidebook

International Nuclear Information System (INIS)

2007-09-01

This publication aims to provide information on the approaches and criteria that would have to be considered for the selection of away-from-reactor (AFR) type spent fuel storage facilities, needs for which have been growing in an increasing number of Member States producing nuclear power. The AFR facilities can be defined as a storage system functionally independent of the reactor operation providing the role of storage until a further destination such as a disposal) becomes available. Initially developed to provide additional storage space for spent fuel, some AFR storage options are now providing additional spaces for extended storage of spent fuel with a prospect for long term storage, which is becoming a progressive reality in an increasing number of Member States due to the continuing debate on issues associated with the endpoints for spent fuel management and consequent delays in the implementation of final steps, such as disposal. The importance of AFR facilities for storage of spent fuel has been recognized for several decades and addressed in various IAEA publications in the area of spent fuel management. The Guidebook on Spent Fuel Storage (Technical Reports Series No. 240 published in 1984 and revised in 1991) discusses factors to be considered in the evaluation of spent fuel storage options. A technical committee meeting (TCM) on Selection of Dry Spent Fuel Storage Technologies held in Tokyo in 1995 also deliberated on this issue. However, there has not been any stand-alone publication focusing on the topic of selection of AFR storage facilities. The selection of AFR storage facilities is in fact a critical step for the successful implementation of spent fuel management programmes, due to the long operational periods required for storage and fuel handling involved with the additional implication of subsequent penalties in reversing decisions or changing the option mid-stream especially after the construction of the facility. In such a context, the long

Micro-Solid Oxide Fuel Cell: A multi-fuel approach for portable applications

International Nuclear Information System (INIS)

Patil, Tarkeshwar C.; Duttagupta, Siddhartha P.

2016-01-01

Highlights: • We report the oxygen ion transport properties at the electrode–electrolyte interface (EEI) of the SOFC for the first time. • This ion transport plays a key role in the overall performance of SOFCs with different fuels. • The GIIB mechanism is also studied for the first time. • GIIB is assumed to be the prime reason for low power density and ion conductivity at the EEI when using hydrocarbon fuels. • Due to its scalability, a fuel cell can serve as a power source for on-chip applications and all portable equipment. - Abstract: The impact of oxygen ion transport at the electrolyte–electrode interface of a micro-solid oxide fuel cell using different fuels is investigated. Model validation is performed to verify the results versus the reported values. Furthermore, as the hydrogen-to-carbon ratio decreases, the diffusivity of the oxygen ion increases. This increase in diffusivity is observed because the number of hydrogen atoms available as the reacting species increases in fuels with lower hydrogen-to-carbon ratios. The oxygen ion conductivity and output power density decrease as the hydrogen-to-carbon ratio of the fuels decreases. The reason behind this impact is the formation of a gas-induced ion barrier at the electrode–electrolyte interface by the CO_2 molecules formed during the reaction at the interface, thus blocking the flow of oxygen ions. As the oxygen ions become blocked, the output current contribution from the reaction also decreases and thereby affects the overall performance of the micro-solid oxide fuel cell. The experimental verification confirms this because of a significant decrease in the output power density. Furthermore, as per the application in portable devices, the appropriate choice of fuel can be chosen so that the micro-solid oxide fuel cell operates at the maximum power density.

Conceptual design and cost estimation of dry cask storage facility for spent fuel

International Nuclear Information System (INIS)

Maki, Yasuro; Hironaga, Michihiko; Kitano, Koichi; Shidahara, Isao; Shiomi, Satoshi; Ohnuma, Hiroshi; Saegusa, Toshiari

1985-01-01

In order to propose an optimum storage method of spent fuel, studies on the technical and economical evaluation of various storage methods have been carried out. This report is one of the results of the study and deals with storage facility of dry cask storage. The basic condition of this work conforms to ''Basic Condition for Spent Fuel Storage'' prepared by Project Group of Spent Fuel Dry Storage at July 1984. Concerning the structural system of cask storage facilities, trench structure system and concrete silo system are selected for storage at reactor (AR), and a reinforced concrete structure of simple design and a structure with membrance roof are selected for away from reactor (AFR) storage. The basic thinking of this selection are (1) cask is put charge of safety against to radioactivity and (2) storage facility is simplified. Conceptual designs are made for the selected storage facilities according to the basic condition. Attached facilities of storage yard structure (these are cask handling facility, cask supervising facility, cask maintenance facility, radioactivity control facility, damaged fuel inspection and repack facility, waste management facility) are also designed. Cost estimation of cask storage facility are made on the basis of the conceptual design. (author)

Universal electrode interface for electrocatalytic oxidation of liquid fuels.

Science.gov (United States)

Liao, Hualing; Qiu, Zhipeng; Wan, Qijin; Wang, Zhijie; Liu, Yi; Yang, Nianjun

2014-10-22

Electrocatalytic oxidations of liquid fuels from alcohols, carboxylic acids, and aldehydes were realized on a universal electrode interface. Such an interface was fabricated using carbon nanotubes (CNTs) as the catalyst support and palladium nanoparticles (Pd NPs) as the electrocatalysts. The Pd NPs/CNTs nanocomposite was synthesized using the ethylene glycol reduction method. It was characterized using transmission electron microscopy, energy dispersive X-ray spectroscopy, X-ray diffraction, voltammetry, and impedance. On the Pd NPs/CNTs nanocomposite coated electrode, the oxidations of those liquid fuels occur similarly in two steps: the oxidations of freshly chemisorbed species in the forward (positive-potential) scan and then, in the reverse scan (negative-potential), the oxidations of the incompletely oxidized carbonaceous species formed during the forward scan. The oxidation charges were adopted to study their oxidation mechanisms and oxidation efficiencies. The oxidation efficiency follows the order of aldehyde (formaldehyde) > carboxylic acid (formic acid) > alcohols (ethanol > methanol > glycol > propanol). Such a Pd NPs/CNTs nanocomposite coated electrode is thus promising to be applied as the anode for the facilitation of direct fuel cells.

Method of manufacturing gadolinium oxide-incorporated nuclear fuel sintering products

International Nuclear Information System (INIS)

Komono, Akira; Seki, Makoto; Omori, Sadayuki.

1987-01-01

Purpose: To manufacture nuclear fuel sintering products excellent in burning property and mechanical property. Constitution: In the manufacturing step for nuclear fuel sintering products, specific metal oxides are added for promoting the growth of crystal grains in the sintering. Those metal oxides melted at a temperature lower than the sintering temperature of a mixture of nuclear fuel oxide powder and oxide power, or those metal oxides causing eutectic reaction are used as the metal oxide. Particularly, those compounds having oxygen atom - metal atom ratio (O/M) of not less than 2 are preferably used. As such metal oxides usable herein transition metal oxides, e.g., Nb 2 O 5 , TiO 2 , MoO 3 and WO 3 are preferred, with Nb 2 O 3 and TiO 2 being preferred particularly. (Seki, T.)

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

Comparison for thorium fuel cycle facilities of two different capacities for implementation of safeguards

International Nuclear Information System (INIS)

Gangotra, Suresh; Grover, R.B.; Ramakumar, K.L.

2013-01-01

Highlights: • Facilities for implementation of safeguards for thorium fuel cycle have been compared. • Two concepts have been compared. • In one concept, the facilities are designed in hub and spoke concept. • In second concept the facilities are designed as self-contained concept. • The comparison is done on a number of factors, which affect safeguardability and proliferation resistance. -- Abstract: Thorium based nuclear fuel cycle has many attractive features, its inherent proliferation resistance being one of them. This is due to the presence of high energy gamma emitting daughter products of U 232 associated with U 233 . This high energy gamma radiation also poses challenges in nuclear material accounting. A typical thorium fuel cycle facility has a number of plants including a fuel fabrication plant for initial and equilibrium core, a reprocessed U 233 fuel fabrication plant, a reprocessing plant, a fuel assembly/disassembly plant and associated waste handling and management plants. A thorium fuel cycle facility can be set up to serve reactors at a site. Alternatively, one can follow a hub and spoke approach with a large thorium fuel cycle facility acting as a hub, catering to the requirements of reactors at several sites as spokes. These two concepts have their respective merits and shortcomings in terms of engineering and economics. The present paper is aimed at comparing the merits and challenges for implementation of safeguards on the two concepts viz. a large fuel cycle hub catering to reactors at several sites versus a small fuel cycle facility dedicated to reactors at a single site

Comparison for thorium fuel cycle facilities of two different capacities for implementation of safeguards

Energy Technology Data Exchange (ETDEWEB)

Gangotra, Suresh, E-mail: sgangotra@yahoo.co.in; Grover, R.B.; Ramakumar, K.L.

2013-09-15

Highlights: • Facilities for implementation of safeguards for thorium fuel cycle have been compared. • Two concepts have been compared. • In one concept, the facilities are designed in hub and spoke concept. • In second concept the facilities are designed as self-contained concept. • The comparison is done on a number of factors, which affect safeguardability and proliferation resistance. -- Abstract: Thorium based nuclear fuel cycle has many attractive features, its inherent proliferation resistance being one of them. This is due to the presence of high energy gamma emitting daughter products of U{sup 232} associated with U{sup 233}. This high energy gamma radiation also poses challenges in nuclear material accounting. A typical thorium fuel cycle facility has a number of plants including a fuel fabrication plant for initial and equilibrium core, a reprocessed U{sup 233} fuel fabrication plant, a reprocessing plant, a fuel assembly/disassembly plant and associated waste handling and management plants. A thorium fuel cycle facility can be set up to serve reactors at a site. Alternatively, one can follow a hub and spoke approach with a large thorium fuel cycle facility acting as a hub, catering to the requirements of reactors at several sites as spokes. These two concepts have their respective merits and shortcomings in terms of engineering and economics. The present paper is aimed at comparing the merits and challenges for implementation of safeguards on the two concepts viz. a large fuel cycle hub catering to reactors at several sites versus a small fuel cycle facility dedicated to reactors at a single site.

Durability of spent nuclear fuels and facility components in wet storage

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-04-01

Wet storage continues to be the dominant option for the management of irradiated fuel elements and assemblies (fuel units). Fuel types addressed in this study include those used in: power reactors, research and test reactors, and defence reactors. Important decisions must be made regarding acceptable storage modes for a broad variety of fuel types, involving numerous combinations of fuel and cladding materials. A broadly based materials database has the following important functions: to facilitate solutions to immediate and pressing materials problems; to facilitate decisions on the most effective long term interim storage methods for numerous fuel types; to maintain and update a basis on which to extend the licenses of storage facilities as regulatory periods expire; to facilitate cost-effective transfer of numerous fuel types to final disposal. Because examinations of radioactive materials are expensive, access to materials data and experience that provide an informed basis to analyse and extrapolate materials behaviour in wet storage environments can facilitate identification of cost-effective approaches to develop and maintain a valuable materials database. Fuel storage options include: leaving the fuel in wet storage, placing the fuel in canisters with cover gases, stored underwater, or transferring the fuel to one of several dry storage modes, involving a range of conditioning options. It is also important to anticipate the condition of the various materials as periods of wet storage are extended or as decisions to transfer to dry storage are implemented. A sound basis for extrapolation is needed to assess fuel and facility component integrity over the expected period of wet storage. A materials database also facilitates assessment of the current condition of specific fuel and facility materials, with minimal investments in direct examinations. This report provides quantitative and semi-quantitative data on materials behaviour or references sources of data to

Durability of spent nuclear fuels and facility components in wet storage

International Nuclear Information System (INIS)

1998-04-01

Wet storage continues to be the dominant option for the management of irradiated fuel elements and assemblies (fuel units). Fuel types addressed in this study include those used in: power reactors, research and test reactors, and defence reactors. Important decisions must be made regarding acceptable storage modes for a broad variety of fuel types, involving numerous combinations of fuel and cladding materials. A broadly based materials database has the following important functions: to facilitate solutions to immediate and pressing materials problems; to facilitate decisions on the most effective long term interim storage methods for numerous fuel types; to maintain and update a basis on which to extend the licenses of storage facilities as regulatory periods expire; to facilitate cost-effective transfer of numerous fuel types to final disposal. Because examinations of radioactive materials are expensive, access to materials data and experience that provide an informed basis to analyse and extrapolate materials behaviour in wet storage environments can facilitate identification of cost-effective approaches to develop and maintain a valuable materials database. Fuel storage options include: leaving the fuel in wet storage, placing the fuel in canisters with cover gases, stored underwater, or transferring the fuel to one of several dry storage modes, involving a range of conditioning options. It is also important to anticipate the condition of the various materials as periods of wet storage are extended or as decisions to transfer to dry storage are implemented. A sound basis for extrapolation is needed to assess fuel and facility component integrity over the expected period of wet storage. A materials database also facilitates assessment of the current condition of specific fuel and facility materials, with minimal investments in direct examinations. This report provides quantitative and semi-quantitative data on materials behaviour or references sources of data to

Seismic design considerations for nuclear fuel cycle facilities

International Nuclear Information System (INIS)

Soni, R.S.; Kushwaha, H.S.; Venkat Raj, V.

2001-01-01

During the last few decades, there have been considerable advances in the field of a seismic design of nuclear structures and components housed inside a Nuclear power Plant (NPP). The seismic design and qualification of theses systems and components are carried out through the use of well proven and established theoretical as well as experimental means. Many of the related research works pertaining to these methods are available in the published literature, codes, guides etc. Contrary to this, there is very little information available with regards to the seismic design aspects of the nuclear fuel cycle facilities. This is probably on account of the little importance attached to these facilities from the point of view of seismic loading. In reality, some of these facilities handle a large inventory of radioactive materials and, therefore, these facilities must survive during a seismic event without giving rise to any sort of undue radiological risk to the plant personnel and the public at large. Presented herein in this paper are the seismic design considerations which are adopted for the design of nuclear fuel cycle facilities in India. (author)

Fuel component of electricity generation cost for the BN-800 reactor with MOX fuel and uranium oxide fuel with increasing of fuel burnup and removing of radial breeding blanket

International Nuclear Information System (INIS)

Raskach, A.

