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Sample records for advanced spent fuel

  1. Advances in HTGR spent fuel treatment technology

    International Nuclear Information System (INIS)

    GA Technologies, Inc. has been investigating the burning of spent reactor graphite under Department of Energy sponsorship since 1969. Several deep fluidized bed burners have been used at the GA pilot plant to develop graphite burning techniques for both spent fuel recovery and volume reduction for waste disposal. Since 1982 this technology has been extended to include more efficient circulating bed burners. This paper includes updates on high-temperature gas-cooled reactor fuel cycle options and current results of spent fuel treatment testing for fluidized and advanced circulating bed burners

  2. Corrosion of spent Advanced Test Reactor fuel

    International Nuclear Information System (INIS)

    The results of a study of the condition of spent nuclear fuel elements from the Advanced Test Reactor (ATR) currently being stored underwater at the Idaho National Engineering Laboratory (INEL) are presented. This study was motivated by a need to estimate the corrosion behavior of dried, spent ATR fuel elements during dry storage for periods up to 50 years. The study indicated that the condition of spent ATR fuel elements currently stored underwater at the INEL is not very well known. Based on the limited data and observed corrosion behavior in the reactor and in underwater storage, it was concluded that many of the fuel elements currently stored under water in the facility called ICPP-603 FSF are in a degraded condition, and it is probable that many have breached cladding. The anticipated dehydration behavior of corroded spent ATR fuel elements was also studied, and a list of issues to be addressed by fuel element characterization before and after forced drying of the fuel elements and during dry storage is presented

  3. Development of advanced spent fuel management process

    International Nuclear Information System (INIS)

    This study is to develop an advanced spent fuel management process for countries which have not yet decided a back-end nuclear fuel cycle policy. The aims of this process development based on the pyroreduction technology of PWR spent fuels with molten lithium, are to reduce the storage volume by a quarter and to reduce the storage cooling load in half by the preferential removal of highly radioactive decay-heat elements such as Cs-137 and Sr-90 only. From the experimental results which confirm the feasibility of metallization technology, it is concluded that there are no problems in aspects of reaction kinetics and equilibrium. However, the operating performance test of each equipment on an engineering scale still remain and will be conducted in 1999. (author). 21 refs., 45 tabs., 119 figs

  4. Development of Advanced Spent Fuel Management Process

    International Nuclear Information System (INIS)

    As a part of research efforts to develop an advanced spent fuel management process, this project focused on the electrochemical reduction technology which can replace the original Li reduction technology of ANL, and we have successfully built a 20 kgHM/batch scale demonstration system. The performance tests of the system in the ACPF hot cell showed more than a 99% reduction yield of SIMFUEL, a current density of 100 mA/cm2 and a current efficiency of 80%. For an optimization of the process, the prevention of a voltage drop in an integrated cathode, a minimization of the anodic effect and an improvement of the hot cell operability by a modulation and simplization of the unit apparatuses were achieved. Basic research using a bench-scale system was also carried out by focusing on a measurement of the electrochemical reduction rate of the surrogates, an elucidation of the reaction mechanism, collecting data on the partition coefficients of the major nuclides, quantitative measurement of mass transfer rates and diffusion coefficients of oxygen and metal ions in molten salts. When compared to the PYROX process of INL, the electrochemical reduction system developed in this project has comparative advantages in its application of a flexible reaction mechanism, relatively short reaction times and increased process yields

  5. Development of Advanced Spent Fuel Management Process

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Chung Seok; Choi, I. K.; Kwon, S. G. (and others)

    2007-06-15

    As a part of research efforts to develop an advanced spent fuel management process, this project focused on the electrochemical reduction technology which can replace the original Li reduction technology of ANL, and we have successfully built a 20 kgHM/batch scale demonstration system. The performance tests of the system in the ACPF hot cell showed more than a 99% reduction yield of SIMFUEL, a current density of 100 mA/cm{sup 2} and a current efficiency of 80%. For an optimization of the process, the prevention of a voltage drop in an integrated cathode, a minimization of the anodic effect and an improvement of the hot cell operability by a modulation and simplization of the unit apparatuses were achieved. Basic research using a bench-scale system was also carried out by focusing on a measurement of the electrochemical reduction rate of the surrogates, an elucidation of the reaction mechanism, collecting data on the partition coefficients of the major nuclides, quantitative measurement of mass transfer rates and diffusion coefficients of oxygen and metal ions in molten salts. When compared to the PYROX process of INL, the electrochemical reduction system developed in this project has comparative advantages in its application of a flexible reaction mechanism, relatively short reaction times and increased process yields.

  6. LWR spent fuel storage technology: Advances and experience

    International Nuclear Information System (INIS)

    By 2003, the year the US Department of Energy (DOE) currently predicts a repository will be available, 58 domestic commercial nuclear-power plant units are expected to run out of wet storage space for LWR spent fuel. To alleviate this problem, utilities implemented advances in storage methods that increased storage capacity as well as reduced the rate of generating spent fuel. Those advances include (1) transhipping spent-fuel assemblies between pools within the same utility system, (2) reracking pools to accommodate additional spent-fuel assemblies, (3) taking credit for fuel burnup in pool storage rack designs, (4) extending fuel burnup, (5) rod consolidation, and (6) dry storage. The focus of this paper is on advances in rod consolidation and dry storage. Wet storage continues to be the predominant US spent-fuel management technology, but as a measure to enhance at-reactor storage capacity, the Nuclear Waste Policy Act of 1982 authorized DOE to assist utilities with licensing at-reactor dry storage. Information exchanges with other nations, laboratory testing and modeling, and cask tests cooperatively funded by US utilities and DOE produced a strong technical basis to develop confidence that LWR spent fuel can be stored safely for several decades in both wet and dry modes. Licensed dry storage of spent fuel in an inert atmosphere was first achieved in the US in 1986. Studies are underway in several countries to determine acceptable conditions for storing LWR spent fuel in air. Rod-consolidation technology is being developed and demonstrated to enhance the capacity for both wet and dry storage. Large-scale commercial implementation is awaiting optimization of practical and economical mechanical systems. 22 refs., 1 fig

  7. Development of Experimental Facilities for Advanced Spent Fuel Management Technology

    International Nuclear Information System (INIS)

    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∼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 α-γ 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 β-γ 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

  8. Development of advanced spent fuel management process. The fabrication and oxidation behavior of simulated metallized spent fuel

    International Nuclear Information System (INIS)

    The simulated metallized spent fuel ingots were fabricated and evaluated the oxidation rates and the activation energies under several temperature conditions to develop an advanced spent fuel management process. It was also checked the alloying characteristics of the some elements with metal uranium. (Author). 3 refs., 1 tab., 36 figs

  9. Development of Advanced Voloxidation Process for Treatment of Spent Fuel

    International Nuclear Information System (INIS)

    Data for evaluation of the effects of advanced voloxidation on pyroprocessing of spent oxide fuel with a determination for a path forward such was produced as follows: effect of particle size and particle structure on oxide reduction, assessment of decladding options for pyroprocessing, effect of removal timing of fission products, analysis of radioactivity and decay heat of advanced voloxidation process, proliferation resistance of advanced voloxidation process, Effect of advanced voloxidation process on shielding. Also, performance objectives for advanced voloxidation with respective to the down stream effects was established. The technology on design and manufacture of voloxidation and off gas treatment equipment was established. The possibility of fabrication of porous granule as a feed material for electro-reduction process was confirmed using rotary voloxidizer and SIMFUEL. The operational conditions for advanced voloxidation process consisting of 4 steps heat treatment was drawn to vaporize fission products and fabricate UO2 granule. The trapping test of Cs and Re(surrogate material of Tc) using newly developed filter were selectively separated at trapping efficiency of 99%, respectively. Data for oxidative decladding, vaporization rate of fission products, and particle size from experiment on voloxidation using spent fuel in ILN hot cell was acquisited including data of off gas trapping characteristics and verification of excellent performance of filter

  10. Radiation Monitoring System in Advanced Spent Fuel Conditioning Process Facility

    International Nuclear Information System (INIS)

    The Advanced spent fuel Conditioning Process is under development for effective management of spent fuel by converting UO2 into U-metal. For demonstration of this process, α-γ type new hot cell was built in the IMEF basement . To secure against radiation hazard, this facility needs radiation monitoring system which will observe the entire operating area before the hot cell and service area at back of it. This system consists of 7 parts; Area Monitor for γ-ray, Room Air Monitor for particulate and iodine in both area, Hot cell Monitor for hot cell inside high radiation and rear door interlock, Duct Monitor for particulate of outlet ventilation, Iodine Monitor for iodine of outlet duct, CCTV for watching workers and material movement, Server for management of whole monitoring system. After installation and test of this, radiation monitoring system will be expected to assist the successful ACP demonstration

  11. Development of Experimental Facilities for Advanced Spent Fuel Management Technology

    International Nuclear Information System (INIS)

    The Advanced spent fuel Conditioning Process Facility(ACPF) and hotcell system technologies were developed in this program for demonstrating safely and effectively the Advanced spent fuel Conditioning Process(ACP) on a lab-scale. With the analysis of work flow and characteristics of the process, ACP was successively demonstrated on a lab-scale experiments and the performance of process was evaluated. The hotcell system was comprehensively evaluated with those results and the design data for the engineering-scale demonstration was derived to propose the direction for the future research and development. The main items performed in this project were as follows. - The reconstruction of ACPF hotcell and demonstration of the ACP - The design and operation technologies for α-γ type nuclear hot cell facility - The overall evaluation of the performance, safety and operation ability of the hotcell system - The acquisition of the government licences for construction and operation and the IAEA licence for using nuclear materials The results of safety analysis and environmental effects assessment and performance data for ACPF had been used for acquiring the government licence for facility operation. The valuable experiences on pyroprocess facility design and operation knowledges would be applied to new Mock-up Facility being scheduled to be a previous stage facility of Integrated Pyroprocess Facility

  12. Homogeneity survey of advanced spent fuel conditioning process hot cell

    International Nuclear Information System (INIS)

    The hot cell facility (ACPF) for research activities related to the advanced spent fuel conditioning process (ACP) is being constructed. The hot cell construction work will be finished in May, 2005. Hot cell is designed to permit safe handling of radioactive materials up to 1,385 TBq and to keep gamma and neutron dose-rate lower than the recommended ones. The dose-rate limit values following the Korean nuclear laws are 0.01 mSv/h at operation area and 0.15 mSv/h at maintenance area. The ACPF is a concrete structure with two rooms, and made its exterior walls of heavy concrete with density of 3.45 g/cm3 and the wall thickness is more than 90 cm

  13. Utilization of spent PWR fuel-advanced nuclear fuel cycle of PWR/CANDU synergism

    Institute of Scientific and Technical Information of China (English)

    HUO Xiao-Dong; XIE Zhong-Sheng

    2004-01-01

    High neutron economy, on line refueling and channel design result in the unsurpassed fuel cycle flexibility and variety for CANDU reactors. According to the Chinese national conditions that China has both PWR and CANDU reactors and the closed cycle policy of reprocessing the spent PWR fuel is adopted, one of the advanced nuclear fuel cycles of PWR/CANDU synergism using the reprocessed uranium of spent PWR fuel in CANDU reactor is proposed, which will save the uranium resource (~22.5%), increase the energy output (~41%), decrease the quantity of spent fuels to be disposed (~2/3) and lower the cost of nuclear power. Because of the inherent flexibility of nuclear fuel cycle in CANDU reactor, and the low radiation level of recycled uranium(RU), which is acceptable for CANDU reactor fuel fabrication, the transition from the natural uranium to the RU can be completed without major modification of the reactor core structure and operation mode. It can be implemented in Qinshan Phase Ⅲ CANDU reactors with little or no requirement of big investment in new design. It can be expected that the reuse of recycled uranium of spent PWR fuel in CANDU reactor is a feasible and desirable strategy in China.

  14. Study on process basic requirements of experimental facility of advanced spent fuel management process

    International Nuclear Information System (INIS)

    The advanced spent fuel management process, which was 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. Hot cell facilities of α-γ type and inert atmosphere are required essentially for safe hot test and verification of this process. In this study, design basic data are established, and these data include process flow, process condition and yields, mass and radioactivity balance of radionuclides, process safety considerations, etc. And also, these data will be utilized for basic and detail design of hot cell facility, secured conservative safety and effective operability

  15. Advanced evaluation of spent fuel in long term wet storage in Slovakia

    International Nuclear Information System (INIS)

    Experiences with an advanced spent nuclear fuel management in Slovakia are presented in this paper. The evaluation and monitoring procedures are based on practices at the Slovak wet interim spent fuel storage facility in NPP Jaslovske Bohunice. Since 1999, leak testing of WWER-440 fuel assemblies are provided by special leak tightness detection system 'Sipping in Pool' delivered by Framatome-ANP with external heating for the precise defects determination. In 2006 a new inspection stand 'SVYP-440' for the monitoring of spent nuclear fuel condition was inserted. This stand has the possibility to open WWER-440 fuel assemblies and examine fuel elements. Optimal ways of spent fuel disposal and monitoring of nuclear fuel condition were designed. With appropriate approach of conservativeness, new factor for specifying of spent fuel leak tightness is introduced in the paper. By using computer simulations (based on SCALE 4.4a code) for fission products creation and measurements by system 'Sipping in Pool', the limit values of leak tightness were established. The mean value of the leak tightness coefficient is expected to be around 10-10

  16. Concept of advanced spent fuel reprocessing based on ion exchange

    International Nuclear Information System (INIS)

    . Furthermore, the ion exchange is appropriate for multi-element mutual separation rather than single element extraction. In the future, ion exchange reprocessing would be expected to be the comprehensive separation process for spent fuels to recover precious and usable elements and to reduce the amount of wastes. (authors)

  17. Safeguardability assessment on pilot-scale advanced spent fuel conditioning facility

    International Nuclear Information System (INIS)

    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

  18. Research on advanced aqueous reprocessing of spent nuclear fuel: literature study

    International Nuclear Information System (INIS)

    The goal of the partitioning and transmutation strategy is to reduce the radiotoxicity of spent nuclear fuel to the level of natural uranium in a short period of time (about 1000 years) and thus the required containment period of radioactive material in a repository. Furthermore, it aims to reduce the volume of waste requiring deep geological disposal and hence the associated space requirements and costs. Several aqueous as well as pyrochemical separation processes have been developed for the partitioning of the long-lived radionuclides from the remaining of the spent fuel. This report aims to describe and compare advanced aqueous reprocessing methods.

  19. Research on advanced aqueous reprocessing of spent nuclear fuel: literature study

    Energy Technology Data Exchange (ETDEWEB)

    Van Hecke, K.; Goethals, P.

    2006-07-15

    The goal of the partitioning and transmutation strategy is to reduce the radiotoxicity of spent nuclear fuel to the level of natural uranium in a short period of time (about 1000 years) and thus the required containment period of radioactive material in a repository. Furthermore, it aims to reduce the volume of waste requiring deep geological disposal and hence the associated space requirements and costs. Several aqueous as well as pyrochemical separation processes have been developed for the partitioning of the long-lived radionuclides from the remaining of the spent fuel. This report aims to describe and compare advanced aqueous reprocessing methods.

  20. Spent fuel reprocessing options

    International Nuclear Information System (INIS)

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

  1. Construction Report of Hot Cell Facility for Demonstration of Advanced Spent Fuel Conditioning Process

    International Nuclear Information System (INIS)

    The advanced spent fuel conditioning process(ACP) was proposed to reduce the overall volume of the PWR spent fuel and improve safety and economy of the long-term storage of spent fuel. The hot cell facilities for demonstration of ACP(ACPF) was consisted of α-γ type heavy concrete hot cell, the auxiliary equipment for hot cell operation, and process equipment. A existing β-γ type hot cell, located in IMEF, was refurbished to minimize construction expenditures for utilization as ACPF. The detail design of hot cell facilities and process was completed, and the safety analysis was performed to substantiate secure of conservative safety. And also, the construction of ACPF and installation of process equipment were completed, and government license for hot cell operation was acquired. In this report, the construction outline and the detail information of hot cell facilities and process equipment s are summarized to utilize for operation and maintenance of hot cell facility and process

  2. Recent advances in hardware and software are to improve spent fuel measurements

    International Nuclear Information System (INIS)

    Vast quantities of spent fuel are available for safeguard measurements, primarily in Commonwealth of Independent States (CIS) of the former Soviet Union. This spent fuel, much of which consists of long-cooling-time material, is going to become less unique in the world safeguards arena as reprocessing projects or permanent repositories continue to be delayed or postponed. The long cooling time of many of the spent fuel assemblies being prepared for intermediate term storage in the CIS countries promotes the possibility of increased accuracy in spent fuel assays. This improvement is made possible through the process of decay of the Curium isotopes and of fission products. An important point to consider for the future that could advance safeguards measurements for reverification and inspection would be to determine what safeguards requirements should be imposed upon this 'new' class of spent fuel, Improvements in measurement capability will obviously affect the safeguards requirements. What most significantly enables this progress in spent fuel measurements is the improvement in computer processing power and software enhancements leading to user-friendly Graphical User Interfaces (GUT's). The software used for these projects significantly reduces the IAEA inspector's time expenditure for both learning and operating computer and data acquisition systems, At the same time, by standardizing the spent fuel measurements, it is possible to increase reproducibility and reliability of the measurement data. Hardware systems will be described which take advantage of the increased computer control available to enable more complex measurement scenarios. A specific example of this is the active regulation of a spent fuel neutron coincident counter's 3He tubes high voltage, and subsequent scaling of measurement results to maintain a calibration for direct assay of the plutonium content of Fast Breeder Reactor spent fuel. The plutonium content has been successfully determined for

  3. A contingency safe, responsible, economic, increased capacity spent nuclear fuel (SNF) advance fuel cycle

    International Nuclear Information System (INIS)

    The purpose of this paper is to have an Advanced Light Water (LWR) fuel cycle and an associated development program to provide a contingency plan to the current DOE effort to license once-through spent Light Water Reactor (LWR) fuel for disposition at Yucca Mountain (YM). The intent is to fully support the forthcoming June 2008 DOE submittal to the Nuclear Regulatory Commission (NRC) based upon the latest DOE draft DOE/EIS-0250F-SID dated October 2007 which shows that the latest DOE YM doses would readily satisfy the anticipated NRC and Environmental Protection Agency (EP) standards. The proposed Advance Fuel Cycle can offer potential resolution of obstacles that might arise during the NRC review and, particularly, during the final hearings process to be held in Nevada. Another reason for the proposed concept is that a substantial capacity growth of the YM repository will be necessary to accommodate the SNF of Advance Light Water Reactors (ALWRs) currently under consideration for United States (U.S.) electricity production (1) and the results of the recently issued study by the Electric Power Research Institute (EPRI) to reduce CO2 emissions (2). That study predicts that by 2030 U.S. nuclear power generation would grow by 64 Gigawatt electrical (GWe) and account for 25.5 percent of the overall U.S. electrical generation. The current annual SNF once-through fuel cycle accumulation would rise from 2000-2100 MT (Metric Tons) to about 3480 MT in 2030 and the total SNF inventory, would reach nearly 500,000 MT by 2100 if U. S. nuclear power continues to grow at 1.1 percent per year after 2030. That last projection does not account for any SNF reduction due to increased fuel burnup or any increased capacity needed 'to establish supply Global Nuclear Energy Partnership (GNEP,) arrangements among nations to provide nuclear fuel and taking back spent fuel for recycling without spreading enrichment and reprocessing technologies' (3). The anticipated capacity of 120 MT planned

  4. Recent advances during the treatment of spent EBR-II fuel

    Energy Technology Data Exchange (ETDEWEB)

    Westphal, B. R.; Mariani, R. D.; Vaden, D. E.; Sherman, S. R.; Li, S. X.; Keiser, D. D., Jr.

    2000-03-20

    Several recent advances have been achieved for the electrometallurgical treatment of spent nuclear fuel. In anticipation of production operations at Argonne National Laboratory-West, development of both electrorefining and metal processing has been ongoing in the post-demonstration phase in order to further optimize the process. These development activities show considerable promise. This paper discusses the results of recent experiments as well as plans for future investigations.

  5. Safety related issues of spent nuclear fuel storage : summary of a NATO advanced research workshop

    International Nuclear Information System (INIS)

    Full text: A NATO Advanced Research Workshop was held in Almaty, Kazakhstan, in September 2005. The Workshop was co-sponsored by the IAEA and was concerned with the safety issues associated with spent fuel and waste from three types of reactor: research reactors with Al alloy-clad dispersion fuels, fast reactors with stainless steel-clad UO2, and commercial light-water reactors with Zr alloy-clad UO2. Fifteen presentations dealt with research reactors, five with the BN-350 fast reactor in Kazakhstan-shut down and in decommissioning, and two with commercial reactors in the U.S. and Ukraine. With 657 research reactors built and 274 still operational, corrosion of Al-clad research reactor spent fuel during wet storage was a major subject for discussion. Programs at the IAEA, in the U.S., and elsewhere, have actively studied corrosion of Al-clad fuel since the 1980s and the major mechanisms for aqueous corrosion of both spent fuel and of spent-fuel-pool structural components appear to be now well understood, as are the procedures required to limit corrosion. Nonetheless, avoiding corrosion requires vigilance in monitoring and controlling water quality. Measures to ensure water quality are now being taken at operating research reactors, but are difficult to impose at reactors that are shutdown, where there is less funding (or staff) for the task. It was noted there are about 62,000 spent research reactor fuel assemblies-most of them in wet storage-at many reactor sites around the world, three-quarters in industrialized nations, the remainder in developing countries. Dry storage of research reactor fuel is also being used or actively considered in France, Poland, Russia and the U.S. A variant of simple dry storage-the 'melt-and-dilute' option-casts the spent research reactor fuel with natural U into steel canisters to produce a corrosion-resistant low-enrichment fuel configuration which is suitable for safe long-term storage. The main safety issue of spent fast reactor

  6. Radio-toxicity of spent fuel of the advanced heavy water reactor.

    Science.gov (United States)

    Anand, S; Singh, K D S; Sharma, V K

    2010-01-01

    The Advanced Heavy Water Reactor (AHWR) is a new power reactor concept being developed at Bhabha Atomic Research Centre, Mumbai. The reactor retains many desirable features of the existing Pressurised Heavy Water Reactor (PHWR), while incorporating new, advanced safety features. The reactor aims to utilise the vast thorium resources available in India. The reactor core will use plutonium as the make-up fuel, while breeding (233)U in situ. On account of this unique combination of fuel materials, the operational characteristics of the fuel as determined by its radioactivity, decay heat and radio-toxicity are being viewed with great interest. Radio-toxicity of the spent fuel is a measure of potential radiological hazard to the members of the public and also important from the ecological point of view. The radio-toxicity of the AHWR fuel is extremely high to start with, being approximately 10(4) times that of the fresh natural U fuel used in a PHWR, and continues to remain relatively high during operation and subsequent cooling. A unique feature of this fuel is the peak observed in its radio-toxicity at approximately 10(5) y of decay cooling. The delayed increase in fuel toxicity has been traced primarily to a build-up of (229)Th, (230)Th and (226)Ra. This phenomenon has been observed earlier for thorium-based fuels and is confirmed for the AHWR fuel. This paper presents radio-toxicity data for AHWR spent fuel up to a period of 10(6) y and the results are compared with the radio-toxicity of PHWR. PMID:19776247

  7. Preliminary assessment of safeguardability on the concepture design of advanced spent fuel conditioning process

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang Yoon; Ha, Jang Ho; Ko, Won Il; Song, Dae Yong; Kim, Ho Dong

    2003-04-01

    In this report, a preliminary study on the safeguardability of ACP (Advanced spent fuel Conditioning Process) was conducted with Los Alamos National Laboratory. The proposed ACP concept is an electrometallurgical treatment technique to convert oxide-type spent nuclear fuels into metal forms, which can achieve significant reduction of the volume and heat load of spent fuel to be stored and disposed of. For the safeguardability analysis of the ACP facility, sub-processes and their KMPs (Key Measurement Points) were defined first, and then their material flows were analyzed. Finally, the standard deviation of the Inventory Difference (ID) value of the facility was estimated with assumption by assuming international target values for the uncertainty of measurement methods and their uncertainty. From the preliminary calculation, we concluded that if the assumptions regarding measurement instruments can be achieved in a safeguards system for the ACP facility, the safeguards goals of International Atomic Energy Agency (IAEA) could be met. In the second phase of this study, further study on sensitivity analyses considering various factors such as measurement errors, facility capacities, MBA periods etc. may be needed.

  8. Design Report of Hot Cell Facilities for Demonstration of Advanced Spent Fuel Conditioning Process

    International Nuclear Information System (INIS)

    The advanced spent fuel conditioning process(ACP) was proposed to reduce the overall volume of the PWR spent fuel and improve safety and economy of the long-term storage of spent fuel. The hot test will be carried out for verification of the ACP in a laboratory scale. For the hot test, the hot cell facilities of α-γ 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 β-γ type will be refurbished to minimize construction expenditures of hot cell facility. The detail design of hot cell facilities and process were completed, and the safety analysis was performed to substantiate secure of conservative safety. This results were utilized for refurbishment of IMEF future hot cell and installation of process equipments, and manufacturing and procurement of hot cell auxiliary equipments. The safety analysis report were submitted to KINS through MOST for license acquisition, the government issued license for construction and operation. And, the hot test for demonstration of the ACP is performing in this hot cell facilities. In this report, the detail design and safety analysis data are summarized to utilize for operation of hot cell facility and process

  9. Establishment of team work system for advanced spent fuel management process

    International Nuclear Information System (INIS)

    The advanced spent fuel management process (ASFMP), which is being developed by KAERI, is now in the 2nd research phase. This phase has a goal to design the total system of active demonstration of ASFMP. It is composed of the core process, remote handling technologies, examination technologies and experimental facilities. For the collaboration of these research fields, a team work system has been established by proper hardware and software selections for use of about 50 project members. This system has been tested by adaptation to the ASFMP project and will be used during the remained project period

  10. Spent fuel management strategies

    International Nuclear Information System (INIS)

    Nuclear fuel cycle is divided into two sections; front end and back end of the fuel cycle. Front end of the fuel cycle, which covers all the activities of the fuel cycle before the fuel goes into the reactor has better developed and well-defined technologies. For storage of the spent fuel which are subjects of the back end of the fuel cycle, the waste management policies are not so well defined. There are three approaches that exist today for management of spent fuel. 1. For once through or open fuel cycles direct disposal of spent fuel in a deep geological repository, 2. For closed fuel cycles reprocessing of spent fuel and recycling of the recovered plutonium and uranium in new mixed oxide (MOX) fuels, 3. The spent fuel is placed in long term interim storage pending a decision as to its ultimate reprocessing or disposal. There are so large scale geological repositories for the final disposal of spent fuel in operation. Studies on suitable site selection, design, construction and licensing take about 30-40 years. Reprocessing, on the other hand, produces plutonium and is therefore under close inspection because of the Non Proliferation Treaty. Today more countries are delaying their final decision about the spent fuel management approach and using the long term interim storage approach

  11. Review of methodological analysis for the nuclear material accounting and control in the advanced spent fuel management process

    International Nuclear Information System (INIS)

    Nuclear materials accounting and verification in radiochemical processing facilities is essential, because it is the first possible time in the nuclear fuel cycle that plutonium can be measured. In these facilities, effective nuclear materials accounting systems and international safeguards inspections rely heavily upon nondestructive assay measurements. Therefore, it is important to know whether the radiation-based nondestructive assay (NDA) techniques for Advanced Spent Fuel Management Process are applicable or not. As a result of reviewing the existing NDA techniques for nuclear material accounting, it was revealed that γ-ray spectrometry, x-ray fluorescence/ densitometry and calorimetry techniques are not applicable to the advanced spent fuel management process because of the size of the measuring devices installed in a hot cell and the samples including some fission products. Therefore, the neutron technique is only applicable to this processing facility. The results reviewed in this study can be used to design a hot cell for the advanced spent fuel management process

  12. The development of technical database of advanced spent fuel management process

    International Nuclear Information System (INIS)

    The purpose of this study is to develop the technical database system to provide useful information to researchers who study on the back end nuclear fuel cycle. Technical database of advanced spent fuel management process was developed for a prototype system in 1997. In 1998, this database system is improved into multi-user systems and appended special database which is composed of thermochemical formation data and reaction data. In this report, the detailed specification of our system design is described and the operating methods are illustrated as a user's manual. Also, expanding current system, or interfacing between this system and other system, this report is very useful as a reference. (Author). 10 refs., 18 tabs., 46 fig

  13. Radiation-resistant requirement analysis of device for the advanced spent fuel conditioning process

    International Nuclear Information System (INIS)

    It is known that high levels of radiation can cause significant damage by altering the properties of materials. A practical understanding of the effects of radiation - how radiation affects various types of materials and components - is required to design equipment to operate reliably in a gamma radiation environment. When designing equipment to operate in a high gamma radiation environment, such as will be present in a nuclear spent fuel handling facility, several important steps should be followed. To do this, the system design process in the radiation environment is reviewed. Also, through the investigation of the foreign literature, threshold values are generally reported. The threshold values are normally the dose required to begin degradation in a particular material property. In order to active test of the Advanced spent fuel Conditioning Process(ACP), the radiation-resistant analysis of the device for active test of the ACP are conducted, which is operated in the radiation environment called a hot-cell. For analyzing the radiation effects of the ACP, a gamma dose rates for a hot-cell environment of the ACP are calculated by using the SCALE 4.4 code. Also, the radiation effect analysis of the main device for the ACP such as vol-oxidation, reduction reactor and servo manipulator is performed

  14. Advanced dry head-end reprocessing of light water reactor spent nuclear fuel

    Science.gov (United States)

    Collins, Emory D.; Delcul, Guillermo D.; Hunt, Rodney D.; Johnson, Jared A.; Spencer, Barry B.