2001-01-01

Nowadays there are two completed design concepts of Nuclear Power Plants (NPPs) with the BN-800 type reactors developed with due regard for advanced safety requirements. One of them is the design of the fourth unit of the Beloyarsk Nuclear Power Plant; the other one is the design of three units of the South Ural Nuclear Power Plant. The both concepts are to use mixed oxide fuel (MOX fuel) based on civil plutonium. Studies on any project include economical analyses and cost of fuel is an essential parameter. In the course of the design works on the both projects such evaluations were done. For BN-800 on the Beloyarsk site nuclear fuel costs were taken from actual expenses of the BN-600 reactor and converted to rated thermal power and design capacity factor of the BN-800 and then increased by 20% in connection with turning to MOX fuel. Then this methodology was rewarding, but the ratio of uranium fuel and MOX fuel costs might change for the last years. For the project of three units of the South Ural Nuclear Power Plant nuclear fuel expenses were calculated from the data on a MOX fuel fabrication production facility (Complex-300). However, investigations performed recently shown that the methodology of economical assessments should be revised, as well as design and technology of MOX fuel fabrication at Complex-300 should be revised to meet all the existing safety requirements. Excepting there is a great bulk of civil plutonium to be reproduced, now we came up against the problem to utilize the exceeding ex-weapons plutonium that obviously can be used for MOX fuel fabrication as well. Construction of the MOX fuel fabrication facility - Complex-300 - was started in 1983. Its design output was planned to provide simultaneously 4 fast reactors of the BN-800 type with MOX fuel. By now about 50% of construction works (taking into account auxiliary buildings and arrangements) and 20% of installation works have been done at Complex-300. Along this, first works to construct

Environmental monitoring standardization of effluent from nuclear fuel cycle facilities in China

International Nuclear Information System (INIS)

Gao Mili

1993-01-01

China has established some environmental monitoring standards of effluent from nuclear fuel cycle facilities. Up to date 33 standards have been issued; 10 to be issued; 11 in drafting. These standards cover sampling, gross activities measurement, analytical methods and management rules and so on. They involve with almost all nuclear fuel cycle facilities and have formed a complete standards system. By the end of the century, we attempt to draft a series of analytical and determination standards in various environmental various medium, they include 36 radionuclides from nuclear fuel cycle facilities. (3 tabs.)

Solid Oxide Fuel Cell

DEFF Research Database (Denmark)

2010-01-01

The solid oxide fuel cell comprising a metallic support material, an active anode layer consisting of a good hydrocarbon cracking catalyst, an electrolyte layer, an active cathode layer, and a transition layer consisting of preferably a mixture of LSM and a ferrite to the cathode current collector...

Development of Demonstration Facility Design Technology for Advanced Nuclear Fuel Cycle Process

International Nuclear Information System (INIS)

Cho, Il Je; You, G. S.; Choung, W. M.

2010-04-01

The main objective of this R and D is to develop the PRIDE (PyRoprocess Integrated inactive DEmonstration) facility for engineering-scale inactive test using fresh uranium, and to establish the design requirements of the ESPF (Engineering Scale Pyroprocess Facility) for active demonstration of the pyroprocess. Pyroprocess technology, which is applicable to GEN-IV systems as one of the fuel cycle options, is a solution of the spent fuel accumulation problems. PRIDE Facility, pyroprocess mock-up facility, is the first facility that is operated in inert atmosphere in the country. By using the facility, the functional requirements and validity of pyroprocess technology and facility related to the advanced fuel cycle can be verified with a low cost. Then, PRIDE will contribute to evaluate the technology viability, proliferation resistance and possibility of commercialization of the pyroprocess technology. The PRIDE evaluation data, such as performance evaluation data of equipment and operation experiences, will be directly utilized for the design of ESPF

Studying international fuel cycle robustness with the GENIUSv2 discrete facilities/materials fuel cycle systems analysis tool

Energy Technology Data Exchange (ETDEWEB)

Wilson, P.H. [Dept. of Engineering Physics, University of Wisconsin-Madison (United States)

2009-06-15

GENIUSv2 (Global Evaluation of Nuclear Infrastructure Utilization Scenarios, hereafter 'GENIUS') is a discrete-facilities/materials nuclear fuel cycle systems analysis tool currently under development at the University of Wisconsin-Madison. For a given scenario, it models nuclear fuel cycle facilities (reactors, fuel fabrication, enrichment, etc.), the institutions that own them (utilities and governments), and the regions in which those institutions operate (sub-national, national, and super-national entities). Facilities work together to provide each other with the materials they need. The results of each simulation include the electricity production in each region as well as operational histories of each facility and isotopic and facility histories of each material object. GENIUS users specify an initial condition and a facility deployment plan. The former describes each region and institution in the scenario as well as facilities that exist at the start. The latter specifies all the facilities that will be built over the course of the simulation (and by which institutions). Each region, institution, and facility can be assigned financial parameters such as tax and interest rates, and facilities also get assigned technical information about how they actually operate. Much of the power of the data model comes from the flexibility to model individual entities to a fine level of detail or to allow them to inherit region-, institution-, or facility-type-specific default parameters. Most importantly to the evaluation of regional, national, and international policies, users can also specify rules that define the affinity (or lack thereof) for trade of particular commodities between particular entities. For instance, these rules could dictate that a particular region or institution always buy a certain commodity (ore, enriched UF{sub 6}, fabricated fuel, etc.) from a particular region or institution, never buy from that region, or merely have a certain

Facile synthesis of porous Pt-Pd nanospheres supported on reduced graphene oxide nanosheets for enhanced methanol electrooxidation

Science.gov (United States)

Li, Shan-Shan; Lv, Jing-Jing; Hu, Yuan-Yuan; Zheng, Jie-Ning; Chen, Jian-Rong; Wang, Ai-Jun; Feng, Jiu-Ju

2014-02-01

In this study, a simple, facile, and effective wet-chemical strategy was developed in the synthesis of uniform porous Pt-Pd nanospheres (Pt-Pd NSs) supported on reduced graphene oxide nanosheets (RGOs) under ambient temperature, where octylphenoxypolye thoxyethanol (NP-40) is used as a soft template, without any seed, organic solvent or special instruments. The as-prepared nanocomposites display enhanced electrocatalytic activity and good stability toward methanol oxidation, compared with commercial Pd/C and Pt/C catalysts. This strategy may open a new route to design and prepare advanced electrocatalysts for fuel cells.

MELCOR modeling of the PBF [Power Burst Facility] Severe Fuel Damage Test 1-4

International Nuclear Information System (INIS)

Madni, I.K.

1990-01-01

This paper describes a MELCOR Version 1.8 simulation of the Power Burst Facility (PBF) Severe Fuel Damage (SFD) Test 1--4. The input data for the analysis were obtained from the Test Results Report and from SCDAP/RELAP5 input. Results are presented for the transient liquid level in the test bundle, clad temperatures, shroud temperatures, clad oxidation and hydrogen generation, bundle geometry changes, fission product release, and heat transfer to the bypass flow. Comparisons are made with experimental data and with SCDAP/RELAP5 calculations. 10 refs., 7 figs

Redox Stable Anodes for Solid Oxide Fuel Cells

Directory of Open Access Journals (Sweden)

Guoliang eXiao

2014-06-01

Full Text Available Solid oxide fuel cells (SOFCs can convert chemical energy from the fuel directly to electrical energy with high efficiency and fuel flexibility. Ni-based cermets have been the most widely adopted anode for SOFCs. However, the conventional Ni-based anode has low tolerance to sulfur-contamination, is vulnerable to deactivation by carbon build-up (coking from direct oxidation of hydrocarbon fuels, and suffers volume instability upon redox cycling. Among these limitations, the redox instability of the anode is particularly important and has been intensively studied since the SOFC anode may experience redox cycling during fuel cell operations even with the ideal pure hydrogen as the fuel. This review aims to highlight recent progresses on improving redox stability of the conventional Ni-based anode through microstructure optimization and exploration of alternative ceramic-based anode materials.

Current developments of fuel fabrication technologies at the plutonium fuel production facility, PFPF

International Nuclear Information System (INIS)

Asakura, K.; Aono, S.; Yamaguchi, T.; Deguchi, M.

2000-01-01

The Japan Nuclear Cycle Development Institute, JNC, designed, constructed and has operated the Plutonium Fuel Production Facility, PFPF, at the JNC Tokai Works to supply MOX fuels to the proto-type Fast Breeder Reactor, FBR, 'MONJU' and the experimental FBR 'JOYO' with 5 tonMOX/year of fabrication capability. Reduction of personal radiation exposure to a large amount of plutonium is one of the most important subjects in the development of MOX fabrication facility on a large scale. As the solution of this issue, the PFPF has introduced automated and/or remote controlled equipment in conjunction with computer controlled operation scheme. The PFPF started its operation in 1988 with JOYO reload fuel fabrication and has demonstrated MOX fuel fabrication on a large scale through JOYO and MONJU fuel fabrication for this decade. Through these operations, it has become obvious that several numbers of equipment initially installed in the PFPF need improvements in their performance and maintenance for commercial utilization of plutonium in the future. Furthermore, fuel fabrication of low density MOX pellets adopted in the MONJU fuel required a complete inspection because of difficulties in pellet fabrication compared with high density pellet for JOYO. This paper describes new pressing equipment with a powder recovery system, and pellet finishing and inspection equipment which has multiple functions, such as grinding measurements of outer diameter and density, and inspection of appearance to improve efficiency in the pellet finishing and inspection steps. Another development of technology concerning an annular pellet and an innovative process for MOX fuel fabrication are also described in this paper. (author)

performance calculations of gadolinium oxide and boron nitride coated fuel

International Nuclear Information System (INIS)

Tanker, E.; Uslu, I.; Disbudak, H.; Guenduez, G.

1997-01-01

A comparative study was performed on the behaviour of natural uranium dioxide-gadolinium oxide mixture fuel and boron nitride coated low enriched fuel in a pressurized water reactor. A fuel element containing one burnable poison fuel pins was modeled with the computer code WIMS, and burn-up dependent critically, fissile isotope inventory and two dimensional power distribution were obtained. Calculations were performed for burnable poison fuels containing 5% and 10% gadolinium oxide and for those coated with 1μ,5μ and 10μ of boron nitride. Boron nitride coating was found superior to gadolinium oxide on account of its smoother criticality curve, lower power peaks and insignificant change in fissile isotope content

Electrode Design for Low Temperature Direct-Hydrocarbon Solid Oxide Fuel Cells

Science.gov (United States)

Chen, Fanglin (Inventor); Zhao, Fei (Inventor); Liu, Qiang (Inventor)

2015-01-01

In certain embodiments of the present disclosure, a solid oxide fuel cell is described. The solid oxide fuel cell includes a hierarchically porous cathode support having an impregnated cobaltite cathode deposited thereon, an electrolyte, and an anode support. The anode support includes hydrocarbon oxidation catalyst deposited thereon, wherein the cathode support, electrolyte, and anode support are joined together and wherein the solid oxide fuel cell operates a temperature of 600.degree. C. or less.

Electrode design for low temperature direct-hydrocarbon solid oxide fuel cells

Science.gov (United States)

Chen, Fanglin; Zhao, Fei; Liu, Qiang

2015-10-06

In certain embodiments of the present disclosure, a solid oxide fuel cell is described. The solid oxide fuel cell includes a hierarchically porous cathode support having an impregnated cobaltite cathode deposited thereon, an electrolyte, and an anode support. The anode support includes hydrocarbon oxidation catalyst deposited thereon, wherein the cathode support, electrolyte, and anode support are joined together and wherein the solid oxide fuel cell operates a temperature of 600.degree. C. or less.

Development of advanced mixed oxide fuels for plutonium management

International Nuclear Information System (INIS)

Eaton, S.; Beard, C.; Buksa, J.; Butt, D.; Chidester, K.; Havrilla, G.; Ramsey, K.

1997-01-01

A number of advanced Mixed Oxide (MOX) fuel forms are currently being investigated at Los Alamos National Laboratory that have the potential to be effective plutonium management tools. Evolutionary Mixed Oxide (EMOX) fuel is a slight perturbation on standard MOX fuel, but achieves greater plutonium destruction rates by employing a fractional nonfertile component. A pure nonfertile fuel is also being studied. Initial calculations show that the fuel can be utilized in existing light water reactors and tailored to address different plutonium management goals (i.e., stabilization or reduction of plutonium inventories residing in spent nuclear fuel). In parallel, experiments are being performed to determine the feasibility of fabrication of such fuels. Initial EMOX pellets have successfully been fabricated using weapons-grade plutonium. (author)

Development of advanced mixed oxide fuels for plutonium management

International Nuclear Information System (INIS)

Eaton, S.; Beard, C.; Buksa, J.; Butt, D.; Chidester, K.; Havrilla, G.; Ramsey, K.