    2014-06-10

    A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450.degree. C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80.degree. C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

  15. Spent nuclear fuel storage

    International Nuclear Information System (INIS)

    When a country becomes self-sufficient in part of the nuclear cycle, as production of fuel that will be used in nuclear power plants for energy generation, it is necessary to pay attention for the best method of storing the spent fuel. Temporary storage of spent nuclear fuel is a necessary practice and is applied nowadays all over the world, so much in countries that have not been defined their plan for a definitive repository, as well for those that already put in practice such storage form. There are two main aspects that involve the spent fuels: one regarding the spent nuclear fuel storage intended to reprocessing and the other in which the spent fuel will be sent for final deposition when the definitive place is defined, correctly located, appropriately characterized as to several technical aspects, and licentiate. This last aspect can involve decades of studies because of the technical and normative definitions at a given country. In Brazil, the interest is linked with the storage of spent fuels that will not be reprocessed. This work analyses possible types of storage, the international panorama and a proposal for future construction of a spent nuclear fuel temporary storage place in the country. (author)

  16. Analysis on the shielding ability of a hot cell to accommodate advanced spent fuel conditioning process

    International Nuclear Information System (INIS)

    A design work is conducting for the IMEF's future cell which located in the basement to use it as a demonstration facility for Advanced Spent Fuel Conditioning Process (ACP). Since the total radiation source which used in ACP is expected as approximately 10 times higher than the design criteria of IMEF, the existing concrete structure cannot meet the shielding requirements. Therefore, shielding design which reinforcing the shielding capability has carried out for the ACP hot cell to satisfy the shielding criteria for the expected maximum radioactivity of ACP. This study presents a shielding analysis results using QADS code for the reinforced shielding wall with heavy concrete, steel or lead, etc. As a results of the analysis, a shielding wall reinforcing method was proposed. Additional shielding analysis was performed for the ACP hot cell with proposed reinforced shielding design using MCNP-4C code, and the validity of radiation shielding design was evaluated

  17. Spent fuel workshop'2002

    International Nuclear Information System (INIS)

    This document gathers the transparencies of the presentations given at the 2002 spent fuel workshop: Session 1 - Research Projects: Overview on the IN CAN PROCESSES European project (M. Cowper), Overview on the SPENT FUEL STABILITY European project (C. Poinssot), Overview on the French R and D project on spent fuel long term evolution, PRECCI (C. Poinssot); Session 2 - Spent Fuel Oxidation: Oxidation of uranium dioxide single crystals (F. Garrido), Experimental results on SF oxidation and new modeling approach (L. Desgranges), LWR spent fuel oxidation - effects of burn-up and humidity (B. Hanson), An approach to modeling CANDU fuel oxidation under dry storage conditions (P. Taylor); Session 3 - Spent Fuel Dissolution Experiments: Overview on high burnup spent fuel dissolution studies at FZK/INE (A. Loida), Results on the influence of hydrogen on spent fuel leaching (K. Spahiu), Leaching of spent UO2 fuel under inert and reducing conditions (Y. Albinsson), Fuel corrosion investigation by electrochemical techniques (D. Wegen), A reanalysis of LWR spent fuel flow through dissolution tests (B. Hanson), U-bearing secondary phases formed during fuel corrosion (R. Finch), The near-field chemical conditions and spent fuel leaching (D. Cui), The release of radionuclides from spent fuel in bentonite block (S.S. Kim), Trace actinide behavior in altered spent fuel (E. Buck, B. Hanson); Session 4 - Radiolysis Issues: The effect of radiolysis on UO2 dissolution determined from electrochemical experiments with 238Pu doped UO2 M. Stroess-Gascoyne (F. King, J.S. Betteridge, F. Garisto), doped UO2 studies (V. Rondinella), Preliminary results of static and dynamic dissolution tests with α doped UO2 in Boom clay conditions (K. Lemmens), Studies of the behavior of UO2 / water interfaces under He2+ beam (C. Corbel), Alpha and gamma radiolysis effects on UO2 alteration in water (C. Jegou), Behavior of Pu-doped pellets in brines (M. Kelm), On the potential catalytic behavior of UO2(s

  18. Spent fuel management

    International Nuclear Information System (INIS)

    The production of nuclear electricity results in the generation of spent fuel that requires safe, secure and efficient management. Appropriate management of the resulting spent fuel is a key issue for the steady and sustainable growth of nuclear energy. Currently about 10,000 tonnes heavy metal (HM) of spent fuel are unloaded every year from nuclear power reactors worldwide, of which 8,500 t HM need to be stored (after accounting for reprocessed fuel). This is the largest continuous source of civil radioactive material generated, and needs to be managed appropriately. Member States have referred to storage periods of 100 years and even beyond, and as storage quantities and durations extend, new challenges arise in the institutional as well as in the technical area. The IAEA gives high priority to safe and effective spent fuel management. As an example of continuing efforts, the 2003 International Conference on Storage of Spent Fuel from Power Reactors gathered 125 participants from 35 member states to exchange information on this important subject. With its large number of Member States, the IAEA is well-positioned to gather and share information useful in addressing Member State priorities. IAEA activities on this topic include plans to produce technical documents as resources for a range of priority topics: spent fuel performance assessment and research, burnup credit applications, cask maintenance, cask loading optimization, long term storage requirements including records maintenance, economics, spent fuel treatment, remote technology, and influence of fuel design on spent fuel storage. In addition to broader topics, the IAEA supports coordinated research projects and technical cooperation projects focused on specific needs

  19. Spent fuel assembly hardware

    International Nuclear Information System (INIS)

    When spent nuclear fuel is disposed of in a repository, the waste package will include the spent fuel assembly hardware, the structural portion of the fuel assembly, and the fuel pins. The spent fuel assembly hardware is the subject of this paper. The basic constituent parts of the fuel assembly will be described with particular attention on the materials used in their construction. The results of laboratory analyses performed to determine radionuclide inventories and trace impurities also will be described. Much of this work has been incorporated into a US Department of Energy (DOE) database maintained by Oak Ridge National Laboratory (ORNL). This database is documented in DOE/RW-0184 and can be obtained from Karl Notz at ORNL. The database provides a single source for information regarding wastes that may be sent to the repository

  20. Spent fuel storage rack

    International Nuclear Information System (INIS)

    Constitution: A square cylinder for containing spent fuels is made of hafnium plates. Welding for the hafnium plates are conducted under vacuum or in inert gases by using electron beams or laser beams. By using hafnium as described above, neutron absorption is improved and square cylinders incorporating the spent fuels can be accumulated at a high density. Furthermore, by welding the hafnium plates under vacuum, embrittlement of the welded portions can be prevented. (Ikeda, J.)

  1. Innovative separation method for advanced spent fuel reprocessing based on tertiary pyridine resin

    International Nuclear Information System (INIS)

    Radiochemical separation experiments have been performed in order to realize a novel reprocessing method based on chromatography techniques using a novel pyridine resin. The newly synthesized tertiary pyridine resin with two functions of ion exchanger and soft-donor was dedicated to the experiments, where highly irradiated mixed oxide fuel from the experimental fast reactor JOYO was used as a reference spent fuel. With a 3-step separation, pure Am and Cm were individually obtained as minor actinide products, and 106Ru group, lanthanides with 137Cs group and Pu group were fractionated, respectively. The decontamination factor of 137Cs and trivalent lanthanides (155Eu, 144Ce) against the Am product exceeded 3.9 x 104 and 1.0 x 105, respectively. The decontamination factor as the mutual separation of 243Cm was larger than 2.2 x 103 against the Am product. Moreover, the content of 137Cs, trivalent lanthanides and 243Cm in Am product did not exceed 2 ppm. The tertiary pyridine resin method is a candidate separation system for an 'advanced ORIENT process', where enhanced separation, transmutation and utilization of actinides, long-lived fission products and rare metal fission product would be oriented. (author)

  2. TRIGA spent fuel storage

    International Nuclear Information System (INIS)

    Storage of spent fuel elements is a step preliminary to final radioactive waste disposal operation. The spent fuel issue will have a common solution for both spent fuel from Cernavoda NPP and research TRIGA reactors currently operated in Romania. For the case of TRIGA reactor spent fuel this will be an alternative solution to the now functioning alternative of 'on site' storing solution adopted so far at INR Pitesti. For the time being the short term storage requirements for TRIGA spent fuel are adequately fulfilled by the pool of a multizonal reactor, the construction of which was definitively stopped. On the other hand the HEU - LEU conversion of the 14 MW TRIGA reactor which will be completed till May 2006, will pose not spent fuel problems as the TRIGA HEU fuel (612 elements) will be transferred in US (not later than May 2009). Consequently, the needs for intermediate storage will be associated only with the LEU spent fuel from TRIGA LEU-SSR and TRIGA LEU-ACPR reactors. In the latter case the maximum number of elements will be 167. For the stationary 14 MW (SSR) reactor but the amount of fuel elements to be stored on a intermediate term will be a function of service span of this reactor as well of the degree of request. Totally, some 1,750 SSR-LEU fuel elements will require intermediate storage. There is a preliminary agreement with 'NUCLEARELECTRICA -S.A.' Company regarding LEU TRIGA spent fuel storage at the intermediate storage facility for spent fuel of Cernavoda NPP.. A safety investigation is underway to determine the impact of LEU spent fuel upon the dry environment containing spent CANDU fuel. To fulfil the requirements imposed by CANDU storage technology the LEU spent fuel will be correspondingly conditioned. Then adequate containers will be used for transportation of fuel to Cernavoda's storage cell. Subcriticality condition in the storage cell loaded with LEU was checked by calculating the multiplication factor for an infinite lattice. The

  3. Status of Spent Fuel Management in Korea

    International Nuclear Information System (INIS)

    Spent fuel generated from nuclear power plants are currently stored in At Reactor water pools. Total spent fuel storage capacity of AERE facilities of 9 nuclear units is about 2,500 MTC. About 500 MTC of spent fuel has been discharged from the reactor since 1980. Existing AERE storage capacity of spent fuel will lose its reserve capacity in the middle of 1990's. Therefore the countermeasure for securing additional storage capacity of spent fuel should be sought. 'Wait and See' was our country's policy for management of spent fuel. But the safe containment and disposal of spent fuel have become much important and imminent issue in Korea now. As a result, Atomic Energy Act was amended in May, 1986. By this amendment, government will take the responsibility of spent fuel management. The Korea Advanced Energy Research Institute will be authorized to carry out the spent fuel management and the related R and D activities. As interim measure for spent fuel management, AFSR facility will be constructed. Spent fuel will be transported to the AFSR facility from the middle of 1990s, and will be stored for some time period. 'Wait and See' is still considered to be more appropriate option for a long-term fuel management plan in Korea, because the introduction of FBRF in this century seems not possible

  4. Advanced spent-fuel waste package fill material: Depleted uranium dioxide

    International Nuclear Information System (INIS)

    The use of depleted uranium dioxide (DUO2) particles has been investigated as fill material inside repository waste packages containing light water reactor (LWR) spent nuclear fuel (SNF). The use of DUO2 fill may eliminate repository criticality concerns, reduce radionuclide release rates from the repository, and dispose of excess depleted uranium

  5. Spent Fuel in Chile

    International Nuclear Information System (INIS)

    The government has made a complete and serious study of many different aspects and possible road maps for nuclear electric power with strong emphasis on safety and energy independence. In the study, the chapter of SFM has not been a relevant issue at this early stage due to the fact that it has been left for later implementation stage. This paper deals with the options Chile might consider in managing its Spent Fuel taking into account foreign experience and factors related to safety, economics, public acceptance and possible novel approaches in spent fuel treatment. The country’s distinctiveness and past experience in this area taking into account that Chile has two research reactors which will have an influence in the design of the Spent Fuel option. (author)

  6. Spent fuel pyroprocessing demonstration

    International Nuclear Information System (INIS)

    A major element of the shutdown of the US liquid metal reactor development program is managing the sodium-bonded spent metallic fuel from the Experimental Breeder Reactor-II to meet US environmental laws. Argonne National Laboratory has refurbished and equipped an existing hot cell facility for treating the spent fuel by a high-temperature electrochemical process commonly called pyroprocessing. Four products will be produced for storage and disposal. Two high-level waste forms will be produced and qualified for disposal of the fission and activation products. Uranium and transuranium alloys will be produced for storage pending a decision by the US Department of Energy on the fate of its plutonium and enriched uranium. Together these activities will demonstrate a unique electrochemical treatment technology for spent nuclear fuel. This technology potentially has significant economic and technical advantages over either conventional reprocessing or direct disposal as a high-level waste option

  7. Advanced techniques for storage and disposal of spent fuel from commercial nuclear power plants

    International Nuclear Information System (INIS)

    Electricity generation using fossil fuel at comparatively low costs forces nuclear energy to explore all economic potentials. The cost advantage of direct disposal of spent nuclear fuel compared to reprocessing gives reason enough to follow that path more and more. The present paper describes components and facilities for long-term storage as well as packaging strategies, developed and implemented under the responsibility of the German utilities operating nuclear power plants. A proposal is made to complement or even to replace the POLLUX cask concept by a system using BSK 3 fuel rod containers together with LB 21 storage casks. (author)

  8. Burnup credit demands for spent fuel management in Ukraine

    International Nuclear Information System (INIS)

    In fact, till now, burnup credit has not be applied in Ukrainian nuclear power for spent fuel management systems (storage and transport). However, application of advanced fuel at VVER reactors, arising spent fuel amounts, represent burnup credit as an important resource to decrease spent fuel management costs. The paper describes spent fuel management status in Ukraine from viewpoint of subcriticality assurance under spent fuel storage and transport. It also considers: 1. Regulation basis concerning subcriticality assurance, 2. Basic spent fuel and transport casks characteristics, 3. Possibilities and demands for burnup credit application at spent fuel management systems in Ukraine. (author)

  9. The advanced fuel cycle initiative: the future path for advanced spent fuel treatment and transmutation research in the United States

    International Nuclear Information System (INIS)

    The U. S. Department of Energy (DOE) has invested over USD 100 million in transmutation research and development over the past three years. The programme has evolved from an accelerator based transmutation programme to a multi-tier reactor and accelerator based programme. These changes have resulted in a significant re-focus of the research and development programme as well as a name change to reflect the new direction. The Advanced Accelerator Application (AAA) programme is now renamed the Advanced Fuel Cycle Initiative (AFCI). Research completed by the AAA programme in Fiscal Year 2002 points to a multi-phased AFCI Programme consisting of two elements that would be conducted in parallel as part of an integrated research effort: an intermediate-term technology element (AFCI Series One), which emphasises advanced technical enhancements to the current commercial nuclear power infrastructure; and a long term technology element (AFCI Series Two), which will require the introduction of next-generation nuclear energy systems to reduce the toxicity of nuclear waste. (author)

  10. Spent fuel management in Argentina

    International Nuclear Information System (INIS)

    The general program on Argentinian Spent Fuel Management has been informed in previous meetings and IAEA publications. This presentation includes an updating of the programs and a short description of the dry storage of Embalse NPP spent fuel. (author)

  11. Spent fuel management in Switzerland

    International Nuclear Information System (INIS)

    Switzerland currently has 3000 MWe being delivered from five nuclear power units. The current spent fuel management and waste disposal programme includes reprocessing of the spent fuel generated up to 1990. The plan for intermediate storage of spent fuel, high-level waste and low/medium level wastes is underway; the main features of the centralized storage facilities are given. (author). 1 fig

  12. Advanced Non-Destructive Assay Systems and Special Instrumentation Requirements for Spent Nuclear Fuel Recycling Facilities

    International Nuclear Information System (INIS)

    The safe and efficient operation of the next generation of Spent Nuclear Fuel (SNF) recycling / reprocessing facilities is dependent upon the availability of high performance real time Non- Destructive Assay (NDA) systems at key in-line points. A diverse variety of such special instrument systems have been developed and commissioned at reprocessing plants worldwide over the past fifty years.. The measurement purpose, technique and plant performance for selected key systems have been reviewed. Obsolescence issues and areas for development are identified in the context of the measurements needs of future recycling facilities and their associated waste treatment plants. Areas of concern include (i) Materials Accountancy and Safeguards, (ii) Head End process control and feed envelope verification, (iii) Real-time monitoring at the Product Finishing Stages, (iv) Criticality safety and (v) Radioactive waste characterization. Common characteristics of the traditional NDA systems in historical recycling facilities are (i) In-house development of bespoke instruments resulting in equipment that if often unique to a given facility and generally not commercially available, (ii) Use of 'novel' techniques - not widely deployed in other applications, (iii) Design features that are tailored to the specific plant requirements of the facility operator, (iv) Systems and software implementation that was not always carried out to modern industry standards and (v) A tendency to be overly complex - refined by on-plant operational usage and experience. Although these systems were 'validated in use' and are generally fit for purpose, there are a number of potential problems in transferring technology that was developed ten or more years ago to the new build SNF recycling facilities of the future. These issues include (i) Obsolescence of components - particularly with respect to computer hardware and data acquisition electronics, (ii) Availability of Intellectual Property and design

  13. Preliminary conceptual designs for advanced packages for the geologic disposal of spent fuel

    International Nuclear Information System (INIS)

    The present study assumes that the spent fuel will be disposed of in mined repositories in continental geologic formations, and that the post-emplacement control of the radioactive species will be accomplished independently by both the natural barrier, i.e., the geosphere, and the engineered barrier system, i.e., the package components consisting of the stabilizer, the canister, and the overpack; and the barrier components external to the package consisting of the hole sleeve and the backfill medium. The present document provides an overview of the nature of the spent fuel waste; the general approach to waste containment, using the defense-in-depth philosophy; material options, both metallic and nonmetallic, for the components of the engineered barrier system; a set of strawman criteria to guide the development of package/engineered barrier systems; and four preliminary concepts representing differing approaches to the solution of the containment problem. These concepts use: a corrosion-resistant meta canister in a special backfill (2 barriers); a mild steel canister in a corrosion-resistant metallic or nonmetallic hole sleeve, surrounded by a special backfill (2 barriers); a corrosion-resistant canister and a corrosion-resistant overpack (or hole sleeve) in a special backfill (3 barriers); and a mild steel canister in a massive corrosion-resistant bore sleeve surrounded by a polymer layer and a special backfill (3 barriers). The lack of definitive performance requirements makes it impossible to evaluate these concepts on a functional basis at the present time

  14. Remote technology applications in spent fuel management

    International Nuclear Information System (INIS)

    Spent fuel management has become a prospective area for application of remote technology in recent years with a steadily growing inventory of spent fuel arising from nuclear power production. A remark that could be made from the review of technical information collected from the IAEA meetings was that remote technology in spent fuel management has matured well through the past decades of industrial experiences. Various remote technologies have been developed and applied in the past for facility operation and maintenance work in spent fuel examination, storage, transportation, reprocessing and radioactive waste treatment, among others, with significant accomplishments in dose reduction to workers, enhancement of reliability, etc. While some developmental activities are continuing for more advanced applications, industrial practices have made use of simple and robust designs for most of the remote systems technology applications to spent fuel management. In the current state of affairs, equipment and services in remote technology are available in the market for applications to most of the projects in spent fuel management. It can be concluded that the issue of critical importance in remote systems engineering is to make an optimal selection of technology and equipment that would best satisfy the as low as reasonably achievable (ALARA) requirements in terms of relevant criteria like dose reduction, reliability, costs, etc. In fact, good selection methodology is the key to efficient implementation of remote systems applications in the modern globalized market. This TECDOC gives a review of the current status of remote technology applications for spent fuel management, based on country reports from some Member States presented at the consultancy meetings, of which updated reports are attached in the annex. The scope of the review covers the series of spent fuel handling operations involved in spent fuel management, from discharge from reactor to reprocessing or

  15. Neutronic Analysis of Advanced SFR Burner Cores using Deep-Burn PWR Spent Fuel TRU Feed

    International Nuclear Information System (INIS)

    In this work, an advanced sodium-cooled fast TRU (Transuranics) burner core using deep-burn TRU feed composition discharged from small LWR cores was neutronically analyzed to show the effects of deeply burned TRU feed composition on the performances of sodium-cooled fast burner core. We consider a nuclear park that is comprised of the commercial PWRs, small PWRs of 100MWe for TRU deep burning using FCM (Fully Ceramic Micro-encapsulated) fuels and advanced sodium-cooled fast burners for their synergistic combination for effective TRU burning. In the small PWR core having long cycle length of 4.0 EFPYs, deep burning of TRU up to 35% is achieved with FCM fuel pins whose TRISO particle fuels contain TRUs in their central kernel. In this paper, we analyzed the performances of the advanced SFR burner cores using TRU feeds discharged from the small long cycle PWR deep-burn cores. Also, we analyzed the effect of cooling time for the TRU feeds on the SFR burner core. The results showed that the TRU feed composition from FCM fuel pins of the small long cycle PWR core can be effectively used into the advanced SFR burner core by significantly reducing the burnup reactivity swing which reduces smaller number of control rod assemblies to satisfy all the conditions for the self controllability than the TRU feed composition discharged from the typical PWR cores

  16. Spent fuel storage at KURRI

    International Nuclear Information System (INIS)

    The Research Reactor Institute, Kyoto University (KURRI) has more than 200 MTR-type spent fuel elements stored in water pools. The longest pool residence time is 21 years at present. The integrity of spent fuel elements have been confirmed by a visual inspection and a sipping test. The spent fuel elements should be reprocessed in accordance with KURRI's policy. KURRI is now negotiating with a reprocessing plant to make a contract, as considering the consequences in U.S. (author)

  17. Spent fuel management in Switzerland

    International Nuclear Information System (INIS)

    Switzerland currently has 3000 MWe being delivered from five nuclear power plants. Two more 1000 MWe power plants are firmly planned. The current spent fuel management and disposal programme including contracts for reprocessing of all spent fuel generated up to 1990 is presented. The plan for intermediate storage of spent nuclear fuel away from the nuclear power plants, as well as the storage of vitrified high-level wastes is given. (author)

  18. Spent Fuel Management in Bulgaria

    International Nuclear Information System (INIS)

    The report presents the legislative framework in the Republic of Bulgaria for spent fuel (SF) management; storage facilities for spent fuel (at reactor spent fuel storage/reactor pond, away from reactor spent fuel storage facility (SFSF) and the dry storage facility), as well as the SF transportation back to Russia. The policy of the Republic of Bulgaria regarding the management of SF and radioactive wastes (RAW) has been based on the moral principle of avoiding to impose undue burdens on future generations. (author)

  19. Spent-fuel-storage alternatives

    International Nuclear Information System (INIS)

    The Spent Fuel Storage Alternatives meeting was a technical forum in which 37 experts from 12 states discussed storage alternatives that are available or are under development. The subject matter was divided into the following five areas: techniques for increasing fuel storage density; dry storage of spent fuel; fuel characterization and conditioning; fuel storage operating experience; and storage and transport economics. Nineteen of the 21 papers which were presented at this meeting are included in this Proceedings. These have been abstracted and indexed

  20. Advances in Development of the Fission Product Extraction Process for the Separation of Cesium and Strontium from Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    JAck D. Law

    2007-09-01

    The Fission Product Extraction (FPEX) Process is being developed as part of the United States Department of Energy Advanced Fuel Cycle Initiative for the simultaneous separation of cesium (Cs) and strontium (Sr) from spent light water reactor (LWR) fuel. Separation of the Cs and Sr will reduce the short-term heat load in a geological repository, and when combined with the separation of americium (Am) and curium (Cm), could increase the capacity of the geological repository by a factor of approximately 100. The FPEX process is based on two highly specific extractants: 4,4',(5')-Di-(t-butyldicyclo-hexano)-18-crown-6 (DtBuCH18C6) and Calix[4]arene-bis-(tert-octylbenzo-crown-6) (BOBCalixC6). The DtBuCH18C6 extractant is selective for strontium and the BOBCalixC6 extractant is selective for cesium. Results of flowsheet testing of the FPEX process with a simulated feed solution in 3.3-cm centrifugal contactors are detailed. Removal efficiencies, distribution coefficient data, coextraction of metals, and process hydrodynamic performance are discussed along with recommendations for future flowsheet testing with actual spent nuclear fuel.

  1. WWER spent fuel storage

    International Nuclear Information System (INIS)

    Selection criteria for PAKS NPP dry storage system are outlined. They include the following: fuel temperature in storage; sub-criticality assurance (avoidance of criticality for fuel in the unirradiated condition without having to take credit for burn-up); assurance of decay heat removal; dose uptake to the operators and public; protection of environment; volume of waste produced during operation and decommissioning; physical protection of stored irradiated fuel assemblies; IAEA safeguards assurance; storage system versus final disposal route; cost of construction and extent of technology transfer to Hungarian industry. Several available systems are evaluated against these criteria, and as a result the GEC ALSTHOM Modular Vault Dry Store (MVDS) system has been selected. The MVDS is a passively cooled dry storage facility. Its most important technical, safety, licensing and technology transfer characteristics are outlined. On the basis of the experience gained some key questions and considerations related to the East European perspective in the field of spent fuel storage are discussed. 8 figs

  2. Spent fuel management in Ukraine and spent fuel data tracking

    International Nuclear Information System (INIS)

    Ukraine has eleven WWER-1000 and two WWER-440 operating reactors at four nuclear plants. These reactors generated almost 45% of Ukraine's electricity. The last of the three RBMK-1000 reactors of Chornobyl NPP was shut down on December 15, 2000. Two WWER-1000 units (one at Khmelnytskyy NPP and another at Rivne NPP) are under construction. According to the Spent Fuel Management Program of Ukraine, which was approved in 2000, the state policy in the spent fuel management field is 'wait and see'. In order to implement this state policy the following problems should be solved: Construction of interim spent fuel storage facilities; Provision of spent fuel transportation from the reactor site to the interim storage facility; Provision of scientific and technical support of the spent fuel management. (author)

  3. Spent fuel management in Japan

    International Nuclear Information System (INIS)

    In Japan 52 commercial nuclear power units are now operated, and the total power generation capacity is about 45 GWe. The cumulative amount of spent fuel arising is about 13,500 tU as of March 1997. Spent fuel is reprocessed, and recovered nuclear materials are to be recycled in LWRs and FBRs. In February 1997 short-term policy measures were announced by the Atomic Energy Commission, which addressed promotion of reprocessing programme in Rokkasho, plutonium utilization in LWRs, spent fuel management, backend measures and FBR development. With regard to the spent fuel management, the policy measures included expansion of spent fuel storage capacity at reactor sites and a study on spent fuel storage away from reactor sites, considering the increasing amount of spent fuel arising. Research and development on spent fuel storage has been carried out, particularly on dry storage technology. Fundamental studies are also conducted to implement the burnup credit into the criticality safety design of storage and transportation casks. Rokkasho reprocessing plant is being constructed towards its commencement in 2003, and Pu utilization in LWRs will be started in 1999. Research and development of future recycling technology are also continued for the establishment of nuclear fuel cycle based on FBRs and LWRs. (author)

  4. Spent fuel storage facility

    International Nuclear Information System (INIS)

    A diffusion-preventive device for the radioactivity of pool water is disposed in a pool chamber for accommodating a spent fuel storage chamber. The diffusion-preventive device comprises an air washer and a recycling blower which discharges air in the air washer to the pool chamber. In this air washer, not-activated pure water, etc. are supplied. The recycling blower is driven to introduce the air in the pool chamber to the air washer, and water is sprayed from a nozzle to moisten the air. In this way, the vapor pressure in the pool chamber can be increased and the amount of vapor generated from the pool can be decreased. The amount of radioactivity transferring from the poor water to the air can thereby be decreased and the amount of radioactivity released to the atmospheric air by means of ventilation air conditioning device can be decreased. (I.N.)

  5. Criticality safety evaluation in Tokai reprocessing plant. High burn up LWR UO2 spent fuel and ATR MOX spent fuel

    International Nuclear Information System (INIS)

    This report presents criticality safety evaluation of each equipment in Tokai reprocessing plant for two types of spent fuels, High burn up 4.2 % enrichment U oxide spent fuel for light water reactor and U-Pu mixed oxide spent fuel for advanced thermal reactor. As a result, it was confirmed that the equipments were safe enough for two types of the spent fuels from view point of criticality safety of single unit and multiple units. (author)

  6. Advanced Reactor Technology Options for Utilization and Transmutation of Actinides in Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    Renewed interest in the potential of nuclear energy to contribute to a sustainable worldwide energy mix is strengthening the IAEA's statutory role in fostering the peaceful uses of nuclear energy, in particular the need for effective exchanges of information and collaborative research and technology development among Member States on advanced nuclear power technologies (Articles III-A.1 and III-A.3). The major challenges facing the long term development of nuclear energy as a part of the world's energy mix are improvement of the economic competitiveness, meeting increasingly stringent safety requirements, adhering to the criteria of sustainable development, and public acceptability. The concern linked to the long life of many of the radioisotopes generated from fission has led to increased R and D efforts to develop a technology aimed at reducing the amount of long lived radioactive waste through transmutation in fission reactors or accelerator driven hybrids. In recent years, in various countries and at an international level, more and more studies have been carried out on advanced and innovative waste management strategies (i.e. actinide separation and elimination). Within the framework of the Project on Technology Advances in Fast Reactors and Accelerator Driven Systems (http://www.iaea.org/inisnkm/nkm/aws/fnss/index.html), the IAEA initiated a number of activities on utilization of plutonium and transmutation of long lived radioactive waste, accelerator driven systems, thorium fuel options, innovative nuclear reactors and fuel cycles, non-conventional nuclear energy systems, and fusion/fission hybrids. These activities are implemented under the guidance and with the support of the IAEA Nuclear Energy Department's Technical Working Group on Fast Reactors (TWG-FR). This publication compiles the analyses and findings of the Coordinated Research Project (CRP) on Studies of Advanced Reactor Technology Options for Effective Incineration of Radioactive Waste (2002

  7. Management of Spent Fuel in Germany

    International Nuclear Information System (INIS)

    This presentation gives an overview on the inventory of radioactive waste and spent fuel in Germany, the state of commissioning of the on-site storages for spent fuel and the balance of reprocessing of spent fuel. (author)

  8. Disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    This report addresses the topic of the mined geologic disposal of spent nuclear fuel from Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). Although some fuel processing options are identified, most of the information in this report relates to the isolation of spent fuel in the form it is removed from the reactor. The characteristics of the waste management system and research which relate to spent fuel isolation are discussed. The differences between spent fuel and processed HLW which impact the waste isolation system are defined and evaluated for the nature and extent of that impact. What is known and what needs to be determined about spent fuel as a waste form to design a viable waste isolation system is presented. Other waste forms and programs such as geologic exploration, site characterization and licensing which are generic to all waste forms are also discussed. R and D is being carried out to establish the technical information to develop the methods used for disposal of spent fuel. All evidence to date indicates that there is no reason, based on safety considerations, that spent fuel should not be disposed of as a waste

  9. Spent fuel treatment in Japan

    International Nuclear Information System (INIS)

    In Japan, 52 nuclear power reactors are operating with a total power generation capacity of 45 GWe. The cumulative amount of spent fuel arising, as of March 1998, is about 14,700 W. Spent fuel is reprocessed and recovered nuclear materials are to be recycled in LWRs and FBRs. Pu utilization in LWRs will commence in 1999. In January 1997, short-term policy measures were announced by the Atomic Energy Commission, which addressed promotion of the reprocessing programme in Rokkasho, plutonium utilization in LWRs, spent fuel management, back-end measures and FBR development. With regard to the spent fuel management, the policy measures included expansion of spent fuel storage capacity at reactor sites and a study on spent fuel storage away-from-reactor sites, considering the increasing amount of spent fuel arising. Valuable experience was been accumulated at the Tokai Reprocessing Plant (TRP), from the start of hot operation in 1977 up to now. The role of the TRP will be changed from an operation-oriented to a more R and D oriented facility, when PNC is reorganized into the new organization JNC. The Rokkasho reprocessing plant is under construction and is expected to commence operation in 2003. R and D of future recycling technologies is also continued for the establishment of a nuclear fuel cycle based on FBRs and LWRs. (author)

  10. Disposal of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    1979-12-01

    This report addresses the topic of the mined geologic disposal of spent nuclear fuel from Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). Although some fuel processing options are identified, most of the information in this report relates to the isolation of spent fuel in the form it is removed from the reactor. The characteristics of the waste management system and research which relate to spent fuel isolation are discussed. The differences between spent fuel and processed HLW which impact the waste isolation system are defined and evaluated for the nature and extent of that impact. What is known and what needs to be determined about spent fuel as a waste form to design a viable waste isolation system is presented. Other waste forms and programs such as geologic exploration, site characterization and licensing which are generic to all waste forms are also discussed. R and D is being carried out to establish the technical information to develop the methods used for disposal of spent fuel. All evidence to date indicates that there is no reason, based on safety considerations, that spent fuel should not be disposed of as a waste.

  11. IAEA spent fuel storage glossary

    International Nuclear Information System (INIS)

    The aim of this glossary is to provide a basis for improved international understanding of terms used in the important area of spent fuel storage technology. The glossary is the product of an IAEA Consultant Group with valuable input from a substantial list of reviewers. The glossary emphasizes fuel storage relevant to power reactors, but is also widely applicable to research reactors. The intention is to define terms from current technologies. Terms are limited to those directly related to spent fuel storage

  12. Spent fuel storage in India

    International Nuclear Information System (INIS)

    Full text: 1. Introduction: Indian Nuclear Power Programme has grown from twin BWR reactors at Tarapur to 12 PHWRs of 220 MW each working at various locations. Additionally six PHWR reactors are in advanced stage of construction. India has gone for closed nuclear fuel cycle option to reprocess the spent fuel for recovery of Uranium and Plutonium to meet ever increasing energy demand. There is a programme to achieve 20,000 MW installed nuclear power capacity by the year 2020. It is planned to construct Spent Fuel Storage Facilities (SFSFs) as the need arises. Wet storage of Spent Fuel has been the main mode of storage in India pending reprocessing. This paper describes various important issues related to design, construction, licensing and operational experience of spent fuel storage facilities at Tarapur and Kalpakkam. 2. Design of Spent Fuel Storage Facility (SFSF): The new SFSFs are located at existing nuclear site to take maximum advantage of existing infrastructure already in place, nearness to reactor and approved site for nuclear facility. Layout: The smooth handling of trailer loaded with shipping cask is ensured by providing two independent airlocks and 7 m wide road with proper turning radius. Location of cask decontamination area, pool water cooling and polishing system and effluent handling system have been suitably decided based on ease of operation and optimum space utilization. The active and in-active services have been suitably located. Seismic design: IAEA TECDOC-250 is followed for seismic design of SFSFs. The independent SFSF are designed for OBE (Operating Basis Earthquake) level of earthquake. The soil-structure interaction has been considered as per ASCE 4-98 standard. The pool structure has been designed for hydrodynamic response during seismic event. The design of various mechanical system and components is carried out as per the respective design codes and standards based on their safety classification and seismic categorization. Fuel Pool

  13. Assessment of spent fuel cooling

    Energy Technology Data Exchange (ETDEWEB)

    Ibarra, J.G.; Jones, W.R.; Lanik, G.F. [and others

    1997-02-01

    The paper presents the methodology, the findings, and the conclusions of a study that was done by the Nuclear Regulatory Commission`s Office for Analysis and Evaluation of Operational Data (AEOD) on loss of spent fuel pool cooling. The study involved an examination of spent fuel pool designs, operating experience, operating practices, and procedures. AEOD`s work was augmented in the area of statistics and probabilistic risk assessment by experts from the Idaho Nuclear Engineering Laboratory. Operating experience was integrated into a probabilistic risk assessment to gain insight on the risks from spent fuel pools.

  14. Spent-fuel-storage alternatives

    Energy Technology Data Exchange (ETDEWEB)

    1980-01-01

    The Spent Fuel Storage Alternatives meeting was a technical forum in which 37 experts from 12 states discussed storage alternatives that are available or are under development. The subject matter was divided into the following five areas: techniques for increasing fuel storage density; dry storage of spent fuel; fuel characterization and conditioning; fuel storage operating experience; and storage and transport economics. Nineteen of the 21 papers which were presented at this meeting are included in this Proceedings. These have been abstracted and indexed. (ATT)

  15. Spent fuel reprocessing. Main operations

    International Nuclear Information System (INIS)

    The behaviour of spent fuels from nuclear reactors is one of the major concern of the nuclear industry. Two alternatives exist: direct disposal or reprocessing, the choice between both strategies depends on political, economical and environmental stakes. This document is devoted to the description of the Purex process (plutonium uranium refining by extraction) used for the reprocessing of spent fuels from water cooled reactors: 1 - Stakes and strategies; 2 - characteristics of fuels: new fuels, conditions of irradiation, characteristics of irradiated fuels (PWR-type fuels, FBR-type fuels, GCR-type fuels, research and propulsion reactor fuels, amount of unloaded spent fuels); 3 - goals and specific constraints: technical goals (efficiency of uranium and plutonium recovery, specification of finite products, limitation of radioactive effluents, wastes conditioning), specific reprocessing constraints; 4 - general considerations about processes: evolution, principal steps; 5 - head-end operations: on reactor site, transport of spent fuel, unloading at the reprocessing plant, storage, mechanical treatments, dissolution (uranium oxide, mixed uranium-plutonium oxides, metal fuels), treatment of gaseous wastes (nitrogen oxides, iodine, filtering), clarifying of solutions (unsoluble particulates, apparatuses), evaluation of the nuclear materials content, adjustment of solutions; 6 - Separation and purification: chemistry (aqueous nitrous solutions, nitrogen compounds, actinides, fission products), extraction mechanisms (solvent properties, nitrogen compounds, actinides, fission products, solvent evolution), extraction cycles (first cycle, plutonium purification cycles, uranium purification cycles), solvent processing, apparatuses (mixers-settlers, pulsed columns, centrifugal extractors, modeling); 7 - Elaboration of finite products: uranium, plutonium (oxalate process, denitration), recycling of out-off specification plutonium oxides. (J.S.)

  16. Iraq spent fuel removal program

    International Nuclear Information System (INIS)

    The paper describes the preparation and operations associated with the removal of the 208 spent fuel assemblies from Iraq, with emphasis on the technical challenges that were overcome during this removal process. (author)

  17. Active Interrogation for Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Swinhoe, Martyn Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dougan, Arden [National Nuclear Security Administration (NNSA), Washington, DC (United States)

    2015-11-05

    The DDA instrument for nuclear safeguards is a fast, non-destructive assay, active neutron interrogation technique using an external 14 MeV DT neutron generator for characterization and verification of spent nuclear fuel assemblies.

  18. Spent fuel management in Argentina

    International Nuclear Information System (INIS)

    The general program on Argentinian Spent Fuel Management has been presented in the previous meeting. This presentation includes an updating of the programs and a short description of the mixed oxide rods pilot plant. (author). 1 fig., 5 photographs

  19. Intermodal transportation of spent fuel

    International Nuclear Information System (INIS)

    Concepts for transportation of spent fuel in rail casks from nuclear power plant sites with no rail service are under consideration by the US Department of Energy in the Commercial Spent Fuel Management program at the Pacific Northwest Laboratory. This report identifies and evaluates three alternative systems for intermodal transfer of spent fuel: heavy-haul truck to rail, barge to rail, and barge to heavy-haul truck. This report concludes that, with some modifications and provisions for new equipment, existing rail and marine systems can provide a transportation base for the intermodal transfer of spent fuel to federal interim storage facilities. Some needed land transportation support and loading and unloading equipment does not currently exist. There are insufficient shipping casks available at this time, but the industrial capability to meet projected needs appears adequate

  20. Development of spent fuel remote handling technology

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Ji Sup; Park, B. S.; Park, Y. S.; Oh, S. C.; Kim, S. H.; Cho, M. W.; Hong, D. H

    1997-12-01

    Since the nation`s policy on spent fuel management is not finalized, the technical items commonly required for safe management and recycling of spent fuel - remote technologies of transportation, inspection, maintenance, and disassembly of spent fuel - are selected and pursued. In this regards, the following R and D activities are carried out : collision free transportation of spent fuel assembly, mechanical disassembly of spent nuclear fuel and graphical simulation of fuel handling / disassembly process. (author). 36 refs., 16 tabs., 77 figs

  1. Development of spent fuel remote handling technology

    International Nuclear Information System (INIS)

    Since the nation's policy on spent fuel management is not finalized, the technical items commonly required for safe management and recycling of spent fuel - remote technologies of transportation, inspection, maintenance, and disassembly of spent fuel - are selected and pursued. In this regards, the following R and D activities are carried out : collision free transportation of spent fuel assembly, mechanical disassembly of spent nuclear fuel and graphical simulation of fuel handling / disassembly process. (author). 36 refs., 16 tabs., 77 figs

  2. Transportation of spent MTR fuels

    Energy Technology Data Exchange (ETDEWEB)

    Raisonnier, D.

    1997-08-01

    This paper gives an overview of the various aspects of MTR spent fuel transportation and provides in particular information about the on-going shipment of 4 spent fuel casks to the United States. Transnucleaire is a transport and Engineering Company created in 1963 at the request of the French Atomic Energy Commission. The company followed the growth of the world nuclear industry and has now six subsidiaries and affiliated companies established in countries with major nuclear programs.

  3. Recent advances in the development of a cobalt dicarbollide based solvent extraction process for the separation of Cs and Sr from spent fuel

    International Nuclear Information System (INIS)

    As part of the Advanced Fuel Cycle Initiative (AFCI), a chlorinated cobalt dicarbollide (CCD)/polyethylene glycol (PEG) based solvent extraction process is being developed for the separation of Cs and Sr from leached spent light water reactor (LWR) fuel. The separation of Cs and Sr would significantly reduce the short-term heat generation of spent nuclear fuel requiring geological disposal. Recent advances in the development of a CCD/PEG process will be presented. The data presented will include acid dependency data, results of batch contact testing using simulant feeds traced with 137Cs, 90Sr and 241Am as well as results of testing to evaluate extractant composition. The impacts of other separation process in an advanced aqueous separation flow sheet on the effectiveness of the CCD/PEG process will be detailed. (authors)

  4. Spent Fuel Management in Slovakia

    International Nuclear Information System (INIS)

    The paper describes the SFM system in the Slovak Republic. In 2008, the Slovak Government accepted in its Decision Nr. 328/2008 “The proposal on the strategy of the back-end of the nuclear power engineering”. The state supervision on nuclear safety of SFM is performed by the Nuclear Regulatory Authority of the Slovak Republik (UJD). The legislative framework in the Slovak Republic is based on acts and regulations. In Slovakia there are four nuclear power units in operation. The spent fuel is stored in at-reactor spent fuel storage pools and cooled by water with presence of the boric acid. After certain cooling time, the spent fuel is removed to the Interim Spent Fuel Storage Facility (ISFSF). For the spent fuel transport transportation container C-30 is used. UJD steers various research tasks under the Research & Development program (R&D). Several years ago we started process of burnup credit (BUC) implementation in Slovakia for VVER-440 reactors. Another R&D project is focused on determination of the relation between the spent fuel residual heat generation and surface temperature of the transport container C-30. By the end of 2009 first two modules — visual inspection and gamma spectroscopy — of inspection stand SVYP-440 at ISFSF were put into operation. (author)

  5. HFIR spent fuel management alternatives

    International Nuclear Information System (INIS)

    The High Flux Isotope Reactor (HFIR) at Martin Marietta Energy Systems' Oak Ridge National Laboratory (ORNL) has been unable to ship its spent fuel to Savannah River Site (SRS) for reprocessing since 1985. The HFIR storage pools are expected to fill up in the February 1994 to February 1995 time frame. If a management altemative to existing HFIR pool storage is not identified and implemented before the HFIR pools are full, the HFIR will be forced to shut down. This study investigated several alternatives for managing the HFIR spent fuel, attempting to identify options that could be implemented before the HFIR pools are full. The options investigated were: installing a dedicated dry cask storage facility at ORNL, increasing HFIR pool storage capacity by clearing the HFIR pools of debris and either close-packing or stacking the spent fuel elements, storing the spent fuel at another ORNL pool, storing the spent fuel in one or more hot cells at ORNL, and shipping the spent fuel offsite for reprocessing or storage elsewhere

  6. Spent Nuclear Fuel project, project management plan

    International Nuclear Information System (INIS)

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

  7. Spent Nuclear Fuel project, project management plan

    Energy Technology Data Exchange (ETDEWEB)

    Fuquay, B.J.

    1995-10-25

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

  8. Spent fuel transportation and storage experience

    International Nuclear Information System (INIS)

    Nuclear Packaging, a Pacific Nuclear Company, is the leading U. S. designer of radioactive material transport packages (casks and overpacks). In 1985, the company designed, fabricated and licensed the first new spent fuel transport container to go into service, in more than a decade. The Model 125-B [USA/9200/B(M)F] rail car mounted, composite lead cask was designed to satisfy unique and demanding requirements associated with transporting damaged nuclear fuel from Three Mile Island Unit 2. Nuclear Packaging has also joined other industry leaders in developing advanced alternative technologies for the interim storage of spent reactor fuel. The storage system design centers around the NuPac CP-9 cask, and emphasizes system economics, operational efficiency, licensability, and shielding effectiveness based on sound ALARA principals. This paper reviews and discusses the basis for these developmental programs, the design considerations and approach, the test program and licensing effort as well as the unique features of both spent fuel container systems

  9. Spent fuel management in Japan

    International Nuclear Information System (INIS)

    Japan has scarce energy resources and depends on foreign resources for 84% of its energy needs. Therefore, Japan has made efforts to utilize nuclear power as a key energy source since mid-1950's. Today, the nuclear energy produced from 49 nuclear power plants is responsible for about 31% of Japan's total electricity supply. The cumulative amount of spent fuel generated as of March 1995 was about 11,600 Mg U. Japan's policy of spent fuel management is to reprocess spent nuclear fuel and recycle recovered plutonium and uranium as nuclear fuel. The Tokai reprocessing plant continues stable operation keeping the annual treatment capacity or around 90 Mg U. A commercial reprocessing plant is under construction at Rokkasho, northern part of Japan. Although FBR is the principal reactor to use plutonium, LWR will be a major power source for some time and recycling of the fuel in LWRs will be prompted. (author). 3 figs

  10. Selection criteria for spent fuel storage technologies

    International Nuclear Information System (INIS)

    Fissile fuel material contained in cladding or encapsulation material that has been irradiated in a power reactor is considered spent fuel. There are several types of spent fuel, including pressurized water reactor (PWR) fuel, boiling water reactor (BWR) fuel, mixed oxide (MOX) fuel, Canada Deuterium uranium (CANDU) fuel, other pressurized heavy water reactor (PHWR) fuels, high temperature reactor (HTR) fuel and advanced gas cooled reactor (AGR) fuel. There are two principal alternatives for managing spent fuel: the direct disposal route: spent fuel, conditioned after a sufficient decay period, is directly disposed of without the separation of fissile components; and the reprocessing route: spent fuel is reprocessed, and high level waste (HLW) containing mostly fission products and a small proportion of the actinides is disposed of after proper conditioning. The initial phase for spent nuclear fuel is storage under wet conditions in reactor pools after discharge from the reactor cores. This cooling phase is necessary for further handling of the spent fuel. After this initial phase sooner or later spent fuel needs to be transferred to another storage facility. This situation occurs for two reasons: first there is not enough capacity in the reactor pool; and second, the reactor has to be decommissioned. Because of these reasons, away from reactor storage (AFR) technology has to be taken into consideration at the beginning. This paper presents the main factors for selection of storage technologies. These factors are important for determination of basic requirements (storage space and storage time). The main factors begin with the spent fuel characteristics. These are: fuel type and dimensions, enrichment ratio, rods, cooling time after discharge, cladding and other materials, fuel integrity and (short and long term) production amounts. Lifetime of the storage facility is another factor to be determined before the design stage. A reasonable lifetime is more than 100

  11. Intermodal transfer of spent fuel

    International Nuclear Information System (INIS)

    As a result of the international standardization of containerized cargo handling in ports around the world, maritime shipment handling is particularly uniform. Thus, handier exposure parameters will be relatively constant for ship-truck and ship-rail transfers at ports throughout the world. Inspectors' doses are expected to vary because of jurisdictional considerations. The results of this study should be applicable to truck-to-rail transfers. A study of the movement of spent fuel casks through ports, including the loading and unloading of containers from cargo vessels, afforded an opportunity to estimate the radiation doses to those individuals handling the spent fuels with doses to the public along subsequent transportation routes of the fuel. A number of states require redundant inspections and for escorts over long distances on highways; thus handlers, inspectors, escort personnel, and others who are not normally classified as radiation workers may sustain doses high enough to warrant concern about occupational safety. This paper addresses the question of radiation safety for these workers. Data were obtained during, observation of the offloading of reactor spent fuel (research reactor spent fuel, in this instance) which included estimates of exposure times and distances for handlers, inspectors and other workers during offloading and overnight storage. Exposure times and distance were also for other workers, including crane operators, scale operators, security personnel and truck drivers. RADTRAN calculational models and parameter values then facilitated estimation of the dose to workers during incident-free ship-to-truck transfer of spent fuel

  12. Nondestructive measurements on spent fuel for the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Nondestructive measurements on spent fuel are being developed to meet safeguards and materials managment requirements at nuclear facilities. Spent-fuel measurement technology and its applications are reviewed

  13. Overview of spent fuel management

    International Nuclear Information System (INIS)

    The annual arising of spent fuel at nuclear power plants worldwide was about 10,000 t HM (tonnes heavy metal) in 1993. The projected cumulative amount of spent fuel generated by the year 2010 will reach 330,000 t HM. Considering that part of it will be reprocessed, the amount of spent fuel to be stored by the year 2010 would be about 215,000 t HM. Since the first large scale geological repositories for the final disposal of spent fuel are not expected to be in operation before the year 2010 (an even greater delay is foreseen) and reprocessing capacity will not be sufficiently expanded, the indications are that interim storage will be the primary option at least for the next 20 years. Continuous attention is being given by international organizations to the collection, analysis and exchange of information on spent fuel storage. Their role in this area is to provide a forum for exchanging information, to co-ordinate and encourage closer co-operation among Member States in certain research and development activities that are of common interest, and to assist countries in responding to problems and finding solutions. (author). 2 refs, 1 fig., 6 tabs

  14. Spent fuel management in Argentina

    International Nuclear Information System (INIS)

    The current Argentine nuclear power programme consists of HWR reactors: two in operation (Atucha-I, 345 MWe and EMBALSE, 600 MWe) one 745 MWe is under construction and another one, 700 MWe will be installed before the end of the century. Plans for spent fuel storage and active programme for the utilization of Mixed Oxide (U-235, Pu-239) fuel which allows the development of technology for reprocessing and MOX fuel fabrication on a pilot plant are described. (author)

  15. Criticality of spent reactor fuel

    International Nuclear Information System (INIS)

    The storage capacity of spent reactor fuel pools can be greatly increased by consolidation. In this process, the fuel rods are removed from reactor fuel assemblies and are stored in close-packed arrays in a canister or skeleton. An earlier study examined criticality consideration for consolidation of Westinghouse fuel, assumed to be fresh, in canisters at the Millstone-2 spent-fuel pool and in the General Electric IF-300 shipping cask. The conclusions were that the fuel rods in the canister are so deficient in water that they are adequately subcritical, both in normal and in off-normal conditions. One potential accident, the water spill event, remained unresolved in the earlier study. A methodology is developed here for spent-fuel criticality and is applied to the water spill event. The methodology utilizes LEOPARD to compute few-group cross sections for the diffusion code PDQ7, which then is used to compute reactivity. These codes give results for fresh fuel that are in good agreement with KENO IV-NITAWL Monte Carlo results, which themselves are in good agreement with continuous energy Monte Carlo calculations. These methodologies are in reasonable agreement with critical measurements for undepleted fuel

  16. Thermalhydraulic analysis of spent fuel baskets

    International Nuclear Information System (INIS)

    This paper presents results from a thermalhydraulic modelling and analysis of the cooling water surrounding spent fuel baskets in the spent fuel bays. The spent fuel basket is a design option to provide more self-shielding features for spent fuel storage. Two CFD models containing the spent fuel baskets are presented using 3D finite volume elements. The first model is for spent fuel baskets in stacks to simulate them in storage bay. The second model is for the “tilter” which is a part of the spent fuel handling equipment in the reception bay. The CFD software Ansys-CFX was used to calculate heat transfer from the spent fuels to the cooling water. The analysis results of test cases demonstrate that these models can be used for detailed design assist for spent fuel handling systems and operations. (author)

  17. Extended storage of spent fuel

    International Nuclear Information System (INIS)

    This document is the final report on the IAEA Co-ordinated Research Programme on the Behaviour of Spent Fuel and Storage Facility Components during Long Term Storage (BEFAST-II, 1986-1991). It contains the results on wet and dry spent fuel storage technologies obtained from 16 organizations representing 13 countries who participated in the co-ordinated research programme. Considerable quantities of spent fuel continue to arise and accumulate. Many countries are investigating the option of extended spent fuel storage prior to reprocessing or fuel disposal. Wet storage continues to predominate as an established technology with the construction of additional away-from-reactor storage pools. However, dry storage is increasingly used with most participants considering dry storage concepts for the longer term. Depending on the cladding type options of dry storage in air or inert gas are proposed. Dry storage is becoming widely used as a supplement to wet storage for zirconium alloy clad oxide fuels. Storage periods as long as under wet conditions appear to be feasible. Dry storage will also continue to be used for Al clad and Magnox type fuel. Enhancement of wet storage capacity will remain an important activity. Rod consolidation to increase wet storage capacity will continue in the UK and is being evaluated for LWR fuel in the USA, and may start in some other countries. High density storage racks have been successfully introduced in many existing pools and are planned for future facilities. For extremely long wet storage (≥50 years), there is a need to continue work on fuel integrity investigations and LWR fuel performance modelling. it might be that pool component performance in some cases could be more limiting than the FA storage performance. It is desirable to make concerted efforts in the field of corrosion monitoring and prediction of fuel cladding and poll component behaviour in order to maintain good experience of wet storage. Refs, figs and tabs

  18. GNS spent fuel cask experience

    International Nuclear Information System (INIS)

    The Gesellschaft fuer Nuklear-Service mbH (GNS), which is owned by German utilities, is responsible for the management of spent fuel and nuclear waste on behalf of the German utilities operating nuclear power plants. This paper describes the spent reactor fuel and waste shipping and/or storage casks that GNS manufacturers for nuclear facilities in Germany, and worldwide. So far more than 30 different casks have been produced in quantities ranging from one to several hundred of each type. GNS participates in the German Support Program to assist the International Atomic Energy Agency (IAEA) in developing verification procedures for dry storage casks containing spent fuel. This activity is also summarized

  19. GNS spent fuel cask experience

    Energy Technology Data Exchange (ETDEWEB)

    Weh, R. (Gesellschaft fuer Nuklear-Service mbH, Hannover (Germany))

    1993-05-01

    The Gesellschaft fuer Nuklear-Service mbH (GNS), which is owned by German utilities, is responsible for the management of spent fuel and nuclear waste on behalf of the German utilities operating nuclear power plants. This paper describes the spent reactor fuel and waste shipping and/or storage casks that GNS manufacturers for nuclear facilities in Germany, and worldwide. So far more than 30 different casks have been produced in quantities ranging from one to several hundred of each type. GNS participates in the German Support Program to assist the International Atomic Energy Agency (IAEA) in developing verification procedures for dry storage casks containing spent fuel. This activity is also summarized.