1997-06-01

A number of advanced Mixed Oxide (MOX) fuel forms are currently being investigated at Los Alamos National Laboratory that have the potential to be effective plutonium management tools. Evolutionary Mixed Oxide (EMOX) fuel is a slight perturbation on standard MOX fuel, but achieves greater plutonium destruction rates by employing a fractional nonfertile component. A pure nonfertile fuel is also being studied. Initial calculations show that the fuel can be utilized in existing light water reactors and tailored to address different plutonium management goals (i.e., stabilization or reduction of plutonium inventories residing in spent nuclear fuel). In parallel, experiments are being performed to determine the feasibility of fabrication of such fuels. Initial EMOX pellets have successfully been fabricated using weapons-grade plutonium

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

Conceptual design report for the away from reactor spent fuel storage facility, Savannah River Plant

International Nuclear Information System (INIS)

1978-12-01

The Department of Energy (DOE) requested that Du Pont prepare a conceptual design and appraisal of cost for Federal budget planning for an away from reactor spent fuel storage facility that could be ready to store fuel by December 1982. This report describes the basis of the appraisal of cost in the amount of $270,000,000 for all facilities. The proposed action is to provide a facility at the Savannah River Plant. The facility will have an initial storage capacity of 5000 metric tons of spent fuel and will be capable of receiving 1000 metric tons per year. The spent fuel will be stored in water-filled concrete basins that are lined with stainless steel. The modular construction of the facility will allow future expansion of the storage basins and auxiliary services in a cost-effective manner. The facility will be designed to receive, handle, decontaminate and reship spent fuel casks; to remove irradiated fuel from casks; to place the fuel in a storage basin; and to cool and control the quality of the water. The facility will also be designed to remove spent fuel from storage basins, load the spent fuel into shipping casks, decontaminated loaded casks and ship spent fuel. The facility requires a license by the Nuclear Regulatory Commission (NRC). Features of the design, construction and operations that may affect the health and safety of the workforce and the public will conform with NRC requirements. The facility would be ready to store fuel by January 1983, based on normal Du Pont design and construction practices for DOE. The schedule does not include the effect of licensing by the NRC. To maintain this option, preparation of the documents and investigation of a site at the Savannah River Plant, as required for licensing, were started in FY '78

Licensed fuel facility status report. Inventory difference data, January-June 1982

International Nuclear Information System (INIS)

1983-02-01

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

A new framework to assess risk for a spent fuel dry storage facility

International Nuclear Information System (INIS)

Ryu, J. H.; Jae, M. S.; Jung, C. W.

2004-01-01

A spent fuel dry storage facility is a dry cooling storage facility for storing irradiated nuclear fuel and associated radioactive materials. It has very small possibilities to release radiation materials. It means a safety analysis for a spent fuel dry storage facility is required before construction. In this study, a new framework for assessing risk associated with a spent fuel dry storage facility is represented. A safety assessment framework includes 3 modules such as assessment of basket/cylinder failure rates, that of overall storage system, and site modeling. A reliability physics model for failure rates, event tree analysis(ETA)/fault tree analysis for system analysis, Bayesian analysis for initial events data, and MACCS code for consequence analysis have been used in this study

Safety issues in urban transit facilities for hydrogen-fueled buses

International Nuclear Information System (INIS)

Hay, R.H.; Ducharme, P.

2004-01-01

'Full text:' The Canadian Transportation Fuel Cell Alliance (CTFCA), created by the Canadian Government as part of its 2000 Climate Change Action Plan, has commissioned MARCON-DDM's Hydrogen Intervention Team (HIT) to provide a roadmap for urban transit systems that wish to move to hydrogen fuel cell-powered bus fleets. HIT is currently in the process of gathering information from hydrogen technology providers, bus manufacturers, fuelling system providers and urban transit systems in Canada, the US and Europe. In September, HIT will be in a position to provide a hands-on perspective of the introduction of fuel-cell buses in the Canadian environment. Part of the process of adding hydrogen-fueled busses to urban transit systems involves phasing in the new technology to minimize the economic cost. This involves substituting hydrogen buses into the normal bus procurement life cycle and maximizing the use of existing facilities for garaging, maintenance and fueling. Using a schematic outline of an urban transit system, this presentation will outline the safety issues specific to hydrogen in such systems, particularly for garaging, maintenance and fueling components. It will then outline how safety of these component is addressed in current and proposed codes, standards and recommended practices. Based on these requirements the impact of the introduction of hydrogen-fueled buses on each component of the transit system will be addressed in terms of the adaptations of current facilities and practices or the requirements for new facilities and practices. (author)

Radiological and environmental surveillance in front-end fuel cycle facilities

International Nuclear Information System (INIS)

Khan, A.H.; Sahoo, S.K.; Tripathi, R.M.

2004-01-01

This paper describes the occupational and environmental radiological safety measures associated with the operations of front end nuclear fuel cycle. Radiological monitoring in the facilities is important to ensure safe working environment, protection of workers against exposure to radiation and comply with regulatory limits of exposure. The radiation exposure of workers in different units of the front end nuclear fuels cycle facilities operated by IREL, UCIL and NFC and environmental monitoring results are summarised

POST CRITICAL HEAT TRANSFER AND FUEL CLADDING OXIDATION

Directory of Open Access Journals (Sweden)

Vojtěch Caha

2016-12-01

Full Text Available The knowledge of heat transfer coefficient in the post critical heat flux region in nuclear reactor safety is very important. Although the nuclear reactors normally operate at conditions where critical heat flux (CHF is not reached, accidents where dryout occur are possible. Most serious postulated accidents are a loss of coolant accident or reactivity initiated accident which can lead to CHF or post CHF conditions and possible disruption of core integrity. Moreover, this is also influenced by an oxide layer on the cladding surface. The paper deals with the study of mathematical models and correlations used for heat transfer calculation, especially in post dryout region, and fuel cladding oxidation kinetics of currently operated nuclear reactors. The study is focused on increasing of accuracy and reliability of safety limit calculations (e.g. DNBR or fuel cladding temperature. The paper presents coupled code which was developed for the solution of forced convection flow in heated channel and oxidation of fuel cladding. The code is capable of calculating temperature distribution in the coolant, cladding and fuel and also the thickness of an oxide layer.

Upgraded Features of Newly Constructed Fuel Assembly Mechanical Characterization Test Facility in KAERI

Energy Technology Data Exchange (ETDEWEB)

Lee, Kang Hee; Kang, Heung Seok; Yoon, Kyung Ho; Kim, Hyung Kyu; Lee, Young Ho; Kim, Soo Ho; Yang, Jae Ho [KAERI, Daejeon (Korea, Republic of)

2016-05-15

Fuel assembly mechanical characterization test facility (FAMeCT) in KAERI is newly constructed with upgraded functional features such as increased loading capacity, under-water vibration testing and severe earthquake simulation for extended fuel design guideline. The facility building is compactly designed in the scale of 3rd floor building and has regions for assembly-wise mechanical test equipment, dynamic load (seismic) simulating test system, small scale hydraulic loop and component wise test equipment. Figure 1 shows schematic regional layout of the facility building. Mechanical test platform and system is designed to increase loading capacity for axial compression test. Structural stability of the support system of new upper core plate simulator is validated through a limit case functional test. Fuel assembly mechanical characterization test facility in KAERI is newly constructed and upgraded with advanced functional features such as uprated loading capacity, under-water vibration testing and severe earthquake simulation for extended fuel design guideline. This paper briefly introduce the test facility construction and scope of the facility and is focused on the upgraded design features of the facility. Authors hope to facilitate the facility more in the future and collaborate with the industry.

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

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.

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

Conceptual design study of a concrete canister spent-fuel storage facility

International Nuclear Information System (INIS)

Lidfors, E.D.; Tabe, T.; Johnson, H.M.

1979-01-01

This report presents a conceptual design study for the interim storage of CANDU spent fuel in concrete canisters. The canisters will be concrete flasks, which contain fuel prepackaged in double steel containment, and will be cooled by natural air convection. This is one of the methods proposed as a potential alternative to water pool storage. A preliminary study of this concept was done by CAFS (Committee Assessing Fuel Storage), and WNRE (Whiteshell Nuclear Research Establishment) is currently conducting a development and demonstration program. This study of a central facility for the storage of all Canadian spent fuel arisings to the year 2000 was completed in 1975. A brief description of the facilities required and the operations involved, a summary of costs, a survey of the monitoring requirements and a prediction of the personnel exposures associated with this method of storing spent fuel are reported here. The estimated total cost of interim storage in cylindrical canisters at a central site is $6.02/kg U (1975 dollars). Approximately half of this cost is incurred in the shipment of fuel from the reactors to the storage facility. (author)

Fuel-sodium reaction product formation in breached mixed-oxide fuel

International Nuclear Information System (INIS)

Bottcher, J.H.; Lambert, J.D.B.; Strain, R.V.; Ukai, S.; Shibahara, S.

1988-01-01

The run-beyond-cladding-breach (RBCB) operation of mixed-oxide LMR fuel pins has been studied for six years in the Experimental Breeder Reactor-II (EBR-II) as part of a joint program between the US Department of Energy and the Power Reactor and Nuclear Fuel Development Corporation of Japan. The formation of fuel-sodium reaction product (FSRP), Na 3 MO 4 , where M = U/sub 1-y/Pu/sub y/, in the outer fuel regions is the major phenomenon governing RBCB behavior. It increases fuel volume, decreases fuel stoichiometry, modifies fission-product distributions, and alters thermal performance of a pin. This paper describes the morphology of Na 3 MO 4 observed in 5.84-mm diameter pins covering a variety of conditions and RBCB times up to 150 EFPD's. 8 refs., 1 fig

Description of the PIE facility for research reactors irradiated fuels in CNEA

International Nuclear Information System (INIS)

Bisca, A.; Coronel, R.; Homberger, V.; Quinteros, A.; Ratner, M.

2002-01-01

The PIE Facility (LAPEP), located at the Ezeiza Atomic Center (CAE), was designed to carry out destructive and non-destructive post-irradiation examinations (PIE) on research and power reactor spent fuels, reactor internals and other irradiated materials, and to perform studies related with: Station lifetime extension; Fuel performance; Development of new fuels; and Failures and determination of their causes. LAPEP is a relevant facility where research and development can be carried out. It is worth mentioning that in this facility the PIE corresponding to the Surveillance Program for the Atucha I Nuclear Power Plant (CNA-1) were successfully performed. Materials testing during the CNA-1 repair and the study of failures in fuel element plugs of the Embalse Nuclear Power Plant (CNE) were also performed. (author)

Transient response and radiation dose estimates for breaches to a spent fuel processing facility

Energy Technology Data Exchange (ETDEWEB)

Solbrig, Charles W., E-mail: soltechco@aol.com; Pope, Chad; Andrus, Jason

2014-08-15

Highlights: • We model doses received from a nuclear fuel facility from boundary leaks due to an earthquake. • The supplemental exhaust system (SES) starts after breach causing air to be sucked into the cell. • Exposed metal fuel burns increasing pressure and release of radioactive contamination. • Facility releases are small and much less than the limits showing costly refits are unnecessary. • The method presented can be used in other nuclear fuel processing facilities. - Abstract: This paper describes the analysis of the design basis accident for Idaho National Laboratory Fuel Conditioning Facility (FCF). The facility is used to process spent metallic nuclear fuel. This analysis involves a model of the transient behavior of the FCF inert atmosphere hot cell following an earthquake initiated breach of pipes passing through the cell boundary. Such breaches allow the introduction of air and subsequent burning of pyrophoric metals. The model predicts the pressure, temperature, volumetric releases, cell heat transfer, metal fuel combustion, heat generation rates, radiological releases and other quantities. The results show that releases from the cell are minimal and satisfactory for safety. This analysis method should be useful in other facilities that have potential for damage from an earthquake and could eliminate the need to back fit facilities with earthquake proof boundaries or lessen the cost of new facilities.

Transient response and radiation dose estimates for breaches to a spent fuel processing facility

International Nuclear Information System (INIS)

Solbrig, Charles W.; Pope, Chad; Andrus, Jason

2014-01-01

Highlights: • We model doses received from a nuclear fuel facility from boundary leaks due to an earthquake. • The supplemental exhaust system (SES) starts after breach causing air to be sucked into the cell. • Exposed metal fuel burns increasing pressure and release of radioactive contamination. • Facility releases are small and much less than the limits showing costly refits are unnecessary. • The method presented can be used in other nuclear fuel processing facilities. - Abstract: This paper describes the analysis of the design basis accident for Idaho National Laboratory Fuel Conditioning Facility (FCF). The facility is used to process spent metallic nuclear fuel. This analysis involves a model of the transient behavior of the FCF inert atmosphere hot cell following an earthquake initiated breach of pipes passing through the cell boundary. Such breaches allow the introduction of air and subsequent burning of pyrophoric metals. The model predicts the pressure, temperature, volumetric releases, cell heat transfer, metal fuel combustion, heat generation rates, radiological releases and other quantities. The results show that releases from the cell are minimal and satisfactory for safety. This analysis method should be useful in other facilities that have potential for damage from an earthquake and could eliminate the need to back fit facilities with earthquake proof boundaries or lessen the cost of new facilities

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

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

Durability of solid oxide fuel cells using sulfur containing fuels

DEFF Research Database (Denmark)

Hagen, Anke; Rasmussen, Jens Foldager Bregnballe; Thydén, Karl Tor Sune

2011-01-01

The usability of hydrogen and also carbon containing fuels is one of the important advantages of solid oxide fuel cells (SOFCs), which opens the possibility to use fuels derived from conventional sources such as natural gas and from renewable sources such as biogas. Impurities like sulfur compounds...... are critical in this respect. State-of-the-art Ni/YSZ SOFC anodes suffer from being rather sensitive towards sulfur impurities. In the current study, anode supported SOFCs with Ni/YSZ or Ni/ScYSZ anodes were exposed to H2S in the ppm range both for short periods of 24h and for a few hundred hours. In a fuel...

Addressing fuel recycling in solid oxide fuel cell systems fed by alternative fuels

DEFF Research Database (Denmark)

Rokni, Masoud

2017-01-01

An innovative study on anode recirculation in solid oxide fuel cell systems with alternative fuels is carried out and investigated. Alternative fuels under study are ammonia, pure hydrogen, methanol, ethanol, DME and biogas from biomass gasification. It is shown that the amount of anode off......%. Furthermore, it is founded that for the case with methanol, ethanol and DME then at high utilization factors, low anode recirculation is recommended while at low utilization factors, high anode recirculation is recommended. If the plant is fed by biogas from biomass gasification then for each utilization...

Identification of a Methane Oxidation Intermediate on Solid Oxide Fuel Cell Anode Surfaces with Fourier Transform Infrared Emission.

Science.gov (United States)

Pomfret, Michael B; Steinhurst, Daniel A; Owrutsky, Jeffrey C

2013-04-18

Fuel interactions on solid oxide fuel cell (SOFC) anodes are studied with in situ Fourier transform infrared emission spectroscopy (FTIRES). SOFCs are operated at 800 °C with CH4 as a representative hydrocarbon fuel. IR signatures of gas-phase oxidation products, CO2(g) and CO(g), are observed while cells are under load. A broad feature at 2295 cm(-1) is assigned to CO2 adsorbed on Ni as a CH4 oxidation intermediate during cell operation and while carbon deposits are electrochemically oxidized after CH4 operation. Electrochemical control provides confirmation of the assignment of adsorbed CO2. FTIRES has been demonstrated as a viable technique for the identification of fuel oxidation intermediates and products in working SOFCs, allowing for the elucidation of the mechanisms of fuel chemistry.

An Approach to Safeguards by Design (SBD) for Fuel Cycle Facilities

International Nuclear Information System (INIS)

Sankaran Nair, P.; Gangotra, S.; Karanam, R.

2015-01-01

Implementation of safeguards in bulk handling facilities such as fuel fabrication facilities and reprocessing facilities are a challenging task. This is attributed to the nuclear material present in the facility in the form of powder, pellet, green pellet, solution and gaseous. Additionally material hold up, material unaccounted for (MUF) and the operations carried out round the clock add to the difficulties in implementing safeguards. In facilities already designed or commissioned or operational, implementation of safeguards measures are relatively difficult. The authors have studied a number of measures which can be adopted at the design stage itself. Safeguard By Design (SBD) measures can help in more effective implementation of safeguards, reduction of cost and reduction in radiological dose to the installation personnel. The SBD measures in the power reactors are comparatively easier to implement than in the fuel fabrication plants, since reactors are item counting facilities while the fuel fabrication plants are bulk handling type of facilities and involves much rigorous nuclear material accounting methodology. The safeguards measures include technical measures like dynamic nuclear material accounting, near real time monitoring, remote monitoring, use of automation, facility imagery, Radio Frequency Identification (RFID) tagging, reduction of MUF in bulk handling facilities etc. These measures have been studied in the context of bulk handling facilities and presented in this paper. Incorporation of these measures at the design stage (SBD) is expected to improve the efficiency of safeguardability in such bulk handling and item counting facilities and proliferation resistance of nuclear material handled in such facilities. (author)

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.

Operation of the nuclear fuel cycle test facilities -Operation of the hot test loop facilities

International Nuclear Information System (INIS)

Chun, S. Y.; Jeong, M. K.; Park, C. K.; Yang, S. K.; Won, S. Y.; Song, C. H.; Jeon, H. K.; Jeong, H. J.; Cho, S.; Min, K. H.; Jeong, J. H.

1997-01-01

A performance and reliability of a advanced nuclear fuel and reactor newly designed should be verified by performing the thermal hydraulics tests. In thermal hydraulics research team, the thermal hydraulics tests associated with the development of an advanced nuclear fuel and reactor haven been carried out with the test facilities, such as the Hot Test Loop operated under high temperature and pressure conditions, Cold Test Loop, RCS Loop and B and C Loop. The objective of this project is to obtain the available experimental data and to develop the advanced measuring techniques through taking full advantage of the facilities. The facilities operated by the thermal hydraulics research team have been maintained and repaired in order to carry out the thermal hydraulics tests necessary for providing the available data. The performance tests for the double grid type bottom end piece which was improved on the debris filtering effectivity were performed using the PWR-Hot Test Loop. The CANDU-Hot Test Loop was operated to carry out the pressure drop tests and strength tests of CANFLEX fuel. The Cold Test Loop was used to obtain the local velocity data in subchannel within HANARO fuel bundle and to study a thermal mixing characteristic of PWR fuel bundle. RCS thermal hydraulic loop was constructed and the experiments have been carried out to measure the critical heat flux. In B and C Loop, the performance tests for each component were carried out. (author). 19 tabs., 78 figs., 19 refs

Operation of the nuclear fuel cycle test facilities -Operation of the hot test loop facilities

Energy Technology Data Exchange (ETDEWEB)

Chun, S. Y.; Jeong, M. K.; Park, C. K.; Yang, S. K.; Won, S. Y.; Song, C. H.; Jeon, H. K.; Jeong, H. J.; Cho, S.; Min, K. H.; Jeong, J. H.

1997-01-01

A performance and reliability of a advanced nuclear fuel and reactor newly designed should be verified by performing the thermal hydraulics tests. In thermal hydraulics research team, the thermal hydraulics tests associated with the development of an advanced nuclear fuel and reactor haven been carried out with the test facilities, such as the Hot Test Loop operated under high temperature and pressure conditions, Cold Test Loop, RCS Loop and B and C Loop. The objective of this project is to obtain the available experimental data and to develop the advanced measuring techniques through taking full advantage of the facilities. The facilities operated by the thermal hydraulics research team have been maintained and repaired in order to carry out the thermal hydraulics tests necessary for providing the available data. The performance tests for the double grid type bottom end piece which was improved on the debris filtering effectivity were performed using the PWR-Hot Test Loop. The CANDU-Hot Test Loop was operated to carry out the pressure drop tests and strength tests of CANFLEX fuel. The Cold Test Loop was used to obtain the local velocity data in subchannel within HANARO fuel bundle and to study a thermal mixing characteristic of PWR fuel bundle. RCS thermal hydraulic loop was constructed and the experiments have been carried out to measure the critical heat flux. In B and C Loop, the performance tests for each component were carried out. (author). 19 tabs., 78 figs., 19 refs.

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.

Comparison of Core Performance with Various Oxide fuels on Sodium Cooled Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Choi, Jin Ha; Kim, Myung Hyun [Kyung Hee University, Yongin (Korea, Republic of)

2016-05-15

The system is called Prototype GenIV Sodium-cooled Fast Reactor (PGSFR). Ultimate goal of PGSFR is test for capability of TRU transmutation. Purpose of this study is test for evaluation of in-core performance and TRU transmutation performance by applying various oxide fuel loaded TRU. Fuel type of reference core is changed to uranium-based oxide fuel. Oxide fuel has a lot of experience through fuel fabrication and reactor operation. This study performed by compared and analyzed a core performance of various oxide fuels. (U,Pu)O{sub 2} and (U,TRU)O{sub 2} which various oxide fuel types are selected as extreme case for comparison with core performance and transmutation capability of TRU isotopes. Thorium-based fuel is known that it has good performance for burner reactor due to low proliferation characteristic. To check the performance of TRU incineration for comparison with uranium-based fuel on prototype SFR, Thorium-based fuel, (Th,U)O{sub 2}, (Th,Pu)O{sub 2} and (Th,TRU)O{sub 2}, is selected. Calculations of core performance for various oxide fuel are performed using the fast calculation tool, TRANSX / DANTSTS / REBUS-3. In this study, comparison of core performance and transmutation performance is conducted with various fuel types in a sodium-cooled fast reactor. Mixed oxide fuel with TRU can produce the energy with small amount of fissile material. However, the TRU fuel is confirmed to bring a potential decline of the safety parameters. In case of (Th,U)O2 fuel, the flux level in thermal neutron region becomes lower because of higher capture cross-section of Th-232 than U-238. However, Th-232 has difficulty in converting to TRU isotopes. Therefore, the TRU consumption mass is relatively high in mixed oxide fuel with thorium and TRU.

Comparison of Core Performance with Various Oxide fuels on Sodium Cooled Fast Reactor

International Nuclear Information System (INIS)

Choi, Jin Ha; Kim, Myung Hyun

2016-01-01

The system is called Prototype GenIV Sodium-cooled Fast Reactor (PGSFR). Ultimate goal of PGSFR is test for capability of TRU transmutation. Purpose of this study is test for evaluation of in-core performance and TRU transmutation performance by applying various oxide fuel loaded TRU. Fuel type of reference core is changed to uranium-based oxide fuel. Oxide fuel has a lot of experience through fuel fabrication and reactor operation. This study performed by compared and analyzed a core performance of various oxide fuels. (U,Pu)O_2 and (U,TRU)O_2 which various oxide fuel types are selected as extreme case for comparison with core performance and transmutation capability of TRU isotopes. Thorium-based fuel is known that it has good performance for burner reactor due to low proliferation characteristic. To check the performance of TRU incineration for comparison with uranium-based fuel on prototype SFR, Thorium-based fuel, (Th,U)O_2, (Th,Pu)O_2 and (Th,TRU)O_2, is selected. Calculations of core performance for various oxide fuel are performed using the fast calculation tool, TRANSX / DANTSTS / REBUS-3. In this study, comparison of core performance and transmutation performance is conducted with various fuel types in a sodium-cooled fast reactor. Mixed oxide fuel with TRU can produce the energy with small amount of fissile material. However, the TRU fuel is confirmed to bring a potential decline of the safety parameters. In case of (Th,U)O2 fuel, the flux level in thermal neutron region becomes lower because of higher capture cross-section of Th-232 than U-238. However, Th-232 has difficulty in converting to TRU isotopes. Therefore, the TRU consumption mass is relatively high in mixed oxide fuel with thorium and TRU.

Analysis of fuel cladding chemical interaction in mixed oxide fuel pins

International Nuclear Information System (INIS)

Weber, J.W.; Dutt, D.S.

1976-01-01

An analysis is presented of the observed interaction between mixed oxide 75 wt percent UO 2 --25 wt percent PuO 2 fuel and 316--20 percent CW stainless steel cladding in LMFBR type fuel pins irradiated in EBR-II. A description is given of the test pins and their operating conditions together with, metallographic observations and measurements of the fuel/cladding reaction, and a correlation equation is developed relating depth of cladding attack to temperature and burnup. Some recent data on cladding reaction in fuel pins with low initial O/M in the fuel are given and compared with the correlation equation curves

The neutronic and fuel cycle performance of interchangeable 3500 MWth metal and oxide fueled LMRs

International Nuclear Information System (INIS)

Fujita, E.K.; Wade, D.C.

1990-01-01

This study summarizes the neutronic and fuel cycle analysis performed at Argonne National Laboratory for an oxide and a metal fueled 3500 MWth LMR. These reactor designs formed the basis for a joint US/European study of LMR ATWS events. The oxide and metal core designs were developed to meet reactor performance specifications that are constrained by requirements for core loading interchangeability and for a small burnup reactivity swing. Differences in the computed performance parameters of the oxide and metal cores, arising from basic differences in their neutronic characteristics, are identified and discussed. It is shown that metal and oxide cores designed to the same ground rules exhibit many similar performance characteristics; however, they differ substantially in reactivity coefficients, control strategies, and fuel cycle options. 12 refs., 2 figs., 12 tabs

CSNI Technical Opinion Papers No. 15 - Ageing management of nuclear fuel cycle facilities

International Nuclear Information System (INIS)

Nocture, Pierre; Daubard, Jean-Paul; Lhomme, Veronique; Martineau, Dominique; Blundell, Neil; Conte, Dorothee; Dobson, Martin; Gmal, Bernhard; Hiltz, Thomas; Ueda, Yoshinori

2012-01-01

Managing the ageing of fuel cycle facilities (FCFs) means, as for other nuclear installations, ensuring the availability of required safety functions throughout their service life while taking into account the changes that occur with time and use. This technical opinion paper identifies a set of good practices by benchmarking strategies and good practices on coping with physical ageing and obsolescence from the facility design stage until decommissioning. It should be of particular interest to nuclear safety regulators, fuel cycle facilities operators and fuel cycle researchers [fr

Facility Operations 1993 fiscal year work plan: WBS 1.3.1

Energy Technology Data Exchange (ETDEWEB)

1992-11-01

The Facility Operations program is responsible for the safe, secure, and environmentally sound management of several former defense nuclear production facilities, and for the nuclear materials in those facilities. As the mission for Facility Operations plants has shifted from production to support of environmental restoration, each plant is making a transition to support the new mission. The facilities include: K Basins (N Reactor fuel storage); N Reactor; Plutonium-Uranium Reduction Extraction (PUREX) Plant; Uranium Oxide (UO{sub 3}) Plant; 300 Area Fuels Supply (N Reactor fuel supply); Plutonium Finishing Plant (PFP).

Facility Operations 1993 fiscal year work plan: WBS 1.3.1

International Nuclear Information System (INIS)

1992-11-01

The Facility Operations program is responsible for the safe, secure, and environmentally sound management of several former defense nuclear production facilities, and for the nuclear materials in those facilities. As the mission for Facility Operations plants has shifted from production to support of environmental restoration, each plant is making a transition to support the new mission. The facilities include: K Basins (N Reactor fuel storage); N Reactor; Plutonium-Uranium Reduction Extraction (PUREX) Plant; Uranium Oxide (UO 3 ) Plant; 300 Area Fuels Supply (N Reactor fuel supply); Plutonium Finishing Plant (PFP)

Oxide thickness measurement for monitoring fuel performance at high burnup

International Nuclear Information System (INIS)

Jaeger, M.A.; Van Swam, L.F.P.; Brueck-Neufeld, K.

1991-01-01

For on-site monitoring of the fuel performance at high burnup, Advanced Nuclear Fuels uses the linear scan eddy current method to determine the oxide thickness of irradiated Zircaloy fuel cans. Direct digital data acquisition methods are employed to collect the data on magnetic storage media. This field-proven methodology allows oxide thickness measurements and rapid interpretation of the data during the reactor outages and makes it possible to immediately reinsert the assemblies for the next operating cycle. The accuracy of the poolside measurements and data acquisition/interpretation techniques have been verified through hot cell metallographic measurements of rods previously measured in the fuel pool. The accumulated data provide a valuable database against which oxide growth models have been benchmarked and allow for effective monitoring of fuel performance. (orig.) [de

The disposal of Canada's nuclear fuel waste: engineering for a disposal facility

International Nuclear Information System (INIS)

Simmons, G.R.; Baumgartner, P.