  20. Review and comment on the advanced spent fuel management process (1): Technical aspects and non-proliferation concerns

    International Nuclear Information System (INIS)

    Efforts are made to analyze the project, the Advanced Spent Fuel Management Technology (ASFMT), which is currently carried out at Korea Atomic Energy Research Institute, on the technical feasibility and validity as well as on the nuclear non-proliferation concerns. The project is a part of a program under the 'Long and Midterm Nuclear Development Program'. On the technical analysis, reviewed the papers presented at the national and international meetings on the subject by KAERI staffs, and also participated to various technical discussions on the 'Mock-up Test', currently in progress. On the non-proliferation concerns, the ASFMT project was reviewed and analyzed in reference to various programs currently in progress or in a formulation stages in US, such as the DOE TOPS and ATW. Further reviewed the past JASNEC process and programs for possible application of the ASFMT project for JASNEC project. Provided a few thoughts for effectively carrying out the ASFMT project, and a plan for the next phase is presented

  1. Review and comment on the advanced spent fuel management process (1): Technical aspects and non-proliferation concerns

    Energy Technology Data Exchange (ETDEWEB)

    Song, Yo Taik

    2001-01-01

    Efforts are made to analyze the project, the Advanced Spent Fuel Management Technology (ASFMT), which is currently carried out at Korea Atomic Energy Research Institute, on the technical feasibility and validity as well as on the nuclear non-proliferation concerns. The project is a part of a program under the 'Long and Midterm Nuclear Development Program'. On the technical analysis, reviewed the papers presented at the national and international meetings on the subject by KAERI staffs, and also participated to various technical discussions on the 'Mock-up Test', currently in progress. On the non-proliferation concerns, the ASFMT project was reviewed and analyzed in reference to various programs currently in progress or in a formulation stages in US, such as the DOE TOPS and ATW. Further reviewed the past JASNEC process and programs for possible application of the ASFMT project for JASNEC project. Provided a few thoughts for effectively carrying out the ASFMT project, and a plan for the next phase is presented.

  2. Advanced head-end alternatives for the processing of US spent fuel

    International Nuclear Information System (INIS)

    The head-end is one of the costlier and probably less-developed components of a reprocessing plant. The overall objective of the advanced head-end is to replace the traditional chop-and-leach approach to control and minimise emissions, allowing for the up-front removal and optimised trapping of key fission products and the recovery of major constituents such as cladding for potential reuse. The basic features of the proposed advanced head-end are sufficiently mature for deployment and will most probably be required to achieve the expected regulatory rules for very low emissions (e.g. 3H, I and Kr) and waste minimization. Most advanced features need further development and can be sequentially implemented with no significant capital and operating costs given that would utilize most of the same equipment and facilities. (authors)

  3. Spent fuel storage

    International Nuclear Information System (INIS)

    To begin with, the author explains the reasons for intermediate storage of fuel elements in nuclear power stations and in a reprocessing plant and gives the temperature and radioactivity curves of LWR fuel elements after removal from the reactor. This is followed by a description of the facilities for fuel element storage in a reprocessing plant and of their functions. Futher topics are criticality and activity control, the problem of cooling time and safety systems. (HR)

  4. Spent fuel management in Poland

    International Nuclear Information System (INIS)

    Full text: The problems with spent nuclear fuel management were beginning in Poland at the moment of discharge of first fuel assembly used research reactor EWA at 1959. It was water-moderated reactor with Ek10 type fuel rods, operated at 2 MW thermal power. In 1966 the EWA reactor was rebuild and new type fuel assemblies: WWR-SM and WWR-M2 were used. The power of reactor was changed up to 10 MW. EWA reactor was operated until February 1995. Spent Ek-10 fuel rods and WWR spent fuel assemblies are stored in water pools in two away from reactor storage facilities. The basic design information on these facilities and experience from their operation are presented in the present presentation. In 1974 the new research reactor MARIA reached criticality and start operation from 1975. This is water beryllium moderated reactor with MR6 or MR5 fuel assemblies. The maximum thermal power is 30 MW. Up to 2000, the fuel assemblies with 80 % enrichment uranium were used. Now, the medium enrichment (36 %) fuel assemblies were used. The spent MR type fuel was stored in at reactor storage tank. The experience with operation of storage pool is presented. Taking into account the fact that the fuel was stored for the long period of time in wet condition, the wide program of physical investigation of spent fuel and storage facilities was carried out. The visual investigation of cladding material was performed. Only initial cladding corrosion was observed for the fuel with 20 years of storage, and the strong corrosion process was visible on elements stored for 30 years. It was observed that corrosion processes is faster if fuel meat is present under cladding material. On the basis of the results of the sipping tests of WWR-SM fuel assemblies with different storage time (from 4 to 31 years), the assessment of the time limit of their storage in wet condition has been made. Based on this assessment, the value of the daily leakage of Cs-137 from WWR-SM and WWR-M2 spent fuel in the wet pool was

  5. Dry spent nuclear fuel transfer

    International Nuclear Information System (INIS)

    Newport News Shipbuilding, (NNS), has been transferring spent nuclear fuel in a dry condition for over 25 years. It is because of this successful experience that NNS decided to venture into the design, construction and operation of a commercial dry fuel transfer project. NNS is developing a remote handling system for the dry transfer of spent nuclear fuel. The dry fuel transfer system is applicable to spent fuel pool-to-cask or cask-to-cask or both operations. It is designed to be compatible with existing storage cask technology as well as the developing multi-purpose canister design. The basis of NNS' design is simple. It must be capable of transferring all fuel designs, it must be capable of servicing 100 percent of the commercial nuclear plants, it must protect the public and nuclear operators, it must be operated cost efficiently and it must be transportable. Considering the basic design parameters, the following are more specific requirements included in the design: (a) Total weight of transfer cask less than 24 tons; (b) no requirement for permanent site modifications to support system utilization; (c) minimal radiation dose to operating personnel; (d) minimal generation of radioactive waste; (e) adaptability to any size and length fuel or cask; (f) portability of system allowing its efficient movement from site to site; (g) safe system; all possible ''off normal'' situations are being considered, and resultant safety systems are being engineered into NNS' design to mitigate problems. The primary focus of this presentation is to provide an overview of NNS' Dry Spent Nuclear Fuel Transfer System. (author). 5 refs

  6. Spent fuel integrity during transportation

    Energy Technology Data Exchange (ETDEWEB)

    Funk, C.W.; Jacobson, L.D.

    1980-01-01

    The conditions of recent shipments of light water reactor spent fuel were surveyed. The radioactivity level of cask coolant was examined in an attempt to find the effects of transportation on LWR fuel assemblies. Discussion included potential cladding integrity loss mechanisms, canning requirements, changes of radioactivity levels, and comparison of transportation in wet or dry media. Although integrity loss or degradation has not been identified, radioactivity levels usually increase during transportation, especially for leaking assemblies.

  7. Worldwide spent fuel transportation logistics

    International Nuclear Information System (INIS)

    This paper presents an overview of the worldwide transportation requirements for spent fuel. Included are estimates of numbers and types of shipments by mode and cask type for 1985 and the year 2000. In addition, projected capital and transportation costs are presented. For the year 1977 and prior years inclusive, there is a cumulative worldwide requirement for approximately 300 MTU of spent fuel storage at away-from-reactor (AFR) facilities. The cumulative requirements for years through 1985 are projected to be nearly 10,000 MTU, and for the years through 2000 the requirements are conservatively expected to exceed 60,000 MTU. These AFR requirements may be related directly to spent fuel transportation requirements. In total nearly 77,000 total cask shipments of spent fuel will be required between 1977 and 2000. These shipments will include truck, rail, and intermodal moves with many ocean and coastal water shipments. A limited number of shipments by air may also occur. The US fraction of these is expected to include 39,000 truck shipments and 14,000 rail shipments. European shipments to regional facilities are expected to be primarily by rail or water mode and are projected to account for 16,000 moves. Pacific basin shipments will account for 4500 moves. The remaining are from other regions. Over 400 casks will be needed to meet the transportation demands. Capital investment is expected to reach $800,000,000 in 1977 dollars. Cumulative transport costs will be a staggering $4.4 billion dollars

  8. NAC's Modular, Advanced Generation, Nuclear All-purpose STORage (MAGNASTOR) system: new generation multipurpose spent fuel storage for global application

    International Nuclear Information System (INIS)

    Multipurpose canister systems (MCS) have been designed, licensed, fabricated, constructed, and loaded over the last decade within the U.S. These systems are characterized as concrete-based storage overpacks containing transportable canisters utilizing redundantly welded closures. Canisters are designed and intended to be transferred into transport packagings for shipment off-site, and canister designs do not preclude their use in waste disposal overpacks. NAC has learned a number of significant lessons in the deployment of its first generation MCS. During this period prior to the next procurement phase, NAC has developed a new generation MCS, incorporating the lessons learned from the first generation while considering the capabilities of the plants populating the next phase. The system is identified as the Modular, Advanced Generation, Nuclear All-purpose STORage (MAGNASTOR) system, and this paper addresses its unique design, fabrication, and operations features. Among these are: a unique developed cell basket design, under patent review, that increases spent fuel capacities and simplifies fabrication while providing high strength and heat removal efficiency: a significantly enhanced canister closure design that improves welding time, personnel dose, and drying performance: a low profile vertical concrete cask design that improves on-site handling and site dose rates, offers tangible threat limitations for beyond-design-basis events, and maintains proven and simple construction/operation features: a simple, proven transfer system that facilitates transfer without excessive dose or handling: a new approach to water removal and canister drying, using a moisture entrainment, gas absorption vacuum (MEGAVAC) system. The paper includes design and licensing status of the MAGNASTOR system, and prototyping development that NAC has performed to date

  9. Advances in applications of burnup credit to enhance spent fuel transportation, storage, reprocessing and disposition. Proceedings of a technical meeting

    International Nuclear Information System (INIS)

    Given a trend towards higher burnup power reactor fuel, the IAEA began an active programme in burnup credit (BUC) with major meetings in 1997 (IAEA-TECDOC-1013), 2000 (IAEA-TECDOC-1241) and 2002 (IAEA-TECDOC-1378) exploring worldwide interest in using BUC in spent fuel management systems. This publication contains the proceedings of the IAEA's 4th major BUC meeting, held in London. Sixty participants from 18 countries addressed calculation methodology, validation and criticality, safety criteria, procedural compliance with safety criteria, benefits of BUC applications, and regulatory aspects in BUC. This meeting encouraged the IAEA to continue its activities on burnup credit including dissemination of related information, given the number of Member States having to deal with increased spent fuel quantities and extended durations. A 5th major meeting on burnup credit is planned 2008. Burnup credit is a concept that takes credit for the reduced reactivity of fuel discharged from the reactor to improve loading density of irradiated fuel assemblies in storage, transportation, and disposal applications, relative to the assumption of fresh fuel nuclide inventories in loading calculations. This report has described a general four phase approach to be considered in burnup credit implementation. Much if not all of the background research and data acquisition necessary for successful burnup credit development in preparation for licensing has been completed. Many fuel types, facilities, and analysis methods are encompassed in the public knowledge base, such that in many cases this guidance will provide a means for rapid development of a burnup credit program. For newer assembly designs, higher enrichment fuels, and more extensive nuclide credit, additional research and development may be necessary, but even this work can build on the foundation that has been established to date. Those, it is hoped that this report will serve as a starting point with sufficient reference to

  10. Spent fuel storage process equipment development

    International Nuclear Information System (INIS)

    Nuclear energy which is a major energy source of national energy supply entails spent fuels. Spent fuels which are high level radioactive meterials, are tricky to manage and need high technology. The objectives of this study are to establish and develop key elements of spent fuel management technologies: handling equipment and maintenance, process automation technology, colling system, and cleanup system. (author)

  11. Total quality in spent fuel pool reracking

    International Nuclear Information System (INIS)

    The nuclear utility environment is one of strict cost control under prescriptive regulations and increasing public scrutiny. This paper presents the results of A Total Quality approach, by a dedicated team, that addresses the need for increased on-site spent fuel storage in this environment. Innovations to spent fuel pool reracking, driven by utilities' specific technical needs and shrinking budgets, have resulted in both product improvements and lower prices. A Total Quality approach to the entire turnkey project is taken, thereby creating synergism and process efficiency in each of the major phases of the project: design and analysis, licensing, fabrication, installation and disposal. Specific technical advances and the proven quality of the team members minimizes risk to the utility and its shareholders and provides a complete, cost effective service. Proper evaluation of spent fuel storage methods and vendors requires a full understanding of currently available customer driven initiatives that reduce cost while improving quality. In all phases of a spent fuel reracking project, from new rack design and analysis through old rack disposal, the integration of diverse experts, at all levels and throughout all phases of a reracking project, better serves utility needs. This Total Quality environment in conjunction with many technical improvements results in a higher quality product at a lower cost

  12. Spent fuel storage criticality safety

    International Nuclear Information System (INIS)

    The safety aspects of the spent fuel storage pool of the Egyptian test and research reactor one (ET-R R-1) has to be assessed as part of a general overall safety evaluation to be included in a safety analysis report (SAR) for this reactor. The present work treats the criticality safety of the spent fuel storage pool. Conservative calculations based on using fresh fuel has been performed, as well as less conservative using burned fuel. The calculations include cross library generation for burned and fresh fuel for the ET-R R-1 fuel type. The WIMS-D 4 code has been used in library generation and burn up calculation the critically calculations are performed using the one dimensional transport code (ANISN) and the two dimensional diffusion code (DIXY2). The possibility of increasing the storage efficiency either by insertion of absorber sheets of soluble boron salts or by reduction of fuel rod separation has been studied. 8 figs., 2 tabs

  13. Interim storage facility for spent fuel

    International Nuclear Information System (INIS)

    The spent fuel generated from the operation of a nuclear power plant is to be treated in the reprocessing plant in Rokkasho, Aomori. At present, spent fuel is stored in the nuclear power plant until it is reprocessed. However the amount of spent fuel generated exceeds the capacity of the reprocessing plant. Hence an additional spent fuel storage facility is needed for the nuclear fuel cycle. The spent fuel interim storage facility is the first institution in Japan that stores spent fuel outside of the nuclear power plant site. Our company has received an order for internal equipment for this facility. This paper introduces an overview of the interim storage facility for spent fuel. (author)

  14. Return of spent TRIGA fuel

    International Nuclear Information System (INIS)

    Spent fuel from J. Stefan Institute TRIGA reactor was successfully shipped to the US in 1999. Totally 219 standard TRIGA fuel rods used in the reactor from 1966 to 1991 were shipped. Together with the experience interesting for other reactors preparing for shipment, the following aspects of the project are explained: training of all persons involved, organization (QA, responsibilities), pre-preparation of the fuel, characterization of the fuel elements (burn-up determination, inspection of physical integrity), technical preparation for the shipment, administrative preparation (environmental impact report, safety report, operating and emergency procedures, qualification of equipment, permit), loading of the shipment containers, transfer of the containers to the port, signing of the bill of lading and transfer of liability. The role of main parties involved (J. Stefan Institute, US-DOE, IAEA, NAC) is explained. According to the contract covering the first shipment, we intend to return also the remaining fuel elements after 2016. (author)

  15. Conceptual evaluation of metal storage cask for conditioned spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Seo, K. S.; Shin, H. S.; Lee, J. C.; Bang, K. S.; Kim, H. D.; Park, S. W. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2003-07-01

    The storage parameters of spent PWR fuel are radioactivity, heat power and its volume. Those values could be reduced to about a quarter by an Advanced spent fuel Conditioning Process (ACP). Firstly, a storage concept and scenario were established considering the characteristics of conditioned spent fuel. If the efficiency of the cooling system is improved and the appropriate quantities of the conditioned spent fuel are stored, the conditioned spent fuels could be stored at a four times higher level of spent fuel storage. One storage unit of conditioned spent fuel was designed to have its capacity equivalent to one PWR spent fuel. It was supposed that a metal storage cask has 7 baskets that can load 28 storage units. Those capacities means that 28 spent PWR fuels in metal casks can be stored. The conceptual evaluations of the critical, shielding, thermal and structural fields were performed. In conclusion, the results of the conceptual evaluations show that the proposed metal cask satisfied with the important design criteria at a smaller size than the existing systems.

  16. Spent fuel storage racks

    International Nuclear Information System (INIS)

    Purpose: To improve bandability and weldability and also increasing neutron absorption capacity, thereby greatly improving the density of fuel storage. Constitution: Cells come in three types: cells formed by square cylinders set on the foundation, staggered by each pitch in longitudinal and lateral directions; cells formed among these square cylinders; and cells formed by enclosing with a vertical plate, one of the outermost sides among the square cylinders. Furthermore, between the corners of each cylinder, angles are longitudinally attached by welding where needed. Each cylinder is constituted by assembling four B-10 austenite stainless steel plates into a square form and attaching B-10 austenite stainless steel patches longitudinally at their angles by electron-beam welding and laser welding. For the abovementioned stainless steel plates, patches and angles, austenite stainless steel added with 0.10 - 0.45 % of B-10 is used. (Kawakami, Y.)

  17. Near surface spent fuel storage: environmental issues

    International Nuclear Information System (INIS)

    Interim storage of spent fuel appears inevitable because of the lack of reprocessing plants and spent fuel repositories. This paper examines the environmental issues potentially associated with management of spent fuel before disposal or reprocessing in a reference scenario. The radiological impacts of spent fuel storage are limited to low-level releases of noble gases and iodine. Water needed for water basin storage of spent fuel and transportation accidents are considered; the need to minimize the distance travelled is pointed out. Resource commitments for construction of the storage facilities are analyzed

  18. Spent fuel corrosion and dissolution

    International Nuclear Information System (INIS)

    This paper presents the current status of the Swedish programme for the study of the corrosion of spent fuel in bicarbonate groundwaters. Results from the on-going experimental programme are presented and compared with the data base accumulated over the past ten years. Release of uranium and the other actinides was solubility-controlled under the semi-static type of experiments performed. The limiting solubility for uranium under oxic conditions was consistent with the hypothesis that the redox potential of the system is assumed to correspond to the U3O7/U3O8 transition. The measured release fractions for 137Cs, 90Sr and 99Tc are discussed and used to exemplify the probable dissolution and corrosion processes involved. A substantial part of the Swedish programme is directed to the characterization of spent fuel before and after corrosion tests. Recent results are presented on the identification of possible corrosion sites. (26 refs.) (au)

  19. Neutron analysis of spent fuel storage installation using parallel computing and advance discrete ordinates and Monte Carlo techniques.

    Science.gov (United States)

    Shedlock, Daniel; Haghighat, Alireza

    2005-01-01

    In the United States, the Nuclear Waste Policy Act of 1982 mandated centralised storage of spent nuclear fuel by 1988. However, the Yucca Mountain project is currently scheduled to start accepting spent nuclear fuel in 2010. Since many nuclear power plants were only designed for -10 y of spent fuel pool storage, > 35 plants have been forced into alternate means of spent fuel storage. In order to continue operation and make room in spent fuel pools, nuclear generators are turning towards independent spent fuel storage installations (ISFSIs). Typical vertical concrete ISFSIs are -6.1 m high and 3.3 m in diameter. The inherently large system, and the presence of thick concrete shields result in difficulties for both Monte Carlo (MC) and discrete ordinates (SN) calculations. MC calculations require significant variance reduction and multiple runs to obtain a detailed dose distribution. SN models need a large number of spatial meshes to accurately model the geometry and high quadrature orders to reduce ray effects, therefore, requiring significant amounts of computer memory and time. The use of various differencing schemes is needed to account for radial heterogeneity in material cross sections and densities. Two P3, S12, discrete ordinate, PENTRAN (parallel environment neutral-particle TRANsport) models were analysed and different MC models compared. A multigroup MCNP model was developed for direct comparison to the SN models. The biased A3MCNP (automated adjoint accelerated MCNP) and unbiased (MCNP) continuous energy MC models were developed to assess the adequacy of the CASK multigroup (22 neutron, 18 gamma) cross sections. The PENTRAN SN results are in close agreement (5%) with the multigroup MC results; however, they differ by -20-30% from the continuous-energy MC predictions. This large difference can be attributed to the expected difference between multigroup and continuous energy cross sections, and the fact that the CASK library is based on the old ENDF

  20. Neutron analysis of spent fuel storage installation using parallel computing and advance discrete ordinates and Monte Carlo techniques

    International Nuclear Information System (INIS)

    In the United States, the Nuclear Waste Policy Act of 1982 mandated centralised storage of spent nuclear fuel by 1988. However, the Yucca Mountain project is currently scheduled to start accepting spent nuclear fuel in 2010. Since many nuclear power plants were only designed for ∼10 y of spent fuel pool storage, >35 plants have been forced into alternate means of spent fuel storage. In order to continue operation and make room in spent fuel pools, nuclear generators are turning towards independent spent fuel storage installations (ISFSIs). Typical vertical concrete ISFSIs are ∼6.1 m high and 3.3 m in diameter. The inherently large system, and the presence of thick concrete shields result in difficulties for both Monte Carlo (MC) and discrete ordinates (SN) calculations. MC calculations require significant variance reduction and multiple runs to obtain a detailed dose distribution. SN models need a large number of spatial meshes to accurately model the geometry and high quadrature orders to reduce ray effects, therefore, requiring significant amounts of computer memory and time. The use of various differencing schemes is needed to account for radial heterogeneity in material cross sections and densities. Two P3, S12, discrete ordinate, PENTRAN (parallel environment neutral-particle Transport) models were analysed and different MC models compared. A multigroup MCNP model was developed for direct comparison to the S N models. The biased A 3MCNP (automated adjoint accelerated MCNP) and unbiased (MCNP) continuous energy MC models were developed to assess the adequacy of the CASK multigroup (22 neutron, 18 gamma) cross sections. The PENTRAN SN results are in close agreement (5%) with the multigroup MC results; however, they differ by ∼20-30% from the continuous-energy MC predictions. This large difference can be attributed to the expected difference between multigroup and continuous energy cross sections, and the fact that the CASK library is based on the old ENDF

  1. Mock-up facilities for the development of an advanced spent fuel management process using molten salt technology

    International Nuclear Information System (INIS)

    The Korea Atomic Energy Research Institute (KAERI) has investigated a new approach to spent fuel storage technology that would reduce the total storage volume and the amount of decay heat. The technology utilizes the reduction of oxide fuel to a metal to reduce the volume and preferentially removing the fission products to reduce the decay heat. The uranium oxide is reduced to uranium metal by Li metal in a molten LiCl salt bath. During the reduction process, fission products are dissolved into the LiCl bath and some of the highly radioactive elements, such as Sr and Cs, are preferentially removed from the bath. The reduced uranium metal is cast into an ingot, put into a storage capsule, and stored using conventional storage methods. The fission products are treated as high level radioactive wastes. Each process of the technology has been studied and analyzed for technical feasibility, and has come to the point for designing and constructing of the mock-up for a demonstration of the technology. This paper presents the detailed design of the mock-up of the system and operational characteristics, along with all the details of the equipment for the system. KAERI plans to use the mock-up for the demonstration using an in-active spent fuel specimen. (authors)

  2. Probability of spent fuel transportation accidents

    Energy Technology Data Exchange (ETDEWEB)

    McClure, J. D.

    1981-07-01

    The transported volume of spent fuel, incident/accident experience and accident environment probabilities were reviewed in order to provide an estimate of spent fuel accident probabilities. In particular, the accident review assessed the accident experience for large casks of the type that could transport spent (irradiated) nuclear fuel. This review determined that since 1971, the beginning of official US Department of Transportation record keeping for accidents/incidents, there has been one spent fuel transportation accident. This information, coupled with estimated annual shipping volumes for spent fuel, indicated an estimated annual probability of a spent fuel transport accident of 5 x 10/sup -7/ spent fuel accidents per mile. This is consistent with ordinary truck accident rates. A comparison of accident environments and regulatory test environments suggests that the probability of truck accidents exceeding regulatory test for impact is approximately 10/sup -9//mile.

  3. Spent Fuel Working Group Report

    International Nuclear Information System (INIS)

    The Department of Energy is storing large amounts of spent nuclear fuel and other reactor irradiated nuclear materials (herein referred to as RINM). In the past, the Department reprocessed RINM to recover plutonium, tritium, and other isotopes. However, the Department has ceased or is phasing out reprocessing operations. As a consequence, Department facilities designed, constructed, and operated to store RINM for relatively short periods of time now store RINM, pending decisions on the disposition of these materials. The extended use of the facilities, combined with their known degradation and that of their stored materials, has led to uncertainties about safety. To ensure that extended storage is safe (i.e., that protection exists for workers, the public, and the environment), the conditions of these storage facilities had to be assessed. The compelling need for such an assessment led to the Secretary's initiative on spent fuel, which is the subject of this report. This report comprises three volumes: Volume I; Summary Results of the Spent Fuel Working Group Evaluation; Volume II, Working Group Assessment Team Reports and Protocol; Volume III; Operating Contractor Site Team Reports. This volume presents the overall results of the Working Group's Evaluation. The group assessed 66 facilities spread across 11 sites. It identified: (1) facilities that should be considered for priority attention. (2) programmatic issues to be considered in decision making about interim storage plans and (3) specific vulnerabilities for some of these facilities

  4. Spent nuclear fuel sampling strategy

    International Nuclear Information System (INIS)

    This report proposes a strategy for sampling the spent nuclear fuel (SNF) stored in the 105-K Basins (105-K East and 105-K West). This strategy will support decisions concerning the path forward SNF disposition efforts in the following areas: (1) SNF isolation activities such as repackaging/overpacking to a newly constructed staging facility; (2) conditioning processes for fuel stabilization; and (3) interim storage options. This strategy was developed without following the Data Quality Objective (DQO) methodology. It is, however, intended to augment the SNF project DQOS. The SNF sampling is derived by evaluating the current storage condition of the SNF and the factors that effected SNF corrosion/degradation

  5. Spent fuel storage requirements 1987

    International Nuclear Information System (INIS)

    Historical inventories of spent fuel and utility estimates of future discharges from US commercial nuclear reactors are presented through the year 2005. The ultimate needs for additional storage capacity are estimated. These estimtes are based on the maximum capacities within current and planned at-reactor facilities and on any planned transshipments of fuel to other reactors or facilities. Historical data through December, 1986, and projected discharges through the end of reactor life are used in this analysis. The source data was supplied by the utilities to the DOE Energy Information Administration (EIA) through the 1987 RW-859 data survey. 14 refs., 4 figs., 9 tabs

  6. Proceedings of spent fuel management technology workshop, 1997. 11. 13 - 11. 14, Taejon, Korea

    International Nuclear Information System (INIS)

    This proceedings cover the advanced spent fuel process technology, the development of a test facility for spent fuel management and remote handling technology, and the characteristics test technology. Fifteen papers are submitted

  7. Proceedings of spent fuel management technology workshop, 1997. 11. 13 - 11. 14, Taejon, Korea

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-31

    This proceedings cover the advanced spent fuel process technology, the development of a test facility for spent fuel management and remote handling technology, and the characteristics test technology. Fifteen papers are submitted.

  8. Sealed can of spent fuel

    International Nuclear Information System (INIS)

    Object: To provide a seal plug cover with a gripping portion fitted to a canning machine and a gripping portion fitted to a gripper of the same configuration as a fuel body for handling the fuel body so as to facilitate the handling work. Structure: A sealed can comprises a vessel and a seal plug cover, said cover being substantially in the form of a bottomed cylinder, which is slipped on the vessel and air-tightly secured by a fastening bolt between it and a flange. The spent fuel body is received into the vessel together with coolant during the step of canning operation. Said seal plug cover has two gripping portions, one for opening and closing the plug cover of the canning machine as an exclusive use member, the other being in the form of a hook-shaped peripheral groove, whereby the gripping portions may be effectively used using the same gripper when the spent fuel body is transported while being received in the sealed can or when the fuel body is removed from the sealed can. (Kawakami, Y.)