1994-01-01

This report presents some general considerations for engineering a nuclear fuel waste disposal facility, alternative disposal-vault concepts and arrangements, and a conceptual design of a used-fuel disposal centre that was used to assess the technical feasibility, costs and potential effects of disposal. The general considerations and alternative disposal-vault arrangements are presented to show that options are available to allow the design to be adapted to actual site conditions. The conceptual design for a used-fuel disposal centre includes descriptions of the two major components of the disposal facility, the Used-Fuel Packaging Plant and the disposal vault; the ancillary facilities and services needed to carry out the operations are also identified. The development of the disposal facility, its operation, its decommissioning, and the reclamation of the site are discussed. The costs, labour requirements and schedules used to assess socioeconomic effects and that may be used to assess the cost burden of waste disposal to the consumer of nuclear energy are estimated. The Canadian Nuclear Fuel Waste Management Program is funded jointly by AECL and Ontario Hydro under the auspices of the CANDU Owners Group. (author)

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

Low temperature spent fuel oxidation under tuff repository conditions

International Nuclear Information System (INIS)

Einziger, R.E.; Woodley, R.E.

1985-01-01

The Nevada Nuclear Waste Storage Investigations Project is studying the suitability of tuffaceous rocks at Yucca Mountain, Nye County, Nevada, for high level waste disposal. The oxidation state of LWR spent fuel in a tuff repository may be a significant factor in determining its ability to inhibit radionuclide migration. Long term exposure at low temperatures to the moist air expected in a tuff repository is expected to increase the oxidation state of the fuel. A program is underway to determine the spent fuel oxidation mechanisms which might be active in a tuff repository. Initial work involves a series of TGA experiments to determine the effectiveness of the technique and to obtain preliminary oxidation data. Tests were run at 200 0 C and 225 0 C for as long as 720 hours. Grain boundary diffusion appears to open up a greater surface area for oxidation prior to onset of bulk diffusion. Temperature strongly influences the oxidation rates. The effect of moisture is small but readily measurable. 25 refs., 7 figs., 4 tabs

Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

Energy Technology Data Exchange (ETDEWEB)

Cowell, B.S.; Fisher, S.E.

1999-02-01

The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option.

Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

International Nuclear Information System (INIS)

Cowell, B.S.; Fisher, S.E.

1999-01-01

The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option

Criticality safety studies for plutonium–uranium metal fuel pin fabrication facility

International Nuclear Information System (INIS)

Stephen, Neethu Hanna; Reddy, C.P.

2013-01-01

Highlights: ► Criticality safety limits for PUMP-F facility is identified. ► The fissile mass which can be handled safely during alloy preparation is 10.5 kg. ► The number of fuel slugs which can be handled safely during injection casting is 53. ► The number of fuel slugs which can be handled safely after fuel fabrication is 71. - Abstract: This study focuses on the criticality safety during the fabrication of fast reactor metal fuel pins comprising of the fuel type U–15Pu, U–19Pu and U–19Pu–6Zr in the Plutonium–Uranium Metal fuel Pin fabrication Facility (PUMP-F). Maximum amount of fissile mass which can be handled safely during master alloy preparation, Injection casting and fuel slug preparation following fuel pin fabrication were identified and fixed based on this study. In the induction melting furnace, the fissile mass can be limited to 10.5 kg. During fuel slug preparation and fuel pin fabrication, fuel slugs and pins were arranged in hexagonal and square lattices to identify the most reactive configuration. The number of fuel slugs which can be handled safely after injection casting can be fixed to be 53, whereas after fuel fabrication it is 71

Radiation protection and environmental surveillance programme in and around Nuclear Fuel Cycle Facilities in India

International Nuclear Information System (INIS)

Tripathi, R.M.

2018-01-01

Radiation safety is an integral part of the operation of the Indian nuclear fuel cycle facilities and safety culture has been inculcated in all the spheres of its operation. Nuclear fuel cycle comprises of mineral exploration, mining, ore processing, fuel fabrication, power plants, reprocessing, waste management and accelerator facilities. Health Physics Division of BARC is entrusted with the responsibility of radiation protection and environmental surveillance in all the nuclear fuel cycle facilities

Dissolution of mixed oxide fuel as a function of fabrication variables

International Nuclear Information System (INIS)

Lerch, R.E.

1979-08-01

Dissolution properties of mechanically blended mixed oxide fuel were very dependent on the six fuel fabrication variables studied. Fuel sintering temperature, source of PuO 2 and PuO 2 content of the fuel had major effects: (1) as the sintering temperature was increased from 1400 to 1700 0 C, pellet dissolution was more complete; (2) pellets made from burned metal derived PuO 2 were more completely dissolved than pellets made from calcined nitrate derived PuO 2 which in turn were more completely dissolved than pellets made from calcined nitrate derived PuO 2 ; (3) as the PuO 2 content decreased from 25 to 15 wt % PuO 2 , pellet dissolution was more complete. Preferential dissolution of uranium occurred in all the mechanically blended mixed oxide. Unirradiated mixed oxide fuel pellets made by the Sol Gel process were generally quite soluble in nitric acid. Unirradiated mixed oxide fuel pellets made by the coprecipitation process dissolved completely and rapidly in nitric acid. Fuel made by the coprecipitation process was more completely dissolved than fuel made by the Sol Gel process which, in turn, was more completely dissolved than fuel made by mechanically blending UO 2 and PuO 2 as shown below. Addition of uncomplexed fluoride to nitric acid during fuel dissolution generally rendered all fuel samples completely dissolvable. In boiling 12M nitric acid, 95 to 99% of the plutonium which was going to dissolve did so in the first hour. Irradiated mechanically blended mixed oxide fuel with known fuel fabrication conditions was also subjected to fuel dissolution tests. While irradiation was shown to increase completeness of plutonium dissolution, poor dissolubility due to adverse fabrication conditions (e.g., low sintering temperature) remained after irradiation

Reversible solid oxide fuel cells (R-SOFCs) with chemically stable proton-conducting oxides

KAUST Repository

Bi, Lei

2015-07-01

Proton-conducting oxides offer a promising way of lowering the working temperature of solid oxide cells to the intermediate temperate range (500 to 700. °C) due to their better ionic conductivity. In addition, the application of proton-conducting oxides in both solid oxide fuel cells (SOFCs) and sold oxide electrolysis cells (SOECs) provides unique advantages compared with the use of conventional oxygen-ion conducting conductors, including the formation of water at the air electrode site. Since the discovery of proton conduction in some oxides about 30. years ago, the development of proton-conducting oxides in SOFCs and SOECs (the reverse mode of SOFCs) has gained increased attention. This paper briefly summarizes the development in the recent years of R-SOFCs with proton-conducting electrolytes, focusing on discussing the importance of adopting chemically stable materials in both fuel cell and electrolysis modes. The development of electrode materials for proton-conducting R-SOFCs is also discussed. © 2015 Elsevier B.V.

Nuclear Solid Waste Processing Design at the Idaho Spent Fuels Facility

International Nuclear Information System (INIS)

Dippre, M. A.

2003-01-01

A spent nuclear fuels (SNF) repackaging and storage facility was designed for the Idaho National Engineering and Environmental Laboratory (INEEL), with nuclear solid waste processing capability. Nuclear solid waste included contaminated or potentially contaminated spent fuel containers, associated hardware, machinery parts, light bulbs, tools, PPE, rags, swabs, tarps, weld rod, and HEPA filters. Design of the nuclear solid waste processing facilities included consideration of contractual, regulatory, ALARA (as low as reasonably achievable) exposure, economic, logistical, and space availability requirements. The design also included non-attended transfer methods between the fuel packaging area (FPA) (hot cell) and the waste processing area. A monitoring system was designed for use within the FPA of the facility, to pre-screen the most potentially contaminated fuel canister waste materials, according to contact- or non-contact-handled capability. Fuel canister waste materials which are not able to be contact-handled after attempted decontamination will be processed remotely and packaged within the FPA. Noncontact- handled materials processing includes size-reduction, as required to fit into INEEL permitted containers which will provide sufficient additional shielding to allow contact handling within the waste areas of the facility. The current design, which satisfied all of the requirements, employs mostly simple equipment and requires minimal use of customized components. The waste processing operation also minimizes operator exposure and operator attendance for equipment maintenance. Recently, discussions with the INEEL indicate that large canister waste materials can possibly be shipped to the burial facility without size-reduction. New waste containers would have to be designed to meet the drop tests required for transportation packages. The SNF waste processing facilities could then be highly simplified, resulting in capital equipment cost savings, operational

Current status of the first interim spent fuel storage facility in Japan

International Nuclear Information System (INIS)

Shinbo, Hitoshi; Kondo, Mitsuru

2008-01-01

In Japan, storage of spent fuels outside nuclear power plants was enabled as a result of partial amendments to the Nuclear Reactor Regulation Law in June 2000. Five months later, Mutsu City in Aomori Prefecture asked the Tokyo Electric Power Company (TEPCO) to conduct technical surveys on siting of the interim spent fuel storage facility (we call it 'Recyclable-Fuel Storage Center'). In April 2003, TEPCO submitted the report on siting feasibility examination, concluded that no improper engineering data for siting, construction of the facility will be possible from engineering viewpoint. Siting Activities for publicity and public acceptance have been continued since then. After these activities, Aomori Prefecture and Mutsu City approved siting of the Recyclable Fuel Storage Center in October 2005. Aomori Prefecture, Mutsu City, TEPCO and Japan Atomic Power Company (JAPC) signed an agreement on the interim spent fuel storage Facility. A month later, TEPCO and JAPC established Recyclable-Fuel Storage Company (RFS) in Mutsu City through joint capital investment, specialized in the first interim spent fuel storage Facility in Japan. In May 2007, we made an application for establishment permit, following safety review by regulatory authorities. In March 2008, we started the preparatory construction. RFS will safely store of spent fuels of TEPCO and JAPC until they will be reprocessed. Final storage capacity will be 5,000 ton-U. First we will construct the storage building of 3,000 ton-U to be followed by second building. We aim to start operation by 2010. (author)

Bench-scale and full-scale studies of nitric oxides reduction by gaseous fuel reburning

International Nuclear Information System (INIS)

Su, S.; Xiang, J.; Sun, L.S.; Hu, S.; Zhu, J.M.

2008-01-01

Nitrogen oxides (NOx) emissions from coal-fired boilers are significant contributors to atmospheric pollution. China has specified more rigorous legal limits for NOx emissions from power plants. As a result of the need to reduce NOx emissions, cost-effective NOx reduction strategies must be explored. This paper presented detailed experimental studies on a gaseous fuel reburning process that was performed in a 36 kilowatt bench-scale down-fired furnace to define the optimal reburning operating conditions when different Chinese coals were fired in the furnace. In addition, the combustion system of a 350 megawatt full-scale boiler was retrofitted according to the experimental results. Finally, the gaseous fuel reburning was applied to the retrofitted full-scale boiler. The purpose of the study was to obtain a better understanding of the influence of the key parameters on nitric oxide (NO) reduction efficiency of the reburning process and demonstrate the gaseous fuel reburning on a 350 MWe coal-fired boiler in China. The paper described the experimental procedure with particular reference to the experimental facility and measurement; a schematic diagram of the experimental system; experimental fuels; and characteristics of coals for the reburning experiments. Results that were presented included influence of reburn zone residence time; influence of gaseous reburn fuel per cent; influence of excess air coefficient; and unburned carbon in fly ash. It was concluded that both an above 50 per cent NO reduction efficiency and low carbon loss can be obtained by the gaseous fuel reburning process under the optimal operating conditions. 20 refs., 5 tabs., 10 figs

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

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

Development of Accident Scenario for Interim Spent Fuel Storage Facility Based on Fukushima Accident

Energy Technology Data Exchange (ETDEWEB)

Lee, Dongjin; Choi, Kwangsoon; Yoon, Hyungjoon; Park, Jungsu [KEPCO-E and C, Yongin (Korea, Republic of)

2014-05-15

700 MTU of spent nuclear fuel is discharged from nuclear fleet every year and spent fuel storage is currently 70.9% full. The on-site wet type spent fuel storage pool of each NPP(nuclear power plants) in Korea will shortly exceed its storage limit. Backdrop, the Korean government has rolled out a plan to construct an interim spent fuel storage facility by 2024. However, the type of interim spent fuel storage facility has not been decided yet in detail. The Fukushima accident has resulted in more stringent requirements for nuclear facilities in case of beyond design basis accidents. Therefore, there has been growing demand for developing scenario on interim storage facility to prepare for beyond design basis accidents and conducting dose assessment based on the scenario to verify the safety of each type of storage.

Oxidative dissolution of ADOPT compared to standard UO2 fuel

International Nuclear Information System (INIS)

Nilsson, Kristina; Roth, Olivia; Jonsson, Mats

2017-01-01

In this work we have studied oxidative dissolution of pure UO 2 and ADOPT (UO 2 doped with Al and Cr) pellets using H 2 O 2 and gammaradiolysis to induce the process. There is a small but significant difference in the oxidative dissolution rate of UO 2 and ADOPT pellets, respectively. However, the difference in oxidative dissolution yield is insignificant. Leaching experiments were also performed on in-reactor irradiated ADOPT and UO 2 pellets under oxidizing conditions. The results indicate that the U(VI) release is slightly slower from the ADOPT pellet compared to the UO 2. This could be attributed to differences in exposed surface area. However, fission products with low UO 2 solubility display a higher relative release from ADOPT fuel compared to standard UO 2 -fuel. This is attributed to a lower matrix solubility imposed by the dopants in ADOPT fuel. The release of Cs is higher from UO 2 which is attributed to the larger grain size of ADOPT. - Highlights: •Oxidative dissolution of ADOPT fuel is compared to standard UO 2 fuel. •Only marginal differences are observed. •The main difference observed is in the relative release rate of fission products. •Differences are claimed to be attributed to a lower matrix solubility imposed by the dopants in ADOPT fuel.

A description of the demonstration Integral Fast Reactor fuel cycle facility

International Nuclear Information System (INIS)

Courtney, J.C.; Carnes, M.D.; Dwight, C.C.; Forrester, R.J.

1991-01-01

A fuel examination facility at the Idaho National Engineering Laboratory is being converted into a facility that will electrochemically process spent fuel. This is an important step in the demonstration of the Integral Fast Reactor concept being developed by Argonne National Laboratory. Renovations are designed to bring the facility up to current health and safety and environmental standards and to support its new mission. Improvements include the addition of high-reliability earthquake hardened off-gas and electrical power systems, the upgrading of radiological instrumentation, and the incorporation of advances in contamination control. A major task is the construction of a new equipment repair and decontamination facility in the basement of the building to support operations

Iron oxide redox chemistry and nuclear fuel disposal

International Nuclear Information System (INIS)

Jobe, D.J.; Lemire, R.J.; Taylor, P.