  9. Spent fuel management in France

    International Nuclear Information System (INIS)

    Spent fuel management in France is characterized by the reprocessing option; reprocessing being one of the major nuclear fuel industries developed to keep pace with the upgrowth of the national nuclear power program based on light water reactors and subsequently on fast breeder reactors. France has launched two industrial projects which will raise the annual reprocessing capacity of La Hague facility to 1600t of water reactor irradiated fuel, based on reprocessing experience and R and D work: - the UP3 plant (800t/year) which will be commissioned in 1989 - the UP2 800 plant, an extension of the existing plant, which will be commissioned in 1991. R and D work is also in progress, together with the extension of the Marcoule Pilot Plant (TOR project), and preparations made for the industrial reprocessing of fuel from the Super Phenix reactor and the first fast breeder reactors (MAR 600 project). 30 refs

  10. Spent fuel management in the Slovak Republic

    International Nuclear Information System (INIS)

    Presentation describes the history, present and future of the spent fuel management in the Slovak Republic. First experiences with spent fuel were gained in the seventies. Spent fuel form A-1 NPP was handled at Jaslovske Bohunice site, in order to prepare the spent fuel for the transport to the former USSR. After shut down of the A-1 NPP, all spent fuel was transported to the USSR. In 1978 first unit of V-1 NPP was set into operation. Actually there are six NPP units of the WWER type at Jaslovske Bohunice and Mochovce sites in operation in the Slovak Republic. These six units produce about 500 spent fuel assemblies per year. In 1988 an Interim spent fuel storage facility was build at Jaslovske Bohunice site. These facility stores spent fuel from four Jaslovske Bohunice units. In 2000 this facility was subject to a reconstruction, seismic upgrade and capacity enlargement. In 2004 Nuclear Regulatory Authority of the Slovak Republic approved transport container C-30 for transport of forty-eight spent fuel assemblies. The transport capacity has risen, so the number of transports could be reduced. In 2006 Slovak Electric Plc. (SE) will start transports of spent fuel from Mochovce site to Interim spent fuel storage facility (ISFSF) Jaslovske Bohunice. In addition, a project of Interim spent fuel storage facility at Mochovce site is going on. In the future Slovakia plans to find definitive solution for the spent fuel. One solution could be reprocessing and further usage in the power reactors, the other solution could be final deposition of spent fuel. (author)

  11. Impacts of the use of spent nuclear fuel burnup credit on DOE advanced technology legal weight truck cask GA-4 fleet size

    International Nuclear Information System (INIS)

    The object of this paper is to study the impact of full and partial spent fuel burnup credit on the capacity of the Legal Weight Truck Spent Fuel Shipping Cask (GA-4) and to determine the numbers of additional spent fuel assemblies which could be accommodated as a result. The scope of the study comprised performing nuclear criticality safety scoping calculations using the SCALE-PC software package and the 1993 spent fuel database to determine logistics for number of spent fuel assemblies to be shipped. The results of the study indicate that more capacity than 2 or 3 pressurized water reactor assemblies could be gained for GA-4 casks when burnup credit is considered. Reduction in GA-4 fleet size and number of shipments are expected to result from the acceptance of spent fuel burnup credit

  12. Intermediate dry storage of the spent fuel of reactor DIORIT

    International Nuclear Information System (INIS)

    The storage of the spent fuel of the reactor DIORIT is required, until it is either reprocessed or conditioned for final disposal. Different storage concepts have been studied as alternative to actual water pools; as a result it has been decided to store the spent fuel in a dry transport and storage casks. The project is well advanced, the cask is under construction and the fuel loading facilities are being implemented; this facilitates the decommissioning of the reactor DIORIT

  13. Evaluation of N,N-dihexyl octanamide as an alternative extractant for the reprocessing of Advanced Heavy Water Reactor spent fuel

    International Nuclear Information System (INIS)

    Advanced Heavy Water Reactor (AHWR) is being developed in India with the specific aim of utilizing thorium for power generation. AHWR sent fuel adds new dimensions to reprocessing by the presence of Pu along with 233U and Th in the spent fuel. This invokes the integration of PUREX and THOREX processes in some combination employing tri-n-butyl phosphate (TBP) as an extractant. However, separation scientists have identified certain problems with the use of TBP as extractant viz. third-phase formation and low separation factor (SF) values of U(VI) and Pu(VI) over Th, and poor decontamination factor (DF) values of U and Pu with respect to fission products. These problems are of particular concern in thorium fuel cycle

  14. Spent fuel management in Belgium

    International Nuclear Information System (INIS)

    Belgium at present operates seven commercial nuclear units (all of the PWR type) distributed over two sites (Tihange and Doel), with a total output of 5600 MW. Until now, the spent fuel of the three oldest units was reprocessed in the Cogema facility at La Hague, in France. The spent fuel of the remaining units is currently stored in the pools of the units. For economic and political reasons, it has been decided to postpone reprocessing of the major part of the fuel. As the existing pools were filling up, it was necessary to provide interim storage facilities. BELGATOM has performed a technical and economic study in order to define the most appropriate solutions. Specific Belgian conditions had to be taken into account for the selection of the solution. The main considerations are the high population density and the relatively high risk of an aircraft crash. Site conditions are such that different solutions were selected at each site. All possibilities were considered and a comparative study was carried out. After the technical and economic study, the choice was made of dry storage in metallic dual purpose casks at the Doel site and of wet storage in pools at the Tihange site. The paper describes the facilities at both sites as well as their design bases. Also addressed are the licensing approach (including the qualification programme) and environmental aspects. Finally, the Belgian approach for further storage is discussed. (author). 2 figs, 1 tab

  15. Spent fuel management in Switzerland

    International Nuclear Information System (INIS)

    Nuclear power was introduced already 18 years ago in Switzerland and it accounts today five nuclear power reactors with a total capacity of about 3000 MWe and supplies about 40% of the electricity production. The nuclear licensing requires that a firm programme of action should show that the management of all radioactive wastes and their final disposal can be accomplished. Different projects and research activities have been presented in the so-called project ''Gewaehr'' - 1985, aiming at demonstration of technical feasibility of radwaste disposal in Switzerland as well as all intermediate steps as required including intermediate storage of spent fuel and radwaste. (author)

  16. Spent fuel management in the Slovak Republic

    International Nuclear Information System (INIS)

    The Slovak Republic has used nuclear energy since 1972. First spent fuel from A-1 NPP was transported to the former Soviet Union for reprocessing. After construction of VVER V-1 and V-2 NPPs the spent fuel was stored in Interim Spent Fuel Storage Facility (ISFSF) Jaslovske Bohunice for 10 years before transportation to the Soviet Union. The Interim Spent Fuel Storage Facility Jaslovske Bohunice was commissioned in 1988. During 1997-2000, the ISFSF was subject to reconstruction and seismic upgrade. Transport container C-30 is used for the transport of spent fuel of VVER 440 type. Basic engineering of the Interim Spent Fuel Storage Facility construction has begun at Mochovce NPP in 2001. Legislative requirements are given in Act No. 130/1998 Coll. and respective regulations. (author)

  17. A central spent fuel storage in Sweden

    International Nuclear Information System (INIS)

    A planned central spent fuel storage facility in Sweden is described. The nuclear power program and quantities of spent fuel generated in Sweden is discussed. A general description of the facility is given with emphasis on the lay-out of the buildings, transport casks and fuel handling. Finally a possible design of a Swedish transportation system is discussed. (author)

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

    International Nuclear Information System (INIS)

    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

  19. Transportation accident scenarios for commercial spent fuel

    International Nuclear Information System (INIS)

    A spectrum of high severity, low probability, transportation accident scenarios involving commercial spent fuel is presented together with mechanisms, pathways and quantities of material that might be released from spent fuel to the environment. These scenarios are based on conclusions from a workshop, conducted in May 1980 to discuss transportation accident scenarios, in which a group of experts reviewed and critiqued available literature relating to spent fuel behavior and cask response in accidents

  20. Spent fuel shipping cask accident evaluation

    International Nuclear Information System (INIS)

    Mathematical models have been developed to simulate the dynamic behavior, following a hypothetical accident and fire, of typical casks designed for the rail shipment of spent fuel from nuclear reactors, and to determine the extent of radioactive releases under postulated conditions. The casks modeled were the IF-300, designed by the General Electric Company for the shipment of spent LWR fuel, and a cask designed by the Aerojet Manufacturing Company for the shipment of spent LMFBR fuel

  1. Overview on spent fuel management strategies

    International Nuclear Information System (INIS)

    This paper presents an overview on spent fuel management strategies which range from reprocessing to interim storage in a centralised facility followed by final disposal in a repository. In either case, more spent fuel storage capacity (wet or dry, at-reactor or away-from-reactor, national or regional) is required as spent fuel is continuously accumulated while most countries prefer to defer their decision to choose between these two strategies. (author)

  2. Transportation accident scenarios for commercial spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Wilmot, E L

    1981-02-01

    A spectrum of high severity, low probability, transportation accident scenarios involving commercial spent fuel is presented together with mechanisms, pathways and quantities of material that might be released from spent fuel to the environment. These scenarios are based on conclusions from a workshop, conducted in May 1980 to discuss transportation accident scenarios, in which a group of experts reviewed and critiqued available literature relating to spent fuel behavior and cask response in accidents.

  3. Spent fuel. Dissolution and oxidation

    International Nuclear Information System (INIS)

    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. U3O7 is formed by diffusion of oxygen into the UO2 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 m2/s. The initial rates of U release from spent fuel and from UO2 appear to be similar. The lowest rates (at 25 degree c >10-4 g/(m2d)) 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/(m2d). 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 UO2 surface rather than the Eh 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 U3O7/U3O8 or the U3O7/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 U3O7/U3O8 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 UO2. (G.B.)

  4. Disposal costs for advanced CANDU fuel cycles

    International Nuclear Information System (INIS)

    The CANDU reactor can 'burn' a wide range of fuels without modification to the reactor system, including natural uranium, slightly enriched uranium, mixed oxide and spent LWR fuels. The economic feasibility of the advanced fuel cycles requires consideration of their disposal costs. Preliminary cost analyses for the disposal of spent CANDU-SEU (Slightly Enriched Uranium) and CANDU-DUPIC (Direct Use of spent PWR fuel In CANDU) fuels have been performed and compared to the internationally published costs for the direct disposal of spent CANDU and LWR fuels. The analyses show significant economic advantages in the disposal costs of CANDU-SEU and CANDU-DUPIC fuels. (author)

  5. Behavior of spent fuel under unsaturated conditions

    International Nuclear Information System (INIS)

    To evaluate the performance of spent fuel in the potential repository at Yucca Mountain, Nevada, spent fuel fragments are being exposed to small and intermittent amounts of simulated groundwater under unsaturated conditions. Both the leachate and the visual appearance of the spent fuel have been characterized for 581 days of testing. The amount of Am and Cm measured in the leachates was one to two orders of magnitude greater than that released from spent fuel under saturated conditions. The cause of this difference has not been firmly identified but may be attributable to the presence of large amounts of actinide-containing colloids in the leachate of the unsaturated tests

  6. Electrochemical methods for leaching of spent fuel

    International Nuclear Information System (INIS)

    Electrochemical methods were used to study the initial dissolution and leaching behavior of spent-fuel fragments. The initial dissolution rate and the nature of the surface film of the spent fuel was shown to be compatible with those of single-crystal UO/sub 2/ surfaces. Thus, studying the behavior of UO/sub 2/ may provide an understanding of spent-fuel leaching mechanisms. Also, the study showed that spent-fuel leach data and dissolution kinetics may be obtained from the electrochemical methods described

  7. Source Term Characterization of HANARO spent fuel for disposal system design

    International Nuclear Information System (INIS)

    Source term characterization for domestic spent fuels has been carried out in a project, 'Development of a Korean Reference disposal System(A-KRS) for the HLW from Advanced Fuel cycles', to propose a high level waste repository accommodating all domestic spent fuels. The domestic spent fuels includes PWR spent fuels, PHWR spent fuels, and HANARO(research reactor) spent fuels. The key input parameters for spent nuclear fuels are total inventory, radioactivity, and decay heat. In this report, a inventory of spent fuels to be generated during entire life-time of HANARO reactor was estimated. Radioactivity and decay heat from HANARO spent fuels were also evaluated by using ORIGEN-ARP. It was expected that about 2 tons of spent fuels would be generated by the time of saturation of storage pool. In case that a capacity of storage pool is extended, about 3 tons of spent fuels were expected. It also revealed that the decay heat of HANARO spent fuel is higher than that of PWR or CANDU spent fuels. It is expected that the source terms calculated in this report would be very useful in the design of disposal system accommodating the HANARO spent fuel

  8. Method of processing spent fuel

    International Nuclear Information System (INIS)

    In a reprocessing process for spent fuels and wet recovery process for scrapped fuels, since an organic solvent used in the extraction step are generally degradated due to the effects of radioactive rays or acids, they are reused after removing degradation products by means of sodium hydroxide and sodium carbonate solution. However, the organic solvents degradated considerably can no more be regenerated and, since they contain much sodium, their de-voluming treatment is restricted to complicate the solidification treatment. In view of the above, the degradation products are removed from the degraded solvents by using a vacuum-freeze-drying method and a vacuum-distillation method for the spent solvents in a solvent cleaning step, and the recovered solutions are reused as well as the most of the radioactive materials are recovered as residues, to thereby reduce the volume of the liquid wastes and simplify the liquid wastes treatment. Further, solutions of plutonium and uranium are powderized by means of the vacuum-freeze-drying method, and resultant nitrates are applied with thermal decomposition, denitration and reduction under calcination into oxide powder. (N.H.)

  9. 78 FR 16601 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System

    Science.gov (United States)

    2013-03-18

    ... License for Storage of Spent Fuel at Power Reactor Sites'' (55 FR 29181; July 18, 1990). This rule also... 3150-AJ22 List of Approved Spent Fuel Storage Casks: MAGNASTOR System AGENCY: Nuclear Regulatory... spent fuel storage regulations by revising the NAC International, Inc. (NAC) Modular Advanced...

  10. Evaluation of a high burnup spent fuel regarding the regulations for a spent fuel dry storage

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Ik Sung; Yang, Young Sik; Bang, Je Geon; Kim, Dae Ho; Kim, Sun Ki; Song, Keun Woo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-05-15

    All nuclear plants have storage pools for spent fuel. These pools are typically 40 or more feet deep. In many countries, the spent fuels are stored under water. The water serves 2 purposes: 1) It serves as a shield to reduce the radiation levels. 2) It cools the fuel assemblies that continue to produce heat (called decay heat). But Korean nuclear plant expects the storage capacity to reach its limit by the year 2016. So, the research for the spent fuel dry storage facilities is necessary. The purpose of this study was to overview the regulatory basis for spent fuel dry storage and to evaluate its applicability for high burnup spent fuel.

  11. Assessment of the storage concept for conditioned spent fuel

    International Nuclear Information System (INIS)

    The spent fuel, the essential by-product of the electricity by the nuclear power reactors, is a highly radioactive waste. Therefore, the development of methods for effective management of this large amount of spent fuel is an important and essential task worldwide. Currently, the Advanced spent fuel Conditioning Process (ACP) is being developed at KAERI as an alternative for effective conditioning of spent fuel for the long-term storage and eventual disposal. This technology involves the process of the reduction of uranium oxide by the lithium metal in a high temperature molten salt bath. In this process, some fission product elements with the high radioactivity and heat load such as cesium and strontium are dissolved in the lithium chloride molten salt. The goals of the ACP is to recover more than 99.8% of the actinide elements and to minimize the radioactivity, heat load and volume of spent fuel to be placed in the interim storage and geological repository. In order to evaluate the storage characteristics of the conditioned spent fuel, a PWR type spent fuel with its initial enrichment of 4.5 wt% of u-235, discharged burn-up of 48 GWd/tU and 10 years of cooling time was selected as a reference base considering the domestic storage status of spent nuclear fuels. As shown in Table 1, the radioactivity and heat power of conditioned spent fuel decrease to 20.7 % and 26.3 % of those of the unconditioned spent fuel, respectively. The volume of the conditioned spent fuel is decreased to about a quarter of the initial spent fuel by removing the structural materials from the spent fuel assemblies. Four types of spent fuel storage systems, such as the metal cask, the concrete cask, the horizontal modular system and the modular vault dry system (MVDS), are currently in use world-wide. As described previously, the maximum storage capacity for the conditioned spent fuel would be extended larger than that of the existing spent fuel storage conditions. In order to confirm the

  12. CANDU spent fuel dry storage interim technique

    International Nuclear Information System (INIS)

    CANDU heavy water reactor is developed by Atomic Energy of Canada (AECL) it has 40 years of design life. During operation, the reactor can discharge a lot of spent fuels by using natural uranium. The spent fuel interim storage should be considered because the spent fuel bay storage capacity is limited with 6 years inventory. Spent fuel wet interim storage technique was adopted by AECL before 1970s, but it is diseconomy and produced extra radiation waste. So based on CANDU smaller fuel bundle dimension, lighter weight, lower burn-up and no-critical risk, AECL developed spent fuel dry interim storage technique which was applied in many CANDU reactors. Spent fuel dry interim storage facility should be designed base on critical accident prevention, decay heat removal, radiation protection and fissionable material containment. According to this introduction, analysis spent fuel dry interim storage facility and equipment design feature, it can be concluded that spent fuel dry interim storage could be met with the design requirement. (author)

  13. Spent fuel management newsletter. No. 2

    International Nuclear Information System (INIS)

    This issue of the newsletter consists of two parts. The first part describes the IAEA Secretariat activities - work and programme of the Nuclear Materials and Fuel Cycle Technology Section of the Division of Nuclear Fuel Cycle and Waste Management, recent and planned meetings and publications, Technical Co-operation projects, Co-ordinated Research programmes. The second part contains country reports - national programmes on spent fuel management: current and planned storage and reprocessing capacities, spent fuel arisings, safety, transportation, storage and treatment of spent fuel

  14. Software to improve spent fuel measurements using the FDET

    International Nuclear Information System (INIS)

    Full text: Vast quantities of spent fuel are available for safeguard measurements, primarily in Commonwealth of Independent States (CIS) of the former Soviet Union. This spent fuel, much of which consists of long cooling time material, is going to become less unique in the world safeguards arena as reprocessing projects or permanent repositories continue to be delayed or postponed. The long cooling time of many of the spent fuel assemblies in the CIS countries being prepared for intermediate term storage promotes the possibility of increased accuracy in spent fuel assays. An important point to consider for the future that could advance safeguards measurements for re-verification and inspection measurements would be to determine what safeguards requirements should be imposed upon this 'new' class of spent fuel. Improvements in measurement capability will obviously affect the safeguards requirements. What most significantly enables this progress in spent fuel measurements is the improvement in computer processing power and software enhancements leading to user-friendly Graphical User Interfaces (GUI's). The software used for these projects significantly reduces the IAEA inspector's time both learning and operating computer and data acquisition systems. While at the same time by standardizing the spent fuel measurements it is possible to increase reproducibility and reliability of the measurement data. The inclusion of various analysis algorithms into the operating environment, which can be performed in real time upon the measurement data, can also lead to increases in safeguard reliability and improvements in efficiency to plant operations. (author)

  15. Spent nuclear fuel reprocessing modeling

    International Nuclear Information System (INIS)

    The long-term wide development of nuclear power requires new approaches towards the realization of nuclear fuel cycle, namely, closed nuclear fuel cycle (CNFC) with respect to fission materials. Plant nuclear fuel cycle (PNFC), which is in fact the reprocessing of spent nuclear fuel unloaded from the reactor and the production of new nuclear fuel (NF) at the same place together with reactor plant, can be one variant of CNFC. Developing and projecting of PNFC is a complicated high-technology innovative process that requires modern information support. One of the components of this information support is developed by the authors. This component is the programme conducting calculations for various variants of process flow sheets for reprocessing SNF and production of NF. Central in this programme is the blocks library, where the blocks contain mathematical description of separate processes and operations. The calculating programme itself has such a structure that one can configure the complex of blocks and correlations between blocks, appropriate for any given flow sheet. For the ready sequence of operations balance calculations are made of all flows, i.e. expenses, element and substance makeup, heat emission and radiation rate are determined. The programme is open and the block library can be updated. This means that more complicated and detailed models of technological processes will be added to the library basing on the results of testing processes using real equipment, in test operating mode. The development of the model for the realization of technical-economic analysis of various variants of technologic PNFC schemes and the organization of 'operator's advisor' is expected. (authors)

  16. The management strategy of spent nuclear fuel

    International Nuclear Information System (INIS)

    The assessment of management strategy of spent nuclear fuel has been carried out. Spent nuclear fuel is one of the by-products of nuclear power plant. The technical operations related to the management of spent fuel discharged from reactors are called the back-end fuel cycle. It can be largely divided into three option s : the once-through cycle, the closed cycle and the so-called ‟wait and see” policy. Whatever strategy is selected for the back-end of the nuclear fuel cycle, Away-from-Reactor (AFR) storage facilities has to be constructed. For the once through cycle, the entire content of spent fuel is considered as waste, and is subject to be disposed of into a deep underground repository. In the closed cycle, however, can be divided into: (1) uranium and plutonium are recovered from spent fuel by reprocessing and recycled to manufacture mixed oxide (MOX) fuel rods, (2) waste transmutation in accelerator-driven subcritical reactors, (3) DUPIC (Direct Use of Spent PWR Fuel In CANDU) concept. In wait and see policy, which means first storing the spent fuel and deciding at a later stage on reprocessing or disposal. (author)

  17. Swelling of spent fuel storage tubes

    International Nuclear Information System (INIS)

    Unexpected swelling phenomena have been reported in the storage racks of the spent fuel pool at several nuclear power plants. Experimental and analytical studies have been carried out in order to identify the governing mechanism and to analyze the interaction of the storage tube and the spent fuel element housed in the tube. (author). 2 refs., 7 figs

  18. Spent Nuclear Fuel (SNF) Project Execution Plan

    International Nuclear Information System (INIS)

    The Spent Nuclear Fuel (SNF) Project supports the Hanford Site Mission to cleanup the Site by providing safe, economic, environmentally sound management of Site spent nuclear fuel in a manner that reduces hazards by staging it to interim onsite storage and deactivates the 100 K Area facilities

  19. Spent Nuclear Fuel (SNF) Project Execution Plan

    Energy Technology Data Exchange (ETDEWEB)

    LEROY, P.G.

    2000-11-03

    The Spent Nuclear Fuel (SNF) Project supports the Hanford Site Mission to cleanup the Site by providing safe, economic, environmentally sound management of Site spent nuclear fuel in a manner that reduces hazards by staging it to interim onsite storage and deactivates the 100 K Area facilities.

  20. Spent nuclear fuel integrity during dry storage

    International Nuclear Information System (INIS)

    Information on spent fuel integrity is of interest in evaluating the impact of long-term dry storage on the behavior of spent fuel rods. Spent fuel used during cask performance tests at the Idaho National Engineering and Environmental Laboratory (INEEL) offers significant opportunities for confirmation of the benign nature of long-term dry storage. The cask performance tests conducted at INEEL between 1984 and 1991 included visual observation and ultrasonic examination of the condition of the cladding, fuel rods, and fuel assembly hardware before dry storage and consolidation of the fuel; and a qualitative determination of the effect of dry storage and fuel consolidation on fission gas release from the spent fuel rods. A variety of cover gases and cask orientations were used during the cask performance tests. Cover gases included vacuum, nitrogen, and helium. The nitrogen and helium backfills were sampled and analyzed to detect leaking spent fuel rods. At the conclusion of each performance test, periodic gas sampling was conducted on each cask as part of a cask surveillance and monitoring activity. A spent fuel behavior project (i.e., enhanced surveillance, monitoring, and gas-sampling activities) was initiated by the U.S. Department of Energy (DOE) in 1994 for intact fuel in a CASTOR V/21 cask and for consolidated fuel in a VSC-17 cask. The results of the on going gas sampling activities are reported in this paper. (author)

  1. Spent Fuel Management in the Slovak Republic

    International Nuclear Information System (INIS)

    The skills in handling spent fuel have been collected in Slovakia for more than 35 years. During this time period a well-established spent fuel management system was created. The Slovak Government established the basic policy of spent fuel management in several resolutions. In 2008, the Slovak Government accepted in its Decision Nr. 328/2008 'The proposal on the strategy of the back-end of the nuclear power engineering'. The state supervision on nuclear safety of spent fuel management is performed by the Nuclear Regulatory Authority of the Slovak Republic (UJD). The legislative framework in the Slovak Republic is based on acts and regulations. Act No. 541/2004 Coll. on Peaceful Use of Nuclear Energy is the main legislative norm. In Slovakia there are four nuclear power units in operation. These units produce about 300 spent fuel assemblies (approximately 36 ton of heavy metal) a year. For temporary storage of the spent fuel after its terminate reloading from the reactor core the at-reactor spent fuel storage pools are used. After at least 2.5 years of storage in the at-reactor pools, the spent fuel is removed to the Interim Spent Fuel Storage Facility (ISFSF). In 2009 the UJD approved the spent fuel transportation container C-30 for next utilization. The license was issued for the transport of spent nuclear fuel from four units in operation as well as from two shut-downed units. UJD supports various research tasks under the Research and Development program (R and D). A methodology on burnup credit application has been developed. Another R and D project is focused on determination of the relation between the spent fuel residual heat generation and surface temperature of the transport container C-30. In 2005 the operator of the ISFSF started installation of an inspection stand. The stand is intended to be used for dismantling of leaky assemblies. By the end of 2009 first two modules - visual inspection and gamma spectroscopy - were put into operation. New

  2. Long term wet spent nuclear fuel storage

    International Nuclear Information System (INIS)

    The meeting showed that there is continuing confidence in the use of wet storage for spent nuclear fuel and that long-term wet storage of fuel clad in zirconium alloys can be readily achieved. The importance of maintaining good water chemistry has been identified. The long-term wet storage behaviour of sensitized stainless steel clad fuel involves, as yet, some uncertainties. However, great reliance will be placed on long-term wet storage of spent fuel into the future. The following topics were treated to some extent: Oxidation of the external surface of fuel clad, rod consolidation, radiation protection, optimum methods of treating spent fuel storage water, physical radiation effects, and the behaviour of spent fuel assemblies of long-term wet storage conditions. A number of papers on national experience are included

  3. Spent fuel response after a postulated loss of spent fuel bay cooling accident

    International Nuclear Information System (INIS)

    A study of the spent fuel behavior in a postulated severe accident is performed to understand the timings of actions and potential consequence associated with an unmitigated loss of cooling for an extended period of time. This study provides input to the 'stress test' for Cernavoda CANDU® 6 plants, requested by WENRA/ENSREG. For extreme situations, in the light of the events which occurred at Fukushima in 2011, this work has assessed the spent fuel response after a postulated loss of spent fuel bay cooling accident, assuming that there is a prolonged loss of all electrical power and water make-up to the spent fuel bay. Assessment results indicate that hydrogen generation is insignificant as long as the spent fuel remains submerged. With a large amount of shield water in the CANDU spent fuel bay, as a passive inherent feature, it is estimated that the onset of spent fuel uncovering takes more than two weeks after loss of the spent fuel bay cooling for the spent fuel bay design with normal load. The potential consequence is also discussed after the water level drops below the first few layers of spent fuel bundles due to boil-off/evaporation. However, there is a significant amount of time to take corrective actions using a number of backup design provisions to prevent spent fuel bundle uncovering. (author)

  4. Spent fuel response after a postulated loss of spent fuel bay cooling accident

    Energy Technology Data Exchange (ETDEWEB)

    Fan, H.Z.; Aboud, R.; Choy, E.; Zhu, W.; Liu, H., E-mail: hazen.fan@candu.com [CANDU Energy Inc., Mississauga, Ontario (Canada)

    2012-07-01

    A study of the spent fuel behavior in a postulated severe accident is performed to understand the timings of actions and potential consequence associated with an unmitigated loss of cooling for an extended period of time. This study provides input to the 'stress test' for Cernavoda CANDU® 6 plants, requested by WENRA/ENSREG. For extreme situations, in the light of the events which occurred at Fukushima in 2011, this work has assessed the spent fuel response after a postulated loss of spent fuel bay cooling accident, assuming that there is a prolonged loss of all electrical power and water make-up to the spent fuel bay. Assessment results indicate that hydrogen generation is insignificant as long as the spent fuel remains submerged. With a large amount of shield water in the CANDU spent fuel bay, as a passive inherent feature, it is estimated that the onset of spent fuel uncovering takes more than two weeks after loss of the spent fuel bay cooling for the spent fuel bay design with normal load. The potential consequence is also discussed after the water level drops below the first few layers of spent fuel bundles due to boil-off/evaporation. However, there is a significant amount of time to take corrective actions using a number of backup design provisions to prevent spent fuel bundle uncovering. (author)

  5. Spent fuel response after a postulated loss of spent fuel bay cooling accident

    Energy Technology Data Exchange (ETDEWEB)

    Fan, H.Z.; Aboud, R.; Choy, E.; Zhu, W.; Liu, H., E-mail: hazen.fan@candu.com [Candu Energy Inc., Mississauga, Ontario (Canada)

    2012-09-15

    A study of the spent fuel behavior in a postulated severe accident is performed to understand the timings of actions and potential consequence associated with an unmitigated loss of cooling for an extended period of time. This study provides input to the 'stress test' for Cernavoda CANDU 6 plants, requested by WENRA/ENSREG. For extreme situations, in the light of the events which occurred at Fukushima in 2011, this work has assessed the spent fuel response after a postulated loss of spent fuel bay cooling accident, assuming that there is a prolonged loss of all electrical power and water make-up to the spent fuel bay. Assessment results indicate that hydrogen generation is insignificant as long as the spent fuel remains submerged. With a large amount of shield water in the CANDU spent fuel bay, as a passive inherent feature, it is estimated that the onset of spent fuel uncovering takes more than two weeks after loss of the spent fuel bay cooling for the spent fuel bay design with normal load. The potential consequence is also discussed after the water level drops below the first few layers of spent fuel bundles due to boil-off/evaporation. However, there is a significant amount of time to take corrective actions using a number of backup design provisions to prevent spent fuel bundle uncovering. (author)

  6. Spent nuclear fuel disposal liability insurance

    International Nuclear Information System (INIS)

    This thesis examines the social efficiency of nuclear power when the risks of accidental releases of spent fuel radionuclides from a spent fuel disposal facility are considered. The analysis consists of two major parts. First, a theoretical economic model of the use of nuclear power including the risks associated with releases of radionuclides from a disposal facility is developed. Second, the costs of nuclear power, including the risks associated with a radionuclide release, are empirically compared to the costs of fossil fuel-fired generation of electricity. Under the provisions of the Nuclear Waste Policy Act of 1982, the federally owned and operated spent nuclear fuel disposal facility is not required to maintain a reserve fund to cover damages from an accidental radionuclide release. Thus, the risks of a harmful radionuclide release are not included in the spent nuclear fuel disposal fee charged to the electric utilities. Since the electric utilities do not pay the full, social costs of spent fuel disposal, they use nuclear fuel in excess of the social optimum. An insurance mechanism is proposed to internalize the risks associated with spent fueled disposal. Under this proposal, the Federal government is required to insure the disposal facility against any liabilities arising from accidental releases of spent fuel radionuclides

  7. The Canadian research reactor spent fuel situation

    International Nuclear Information System (INIS)

    This paper summarizes the present research reactor spent fuel situation in Canada. The research reactors currently operating are listed along with the types of fuel that they utilize. Other shut down research reactors contributing to the storage volume are included for completeness. The spent fuel storage facilities associated with these reactors and the methods used to determine criticality safety are described. Finally the current inventory of spent fuel and where it is stored is presented along with concerns for future storage. (author). 3 figs

  8. Proceedings of the third spent fuel workshop

    International Nuclear Information System (INIS)

    The third workshop, held in Boston, Mass. November 10-11, 1983 was organized by Battelle PNL. Questions concerning spent fuel behaviour in nuclear waste repositories were discussed. The following three lectures were presented. The corrosion of Spent UO2-Fuel in Synthetic Groundwater, R.S. Forsyth, K. Svanberg and L.O. Werme. Leaching and Radiolysis Studies on UO2 Fuel, L.H. Johnson, S. Stroes-Gascoyne, D.W. Shoesmith, M.G. Bailey and D.M. Sellinger. Comparison of Spent Fuel and UO2 Release in Salt Brines, W.J. Gray and G.L. McVay. (G.B.)