1997-04-01

Solubility and stability data for iron (III) oxides and aqueous Fe(II) and Fe(III) species are reviewed, and selected values are used to calculate potential-pH diagrams for the iron system at temperatures of 25 and 100 deg C, chloride activities {C1 - } = 10 -2 and 1 mol/kg, total carbonate activity {C T } = 10 -3 mol/kg, and iron(III) oxide/oxyhydroxide solubility products (25 deg C values) K sp = {Fe 3+ }{OH - } 3 = 10 -38.5 , 10 -40 and 10 -42 . The temperatures and anion concentrations bracket the range of conditions expected in a Canadian nuclear fuel waste disposal vault. The three solubility products represent a conservative upper limit, a most probable value, and a minimum credible value, respectively, for the iron oxides likely to be important in controlling redox conditions in a disposal vault for CANDU nuclear reactor fuel. Only in the first of these three cases do the calculated redox potentials significantly exceed values under which oxidative dissolution of the fuel may occur. (author)

Study for the selection of a supplementary spent fuel storage facility for KANUPP

International Nuclear Information System (INIS)

Ahmed, W.; Iqbal, M.J.; Arshad, M.

1999-01-01

Steps taken for construction of the spent fuel facility of Karachi Nuclear Power Plant (KANUPP) are the following: choice of conceptual design and site selection; preliminary design and preparation of Preliminary Safety Analysis Report (PSAR); Construction of the facility and preparation of PSAR; testing/commissioning and loading of the storage facility. Characterisation of the spent fuel is essential for design of the storage facility. After comparison of various storage types, it seems that construction of dry storage facility based on concrete canisters at KANUPP site is a suitable option to enhance the storage capacity

Licensing of spent fuel storage facility including its physical protection in the Czech Republic

International Nuclear Information System (INIS)

Fajman, V.; Sedlacek, J.

1992-01-01

The current spent fuel management policies as practised in the Czech Republic are described, and the conception of the fuel cycle back end is outlined. The general principles and the legislative framework are explained of the licensing process concerning spent fuel interim storage facilities, including the environmental impact assessment component. The history is outlined of the licensing process for the spent fuel storage facility at the Dukovany NPP site, including the licensing of the transport and storage cask. The basic requirements placed on the physical safeguarding of the facility and on the licensing process are given. (J.B.). 13 refs

Performance of advanced oxide fuel pins in EBR-II

International Nuclear Information System (INIS)

Lawrence, L.A.; Jensen, S.M.; Hales, J.W.; Karnesky, R.A.; Makenas, B.J.

1986-05-01

The effects of design and operating parameters on mixed-oxide fuel pin irradiation performance were established for the Hanford Engineering Development Laboratory (HEDL) advanced oxide EBR-II test series. Fourteen fuel pins breached in-reactor with reference 316 SS cladding. Seven of the breaches are attributed to FCMI. Of the remaining seven breached pins, three are attributed to local cladding over-temperatures similar to the breach mechanism for the reference oxide pins irradiated in EBR-II. FCCI was found to be a contributing factor in two high burnup, i.e., 11.7 at. % breaches. The remaining two breaches were attributed to mechanical interaction of UO 2 fuel and fission products accumulated in the lower cladding insulator gap, and a loss of cladding ductility possibly due to liquid metal embrittlement. Fuel smear density appears to have the most significant impact on lifetime. Quantitative evaluations of cladding diameter increases attributed to FCMI, established fuel smear density, burnup, and cladding thickness-to-diameter ratio as the major parameters influencing the extent of cladding strain

Simulation with GOTHIC of experiments Oxidation of fuel in Air; Simulacion con GOTHIC de Experimentos de Oxidacion de Combustible en Aire

Energy Technology Data Exchange (ETDEWEB)

Martinez-Murillo Mendez, J. C.

2012-07-01

In the present work has been addressed for the first time la simulation with the GOTHIC code, experiments oxidation and ignition of SFP in phase 1. This work represents a solid starting point for analysis of specific degradation of fuel in the pools of our facilities.

Licensed-fuel-facility status report: inventory difference data, July 1981-December 1981

International Nuclear Information System (INIS)

1982-10-01

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

Licensed fuel facility status report: Inventory difference data, January 1986-June 1986

International Nuclear Information System (INIS)

1987-02-01

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

Licensed fuel facility status report: Inventory difference data, January 1987-June 1987

International Nuclear Information System (INIS)

1988-03-01

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, plutonium, or uranium-233

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

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.)

Some technical aspects of the nuclear material accounting and control at nuclear fuel cycle facilities

International Nuclear Information System (INIS)

Miller, O.A.; Babaev, N.S.; Gryazev, V.M.; Gadzhiev, G.I.; Gabeskiriya, V.Ya.

1977-01-01

The possibilities of nuclear material accounting and control are discussed at nuclear facilities of fuel cycle (WWER-type reactor, fuel fabrication plant, reprocessing plant and uranium enrichment facility) and zero energy fast reactor facility. It is shown that for nuclear material control the main method is the accounting with the application isotopic correlations at the reprocessing plant and enrichment facility. Possibilities and limitations of the application of destructive and non-destructive methods are discussed for nuclear material determinations at fuel facilities and their role in the accounting and safeguards systems as well as possibilities of the application of neutron method at a zero energy fast reactor facility [ru

Hanford Site existing irradiated fuel storage facilities description

Energy Technology Data Exchange (ETDEWEB)

Willis, W.L.

1995-01-11

This document describes facilities at the Hanford Site which are currently storing spent nuclear fuels. The descriptions provide a basis for the no-action alternatives of ongoing and planned National Environmental Protection Act reviews.

Thermal stress analysis of the fuel storage facility

International Nuclear Information System (INIS)

Chen, W.W.

1991-12-01

This paper presents the results of a nonlinear finite-element analysis to determine the structural integrity of the walls of the nuclear fuel storage room in the Radio Isotope Power System Facility of the Fuels and Materials Examination Facility (FMEF) Project. The analysis was performed to assess the effects of thermal loading on the walls that would result from a loss-of-cooling accident. The results obtained from using the same three-dimensional finite-element model with different types of elements, the eight-node brick element and the nonlinear concrete element, and the calculated results using the analytical solutions, are compared. The concrete responses in terms of octahedral normal and shearing stresses are described. The crack and crush states of the concrete were determined on the basis of multiaxial failure criteria

Conversion and Blending Facility highly enriched uranium to low enriched uranium as oxide. Revision 1

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-07-05

This Conversion and Blending Facility (CBF) will have two missions: (1) convert HEU materials into pure HEU oxide and (2) blend the pure HEU oxide with depleted and natural uranium oxide to produce an LWR grade LEU product. The primary emphasis of this blending operation will be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. To the extent practical, the chemical and isotopic concentrations of blended LEU product will be held within the specifications required for LWR fuel. Such blended LEU product will be offered to the United States Enrichment Corporation (USEC) to be sold as feed material to the commercial nuclear industry. Otherwise, blended LEU will be produced as a waste suitable for storage or disposal.

Breaking up of pure and simulated 'burnt' mixed oxide fuel by chemical interaction with oxidized sodium

International Nuclear Information System (INIS)

Besnard, R.; Chaudat, J.P.

1983-01-01

A large experimental program have permitted to investigate the behaviour of mixed oxide fuel coming in contact with hot oxidized sodium. The kinetic of the reaction, the size and the chemical nature of the particules after interaction have been studied. The main part of experiments have been performed using mixed oxide fuel non irradiated at first and with simulated fission products afterwards. Complementary informations have been obtained with UO 2 fuel pellets. After description of the experimental devices, the results are discussed and the importance of the main parameters, like temperature and fission products effect, are pointed out. (orig.)

Interconnection of bundled solid oxide fuel cells

Science.gov (United States)

Brown, Michael; Bessette, II, Norman F; Litka, Anthony F; Schmidt, Douglas S

2014-01-14

A system and method for electrically interconnecting a plurality of fuel cells to provide dense packing of the fuel cells. Each one of the plurality of fuel cells has a plurality of discrete electrical connection points along an outer surface. Electrical connections are made directly between the discrete electrical connection points of adjacent fuel cells so that the fuel cells can be packed more densely. Fuel cells have at least one outer electrode and at least one discrete interconnection to an inner electrode, wherein the outer electrode is one of a cathode and and anode and wherein the inner electrode is the other of the cathode and the anode. In tubular solid oxide fuel cells the discrete electrical connection points are spaced along the length of the fuel cell.

MTR radiological database for SRS spent nuclear fuel facilities

International Nuclear Information System (INIS)

Blanchard, A.

2000-01-01

A database for radiological characterization of incoming Material Test Reactor (MTR) fuel has been developed for application to the Receiving Basin for Offsite Fuels (RBOF) and L-Basin spent fuel storage facilities at the Savannah River Site (SRS). This database provides a quick quantitative check to determine if SRS bound spent fuel is radiologically bounded by the Reference Fuel Assembly used in the L-Basin and RBOF authorization bases. The developed database considers pertinent characteristics of domestic and foreign research reactor fuel including exposure, fuel enrichment, irradiation time, cooling time, and fuel-to-moderator ratio. The supplied tables replace the time-consuming studies associated with authorization of SRS bound spent fuel with simple hand calculations. Additionally, the comprehensive database provides the means to overcome resource limitations, since a series of simple, yet conservative, hand calculations can now be performed in a timely manner and replace computational and technical staff requirements

High temperature transient deformation of mixed oxide fuels

International Nuclear Information System (INIS)

Slagle, O.D.

1986-01-01

The purpose of this paper is to present recent experimental results on fuel creep under transient conditions at high temperatures. The effect of temperature, stress, heating rate, density and grain size were considered. An empirical formulation is derived for the relationship between strain, stress, temperature and heating rate. This relationship provides a means for incorporating stress relief into the analysis of fuel-cladding interaction during an overpower transient. The effect of sample density and initial grain size is considered by varying the sample parameters. Previously derived steady-state creep relationships for the high temperature creep of mixed oxide fuel were combined with the time dependency of creep found for UO 2 to calculate a transient creep relationship for mixed oxide fuel. These calculated results were found to be in good agreement with the measured high temperature transient creep results

Los Alamos Hot-Cell-Facility modifications for examining FFTF fuel pins

International Nuclear Information System (INIS)

Campbell, B.M.; Ledbetter, J.M.

1982-01-01

Commissioned in 1960, the Wing 9 Hot Cell Facility at Los Alamos was recently modified to meet the needs of the 1980s. Because fuel pins from the Fast Flux Test Facility (FFTF) at the Hanford Engineering Development Laboratory (HEDL) are too long for examination in the original hot cells, we modified cells to accommodate longer fuel pins and to provide other capabilities as well. For instance, the T-3 shipping cask now can be opened in an inert atmosphere that can be maintained for all nondestructive and destructive examinations of the fuel pins. The full-length pins are visually examined and photographed, the wire wrap is removed, and fission gas is sampled. After the fuel pin is cropped, a cap is seal-welded on the section containing the fuel column. This section is then transferred to other cells for gamma-scanning, radiography, profilometry, sectioning for metallography, and chemical analysis

Fuel-cladding mechanical interaction effects in fast reactor mixed oxide fuel

Energy Technology Data Exchange (ETDEWEB)

Boltax, A [Westinghouse Electric Corporation, Advanced Reactor Division, Madison, PA (United States); Biancheria, A

1977-04-01

Thermal and fast reactor irradiation experiments on mixed oxide fuel pins under steady-state and power change conditions reveal evidence for significant fuel-cladding mechanical interaction (FCMI) effects. Analytical studies with the LIFE-III fuel performance code indicate that high cladding stresses can be produced by general and local FCMI effects. Also, evidence is presented to show that local cladding strains can be caused by the accumulation of cesium at the fuel-cladding interface. Although it is apparent that steady-state FCMI effects have not given rise to cladding breaches in current fast reactors, it is anticipated that FCMI may become more important in the future because of interest in: higher fuel burnups; increased power ramp rates; load follow operation; and low swelling cladding alloys. (author)

Fuel-cladding mechanical interaction effects in fast reactor mixed oxide fuel

International Nuclear Information System (INIS)

Boltax, A.; Biancheria, A.

1977-01-01

Thermal and fast reactor irradiation experiments on mixed oxide fuel pins under steady-state and power change conditions reveal evidence for significant fuel-cladding mechanical interaction (FCMI) effects. Analytical studies with the LIFE-III fuel performance code indicate that high cladding stresses can be produced by general and local FCMI effects. Also, evidence is presented to show that local cladding strains can be caused by the accumulation of cesium at the fuel-cladding interface. Although it is apparent that steady-state FCMI effects have not given rise to cladding breaches in current fast reactors, it is anticipated that FCMI may become more important in the future because of interest in: higher fuel burnups; increased power ramp rates; load follow operation; and low swelling cladding alloys. (author)

Pilot-scale equipment development for pyrochemical treatment of spent oxide fuel

International Nuclear Information System (INIS)

Herrmann, S. D.

1999-01-01

Fundamental objectives regarding spent nuclear fuel treatment technologies include, first, the effective distribution of spent fuel constituents among product and stable waste forms and, second, the minimization and standardization of waste form types and volumes. Argonne National Laboratory (ANL) has developed and is presently demonstrating the electrometallurgical treatment of sodium-bonded metal fuel from Experimental Breeder Reactor II, resulting in an uranium product and two stable waste forms, i.e. ceramic and metallic. Engineering efforts are underway at ANL to develop pilot-scale equipment which would precondition irradiated oxide fuel via pyrochemical processing and subsequently allow for electrometallurgical treatment of such non-metallic fuels into standard product and waste forms. This paper highlights the integration of proposed spent oxide fuel treatment with existing electrometallurgical processes. System designs and technical bases for development of pilot-scale oxide reduction equipment are also described

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

Development of moderated neutron calibration fields simulating workplaces of MOX fuel facilities

International Nuclear Information System (INIS)

Tsujimura, Norio; Yoshida, Tadayoshi; Takada, Chie

2005-01-01

It is important for the MOX fuel facilities to control neutrons produced by the spontaneous fission of plutonium isotopes and those from (α,n) reactions between 18 O and α particles emitted by 238 Pu. Neutron dose meters should be calibrated for measuring these neutrons. We have developed moderated-neutron calibration fields employing a 252 Cf neutron source and moderators mainly for the characteristics evaluation and the calibration of neutron detectors used in MOX fuel facilities. Neutron energy spectrum can be adjusted by changing the position of the 252 Cf neutron source and combining different moderators to simulate the neutron field of the MOX fuel facility. This performance is realized owing to using an existing neutron irradiation room. (K. Yoshida)

Automated manufacturing of breeder reactor fuels

International Nuclear Information System (INIS)

Nyman, D.H.; Benson, E.M.; Bennett, D.W.

1983-09-01

The Secure Automated Fabrication (SAF) line is an automated, remotely controlled breeder fuel pin fabrication process which is to be installed in the Fuels and Materials Examination Facility (FMEF). The FMEF is presently under construction at Hanford and is scheduled for completion in 1984. The SAF line is scheduled for startup in 1987 and will produce mixed uranium-plutonium oxide fuel pins for the Fast Flux Test Facility (FFTF). Radiological protection requirements, computer control equipment, use of robotics, and the fabrication process is described

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.

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

Transmutation of minor actinide using BWR fueled mixed oxide

International Nuclear Information System (INIS)

Susilo, Jati

2000-01-01

Nuclear spent fuel recycle has a strategic importance in the aspect of nuclear fuel economy and prevention of its spread-out. One among other application of recycle is to produce mixed oxide fuel (Mo) namely mixed Plutonium and uranium oxide. As for decreasing the burden of nuclear high level waste (HLW) treatment, transmutation of minor actinide (MA) that has very long half life will be carried out by conversion technique in nuclear reactor. The purpose of this study was to know influence of transition fuel cell regarding the percent weight of transmutation MA in the BWR fueled MOX. Calculation of cell BWR was used SRAC computer code, with assume that the reactor in equilibrium. The percent weight of transmutation MA to be optimum by increasing the discharge burn-up of nuclear fuel, raising ratio of moderator to fuel volume (Vm/Vf), and loading MA with percent weight about 3%-6% and also reducing amount of percent weight Pu in MOX fuel. For mixed fuel standard reactor, reactivity value were obtained between about -50pcm ∼ -230pcm for void coefficient and -1.8pcm ∼ -2.6pcm for fuel temperature coefficient

Trends for Methane Oxidation at Solid Oxide Fuel Cell Conditions

DEFF Research Database (Denmark)

Kleis, Jesper; Jones, Glenn; Abild-Pedersen, Frank

2009-01-01

First-principles calculations are used to predict a plausible reaction pathway for the methane oxidation reaction. In turn, this pathway is used to obtain trends in methane oxidation activity at solid oxide fuel cell (SOFC) anode materials. Reaction energetics and barriers for the elementary...... the Ni surfaces to other metals of interest. This allows the reactivity over the different metals to be understood in terms of two reactivity descriptors, namely, the carbon and oxygen adsorption energies. By combining a simple free-energy analysis with microkinetic modeling, activity landscapes of anode...

Dry storage of spent fuel elements: interim facility

International Nuclear Information System (INIS)

Quihillalt, O.J.

1993-01-01

Apart from the existing facilities to storage nuclear fuel elements at Argentina's nuclear power stations, a new interim storage facility has been planned and projected by the Argentinean Atomic Energy Commission (CNEA) that will be constructed by private group. This article presents the developments and describes the activities undertaken until the national policy approach to the final decision for the most suitable alternative to be adopted. (B.C.A.). 09 refs, 01 fig, 09 tabs

Accidents and troubles in nuclear fuel facilities in fiscal year 1987

International Nuclear Information System (INIS)

1988-01-01

The number of the accidents and troubles reported in fiscal year 1987 in relation to nuclear fuel facilities based on the stipulation of the law on the regulation of nuclear raw materials, nuclear fuel materials and nuclear reactors was two. In Tokai Works, Power Reactor and Nuclear Fuel Development Corp., on September 17, 1987, the conveyor for transporting spent fuel in the separation and refining shop of the reprocessing plant broke down, consequently, the operation of the reprocessing plant was stopped for about five months. In Tokai Testing Works, Mitsubishi Heavy Industries Ltd., on February 7, 1988, a worker who was putting up posters in the control area of the uranium experiment facilities fell from a stepladder, and required treatment by entering a hospital for about one month, suffering bone fracture. (K.I.)

Technical, economic and institutional aspects of regional spent fuel storage facilities

International Nuclear Information System (INIS)

2005-11-01

A particular challenge facing countries with small nuclear programmes is the preparation for extended interim storage and then disposal of their spent nuclear fuel. The costs and complications of providing for away-from-reactor storage facilities and/or geological repositories for relatively small amounts of spent fuel may be prohibitively high, motivating interest in regional solutions. This publication addresses the technical, economic and institutional aspects of regional spent fuel storage facilities (RSFSF) and is based on the results of a series of meetings on this topic with participants from IAEA Member States. Topics discussed include safety criteria and standards, safeguards and physical protection, fuel acceptance criteria, long term stability of systems and stored fuel, selection of site, infrastructure aspects, storage technology, licensing, operations, transport, decommissioning, as well as research and development. Furthermore the publication comprises economic, financial and institutional considerations including organizations and legal aspects followed by political and public acceptance and ethical considerations. Approaches and processes for implementation are discussed, as well as the overall benefits and risks of implementing a regional facility. It is illustrated that implementing a RSFSF facility would involve simultaneously addressing a wide range of diverse challenges. The appendix to this report tabulates the numerous issues that have been touched upon in the study. It appears, however, from the discussions that the challenges can in principle be met; the RSFSF concept is technically feasible and potentially economically viable. The technical committees producing this report did not identify any obvious institutional deficiencies that would prevent completion of such a project. Storing spent fuel in a few safe, reliable, secure facilities could enhance safeguards, physical protection and non-proliferation benefits. The committee also

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

Licensed fuel facility status report: Inventory difference data, January 1988--June 1988

International Nuclear Information System (INIS)

1989-03-01

NRC is committed to the periodic publication of licensed fuel facilities' inventory difference data, after 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 (i) one effective kilogram of special nuclear material of low strategic significance, (ii) one effective kilogram of special nuclear material of moderate strategic significance, (iii) one effective kilogram of strategic special nuclear material contained in irradiated fuel reprocessing operations, or (iv) five formula kilograms of strategic special nuclear material

Reirradiation of mixed-oxide fuel pins at increased temperatures

International Nuclear Information System (INIS)

Lawrence, L.A.; Weber, E.T.

1976-05-01

Mixed-oxide fuel pins from EBR-II irradiations were reirradiated in the General Electric Test Reactor (GETR) at higher temperatures than experienced in EBR-II to study effects of the increased operating temperatures on thermal/mechanical and chemical behavior. The response of a mixed-oxide fuel pin to a power increase after having operated at a lower power for a significant portion of its life-time is an area of performance evaluation where little information currently exists. Results show that the cladding diameter changes resulting from the reirradiation are strongly dependent upon both prior burnup level and the magnitude of the temperature increase. Results provide the initial rough outlines of boundaries within which mixed-oxide fuel pins can or cannot tolerate power increases after substantial prior burnup at lower powers

Quality assurance for breeder reactor fuel fabrication

International Nuclear Information System (INIS)

Marx, E.R.

1978-01-01

Fuel performance in the Fast Flux Test Facility (FFTF) depends on fabrication of fuel to rigorous quality standards. The quality program including Management, Procurement, Fabrication, Inspection, Records, and Audits is discussed as well as unique mixed oxide fuel inspections such as homogeneity inspection, analytical chemistry, and nondestructive fissile assay

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%

PETER loop. Multifunctional test facility for thermal hydraulic investigations of PWR fuel elements

International Nuclear Information System (INIS)

Ganzmann, I.; Hille, D.; Staude, U.

2009-01-01

The reliable fuel element behavior during the complete fuel cycle is one of the fundamental prerequisites of a safe and efficient nuclear power plant operation. The fuel element behavior with respect to pressure drop and vibration impact cannot be simulated by means of fluid-structure interaction codes. Therefore it is necessary to perform tests using fuel element mock-ups (1:1). AREVA NP has constructed the test facility PETER (PWR fuel element tests in Erlangen) loop. The modular construction allows maximum flexibility for any type of fuel elements. Modern measuring instrumentation for flow, pressure and vibration characterization allows the analysis of cause and consequences of thermal hydraulic phenomena. PETER loop is the standard test facility for the qualification of dynamic fuel element behavior in flowing fluid and is used for failure mode analysis.

Predisposal Management of Radioactive Waste from Nuclear Fuel Cycle Facilities. Specific Safety Guide

International Nuclear Information System (INIS)

2016-01-01

This Safety Guide provides guidance on the predisposal management of all types of radioactive waste (including spent nuclear fuel declared as waste and high level waste) generated at nuclear fuel cycle facilities. These waste management facilities may be located within larger facilities or may be separate, dedicated waste management facilities (including centralized waste management facilities). The Safety Guide covers all stages in the lifetime of these facilities, including their siting, design, construction, commissioning, operation, and shutdown and decommissioning. It covers all steps carried out in the management of radioactive waste following its generation up to (but not including) disposal, including its processing (pretreatment, treatment and conditioning). Radioactive waste generated both during normal operation and in accident conditions is considered

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

Seismic design and analysis of nuclear fuel cycle facilities in France

International Nuclear Information System (INIS)

Sollogoub, P.

2001-01-01

Methodology for seismic design of nuclear fuel facilities and power plants in France is described. After the description of regulatory and normative texts for seismic design, different elements are examined: definition of ground motion, analysis methods, new trends, reevaluation and specificity of Fuel Cycle Facilities. R/D developments are explicated in each part. Their final objective are to better quantify the margins of each step which, in relation with safety analysis,lead to balanced design, analysis and retrofit rules. (author)

High burnup performance of an advanced oxide fuel assembly in FFTF [Fast Flux Test Facility] with ferritic/martensitic materials

International Nuclear Information System (INIS)

Bridges, A.E.; Saito, G.H.; Lovell, A.J.; Makenas, B.J.

1986-05-01

An advanced oxide fuel assembly with ferritic/martensitic materials has successfully completed its sixth cycle of irradiation in the FFTF, reaching a peak pellet burnup greater than 100 MWd/KgM and a peak fast fluence greater than 15 x 10 22 n/cm 2 . The cladding, wire-wrap, and duct material for the ACO-1 test assembly is the ferritic/martensitic alloy, HT9, which was chosen for use in long-lifetime fuel assemblies because of its good nominal temperature creep strength and low swelling rate. Valuable experience on the performance of HT9 materials has been gained from this test, advancing our quest for long-lifetime fuel. Pertinent data, obtained from the ACO-1 test assembly, will support the irradiation of the Core Demonstration Experiment in FFTF

LG Solid Oxide Fuel Cell (SOFC) Model Development

Energy Technology Data Exchange (ETDEWEB)

Haberman, Ben [LG Fuel Cell Systems Inc., North Canton, OH (United States); Martinez-Baca, Carlos [LG Fuel Cell Systems Inc., North Canton, OH (United States); Rush, Greg [LG Fuel Cell Systems Inc., North Canton, OH (United States)

2013-05-31

This report presents a summary of the work performed by LG Fuel Cell Systems Inc. during the project LG Solid Oxide Fuel Cell (SOFC) Model Development (DOE Award Number: DE-FE0000773) which commenced on October 1, 2009 and was completed on March 31, 2013. The aim of this project is for LG Fuel Cell Systems Inc. (formerly known as Rolls-Royce Fuel Cell Systems (US) Inc.) (LGFCS) to develop a multi-physics solid oxide fuel cell (SOFC) computer code (MPC) for performance calculations of the LGFCS fuel cell structure to support fuel cell product design and development. A summary of the initial stages of the project is provided which describes the MPC requirements that were developed and the selection of a candidate code, STAR-CCM+ (CD-adapco). This is followed by a detailed description of the subsequent work program including code enhancement and model verification and validation activities. Details of the code enhancements that were implemented to facilitate MPC SOFC simulations are provided along with a description of the models that were built using the MPC and validated against experimental data. The modeling work described in this report represents a level of calculation detail that has not been previously available within LGFCS.