  9. Dry refabrication technology development of spent nuclear fuel

    International Nuclear Information System (INIS)

    Key technologies highly applicable to the development of advanced nuclear fuel cycle for the spent fuel recycling were developed using spent fuel and simulated spent fuel (SIMFUEL). In the frame work of dry process oxide products fabrication and the property characteristics of dry process products, hot cell experimental data for decladding, powdering and oxide product fabrication from low and high burnup spent fuel have been produced, basic technology for fabrication of spent fuel standard material has been developed, and remotely modulated welding equipment has been designed and fabricated. Also, fabrication technology of simulated dry process products was established and property models were developed based on reproducible property measurement data. In the development of head-end technology for dry refabrication of spent nuclear fuel and key technologies for volume reduction of head-end process waste which are essential in back-end fuel cycle field including pyro-processing, advanced head-end unit process technology development includes the establishment of experimental conditions for synthesis of porous fuel particles using a granulating furnace and for preparation of UO2 pellets, and fabrication and performance demonstration of engineering scale equipment for off-gas treatment of semi-volatile nuclides, and development of phosphate ceramic technology for immobilization of used filters. Radioactivation characterization and treatment equipment design of metal wastes from pretreatment process was conducted, and preliminary experiments of chlorination/electrorefining techniques for the treatment of hull wastes were performed. Based on the verification of the key technologies for head-end process via the hot-cell tests using spent nuclear fuel, pre-conceptual design for the head-end equipments was performed

  10. Inventories of high burn up LWR UO2 spent fuel and ATR MOX spent fuel in Tokai reprocessing plant

    International Nuclear Information System (INIS)

    This report describes calculated results of inventory of radioactivity in the Tokai Reprocessing Plant with calculation code, based on initial conditions and nuclear data library. The inventories were compared with three types of spent fuels, High burn up U oxide for light water reactor, U-Pu mixed oxide for advanced thermal reactor and the design based fuel for Tokai Reprocessing Plant. (author)

  11. Nuclear spent fuel management. Experience and options

    International Nuclear Information System (INIS)

    Spent nuclear fuel can be stored safely for long periods at relatively low cost, but some form of permanent disposal will eventually be necessary. This report examines the options for spent fuel management, explores the future prospects for each stage of the back-end of the fuel cycle and provides a thorough review of past experience and the technical status of the alternatives. Current policies and practices in twelve OECD countries are surveyed

  12. German Approach to Spent Fuel Management

    International Nuclear Information System (INIS)

    The management of spent fuel was based on two powerful columns until 30 June 2005, i. e. reprocessing and direct disposal. After this date any delivery of spent fuel to reprocessing plants was prohibited so that the direct disposal of unreprocessed spent fuel is the only available option in Germany today. The main steps of the current concept are: (i) Intermediate storage of spent fuel, which is the only step in practice. After the first cooling period in spent fuel storage pools it continues into cask-receiving dry storage facilities. Identification of casks, 'freezing' of inventories in terms of continuity of knowledge, monitoring the access to spent fuel, verifying nuclear material movements in terms of cask transfers and ensurance against diversion of nuclear material belong to the fundamental safeguards goals which have been achieved in the intermediate storage facilities by containment and surveillance techniques in unattended mode. (ii) Conditioning of spent fuel assemblies by separating the fuel rods from structural elements. Since the pilot conditioning facility in Gorleben has not yet come into operation, the underlying safeguards approach which focuses on safeguarding the key measurement points - the spent fuel related way in and out of the facility - has not been applied yet. (iii) Disposal in deep geological formations, but no decision has been made so far neither regarding the location of a geological repository nor regarding the safeguards approach for the disposal concept of spent fuel. The situation was complicated by a moratorium which suspended the underground exploration of the Gorleben salt dome as potential geological repository for spent fuel. The moratorium expires in October 2010. Nevertheless, considerable progress has been made in the development of disposal concepts. According to the basic, so-called POLLUX (registered) -concept spent fuel assemblies are to be conditioned after dry storage and reloaded into the POLLUX (registered) -cask

  13. Nuclear criticality safety studies applicable to spent fuel shipping cask designs and spent fuel storage

    International Nuclear Information System (INIS)

    Criticality analyses of water-moderated and reflected arrays of LWR fresh and spent fuel assemblies were carried out in this study. The calculated results indicate that using the assumption of fresh fuel loading in spent fuel shipping cask design leads to assembly spacings which are about twice the spacings of spent fuel loadings. Some shipping cask walls of composite lead and water are more effective neutron reflectors than water of 30.48 cm

  14. Dry Refabrication Technology Development of Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    Key technical data on advanced nuclear fuel cycle technology development for the spent fuel recycling have been produced in this study. In the frame work of DUPIC, dry process oxide products fabrication, hot cell experimental data for decladding, powdering and oxide product fabrication from low and high burnup spent fuel have been produced, basic technology for fabrication of spent fuel standard material has been developed, and remote modulated welding equipment has been designed and fabricated. In the area of advanced pre-treatment process development, a rotary-type oxidizer and spherical particle fabrication process were developed by using SIMFUEL and off-gas treatment technology and zircalloy tube treatment technology were studied. In the area of the property characteristics of dry process products, fabrication technology of simulated dry process products was established and property models were developed based on reproducible property measurement data

  15. Reactor TRIGA PUSPATI (RTP) spent fuel pool conceptual design

    International Nuclear Information System (INIS)

    Reactor TRIGA PUSPATI (RTP) is the one and only research reactor in Malaysia that has been safely operated and maintained since 1982. In order to enhance technical capabilities and competencies especially in nuclear reactor engineering a feasibility study on RTP power upgrading was proposed to serve future needs for advance nuclear science and technology in the country with the capability of designing and develop reactor system. The need of a Spent Fuel Pool begins with the discharge of spent fuel elements from RTP for temporary storage that includes all activities related to the storage of fuel until it is either sent for reprocessed or sent for final disposal. To support RTP power upgrading there will be major RTP systems replacement such as reactor components and a new temporary storage pool for fuel elements. The spent fuel pool is needed for temporarily store the irradiated fuel elements to accommodate a new reactor core structure. Spent fuel management has always been one of the most important stages in the nuclear fuel cycle and considered among the most common problems to all countries with nuclear reactors. The output of this paper will provide sufficient information to show the Spent Fuel Pool can be design and build with the adequate and reasonable safety assurance to support newly upgraded TRIGA PUSPATI TRIGA Research Reactor. (author)

  16. Spent Fuel Management Newsletter. No. 1

    International Nuclear Information System (INIS)

    This Newsletter has been prepared in accordance with the recommendations of the International Regular Advisory Group on Spent Fuel Management and the Agency's programme (GC XXXII/837, Table 76, item 14). The main purpose of the Newsletter is to provide Member States with new information about the state-of-the-art in one of the most important parts of the nuclear fuel cycle - Spent Fuel Management. The contents of this publication consists of two parts: (1) IAEA Secretariat contribution -work and programme of the Nuclear Materials and Fuel Cycle Technology Section of the Division of Nuclear Fuel Cycle and Waste Management, recent and planned meetings and publications, Technical Co-operation projects, Co-ordinated Research programmes, etc. (2) Country reports - national programmes on spent fuel management: current and planned storage and reprocessing capacities, spent fuel arisings, safety, transportation, storage, treatment of spent fuel, some aspects of uranium and plutonium recycling, etc. The IAEA expects to publish the Newsletter once every two years between the publications of the Regular Advisory Group on Spent Fuel Management. Figs and tabs

  17. Status report on spent fuel management

    International Nuclear Information System (INIS)

    France, Great Britain, the Soviet Union, the Federal Republic of Germany and Japan are today the countries with the most large-scale projects underway to construct reprocessing plants for spent oxide fuel from light water reactors. In some countries, including the United States, Canada and Sweden, the spent fuel management is planned to be based on the direct disposal. Except for reprocessing little services for spent fuel management are made available to foreign power companies. The main option in TVO's long-range programme for spent fuel management is based on direct disposal in the Finnish bedrock. The reposiory site will be selected in the year 2000 and the final disposal is scheduled to begin in 2020. At the same time, TVO continues monitoring the international market for related services and will individually assess their suitability to TVO's spent fuel management. A serious alternative for a domestic direct disposal would be that of sending the spent fuel out of Finland either for reprocessing or final disposal without a duty to take back any waste. For such a concept, it would be essential to cover the overall quantity of TVO's spent fuel and also to be competetive in financial terms. At the moment, there is none such foreign service available. Regarding the domestic scheme for direct disposal the R and D work will for some years be focused on the encapsulating process and canister materials, as well as the alternative repository concepts

  18. Safeguards for spent fuels: Verification problems

    International Nuclear Information System (INIS)

    The accumulation of large quantities of spent nuclear fuels world-wide is a serious problem for international safeguards. A number of International Atomic Energy Agency (IAEA) member states, including the US, consider spent fuel to be a material form for which safeguards cannot be terminated, even after permanent disposal in a geologic repository. Because safeguards requirements for spent fuels are different from those of conventional bulk-handling and item-accounting facilities, there is room for innovation to design a unique safeguards regime for spent fuels that satisfies the goals of the nuclear nonproliferation treaty at a reasonable cost to both the facility and the IAEA. Various strategies being pursued for long-term management of spent fuels are examined with a realistic example to illustrate the problems of verifying safeguards under the present regime. Verification of a safeguards regime for spent fuels requires a mix of standard safeguards approaches, such as quantitative verification and use of seals, with other measures that are unique to spent fuels. 17 refs

  19. Modeling the highway transportation of spent fuel

    International Nuclear Information System (INIS)

    There will be a substantial increase in the number of spent fuel shipments on the nation's highway system in the next thirty years. Most of the spent fuel will be moving from reactors to a spent fuel repository. This study develops two models that evaluate the risk and cost of moving the spent fuel. The Minimum Total Transport Risk Model (MTTRM) seeks an efficient solution for this problem by finding the minimum risk path through the network and sending all the spent fuel shipments over this one path. The Equilibrium Transport Risk Model (ETRM) finds an equitable solution by distributing the shipments over a number of paths in the network. This model decreases the risk along individual paths, but increases society's risk because the spent fuel shipments are traveling over more links in the network. The study finds that there is a trade off between path risk and societal risk. As path risk declines, societal risk rises. The cost of shipping also increases as the number of paths expand. The cost and risk of shipping spent fuel from ten reactors to four potential repository sites are evaluated using the MTTRM. The temporary monitored retrievable storage (MRS) facility in Tennessee is found to be the minimum cost and minimum risk solution. When direct shipment to the permanent sites is considered, Deaf Smith, Texas is the least cost and least incident free transport risk location. Yucca Mountain, Nevada is the least risk location when the focus is placed on the potential consequences of an accident

  20. BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

    International Nuclear Information System (INIS)

    The objective of this report is to identify issues and support documentation and identify and detail existing research on spent fuel dry storage; provide information to support potential R&D for the UKABWR (United Kingdom Advanced Boiling Water Reactor) Spent Fuel Interim Storage (SFIS) Pre-Construction Safety Report; and support development of answers to questions developed by the regulator. Where there are gaps or insufficient data, Oak Ridge National Laboratory (ORNL) has summarized the research planned to provide the necessary data along with the schedule for the research, if known. Spent nuclear fuel (SNF) from nuclear power plants has historically been stored on site (wet) in spent fuel pools pending ultimate disposition. Nuclear power users (countries, utilities, vendors) are developing a suite of options and set of supporting analyses that will enable future informed choices about how best to manage these materials. As part of that effort, they are beginning to lay the groundwork for implementing longer-term interim storage of the SNF and the Greater Than Class C (CTCC) waste (dry). Deploying dry storage will require a number of technical issues to be addressed. For the past 4-5 years, ORNL has been supporting the U.S. Department of Energy (DOE) in identifying these key technical issues, managing the collection of data to be used in issue resolution, and identifying gaps in the needed data. During this effort, ORNL subject matter experts (SMEs) have become expert in understanding what information is publicly available and what gaps in data remain. To ensure the safety of the spent fuel under normal and frequent conditions of wet and subsequent dry storage, intact fuel must be shown to: 1.Maintain fuel cladding integrity; 2.Maintain its geometry for cooling, shielding, and subcriticality; 3.Maintain retrievability, and damaged fuel with pinhole or hairline cracks must be shown not to degrade further. Where PWR (pressurized water reactor) information is

  1. BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

    Energy Technology Data Exchange (ETDEWEB)

    Bevard, Bruce Balkcom [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mertyurek, Ugur [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Scaglione, John M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-10-01

    The objective of this report is to identify issues and support documentation and identify and detail existing research on spent fuel dry storage; provide information to support potential R&D for the UKABWR (United Kingdom Advanced Boiling Water Reactor) Spent Fuel Interim Storage (SFIS) Pre-Construction Safety Report; and support development of answers to questions developed by the regulator. Where there are gaps or insufficient data, Oak Ridge National Laboratory (ORNL) has summarized the research planned to provide the necessary data along with the schedule for the research, if known. Spent nuclear fuel (SNF) from nuclear power plants has historically been stored on site (wet) in spent fuel pools pending ultimate disposition. Nuclear power users (countries, utilities, vendors) are developing a suite of options and set of supporting analyses that will enable future informed choices about how best to manage these materials. As part of that effort, they are beginning to lay the groundwork for implementing longer-term interim storage of the SNF and the Greater Than Class C (CTCC) waste (dry). Deploying dry storage will require a number of technical issues to be addressed. For the past 4-5 years, ORNL has been supporting the U.S. Department of Energy (DOE) in identifying these key technical issues, managing the collection of data to be used in issue resolution, and identifying gaps in the needed data. During this effort, ORNL subject matter experts (SMEs) have become expert in understanding what information is publicly available and what gaps in data remain. To ensure the safety of the spent fuel under normal and frequent conditions of wet and subsequent dry storage, intact fuel must be shown to: 1.Maintain fuel cladding integrity; 2.Maintain its geometry for cooling, shielding, and subcriticality; 3.Maintain retrievability, and damaged fuel with pinhole or hairline cracks must be shown not to degrade further. Where PWR (pressurized water reactor) information is

  2. Onsite storage of spent nuclear fuel in metalic spent fuel storage casks

    International Nuclear Information System (INIS)

    Virginia Electric and Power Company (Vepco) owns and operates two nuclear power stations within its system: the North Anna Power Station located in Louisa County, Virginia; and the Surry Power Station located in Surry County, Virginia. Each of these power stations has two pressurized water reactor operating units which share a common spent fuel pool. Under the Nuclear Waste Policy Act of 1982, Vepco is responsible for providing interim spent fuel storage until availability of the Federal Repository. Vepco has studied a number of options and has developed a program to provide the required onsite interim spent fuel storage. Options considered by Vepco included reracking, pin consolidation, dry storage and construction of a new spent fuel pool to provide the increased spent fuel storage capacity required. Vepco has selected reracking at North Anna combined with dry storage in metal spent fuel storage casks at Surrey to provide the required onsite spent fuel storage. A dry cask storage facility design and license application were developed and the license application was submitted to the NRC in October, 1982. The selection of the option to use dry cask storage of spent fuel at Surry represents the first attempt to license dry storage of spent nuclear fuel in the United States. This storage option is expected to provide an effective option for utilities without adequate storage space in their existing spent fuel pools

  3. Spent fuel transportation in the United States: commercial spent fuel shipments through December 1984

    International Nuclear Information System (INIS)

    This report has been prepared to provide updated transportation information on light water reactor (LWR) spent fuel in the United States. Historical data are presented on the quantities of spent fuel shipped from individual reactors on an annual basis and their shipping destinations. Specifically, a tabulation is provided for each present-fuel shipment that lists utility and plant of origin, destination and number of spent-fuel assemblies shipped. For all annual shipping campaigns between 1980 and 1984, the actual numbers of spent-fuel shipments are defined. The shipments are tabulated by year, and the mode of shipment and the casks utilized in shipment are included. The data consist of the current spent-fuel inventories at each of the operating reactors as of December 31, 1984. This report presents historical data on all commercial spent-fuel transportation shipments have occurred in the United States through December 31, 1984

  4. Spent fuel management of NPPs in Argentina

    International Nuclear Information System (INIS)

    There are two Nuclear Power Plants in operation in Argentina: 'Atucha I' (unique PHWR design) in operation since 1974, and 'Embalse' (typical Candu reactor) which started operation in 1984. Both NPPs are operated by 'Nucleoelectrica Argentina S.A' which is responsible for the management and interim storage of spent fuel till the end of the operative life of the plants. A third NPP, 'Atucha II' is under construction, with a similar design of Atucha I. The legislative framework establishes that after final shutdown of a NPP the spent fuel will be transferred to the 'National Atomic Energy Commission', which is also responsible for the decommissioning of the Plants. In Atucha I, the spent fuel is stored underwater, until another option is implemented meanwhile in Embalse the spent fuel is stored during six years in pools and then it is moved to a dry storage. A decision about the fuel cycle back-end strategy will be taken before year 2030. (authors)

  5. International experience in conditioning spent fuel elements

    International Nuclear Information System (INIS)

    The purpose of this report is to compile and present in a clear form international experience (USA, Canada, Sweden, FRG, UK, Japan, Switzerland) gained to date in conditioning spent fuel elements. The term conditioning is here taken to mean the handling and packaging of spent fuel elements for short- or long-term storage or final disposal. Plants of a varying nature fall within this scope, both in terms of the type of fuel element treated and the plant purpose eg. experimental or production plant. Emphasis is given to plants which bear some similarity to the concept developed in Germany for direct disposal of spent fuel elements. Worldwide, however, relatively few conditioning plants are in existence or have been conceived. Hence additional plants have been included where aspects of the experience gained are also of relevance eg. plants developed for the consolidation of spent fuel elements. (orig./HP)

  6. Spent Nuclear Fuel Project dose management plan

    International Nuclear Information System (INIS)

    This dose management plan facilitates meeting the dose management and ALARA requirements applicable to the design activities of the Spent Nuclear Fuel Project, and establishes consistency of information used by multiple subprojects in ALARA evaluations. The method for meeting the ALARA requirements applicable to facility designs involves two components. The first is each Spent Nuclear Fuel Project subproject incorporating ALARA principles, ALARA design optimizations, and ALARA design reviews throughout the design of facilities and equipment. The second component is the Spent Nuclear Fuel Project management providing overall dose management guidance to the subprojects and oversight of the subproject dose management efforts

  7. Spent fuel disposal impact on plant decommissioning

    International Nuclear Information System (INIS)

    Regardless of the decommissioning option selected (DECON, SAFSTOR, or ENTOMB), a 10 CFR 50 license cannot be terminated until the spent fuel is either removed from the site or stored in a separately 10 CFR 72 licensed Independent Spent Fuel Storage Installation (ISFSI). Humboldt Bay is an example of a plant which has selected the SAFSTOR option. Its spent fuel is currently in wet storage in the plant's spent fuel pool. When it completes its dormant period and proceeds with dismantlement, it will have to dispose of its fuel or license an ISFSI. Shoreham is an example of a plant which has selected the DECON option. Fuel disposal is currently critical path for license termination. In the event an ISFSI is proposed to resolve the spent fuel removal issue, whether wet or dry, utilities need to properly determine the installation, maintenance, and decommissioning costs for such a facility. In considering alternatives for spent fuel removal, it is important for a utility to properly account for ISFSI decommissioning costs. A brief discussion is presented on one method for estimating ISFSI decommissioning costs

  8. Spent fuel workshop'2002

    Energy Technology Data Exchange (ETDEWEB)

    Poinssot, Ch

    2002-07-01

    This document gathers the transparencies of the presentations given at the 2002 spent fuel workshop: Session 1 - Research Projects: Overview on the IN CAN PROCESSES European project (M. Cowper), Overview on the SPENT FUEL STABILITY European project (C. Poinssot), Overview on the French R and D project on spent fuel long term evolution, PRECCI (C. Poinssot); Session 2 - Spent Fuel Oxidation: Oxidation of uranium dioxide single crystals (F. Garrido), Experimental results on SF oxidation and new modeling approach (L. Desgranges), LWR spent fuel oxidation - effects of burn-up and humidity (B. Hanson), An approach to modeling CANDU fuel oxidation under dry storage conditions (P. Taylor); Session 3 - Spent Fuel Dissolution Experiments: Overview on high burnup spent fuel dissolution studies at FZK/INE (A. Loida), Results on the influence of hydrogen on spent fuel leaching (K. Spahiu), Leaching of spent UO{sub 2} fuel under inert and reducing conditions (Y. Albinsson), Fuel corrosion investigation by electrochemical techniques (D. Wegen), A reanalysis of LWR spent fuel flow through dissolution tests (B. Hanson), U-bearing secondary phases formed during fuel corrosion (R. Finch), The near-field chemical conditions and spent fuel leaching (D. Cui), The release of radionuclides from spent fuel in bentonite block (S.S. Kim), Trace actinide behavior in altered spent fuel (E. Buck, B. Hanson); Session 4 - Radiolysis Issues: The effect of radiolysis on UO{sub 2} dissolution determined from electrochemical experiments with {sup 238}Pu doped UO{sub 2} M. Stroess-Gascoyne (F. King, J.S. Betteridge, F. Garisto), doped UO{sub 2} studies (V. Rondinella), Preliminary results of static and dynamic dissolution tests with {alpha} doped UO{sub 2} in Boom clay conditions (K. Lemmens), Studies of the behavior of UO{sub 2} / water interfaces under He{sup 2+} beam (C. Corbel), Alpha and gamma radiolysis effects on UO{sub 2} alteration in water (C. Jegou), Behavior of Pu-doped pellets in brines

  9. Development of Spent Fuel Examination Technology

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ho Dong; Park, K. J.; Shin, H. S. (and others)

    2007-04-15

    For the official operation of ACPF Facility Attachment based on facility declared DIQ was issued by IAEA and officialized upon ROK government approval. This procedure gives an essential ground to negotiate Joint Determination between governments of ROK and US. For ACPF process material accountability a neutron coincidence counting system was developed and calibrated with Cf-252 source. Its performance test demonstrated that over-all counting efficiency was about 21% with random error, 1.5% against calibration source, which found to be satisfactory to the expected design specification. A calibration curve derived by MCNP code with relationship between ASNC doublet counts vs. neutron activity of Cm-244 showed calibration constant to be 2.78x10E5 counts/s.g which would be used for initial ACP hot operation test. Nuclear material transportation and temporary storage system was established for active demonstration of advanced spent fuel management process line and would be directly applied to the effective management of wastes arising from active demonstration and would later contribute as a base data to development of inter hot-cell movement system in pyro-processing line. In addition, an optimal spent fuel for the ACP demonstration was selected and a computer code was developed as a tool to estimate the expected source term at each key measurement point of ACP.