Radioactive waste management in a fuel reprocessing facility in fiscal 1982

International Nuclear Information System (INIS)

1984-01-01

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

Fuels and Materials Examination Facility: Environmental assessment, Hanford site, Richland, Washington: Environmental assessment

International Nuclear Information System (INIS)

1980-07-01

The Fuels and Materials Examination Facility (FMEF) and the High Performance Fuel Laboratory (HPFL) were originally proposed to be constructed as separate facilities in the 400 Area of the Hanford Site near Richland, Washington. The environmental effects of these two facilities were described and evaluated in the FMEF Environmental Assessment and the HPFL Final Environmental Impact Statement, ERDA-1550. For economic reasons, the two facilities will no longer be built as separate facilities. The FMEF facility plans have been modified to incorporate some of the features of the proposed HPFL facility while retaining essentially all of the capabilities of the original FMEF proposal. The purpose of this document is to update the FMEF Environmental Assessment to appropriately reflect addition of certain HPFL features into the FMEF facility and to assess the environmental affects of the facility which resulted from inclusion of HPFL features into the FMEF facility

Oxidative dissolution of ADOPT compared to standard UO{sub 2} fuel

Energy Technology Data Exchange (ETDEWEB)

Nilsson, Kristina [School of Chemical Science and Engineering, Applied Physical Chemistry, KTH Royal Institute of Technology, SE-100 44 Stockholm (Sweden); Roth, Olivia [Studsvik Nuclear AB, SE-611 82 Nyköping (Sweden); Jonsson, Mats, E-mail: matsj@kth.se [School of Chemical Science and Engineering, Applied Physical Chemistry, KTH Royal Institute of Technology, SE-100 44 Stockholm (Sweden)

2017-05-15

In this work we have studied oxidative dissolution of pure UO{sub 2} and ADOPT (UO{sub 2} doped with Al and Cr) pellets using H{sub 2}O{sub 2} and gammaradiolysis to induce the process. There is a small but significant difference in the oxidative dissolution rate of UO{sub 2} and ADOPT pellets, respectively. However, the difference in oxidative dissolution yield is insignificant. Leaching experiments were also performed on in-reactor irradiated ADOPT and UO{sub 2} pellets under oxidizing conditions. The results indicate that the U(VI) release is slightly slower from the ADOPT pellet compared to the UO{sub 2.} This could be attributed to differences in exposed surface area. However, fission products with low UO{sub 2} solubility display a higher relative release from ADOPT fuel compared to standard UO{sub 2}-fuel. This is attributed to a lower matrix solubility imposed by the dopants in ADOPT fuel. The release of Cs is higher from UO{sub 2} which is attributed to the larger grain size of ADOPT. - Highlights: •Oxidative dissolution of ADOPT fuel is compared to standard UO{sub 2} fuel. •Only marginal differences are observed. •The main difference observed is in the relative release rate of fission products. •Differences are claimed to be attributed to a lower matrix solubility imposed by the dopants in ADOPT fuel.

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.

Pilot-scale equipment development for lithium-based reduction of spent oxide fuel

International Nuclear Information System (INIS)

Herrmann, S. D.

1998-01-01

An integral function of the electrometallurgical conditioning of DOE spent nuclear fuel is the standardization of waste forms. Argonne National Laboratory (ANL) has developed and is presently demonstrating the electrometallurgical conditioning of sodium-bonded metal fuel from Experimental Breeder Reactor II, resulting in uranium, ceramic waste, and metal waste forms. Engineering studies are underway at ANL in support of pilot-scale equipment development, which would precondition irradiated oxide fuel and likewise demonstrate the application of electrometallurgical conditioning to such non-metallic fuels. This paper highlights the integration of proposed spent oxide fuel conditioning with existing electrometallurgical processes. Additionally, technical bases for engineering activities to support a scale up of an oxide reduction process are described

Refueling the RPI reactor facility with low-enrichment fuel

International Nuclear Information System (INIS)

Harris, D.R.; Rodriguez-Vera, F.; Wicks, F.E.

1985-01-01

The RPI Critical Facility has operated since 1963 with a core of thin, highly enriched fuel plates in twenty-five fuel assembly boxes. A program is underway to refuel the reactor with 4.81 w/o enriched SPERT (F-1) fuel rods. Use of these fuel rods will upgrade the capabilities of the reactor and will eliminate a security risk. Adequate quantities of SPERT (F-1) fuel rods are available, and their use will result in a great cost saving relative to manufacturing new low-enrichment fuel plates. The SPERT fuel rods are 19 inches longer than are the present fuel plates, so a modified core support structure is required. It is planned to support and position the SPERT fuel pins by upper and lower lattice plates, thus avoiding the considerable cost of new fuel assembly boxes. The lattice plates will be secured to the existing top and bottom plates. The design permits the fabrication and use of other lattice plates for critical experiment research programs in support of long-lived full development for power reactors. (author)

Licensed-fuel-facility status report, inventory difference data January 1981-June 1981

International Nuclear Information System (INIS)

1982-07-01

NRC is committed to the periodic release of inventory difference data from the licensed fuel facilities after the agency has had an opportunity to review the data and has performed any related investigations associated with the data. Information included 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

Steam and partial oxidation reforming options for hydrogen production from fossil fuels for PEM fuel cells

Directory of Open Access Journals (Sweden)

Yousri M.A. Welaya

2012-06-01

Full Text Available Proton exchange membrane fuel cell (PEM generates electrical power from air and from hydrogen or hydrogen rich gas mixtures. Therefore, there is an increasing interest in converting current hydrocarbon based marine fuels such as natural gas, gasoline, and diesel into hydrogen rich gases acceptable to the PEM fuel cells on board ships. Using chemical flow sheeting software, the total system efficiency has been calculated. Natural gas appears to be the best fuel for hydrogen rich gas production due to its favorable composition of lower molecular weight compounds. This paper presents a study for a 250 kW net electrical power PEM fuel cell system utilizing a partial oxidation in one case study and steam reformers in the second. This study has shown that steam-reforming process is the most competitive fuel processing option in terms of fuel processing efficiency. Partial oxidation process has proved to posses the lowest fuel processing efficiency. Among the options studied, the highest fuel processing efficiency is achieved with natural gas steam reforming system.

Conceptual structure design of experimental facility for advanced spent fuel conditioning process

International Nuclear Information System (INIS)

Joo, J. S.; Koo, J. H.; Jung, W. M.; Jo, I. J.; Kook, D. H.; Yoo, K. S.

2003-01-01

A study on the advanced spent fuel conditioning process (ACP) is carring out for the effective management of spent fuels of domestic nuclear power plants. This study presents basic shielding design, modification of IMEF's reserve hot cell facility which reserved for future usage, conceptual and structural architecture design of ACP hot cell and its contents, etc. considering the characteristics of ACP. The results of this study will be used for the basic and detail design of ACP demonstration facility, and utilized as basic data for the safety evaluation as essential data for the licensing of the ACP facility

Durable and Robust Solid Oxide Fuel Cells

DEFF Research Database (Denmark)

Hjalmarsson, Per; Knibbe, Ruth; Hauch, Anne

project had as one of its’ overarching goals to improve durability and robustness of the Danish solid oxide fuel cells. The project focus was on cells and cell components suitable for SOFC operation in the temperature range 600 – 750 °C. The cells developed and/or studied in this project are intended......The solid oxide fuel cell (SOFC) is an attractive technology for the generation of electricity with high efficiency and low emissions. Risø DTU (now DTU Energy Conversion) works closely together with Topsoe Fuel Cell A/S in their effort to bring competitive SOFC systems to the market. This 2-year...... for use within the CHP (Combined Heat and Power) market segment with stationary power plants in the range 1 – 250 kWe in mind. Lowered operation temperature is considered a good way to improve the stack durability since corrosion of the interconnect plates in a stack is lifetime limiting at T > 750 °C...

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.

Facility for in-reactor creep testing of fuel cladding

International Nuclear Information System (INIS)

Kohn, E.; Wright, M.G.

1976-11-01

A biaxial stress creep test facility has been designed and developed for operation in the WR-1 reactor. This report outlines the rationale for its design and describes its construction and the operating experience with it. The equipment is optimized for the determination of creep data on CANDU fuel cladding. Typical results from Zr-2.5 wt% Nb fuel cladding are used to illustrate the accuracy and reliability obtained. (author)

Czech interim spent fuel storage facility: operation experience, inspections and future plans

International Nuclear Information System (INIS)

Fajman, V.; Bartak, L.; Coufal, J.; Brzobohaty, K.; Kuba, S.

1999-01-01

The paper describes the situation in the spent fuel management in the Czech Republic. The interim Spent Fuel Storage Facility (ISFSF) at Dukovany, which was commissioned in January 1997 and is using dual transport and storage CASTOR - 440/84 casks, is briefly described. The authors deal with their experience in operating and inspecting the ISFSF Dukovany. The structure of the basic safety document 'Limits and Conditions of Normal Operation' is also mentioned, including the experience of the performance. The inspection activities focused on permanent checking of the leak tightness of the CASTOR 440/84 casks, the maximum cask temperature and inspections monitoring both the neutron and gamma dose rate as well as the surface contamination. The results of the inspections are mentioned in the presentation as well. The operator's experience with re-opening partly loaded and already dried CASTOR-440/84 cask, after its transport from NPP Jaslovske Bohunice to the NPP Dukovany is also described. The paper introduces briefly the concept of future spent fuel storage both from the NPP Dukovany and the NPP Temelin, as prepared by the CEZ. The preparatory work for the Central Interim Spent Nuclear Fuel Storage Facility (CISFSF) in the Czech Republic and the information concerning the planned storage technology for this facility is discussed in the paper as well. The authors describe the site selection process and the preparatory steps concerning new spent fuel facility construction including the Environmental Impact Assessment studies. (author)

Performance Evaluation of Solid Oxide Fuel Cell by Computer ...

African Journals Online (AJOL)

The search for sustainable energy source that can compete with the existing one led to the discovery and acceptance of fuel cell technologies as a perfect replacement for fossil fuel. The ability of Solid Oxide Fuel Cells (SOFC) to capture the heat generation during the process of energy generation from electrochemical ...

Radiological and environmental safety in front-end fuel cycle facilities

International Nuclear Information System (INIS)

Puranik, V.D.

2011-01-01

The front end nuclear fuel cycle comprises of mining and processing of beach mineral sands along the southern coast of Kerala, Tamilnadu and Orissa, mining and processing of uranium ore in Singhbhum-East in Jharkhand and refining and fuel fabrication at Hyderabad. The Health Physics Units (HPUs)/Environmental Survey Laboratories (ESLs) set up at each site from inception of operation to carry out regular in-plant, personnel monitoring and environmental surveillance to ensure safe working conditions, evaluate radiation exposure of workers, ensure compliance with statutory norms, help in keeping the environmental releases well within the limits and advise appropriate control measures. This paper describes the occupational and environmental radiological safety measures associated with the operations of front end of nuclear fuel cycle. Radiological monitoring in these facilities is important to ensure safe working environment, protection of workers against exposure to radiation and comply with regulatory limits of exposure. The radiation exposure of workers in different units of the front end nuclear fuels cycle facilities operated by IREL, UCIL and NFC and environmental monitoring results are summarised in this paper

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

Temporary storage facility for spent nuclear fuels at the Atucha I nuclear power station (CNA)

International Nuclear Information System (INIS)

Wasinger, K.

1983-01-01

According to plans of the Argentine Atomic Energy Commission (CNEA), the spent nuclear fuel elements of the Atucha I Nuclear Power Station are to be stored temporarily pending a decision about the ultimate disposal concept. The holding capacity of the first fuel storage facility built by the German KWU together with the whole power plant had been expanded in 1978 to a level good until mid-1982. In 1977, KWU drafted the concept of another fuel storage facility. Like the first one, it was designed as a wet storage system attached to the power plant installations and had a holding capacity of 6944 fuel elements, which corresponds to some 1100 te of uranium. This extends the storage capacity up until 1996. In 1978, KWU was commissioned by CNEA to plan the whole facility and deliver the mechanical and electrical equipment. CNEA themselves assumed responsibility for the construction work. The second fuel storage facility was commissioned three years after the start of construction. (orig.) [de

Rational Design of a Water-Storable Hierarchical Architecture Decorated with Amorphous Barium Oxide and Nickel Nanoparticles as a Solid Oxide Fuel Cell Anode with Excellent Sulfur Tolerance.

Science.gov (United States)

Song, Yufei; Wang, Wei; Ge, Lei; Xu, Xiaomin; Zhang, Zhenbao; Julião, Paulo Sérgio Barros; Zhou, Wei; Shao, Zongping

2017-11-01

Solid oxide fuel cells (SOFCs), which can directly convert chemical energy stored in fuels into electric power, represent a useful technology for a more sustainable future. They are particularly attractive given that they can be easily integrated into the currently available fossil fuel infrastructure to realize an ideal clean energy system. However, the widespread use of the SOFC technology is hindered by sulfur poisoning at the anode caused by the sulfur impurities in fossil fuels. Therefore, improving the sulfur tolerance of the anode is critical for developing SOFCs for use with fossil fuels. Herein, a novel, highly active, sulfur-tolerant anode for intermediate-temperature SOFCs is prepared via a facile impregnation and limited reaction protocol. During synthesis, Ni nanoparticles, water-storable BaZr 0.4 Ce 0.4 Y 0.2 O 3- δ (BZCY) perovskite, and amorphous BaO are formed in situ and deposited on the surface of a Sm 0.2 Ce 0.8 O 1.9 (SDC) scaffold. More specifically, a porous SDC scaffold is impregnated with a well-designed proton-conducting perovskite oxide liquid precursor with the nominal composition of Ba(Zr 0.4 Ce 0.4 Y 0.2 ) 0.8 Ni 0.2 O 3- δ (BZCYN), calcined and reduced in hydrogen. The as-synthesized hierarchical architecture exhibits high H 2 electro-oxidation activity, excellent operational stability, superior sulfur tolerance, and good thermal cyclability. This work demonstrates the potential of combining nanocatalysts and water-storable materials in advanced electrocatalysts for SOFCs.

Experimental and thermodynamic evaluation of the melting behavior of irradiated oxide fuels

International Nuclear Information System (INIS)

Adamson, M.G.; Aitken, E.A.; Caputi, R.W.

1985-01-01

Onset of melting is an important performance limit for irradiated UO 2 and UO 2 -based nuclear reactor fuels. Melting (solidus) temperatures are reasonably well known for starting fuel materials such as UO 2 and (U,PU)O 2 , however the influence of burnup on oxide fuel melting behavior continues to represent an area of considerable uncertainty. In this paper we report the results of a variety of melting temperature measurements on pseudo-binary fuel-fissia mixtures such as UO 2 -PUO 2 , UO 2 -CeO 2 , UO 2 -BaO, UO 2 -SrO, UO 2 -BaZrO 3 and UO 2 -SrZrO 3 . These measurements were performed using the thermal arrest technique on tungsten-encapsulated specimens. Several low melting eutectics, the existence of which had previously been inferred from post-irradiation examinations of high burnup mixed oxide fuels, were characterized in the course of the investigation. Also, an assessment of melting temperature changes in irradiated oxide fuels due to the production and incorporation of soluble oxidic fission products was performed by application of solution theory to the available pseudo-binary phase diagram data. The results of this assessment suggest that depression of oxide fuel solidus temperatures by dissolved fission products is substantially less than that indicated by earlier experimental studies. (orig.)