  10. Development of Spent Fuel Examination Technology

    International Nuclear Information System (INIS)

    For the official operation of ACPF Facility Attachment based on facility declared DIQ was issued by IAEA and officialized upon ROK government approval. This procedure gives an essential ground to negotiate Joint Determination between governments of ROK and US. For ACPF process material accountability a neutron coincidence counting system was developed and calibrated with Cf-252 source. Its performance test demonstrated that over-all counting efficiency was about 21% with random error, 1.5% against calibration source, which found to be satisfactory to the expected design specification. A calibration curve derived by MCNP code with relationship between ASNC doublet counts vs. neutron activity of Cm-244 showed calibration constant to be 2.78x10E5 counts/s.g which would be used for initial ACP hot operation test. Nuclear material transportation and temporary storage system was established for active demonstration of advanced spent fuel management process line and would be directly applied to the effective management of wastes arising from active demonstration and would later contribute as a base data to development of inter hot-cell movement system in pyro-processing line. In addition, an optimal spent fuel for the ACP demonstration was selected and a computer code was developed as a tool to estimate the expected source term at each key measurement point of ACP

  11. Programme on spent fuel management in India

    International Nuclear Information System (INIS)

    The Indian Atomic Energy Programme aims at harnessing the natural resources in the most optimal manner. To achieve this end, a three phase Nuclear Power Programme was devised in the early days of establishment of the Department of Atomic Energy. It envisages the utilization of the modest uranium reserves and rich thorium deposits in the country. The limited natural reserves of fuel materials have prompted India to pay increased attention to the back end of the fuel cycle, which consists of Reprocessing and Waste Management. The spent fuel is valued in India as a source for fuel and is treated accordingly, to recover the important fissile materials. Thus, the route of reprocessing of spent fuel in order to recycle uranium and plutonium in future reactors was opted for. Today, India possesses the capability and facilities, catering to the entire fuel cycle, i.e., starting from the mining of the ore, through fabrication of fuel and its application in reactors, to reprocessing of the spent fuel and appropriate waste management. With the third Reprocessing Plant almost ready for commissioning and second and third waste Immobilization Plants under constructions, the Spent Fuel Management Programme has come of age in India. This paper presents an overview of the status of the spent fuel management programme and its future perspective. (author)

  12. Release of segregated nuclides from spent fuel

    International Nuclear Information System (INIS)

    The potential release of fission and activation products from spent nuclear fuel into groundwater after container failure in the Swedish deep repository is discussed. Data from studies of fission gas release from representative Swedish BWR fuel are used to estimate the average fission gas release for the spent fuel population. Information from a variety of leaching studies on LWR and CANDU fuel are then reviewed as a basis for estimating the fraction of the inventory of key radionuclides that could be released preferentially (the Instant Release Fraction of IRF) upon failure of the fuel cladding. The uncertainties associated with these estimates are discussed

  13. Spent Fuel Background Report Volume I

    International Nuclear Information System (INIS)

    This report is an overview of current spent nuclear fuel management in the DOE complex. Sources of information include published literature, internal DOE documents, interviews with site personnel, and information provided by individual sites. Much of the specific information on facilities and fuels was provided by the DOE sites in response to the questionnaire for data for spent fuels and facilities data bases. This information is as accurate as is currently available, but is subject to revision pending results of further data calls. Spent fuel is broadly classified into three categories: (a) production fuels, (b) special fuels, and (c) naval fuels. Production fuels, comprising about 80% of the total inventory, are those used at Hanford and Savannah River to produce nuclear materials for defense. Special fuels are those used in a wide variety of research, development, and testing activities. Special fuels include fuel from DOE and commercial reactors used in research activities at DOE sites. Naval fuels are those developed and used for nuclear-powered naval vessels and for related research and development. Given the recent DOE decision to curtail reprocessing, the topic of main concern in the management of spent fuel is its storage. Of the DOE sites that have spent nuclear fuel, the vast majority is located at three sites-Hanford, INEL, and Savannah River. Other sites with spent fuel include Oak Ridge, West Valley, Brookhaven, Argonne, Los Alamos, and Sandia. B ampersand W NESI Lynchburg Technology Center and General Atomics are commercial facilities with DOE fuel. DOE may also receive fuel from foreign research reactors, university reactors, and other commercial and government research reactors. Most DOE spent fuel is stored in water-filled pools at the reactor facilities. Currently an engineering study is being performed to determine the feasibility of using dry storage for DOE-owned spent fuel currently stored at various facilities. Delays in opening the deep

  14. Development of CANDU Spent Fuel Sipping System

    International Nuclear Information System (INIS)

    As the tendency is toward radioactivity zero-leakage on the reactor core for the safe operation of nuclear power plants, the importance of detecting radioactivity leaking from fuel assemblies irradiated in the core is being on the rise. Nuclear fuel, even though it is designed and fabricated in terms of excellent thermal performance and mechanical integrity, can be damaged under unexpected circumstances. An excessive hydriding on fuel rods and pellet-to-clad interaction., etc. can result in failed fuel rod. It is, thus, considered that a inspection process is prerequisite procedure to identify causes of such failed fuel rods for the safe operation of nuclear power plants. If a fuel rod failure occurs during the operation of a nuclear power plant, the coolant water becomes contaminated by leaked fission products, and the power level of the plant has to be lowered or the operation to be stopped. In addition, the spent fuel that have been stored in a spent fuel storage pool for a long time is now transferring to a dry storage. To maximize the integrity of the dry storage, all the fuels transferring to a dry storage should be examined their integrities exactly and efficiently. Therefore, the ultimate purpose of this study is to develop a system capable of judging whether the long-term stored fuel in spent fuel storage pool is failed or not. In this study, a spent fuel sipping system with wet leakage detection technology is developed to make it possible

  15. LWR spent fuel management in Germany

    International Nuclear Information System (INIS)

    The spent fuel management strategy in the Federal Republic of Germany is based alternatively on interim storage and subsequent reprocessing of spent fuel or on extended storage and direct disposal of spent fuel. By economic and strategic reasons the spent fuel burnup is presently achieving 50 GWd/tHM and will targeting 55 GWd/tHM batch average. Recently the CASTOR V/19 license is issued to store spent fuel assemblies (SFAs) with up to 55 GWd/tU burnup (batch average) for 40 years. The integral pool storage capacity in Germany is 5600 tHM without the necessary full core reserve. The AFR spent fuel storage sites of Ahaus ( 4200 tHM) and Gorleben (3800 tHM) are in operation. The PKA pilot-facility to condition the SFAs is in the final state of erection and alternative approaches for SFAs with a higher burnup and/or MOX fuel are under investigation. The underground exploration of the Gorleben salt dome is in progress. Presently the non heat generating waste is disposed in the former Morsleben salt mine. Licensing of the larger Konrad iron mine for that purpose is under treatment. (author)

  16. Safeguards issues in spent fuel consolidation facilities

    Energy Technology Data Exchange (ETDEWEB)

    Belew, W.L.; Moran, B.W.

    1991-01-01

    In the nuclear power industry, the fuel assembly is the basic unit for nuclear material accountancy. The safeguards procedures for the spent fuel assemblies, therefore, are based on an item accountancy approach. When fuel consolidation occurs in at-reactor'' or away-from-reactor'' facilities, the fuel assemblies are disassembled and cease to be the basic unit containing nuclear material. Safeguards can no longer be based on item accountancy of fuel assemblies. The spent fuel pins containing plutonium are accessible, and the possibilities for diversion of spent fuel for clandestine reprocessing to recover the plutonium are increased. Thus, identifying the potential safeguards concerns created by operation of these facilities is necessary. Potential safeguards techniques to address these concerns also must be identified so facility designs may include the equipment and systems required to provide an acceptable level of assurance that the international safeguards objectives can be met when these facilities come on-line. The objectives of this report are (1) to identify the safeguards issues associated with operation of spent fuel consolidation facilities, (2) to provide a preliminary assessment of the assessment of the safeguards vulnerabilities introduced, and (3) to identify potential safeguards approaches that could meet international safeguards requirements. The safeguards aspects of spent fuel consolidation are addressed in several recent reports and papers. 11 refs., 3 figs., 3 tabs.

  17. Development of spent fuel storage process equipment

    International Nuclear Information System (INIS)

    The scope of the research and development project covers the development of various remote operation technologies which are important assets for the repairment and maintenance of spent fuel handling facilities as well as the actual handling of spent fuels. As a key technology pertaining to such an objective, an anti-swing overhead crane system is developed. The anti-swing crane system is designed to provide oscillation free transportation of heavy equipments and materials such as spent fuel casks in nuclear facilities, therefore, an increased level of safety may be achieved. Also a teleoperated robotic impact wrench system is developed by adopting multi-sensor integration and suitably designed impact wrench module. The performance of the impact wrench system is tested by opening the spent fuel cask lid. Other related efforts in technological innovations are also made in the development of fuzzy logic controller for a tele-visual surveillance system and the design of a three-dimensional range finder. (Author)

  18. Choosing a spent fuel interim storage system

    International Nuclear Information System (INIS)

    The Transnucleaire Group has developed different modular solutions to address spent fuel interim storage needs of NPP. These solutions, that are present in Europe, USA and Asia are metal casks (dual purpose or storage only) of the TN 24 family and the NUHOMS canister based system. It is not always simple for an operator to sort out relevant choice criteria. After explaining the basic designs involved on the examples of the TN 120 WWER dual purpose cask and the NUHOMS 56 WWER for WWER 440 spent fuel, we shall discuss the criteria that govern the choice of a given spent fuel interim storage system from the stand point of the operator. In conclusion, choosing and implementing an interim storage system is a complex process, whose implications can be far reaching for the long-term success of a spent fuel management policy. (author)

  19. TRIGA Mark II Ljubljana - spent fuel transportation

    International Nuclear Information System (INIS)

    The most important activity in 1999 was shipment of the spent fuel elements back to the United States for final disposal. This activity started already in 1998 with some governmental support. In July 1999 all spent fuel elements (219 pieces) from the TRIGA research reactor in Ljubljana were shipped back to the United Stated by the ship from the port Koper in Slovenia. At the same time shipment of the spent fuel from the research reactor in Pitesti, Romania, and the research reactor in Rome, Italy, was conducted. During the loading the radiation exposure to the workers was rather low. The loading and shipment of the spent nuclear fuel went very smoothly and according the accepted time table. During the last two years the TRIGA research reactor in Ljubljana has been in operation about 1100 hours per year and without any undesired shut-down. (authors)

  20. Operating Experience in Spent Fuel Storage Casks

    International Nuclear Information System (INIS)

    A safe storage of spent fuels has been considered as one of the inevitable tasks for TEPCO for the last few decades. In order to increase flexibility for the fuel storage measures, TEPCO has been storing spent fuels in an on-site dry storage facility at Fukushima-Daiichi Nuclear Power Station. Since 1995, more than 400 fuel assemblies have been safely store. Integrity of storage casks and fuels were carefully checked by periodical inspections, which were conducted in 2000 and 2005. The next investigation will be held within a few years in order to verify the safety conditions even after a 15-year storage. These series of inspections will give plenty of useful data for the design and operation of the Mutsu facility, which will be the first off-site interim spent fuel storage facility away from any reactor site in Japan. (author)

  1. Timely topics on spent fuel storage

    Energy Technology Data Exchange (ETDEWEB)

    Selin, I. (Nuclear Regulatory Commission, Washington, DC (United States))

    1994-10-01

    For those plants in premature or extended shutdown, the NRC finds several strong reasons why the interim, on-site storage of spent fuel should often be shifted from the existing fuel pool to a dry storage system. These reasons include the continuing operational support activities needed to keep a fuel pool operating properly. Water chemistry and cleanliness, surveillance of rack and fuel conditions, and maintenance and surveillance of support systems are all activities that are second nature to an operating plant, but may not always receive adequate attention in a plant permanently shut down. Therefore, the NRC increasingly views dry storage as the preferred method of interim storage of mature spent fuels for plants in permanent shut-down, as well as for supplementary storage in many operating plants. The author discusses regulatory policies and practices concerning the interim storage of spent fuel and the increasing use of dry cask systems in this paper.

  2. Timely topics on spent fuel storage

    International Nuclear Information System (INIS)

    For those plants in premature or extended shutdown, the NRC finds several strong reasons why the interim, on-site storage of spent fuel should often be shifted from the existing fuel pool to a dry storage system. These reasons include the continuing operational support activities needed to keep a fuel pool operating properly. Water chemistry and cleanliness, surveillance of rack and fuel conditions, and maintenance and surveillance of support systems are all activities that are second nature to an operating plant, but may not always receive adequate attention in a plant permanently shut down. Therefore, the NRC increasingly views dry storage as the preferred method of interim storage of mature spent fuels for plants in permanent shut-down, as well as for supplementary storage in many operating plants. The author discusses regulatory policies and practices concerning the interim storage of spent fuel and the increasing use of dry cask systems in this paper

  3. Spent-fuel-storage studies at the Barnwell Nuclear Fuel Plant. Studies and research concerning BNFP

    International Nuclear Information System (INIS)

    This report contains the results of various studies and demonstrations related to advanced spent-fuel-storage techniques which were performed at the Barnwell Nuclear Fuel Plant (BNFP) in 1982. The demonstrations evaluated various technical aspects of fuel disassembly and canning and dry-storage techniques. The supporting studies examined thermal limitations and criticality concerns

  4. An intelligent spent fuel database for BWR fuels

    International Nuclear Information System (INIS)

    The present aim is to establish an intelligent database of Spent Fuel Data (including physical fuel data and reactor operating history information) to support burnup credit analyses for Boiling Water Reactor Fuel. At a later date, information of Pressurized Water Reactor Fuel and existing Post-Irradiation Examination (PIE) data for benchmarking fuel composition calculations may be integrated into the database. (author)

  5. Antineutrino monitoring of spent nuclear fuel

    OpenAIRE

    Brdar, Vedran; Huber, Patrick; Kopp, Joachim

    2016-01-01

    Military and civilian applications of nuclear energy have left a significant amount of spent nuclear fuel over the past 70 years. Currently, in many countries world wide, the use of nuclear energy is on the rise. Therefore, the management of highly radioactive nuclear waste is a pressing issue. In this letter, we explore antineutrino detectors as a tool for monitoring and safeguarding nuclear waste material. We compute the flux and spectrum of antineutrinos emitted by spent nuclear fuel eleme...

  6. Investigation of Spent Nuclear Fuel Pool Coolability

    OpenAIRE

    Nimander, Fredrik

    2011-01-01

    The natural catastrophe at Fukushima Dai-ichi 2011 enlightened the nuclear community. This master thesis reveals the non-negligible risks regarding the short term storage of spent nuclear fuel. The thesis has also investigated the possibility of using natural circulation of air in a passive safety system to cool the spent nuclear fuel pools. The results where conclusive: The temperature difference between the heated air and ambient air is far too low for natural circulation of air to remove a...

  7. Status on spent fuel management in Spain

    International Nuclear Information System (INIS)

    To confront the lack of spent fuel storage locations at the pools of the nuclear power plants, different actions have been undertaken by Enresa in conjunction with the Plant utilities. Basically, these measures have consisted in expanding the capacities of the spent fuel storage pools to their maximum capacity by exchanging their racks and in those cases where reracking is no further possible, dry storage will be provided, initially by means of dual purpose metallic casks. (author)

  8. Probable leaching mechanisms for spent fuel

    International Nuclear Information System (INIS)

    At the Pacific Northwest Laboratory, researchers in the Waste/Rock Interaction Technology Program are studying spent fuel as a possible waste form for the Office of Nuclear Waste Isolation. This paper presents probable leaching mechanisms for spent fuel and discusses current progress in identifying and understanding the leaching process. During the past year, experiments were begun to study the complex leaching mechanism of spent fuel. The initial work in this investigation was done with UO2, which provided the most information possible on the behavior of the spent-fuel matrix without encountering the very high radiation levels associated with spent fuel. Both single-crystal and polycrystalline UO2 samples were used for this study, and techniques applicable to remote experimentation in a hot cell are being developed. The effects of radiation are being studied in terms of radiolysis of water and surface activation of the UO2. Dissolution behavior and kinetics of UO2 were also investigated by electrochemical measurement techniques. These data will be correlated with those acquired when spent fuel is tested in a hot cell. Oxidation effects represent a major area of concern in evaluating the stability of spent fuel. Dissolution of UO2 is greatly increased in an oxidizing solution because the dissolution is then controlled by the formation of hexavalent uranium. In solutions containing very low oxygen levels (i.e., reducing solutions), oxidation-induced dissolution may be possible via a previously oxidized surface, through exposure to air during storage, or by local oxidants such as O2 and H2O2 produced from radiolysis of water and radiation-activated UO2 surfaces. The effects of oxidation not only increase the dissolution rate, but could lead to the disintegration of spent fuel into fine fragments

  9. Effects of alpha-decay on spent fuel corrosion behaviour

    International Nuclear Information System (INIS)

    An overview of results in the area of spent fuel characterization as nuclear waste is presented. These studies are focused on primary aspects of spent fuel corrosion, by considering different fuel compositions and burn ups, as well as a wide set of environmental conditions. The key parameter is the storage time of the fuel e.g. in view of spent fuel retrieval or in view of its final disposal. To extrapolate data obtainable from a laboratory-acceptable timescale to those expected after storage periods of interest have elapsed (amounting in the extreme case to geological ages) is a tough challenge. Emphasis is put on key aspects of fuel corrosion related to fuel properties at a given age and environmental conditions expected in the repository: e.g. the fuel activity (radiolysis effects), the effects of helium build-up and of groundwater composition. A wide range of techniques, from traditional leaching experiments to advanced electrochemistry, and of materials, including spent fuel with different compositions/burnups and analogues like the so-called alpha-doped UO2, are employed for these studies. The results confirm the safety of European underground repository concepts. (authors)

  10. Spent fuel storage practices and perspectives for WWER fuel in Eastern Europe

    International Nuclear Information System (INIS)

    In this lecture the general issues and options in spent fuel management and storage are reviewed. Quantities of spent fuel world-wide and spent fuel amounts in storage as well as spent fuel capacities are presented. Selected examples of typical spent fuel storage facilities are discussed. The storage technologies applied for WWER fuel is presented. Description of other relevant storage technologies is included

  11. Nuclear Spent Fuel Management in Spain

    International Nuclear Information System (INIS)

    The radioactive waste management policy is established by the Spanish Government through the Ministry of Industry, Tourism and Commerce. This policy is described in the Cabinet-approved General Radioactive Waste Plan. ENRESA is the Spanish organization in charge of radioactive waste and nuclear SFM and nuclear installations decommissioning. The priority goal in SFM is the construction of the centralized storage facility named Almacén Temporal Centralizado (ATC), whose generic design was approved by the safety authority, Consejo de Seguridad Nuclear. This facility is planned for some 6.700 tons of heavy metal. The ATC site selection process, based on a volunteer community’s scheme, has been launched by the Government in December 2009. After the selection of a site in a participative and transparent process, the site characterization and licensing activities will support the construction of the facility. Meanwhile, extension of the on-site storage capacity has been implemented at the seven nuclear power plants sites, including past reracking at all sites. More recent activities are: reracking performed at Cofrentes NPP; dual purpose casks re-licensing for higher burnup at Trillo NPP; transfer of the spent fuel inventory at Jose Cabrera NPP to a dry-storage system, to allow decommissioning operations; and licence application of a dry-storage installation at Ascó NPP, to provide the needed capacity until the ATC facility operation. For financing planning purposes, the long-term management of spent fuel is based on direct disposal. A final decision about major fuel management options is not made yet. To assist the decision makers a number of activities are under way, including basic designs of a geological disposal facility for clay and granite host rocks, together with associated performance assessment, and supported by a R&D programme, which also includes research projects in other options like advanced separation and transmutation. (author)

  12. Spent fuel storage and transportation - ANSTO experience

    International Nuclear Information System (INIS)

    The Australian Nuclear Science and Technology Organisation (ANSTO) has operated the 10 MW DIDO class High Flux Materials Test Reactor (HIFAR) since 1958. Refuelling the reactor produces about 38 spent fuel elements each year. Australia has no power reactors and only one operating research reactor so that a reprocessing plant in Australia is not an economic proposition. The HEU fuel for HIFAR is manufactured at Dounreay using UK or US origin enriched uranium. Spent fuel was originally sent to Dounreay, UK for reprocessing but this plant was shutdown in 1998. ANSTO participates in the US Foreign Research Reactor Spent Fuel Return program and also has a contract with COGEMA for the reprocessing of non-US origin fuel

  13. Storage of spent fuel from power reactors. 2003 conference proceedings

    International Nuclear Information System (INIS)

    An International Conference on Storage of Spent Fuel from Power Reactors was organized by the IAEA in co-operation with the OECD Nuclear Energy Agency. The conference gave an opportunity to exchange information on the state of the art and prospects of spent fuel storage, to discuss the worldwide situation and the major factors influencing the national policies in this field and to identify the most important directions that national efforts and international co-operation in this area should take. The conference confirmed that the primary spent fuel management solution for the next decades will be interim storage. While the next step can be reprocessing or disposal, all spent fuel or high level waste from reprocessing must sooner or later be disposed of. The duration of interim storage is now expected to be much longer than earlier projections (up to 100 years and beyond). The storage facilities will have to be designed for these longer storage times and also for receiving spent fuel from advanced fuel cycle practices (i.e. high burnup and MOX spent fuel). It was noted that the handling and storage of spent fuel is a mature technology and meets the stringent safety requirements applicable in the different countries. The changes in nuclear policy and philosophy across the world, and practical considerations, have made storage a real necessity in the nuclear power industry. Utilities, vendors and regulators alike are addressing this adequately. The IAEA wishes to express appreciation to all chairs and co-chairs as well as all authors for their presentations to the conference and papers included in these proceedings

  14. Fact sheet on spent fuel management

    International Nuclear Information System (INIS)

    The IAEA gives high priority to safe and effective spent fuel management. As an example of continuing efforts, the 2003 International Conference on Storage of Spent Fuel from Power Reactors gathered 125 participants from 35 member states to exchange information on this important subject. With its large number of Member States, the IAEA is well-positioned to gather and share information useful in addressing Member State priorities. IAEA activities on this topic include plans to produce technical documents as resources for a range of priority topics: spent fuel performance assessment and research, burnup credit applications, cask maintenance, cask loading optimization, long term storage requirements including records maintenance, economics, spent fuel treatment, remote technology, and influence of fuel design on spent fuel storage. In addition to broader topics, the IAEA supports coordinated research projects and technical cooperation projects focused on specific needs. The proceedings of the 2003 IAEA conference on storage of spent fuel from power reactors has been ranked in the top twenty most accessed IAEA publications. These proceedings are available for free downloads at http://www-pub.iaea.org/MTCD/publications/PubDetails.asp?pubId=6924]. The IAEA organized and held a 2004 meeting focused on long term spent fuel storage provisions in Central and Eastern Europe, using technical cooperation funds to support participation by these Member States. Over ninety percent of the participants in this meeting rated its value as good or excellent, with participants noting that the IAEA is having a positive effect in stimulating communication, cooperation, and information dissemination on this important topic. The IAEA was advised in 2004 that results from a recent coordinated research project (IAEA-TECDOC-1343) were used by one Member State to justify higher clad temperatures for spent fuel in dry storage, leading to more efficient storage and reduced costs. Long term

  15. 78 FR 3853 - Retrievability, Cladding Integrity and Safe Handling of Spent Fuel at an Independent Spent Fuel...

    Science.gov (United States)

    2013-01-17

    ... Independent Spent Fuel Storage Installation and During Transportation AGENCY: Nuclear Regulatory Commission... transport of spent nuclear fuel are separate from requirements for storage of spent nuclear fuel. Because... transition from storage to transport by potentially minimizing future handling of spent fuel and......

  16. Reprocessing of spent fuel from AHWR: preliminary batch studies

    International Nuclear Information System (INIS)

    Scheme for the reprocessing of spent fuel from advanced heavy water reactor (AHWR) is under development. Present paper describes the results of some of the preliminary batch studies carried out to collect the data required for conducting counter-current studies. Studies are carried out using simulated solutions and include data on extraction as well as stripping. (author)

  17. Regeneration of ammonia borane spent fuel

    International Nuclear Information System (INIS)

    A necessary target in realizing a hydrogen (H2) economy, especially for the transportation sector, is its storage for controlled delivery, presumably to an energy producing fuel cell. In this vein, the U.S. Department of Energy's Centers of Excellence (CoE) in Hydrogen Storage have pursued different methodologies, including metal hydrides, chemical hydrides, and sorbents, for the expressed purpose of supplanting gasoline's current > 300 mile driving range. Chemical H2 storage has been dominated by one appealing material, ammonia borane (H3N-BH3, AB), due to its high gravimetric capacity of H2 (19.6 wt %) and low molecular weight (30.7 g mol-1). In addition, AB has both hydridic and protic moieties, yielding a material from which H2 can be readily released in contrast to the loss of H2 from C2H6 which is substantially endothermic. As such, a number of publications have described H2 release from amine boranes, yielding various rates depending on the method applied. The viability of any chemical H2 storage system is critically dependent on efficient recyclability, but reports on the latter subject are sparse, invoke the use of high energy reducing agents, and suffer from low yields. Our group is currently engaged in trying to find and fully demonstrate an energy efficient regeneration process for the spent fuel from H2 depleted AB with a minimum number of steps. Although spent fuel composition depends on the dehydrogenation method, we have focused our efforts on the spent fuel resulting from metal-based catalysis, which has thus far shown the most promise. Metal-based catalysts have produced the fastest rates for a single equivalent of H2 released from AB and up to 2.5 equiv. of H2 can be produced within 2 hours. While ongoing work is being carried out to tailor the composition of spent AB fuel, a method has been developed for regenerating the predominant product, polyborazylene (PB) which can be obtained readily from the decomposition of borazine or from nickel

  18. Spent Nuclear Fuel Project Technical Databook

    International Nuclear Information System (INIS)

    The Spent Nuclear Fuel (SNF) Project Technical Databook is developed for use as a common authoritative source of fuel behavior and material parameters in support of the Hanford SNF Project. The Technical Databook will be revised as necessary to add parameters as their Databook submittals become available

  19. Dissolution studies of spent nuclear fuels

    International Nuclear Information System (INIS)

    To obtain quantitative data on the dissolution of high burnup spent nuclear fuel, dissolution study have been carried out at the Department of Chemistry, JAERI, from 1984 under the contract with STA entitled 'Reprocessing Test Study of High Burnup Fuel'. In this study PWR spent fuels of 8,400 to 36,100 MWd/t in averaged burnup were dissolved and the chemical composition and distribution of radioactive nuclides were measured for insoluble residue, cladding material (hull), off-gas and dissolved solution. With these analyses basic data concerning the dissolution and clarification process in the reprocessing plant were accumulated. (author)

  20. Neutron intensity of fast reactor spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Takamatsu, Misao; Aoyama, Takafumi [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-03-01

    Neutron intensity of spent fuel of the JOYO Mk-II core with a burnup of 62,500 MWd/t and cooling time of 5.2 years was measured at the spent fuel storage pond. The measured data were compared with the calculated values based on the JOYO core management code system `MAGI`, and the average C/E approximately 1.2 was obtained. It was found that the axial neutron intensity didn`t simply follow the burnup distribution, and the neutron intensity was locally increased at the bottom end of the fuel region due to an accumulation of {sup 244}Cm. (author)

  1. Integrated spent nuclear fuel database system

    International Nuclear Information System (INIS)

    The Distributed Information Systems software Unit at the Idaho National Engineering Laboratory has designed and developed an Integrated Spent Nuclear Fuel Database System (ISNFDS), which maintains a computerized inventory of all US Department of Energy (DOE) spent nuclear fuel (SNF). Commercial SNF is not included in the ISNFDS unless it is owned or stored by DOE. The ISNFDS is an integrated, single data source containing accurate, traceable, and consistent data and provides extensive data for each fuel, extensive facility data for every facility, and numerous data reports and queries

  2. Status and prospects for spent fuel management in France

    International Nuclear Information System (INIS)

    The spent fuel arisings and storage capacities, the interface between fuel storage and transportation activities, the spent fuel storage technology, the reprocessing and recycling industrial activities in France are described in the paper. (author). 6 figs, 8 tabs

  3. Development and engineering plan for graphite spent fuels conditioning program

    International Nuclear Information System (INIS)

    Irradiated (or spent) graphite fuel stored at the Idaho Chemical Processing Plant (ICPP) includes Fort St. Vrain (FSV) reactor and Peach Bottom reactor spent fuels. Conditioning and disposal of spent graphite fuels presently includes three broad alternatives: (1) direct disposal with minimum fuel packaging or conditioning, (2) mechanical disassembly of spent fuel into high-level waste and low-level waste portions to minimize geologic repository requirements, and (3) waste-volume reduction via burning of bulk graphite and other spent fuel chemical processing of the spent fuel. A multi-year program for the engineering development and demonstration of conditioning processes is described. Program costs, schedules, and facility requirements are estimated

  4. Spent fuel management for research reactors

    International Nuclear Information System (INIS)

    There are six research reactors in Argentina using fuel elements uranium enriched from 20 to 90%. Spent fuel elements management is limited to RA-1 (Argonaut type used for training) and RA-3 (for experimentation and radioisotope production), as for the others no changes have been carried out over the cores. The first core of RA-1 was reprocessed and the second core was manufactured with U3O8 obtained from reprocessing, and once spent was transferred to a dry storage. RA-3 is pool type and fuel elements are MTR enriched at 90% 24 fuel elements in the core. Up to now 238 fuel elements have been used with burn-up from 16 to 40% and about 8 gU per plate. At present RA-3 is being remodelled in order to be able to use fuel elements enriched to 20%. After decay in a pool at reactor building, they are transported to a wet storage facility for spent fuel elements. Control rods are treated in the same way. From 15 years of storage facility operation, it follows as a consequence that the proposed objectives have been fulfilled sufficiently well, as shown by radioactivity measurements on the water hole lines (5 x 10-4 Ci/m3 of Cs-137) and the neutron interrogation method on the spent fuel elements tested presented values less than 10-8 gU/cm3, that was considered acceptable). Nevertheless, some corrosion areas were observed. Program activities involve on one hand mainly new chemical and physical controls in order to optimize spent fuel storage and, on the other hand, the study of the corrosion observed with the aim to minimize it and to ensure safe storage. (author). 6 figs, 2 tabs

  5. Laser surveillance system for spent fuel

    International Nuclear Information System (INIS)

    A laser surveillance system installed at spent fuel storage pools will provide the safeguard inspector with specific knowledge of spent fuel movement that cannot be obtained with current surveillance systems. The laser system will allow for the division of the pool's spent fuel inventory into two populations - those assemblies which have been moved and those which haven't - which is essential for maximizing the efficiency and effectiveness of the inspection effort. We have designed, constructed, and tested a laser system and have used it with a simulated BWR assembly. The reflected signal from the zircaloy rods depends on the position of the assembly, but in all cases is easily discernable from the reference scan of background with no assembly

  6. Spent Nuclear Fuel Alternative Technology Decision Analysis

    International Nuclear Information System (INIS)

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

  7. Spent Nuclear Fuel Alternative Technology Decision Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Shedrow, C.B.

    1999-11-29

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

  8. Taipower's spent fuel interim storage program

    International Nuclear Information System (INIS)

    Taipower has three twin-unit nuclear power stations in operation-2 BWR stations and one PWR station. The spent fuel pool reracking project at the Maanshan Nuclear Power Station (PWR) is scheduled to be completed by 1993. Its pool storage capacity will be expanded to accommodate storage of spent fuel generated throughout the plant's 40-year operation. For the Chinshan and Kuosheng Nuclear Power Stations (all BWRS), reracking operation at Chinshan spent fuel pools was completed in 1987 while the other is scheduled to be completed by end 1991. In summary, after reracking, the time of losing the full core dump (FCR) capacity will be extended to about the year 1999, 2004 and 2016 for Chinshan, Kuosheng and Maanshan plants respectively

  9. Development of spent fuel remote handling technology

    International Nuclear Information System (INIS)

    Since the amount of the spent fuel rapidly increases, the current R and D activities are focused on the technology development related with the storage and utilization of the spent fuel. In this research, to provide such a technology, the mechanical head-end process has been developed. In detail, the swing and shock-free crane and the RCGLUD(Remote Cask Grappling and Lid Unbolting Device) were developed for the safe transportation of the spent fuel assembly, the LLW drum and the transportation cask. Also, the disassembly devices required for the head-end process were developed. This process consists of an assembly downender, a rod extractor, a rod cutter, a fuel decladding device, a skeleton compactor, a force-rectifiable manipulator for the abnormal spent fuel disassembly, and the gantry type telescopic transporter, etc. To provide reliability and safety of these devices, the 3 dimensional graphic design system is developed. In this system, the mechanical devices are modelled and their operation is simulated in the virtual environment using the graphic simulation tools. So that the performance and the operational mal-function can be investigated prior to the fabrication of the devices. All the devices are tested and verified by using the fuel prototype at the mockup facility

  10. Spent fuel management: Current status and prospects

    International Nuclear Information System (INIS)

    The main objective of the Advisory Group on Spent Fuel Management is to review the world-wide situation in Spent Fuel Management, to define the most important directions of national efforts and international cooperation in this area, to exchange information on the present status and progress in performing the back-end of Nuclear Fuel Cycle and to elaborate the general recommendations for future Agency programmes in the field of spent fuel management. This report which is a result of the third IAEA Advisory Group Meeting (the first and second were held in 1984 and 1986) is intended to provide the reader with an overview of the status of spent fuel management programmes in a number of leading countries, with a description of the past and present IAEA activities in this field of Nuclear Fuel Cycle and with the Agency's plans for the next years, based on the proposals and recommendations of Member States. A separate abstract was prepared for each of 14 papers presented at the advisory group meeting. Refs, figs and tabs

  11. Spent fuel storage system for LMFBR fuel experiments

    International Nuclear Information System (INIS)

    Fuel that had been irradiated in the Argonne National Laboratory Experimental Breeder Reactor II (EBR-II) at Idaho Falls, Idaho, and examined at the Hanford Engineering Development Laboratory at Richland, Washington, was placed in long term retrievable storage utilizing a system designed at Hanford. The Spent Fuel Storage Cask system was designed for transport and storage of a large quantity of spent fuel at the Hanford 200 Area transuranic (TRU) asphalt storage pad. The entire system is designed for long term retrievable storage to allow future reprocessing of the fuel. The system was designed to meet the criticality, shielding, and thermal requirements for a maximum fuel load of four kilograms fissile. The Spent Fuel Storage Cask was built to transport and store the fuel from EBR-II on the TRU asphalt storage pad

  12. SEU43 fuel bundle shielding analysis during spent fuel transport

    Energy Technology Data Exchange (ETDEWEB)

    Margeanu, C. A.; Ilie, P.; Olteanu, G. [Inst. for Nuclear Research Pitesti, No. 1 Campului Street, Mioveni 115400, Arges County (Romania)

    2006-07-01

    The basic task accomplished by the shielding calculations in a nuclear safety analysis consist in radiation doses calculation, in order to prevent any risks both for personnel protection and impact on the environment during the spent fuel manipulation, transport and storage. The paper investigates the effects induced by fuel bundle geometry modifications on the CANDU SEU spent fuel shielding analysis during transport. For this study, different CANDU-SEU43 fuel bundle projects, developed in INR Pitesti, have been considered. The spent fuel characteristics will be obtained by means of ORIGEN-S code. In order to estimate the corresponding radiation doses for different measuring points the Monte Carlo MORSE-SGC code will be used. Both codes are included in ORNL's SCALE 5 programs package. A comparison between the considered SEU43 fuel bundle projects will be also provided, with CANDU standard fuel bundle taken as reference. (authors)

  13. Impact Analysis for Fuel Assemblies in Spent Fuel Storage Rack

    International Nuclear Information System (INIS)

    The design and structural integrity evaluation of a spent fuel storage rack (SFSR) utilized for storing and protecting the spent fuel assemblies generated during the operation of a reactor are very important in terms of nuclear safety and waste management. The objective of this study is to show the validity of the SFSR design as well as fuel assembly through a structural integrity evaluation based on a numerical analysis. In particular, a dynamic time history analysis considering the gaps between the fuel assemblies and the walls of the storage cell pipes in the SFSR was performed to check the structural integrity of the fuel assembly and storage cell pipe

  14. Concrete Spent Fuel Cask Criticality Calculation

    International Nuclear Information System (INIS)

    A preliminary analysis of the concrete cask for the intermediate dry storage of the spent fuel of NPP Krsko should include an estimation of the effective multiplication factor. Assuming 16x16 fuel elements, 4.3% initial enrichment, 45 GWd/tU burnup and 10 years cooling time, a concrete spent fuel capacity of 10 spent fuel assemblies is proposed. Fuel assemblies are placed inside inner cavity in a 'basket' - a boron (1%) doped steel structure. Heavy concrete (25% Fe), 45 cm thick, is enclosed in a carbon steel shell. There is also a stainless steel (SS304) lining of the storage cavity. Isotope inventory of the spent fuel after a 10 years cooling time is calculated using ORIGEN-S functional module of the SCALE-4.2 code package. The effective multiplication factor keff of dry (helium filled) and wet (water filled) cask for fresh and used fuel is calculated using CSAS4 Monte Carlo method based control module of the same SCALE-4.2 code package. The obtained results of keff of the dry cask for fresh and spent fuel are well below the required 0.95 value, but those for the water filled cask are above this value. Therefore, several additional calculations of the keff varying the thickness of a boral basket structure which had replaced the stainless steel one were done. It turned out that at least a 1.5 cm thick boral wall was needed to meet the required 0.95 value for keff. (author)

  15. Spent Fuel Behaviour During Interim Storage

    International Nuclear Information System (INIS)

    Objective: Review of spent fuel data relevant for future storage in Spain Perform destructive and non-destructive examinations on irradiated and non-irradiated fuel rods relevant to Spanish spent fuel management. Research approach: Among the programmes initiated in the last years (finished or about to be finished) one may highlight the following ones: • Isotopic measurements on high burnup fuels: up to 75 GW·d·t(U)-1 PWR and 53 GW·d·t(U)-1 BWR peak values; • Mechanical tests on high burnup PWR (ZIRLO) cladding and BWR (Zry-2) cladding samples; • Mechanical tests on unirradiated ZIRLO rods. Influence of hydrides content; • Modelling of mechanical tests with unirradiated claddings; • Interim storage creep modelling; • Burnup measurement equipment; • Fuel database

  16. New Methods for Evaluation of Spent Fuel Condition during Long-Term Storage in Slovakia

    International Nuclear Information System (INIS)

    Experiences with an advanced spent nuclear fuel management in Slovakia are presented in this paper. The evaluation and monitoring procedures are based on practices at the Slovak wet interim spent fuel storage facility in NPP Jaslovske Bohunice. Since 1999, leak testing of WWER-440 fuel assemblies are provided by special leak tightness detection system Sipping in pool delivered by Framatomeanp with external heating for the precise defects determination. In 2006, a new inspection stand SVYP-440 for monitoring of spent nuclear fuel condition was inserted. This stand has the possibility to open WWER-440 fuel assemblies and examine fuel elements. Optimal ways of spent fuel disposal and monitoring of nuclear fuel condition were designed. With appropriate approach of conservativeness, new factor for specifying spent fuel leak tightness is introduced in the paper. By using computer simulations (based on SCALE 4.4a code) for fission products creation and measurements by system Sipping in pool the limit values of leak tightness were established.

  17. Hanford spent fuel inventory baseline

    International Nuclear Information System (INIS)

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

  18. Hanford spent fuel inventory baseline

    Energy Technology Data Exchange (ETDEWEB)

    Bergsman, K.H.

    1994-07-15

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

  19. Classification of spent nuclear fuel (SNF)

    International Nuclear Information System (INIS)

    This report is one of a series of eight prepared by E. R. Johnson Associates, Inc. (JAI) under ORNL's contract with DOE's OCRWM Systems Integration Program and in support of the Annual Capacity Report (ACR) Issue Resolution Process. The report topics relate specifically to the list of high-priority technical waste acceptance issues developed jointly by DOE and a utility-working group. JAI performed various analyses and studies on each topic to serve as starting points for further discussion and analysis leading eventually to finalizing the process by which DOE will accept spent fuel and waste into its waste management system. This document discusses the classification of spent nuclear fuels

  20. Dry spent fuel transfer system design

    International Nuclear Information System (INIS)

    The design of a system for the transfer of spent fuel outside the spent fuel pool is being developed by the Electric Power Research Instiute and the U.S. Department of Energy. The design approach uses proven equipment design concepts for simplicity and flexibility. The design appears to be technically, operationally, and economically feasible. In addition, U.S. Nuclear Regulatory Commission (NRC) approval under 10CFR72 appears feasible. The final design will be considered for submittal to the NRC for review. A demonstration at an existing DOE facility is being considered

  1. Thermal-hydraulic modeling of reracked spent fuel pool

    International Nuclear Information System (INIS)

    The simple model of the spent fuel pool for computer code GOTHIC, which enables calculation of thermal-hydraulic parameters of the reracked spent fuel pool of NPP Krsko, has been developed. This model encompasses all basic characteristics of spent fuel pool, which are necessary to simulate a global behavior of spent fuel pool cooling. Within this model, the temperatures of the spent fuel pool for steady state, as well as temperature increases after loss of cooling were calculated for NPP Krsko reracked spent fuel pool. (author)

  2. Spent Fuel Management of NPPs in Argentina

    International Nuclear Information System (INIS)

    There are two Nuclear Power Plants in operation in Argentina: “Atucha I” (unique PHWR design) in operation since 1974, and “Embalse” (typical CANDU reactor) which started operation in 1984. Both NPPs are operated by “Nucleoeléctrica Argentina S.A” which is responsible for the management and interim storage of spent fuel till the end of the operative life of the plants. A third NPP, “Atucha II” is under construction, with a similar design of Atucha I. The legislative framework establishes that after final shutdown of a NPP the spent fuel will be transferred to the “National Atomic Energy Commission”, which is also responsible for the decommissioning of the Plants. In Atucha I, the spent fuel is stored underwater, until another option is implemented meanwhile in Embalse the spent fuel is stored during six years in pools and then it is moved to a dry storage. A decision about the fuel cycle back-end strategy will be taken before year 2030. (author)

  3. Spent fuel surveillance and monitoring methods

    International Nuclear Information System (INIS)

    The Technical Committee Meeting on ''Spent Fuel Surveillance and Monitoring Methods'' (27-30 October 1987) has been organized in accordance with recommendations of the International Standing Advisory Group on Spent Fuel Management during its second meeting in 1986. The aim of the meeting was to discuss the above questions with emphasis on current design and operation criteria, safety principles and licensing requirements and procedures in order to prevent: inadvertent criticality, undue radiation exposure, unacceptable release of radioactivity as well as control for loss of storage pool water, crud impact, water chemistry, distribution and behaviour of particulates in cooling water, oxidation of intact and failed fuel rods as a function of temperature and burnup; distribution of radiation and temperature through dry cask wall, monitoring of leakages from pools and gas escapes from dry storage facilities, periodical integrity tests of the containment barriers, responsibilities of organizations for the required operation, structure, staff and subordination, etc. The presentations of the Meeting were divided into two sessions: Spent fuel surveillance programmes and practice in Member States (4 papers); Experimental methods developed in support of spent fuel surveillance programmes (5 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs

  4. Spent Fuel Storage Operation - Lessons Learned

    International Nuclear Information System (INIS)

    Experience gained in planning, constructing, licensing, operating, managing and modifying spent fuel storage facilities in some Member States now exceeds 50 years. Continual improvement is only achieved through post-project review and ongoing evaluation of operations and processes. This publication is aimed at collating and sharing lessons learned. Hopefully, the information provided will assist Member States that already have a developed storage capability and also those considering development of a spent nuclear fuel storage capability in making informed decisions when managing their spent nuclear fuel. This publication is expected to complement the ongoing Coordinated Research Project on Spent Fuel Performance Assessment and Research (SPAR-III); the scope of which prioritizes facility operational practices in lieu of fuel and structural components behaviour over extended durations. The origins of the current publication stem from a consultants meeting held on 10-12 December 2007 in Vienna, with three participants from the IAEA, Slovenia and USA, where an initial questionnaire on spent fuel storage was formulated (Annex I). The resultant questionnaire was circulated to participants of a technical meeting, Spent Fuel Storage Operations - Lessons Learned. The technical meeting was held in Vienna on 13-16 October 2008, and sixteen participants from ten countries attended. A consultants meeting took place on 18-20 May 2009 in Vienna, with five participants from the IAEA, Slovenia, UK and USA. The participants reviewed the completed questionnaires and produced an initial draft of this publication. A third consultants meeting took place on 9-11 March 2010, which six participants from Canada, Hungary, IAEA, Slovenia and the USA attended. The meeting formulated a second questionnaire (Annex II) as a mechanism for gaining further input for this publication. A final consultants meeting was arranged on 20-22 June 2011 in Vienna. Six participants from Hungary, IAEA, Japan

  5. Commercial waste and spent fuel packaging program. Annual report

    International Nuclear Information System (INIS)

    This document is a report of activities performed by Westinghouse Advanced Energy Systems Division - Nevada Operations in meeting subtask objectives described in the Nevada Nuclear Waste Storage Investigations (NNWSI) Project Plan and revised planning documentation for Fiscal Year (FY) 1981. Major activities included: completion of the first fuel exchange in the Spent Fuel Test - Climax program; plasma arc welder development; modification and qualification of a canister cutter; installation, and activation of a remote area monitor, constant air monitor and an alpha/beta/gamma counting system; qualification of grapples required to handle pressurized water reactor or boiling water reactor fuel and high level waste (HLW) logs; data acquisition from the 3 kilowatt soil temperature test, 2 kw fuel temperature test, and 2 kw drywell test; calorimetry of the fuel assembly used in the fuel temperature test; evaluation of moisture accumulation in the drywells and recommendations for proposed changes; revision of safety assessment document to include HLW log operations; preparation of quality assurance plan and procedures; development and qualification of all equipment and procedures to receive, handle and encapsulate both the HLW log and spent fuel for the basalt waste isolation program/near surface test facility program; preliminary studies of both the requirements to perform waste packaging for the test and evaluation facility and a cask storage program for the DOE Interim Spent Fuel Management program; and remote handling operations on radioactive source calibration in support of other contractors

  6. Advanced Fuel Bundles for PHWRS

    International Nuclear Information System (INIS)

    The fuel used by NPCIL presently is natural uranium dioxide in the form of 19- element fuel bundles for 220 MWe PHWRs and 37-element fuel bundles for the TAPP-3&4 540 MWe units. The new 700 MWe PHWRs also use 37-element fuel bundles. These bundles are of short 0.5 m length of circular geometry. The cladding is of collapsible type made of Zircaloy-4 material. PHWRs containing a string of short length fuel bundles and the on-power refueling permit flexibility in using different advanced fuel designs and in core fuel management schemes. Using this flexibility, alternative fuel concepts are tried in Indian PHWRs. The advances in PHWR fuel designs are governed by the desire to use resources other than uranium, improve fuel economics by increasing fuel burnup and reduce overall spent nuclear fuel waste and improve reactor safety. The rising uranium prices are leading to a relook into the Thorium based fuel designs and reprocessed Uranium based and Plutonium based MOX designs and are expected to play a major role in future. The requirement of synergism between different type of reactors also plays a role. Increase in fuel burnup beyond 15 000 MW∙d/TeU in PHWRs, using higher fissile content materials like slightly enriched uranium, Mixed Oxide and Thorium Oxide in place of natural uranium in fuel elements, was studied many PHWR operating countries. The work includes reactor physics studies and test irradiation in research reactors and power reactors. Due to higher fissile content these bundles will be capable of delivering higher burnup than the natural uranium bundles. In India the fuel cycle flexibility of PHWRs is demonstrated by converting this type of technical flexibility to the real economy by irradiating these different types of advanced fuel materials namely Thorium, MOX, SEU, etc. The paper gives a review of the different advanced fuel design concepts studied for Indian PHWRs. (author)

  7. Radioactivity of spent TRIGA fuel

    International Nuclear Information System (INIS)

    Some of the oldest TRIGA fuel in the Malaysian Reaktor TRIGA PUSPATI (RTP) is approaching the limit of its end of life with burn-up of around 20%. Hence it is prudent for us to start planning on the replacement of the fuel in the reactor and other derivative activities associated with it. In this regard, we need to understand all of the risk associated with such operation and one of them is to predict the radioactivity of the fuel, so as to estimate the safety of our working conditions. The radioactivity of several fuels are measured and compared with simulation results to confirm the burnup levels of the selected fuels. The radioactivity measurement are conducted inside the water tank to reduce the risk of exposure and in this case the detector wrapped in plastics are lowered under water. In nuclear power plant, the general practice was to continuously burn the fuel. In research reactor, most operations are based on the immediate needs of the reactor and our RTP for example operate periodically. By integrating the burnup contribution for each core configuration, we simplify the simulation of burn up for each core configuration. Our results for two (2) fuel however indicates that the dose from simulation underestimate the actual dose from our measurements. Several postulates are investigated but the underlying reason remain inconclusive

  8. Radioactivity of spent TRIGA fuel

    Energy Technology Data Exchange (ETDEWEB)

    Usang, M. D., E-mail: mark-dennis@nuclearmalaysia.gov.my; Nabil, A. R. A.; Alfred, S. L.; Hamzah, N. S.; Abi, M. J. B.; Rawi, M. Z. M.; Abu, M. P. [Reactor Department, Malaysian Nuclear Agency, Bangi, 43000 Kajang, Selangor (Malaysia)

    2015-04-29

    Some of the oldest TRIGA fuel in the Malaysian Reaktor TRIGA PUSPATI (RTP) is approaching the limit of its end of life with burn-up of around 20%. Hence it is prudent for us to start planning on the replacement of the fuel in the reactor and other derivative activities associated with it. In this regard, we need to understand all of the risk associated with such operation and one of them is to predict the radioactivity of the fuel, so as to estimate the safety of our working conditions. The radioactivity of several fuels are measured and compared with simulation results to confirm the burnup levels of the selected fuels. The radioactivity measurement are conducted inside the water tank to reduce the risk of exposure and in this case the detector wrapped in plastics are lowered under water. In nuclear power plant, the general practice was to continuously burn the fuel. In research reactor, most operations are based on the immediate needs of the reactor and our RTP for example operate periodically. By integrating the burnup contribution for each core configuration, we simplify the simulation of burn up for each core configuration. Our results for two (2) fuel however indicates that the dose from simulation underestimate the actual dose from our measurements. Several postulates are investigated but the underlying reason remain inconclusive.

  9. Radioactivity of spent TRIGA fuel

    Science.gov (United States)

    Usang, M. D.; Nabil, A. R. A.; Alfred, S. L.; Hamzah, N. S.; Abi, M. J. B.; Rawi, M. Z. M.; Abu, M. P.

    2015-04-01

    Some of the oldest TRIGA fuel in the Malaysian Reaktor TRIGA PUSPATI (RTP) is approaching the limit of its end of life with burn-up of around 20%. Hence it is prudent for us to start planning on the replacement of the fuel in the reactor and other derivative activities associated with it. In this regard, we need to understand all of the risk associated with such operation and one of them is to predict the radioactivity of the fuel, so as to estimate the safety of our working conditions. The radioactivity of several fuels are measured and compared with simulation results to confirm the burnup levels of the selected fuels. The radioactivity measurement are conducted inside the water tank to reduce the risk of exposure and in this case the detector wrapped in plastics are lowered under water. In nuclear power plant, the general practice was to continuously burn the fuel. In research reactor, most operations are based on the immediate needs of the reactor and our RTP for example operate periodically. By integrating the burnup contribution for each core configuration, we simplify the simulation of burn up for each core configuration. Our results for two (2) fuel however indicates that the dose from simulation underestimate the actual dose from our measurements. Several postulates are investigated but the underlying reason remain inconclusive.

  10. TRIGA spent-fuel storage criticality analysis

    International Nuclear Information System (INIS)

    A criticality safety analysis of a pool-type storage for spent TRIGA Mark II reactor fuel is presented. Two independent computer codes are applied: the MCNP Monte Carlo code and the WIMS lattice cell code. Two types of fuel elements are considered: standard fuel elements with 12 wt% uranium concentration and FLIP fuel elements. A parametric study of spent-fuel storage lattice pitch, fuel element burnup, and water density is presented. Normal conditions and postulated accident conditions are analyzed. A strong dependence of the multiplication factor on the distance between the fuel elements and on the effective water density is observed. A multiplication factor 6.5 cm, regardless of the fuel element type and burnup. At shorter distances, the subcriticality can be ensured only by adding absorbers to the array of fuel rods even if the fuel rods were burned to ∼20% burnup. The results of both codes agree well for normal conditions. The results show that WIMS may be used as a complement to the Monte Carlo code in some parts of the criticality analysis

  11. Spent fuel treatment at ANL-West

    International Nuclear Information System (INIS)

    At Argonne National Laboratory-West (ANL-West) there are several thousand kilograms of metallic spent nuclear fuel containing bond sodium. This fuel will be treated in the Fuel Cycle Facility at ANL-West to produce stable waste forms for storage and disposal. The treatment operations will employ a pyrochemical process that also has applications for treating most of the fuel types within the Department of Energy complex. The treatment equipment is in its last stage of readiness, and operations will begin in the Fall of 1994

  12. BR-100 spent fuel shipping cask development

    International Nuclear Information System (INIS)

    Continued public acceptance of commercial nuclear power is contingent to a large degree on the US Department of Energy (DOE) establishing an integrated waste management system for spent nuclear fuel. As part of the from-reactor transportation segment of this system, the B ampersand W Fuel Company (BWFC) is under contract to the DOE to develop a spent-fuel cask that is compatible with both rail and barge modes of transportation. Innovative design approaches were the keys to achieving a cask design that maximizes payload capacity and cask performance. The result is the BR-100, a 100-ton rail/barge cask with a capacity of 21 PWR or 52 BWR ten-year cooled, intact fuel assemblies. 3 figs

  13. KUR fuels: Spent fuel return and reduced enrichment program

    International Nuclear Information System (INIS)

    The Research Reactor Institute of Kyoto University (KURRI) has more than 250 MTR-type HEU spent fuel elements. They have been stored in water pools after irradiation in the Kyoto University Research Reactor (KUR) core. The longest pool residence time is 25 years. In accordance with the Foreign Research Reactor Spent Nuclear Fuel Receipt Program of the United States, sixty KUR spent fuel elements were shipped from KURRI to the Savannah River Site of the US DOE in August 1999. This shipment was done successfully through a public port in Osaka Prefecture, Japan. This is the first shipment in the past twenty-six years after the last shipment through the Yokohama Port. Concerning the use of a public port, we had to solve many issues for public acceptance. In this paper, we describe how we have stored the spent fuels for a long time with high integrity and how we have obtained public acceptance for the transport. So far we have HEU fuels to be used until March 2004, which is already agreed by US DOE. We are looking for candidate LEU fuel materials after HEU, and also spent fuel handling of the new LEU fuel. (author)

  14. Spent nuclear fuel project integrated schedule plan

    International Nuclear Information System (INIS)

    The Spent Nuclear Fuel Integrated Schedule Plan establishes the organizational responsibilities, rules for developing, maintain and status of the SNF integrated schedule, and an implementation plan for the integrated schedule. The mission of the SNFP on the Hanford site is to provide safe, economic, environmentally sound management of Hanford SNF in a manner which stages it to final disposition. This particularly involves K Basin fuel

  15. Regional spent fuel storage facility (RSFSF)

    International Nuclear Information System (INIS)

    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)

  16. Self-interrogation of spent fuel

    International Nuclear Information System (INIS)

    A new method for the assay of spent-fuel assemblies has been developed that eliminates the need for external isotopic neutron sources, yet retains the advantages of an active interrogation system. The assay is accomplished by changing the reactivity of the system and correlating the measurements to burnup

  17. Spent nuclear fuel project product specification

    International Nuclear Information System (INIS)

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

  18. Spent Nuclear Fuel Storage Program user's guide

    International Nuclear Information System (INIS)

    The purpose of this manual is to present procedures to execute the Spent Nuclear Fuel Storage Model (SNFSM) program. This manual includes an overview of the model, operating environment, input and output specifications and user procedures. An example of the execution of the program is included to assist potential users

  19. Operation of spent fuel storage facilities

    International Nuclear Information System (INIS)

    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

  20. Foreign encapsulation concepts of spent fuel

    International Nuclear Information System (INIS)

    The foreign encapsulation concepts of spent fuel in countries, which have geologies similar to that in Finland are reviewed. The main interests in this report were alternative concepts of canister materials as well as the design, fabrication and testing programs planned to evaluate the canister performance. Also present concepts of mined geological repositories and facilities for waste handling before final disposal are reviewed. (author)

  1. Spent Nuclear Fuel (SNF) Project Product Specification

    Energy Technology Data Exchange (ETDEWEB)

    PAJUNEN, A.L.

    2000-01-20

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

  2. The cascad spent fuel dry storage facility

    International Nuclear Information System (INIS)

    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

  3. Reactor BN-350 spent fuel handling

    International Nuclear Information System (INIS)

    In pursuance with the Decree No. 456 of the Government of Kazakhstan, dated 22 April of 1999, BN-350 reactor shall be converted to SAFSTOR state for 50 years period followed by dismantling and disposal. Nuclear fuel unloading and safe arrangement for long-term storage in a specially constructed storage facility outside the reactor plant is one of the main criteria of reactor conversion of SAFSTOR state. In accordance with principles of nonproliferation and cancellation of 'nuclear test sites' the 'Baikal-1' bench-top complex located at National Nuclear Center of the Republic of Kazakhstan site is defined by Kazakhstan side decision as a location for long-term storage of BN-350 spent fuel. Project of BN-350 spent fuel transportation and arrangement for long-term storage includes several stages for completion. Currently the spent fuel is unloaded and packed into sealed jackets filled with inert gas. Thus the first Project stage - spent fuel preparation for transportation and provision of necessary temporary storage condition in BN-350 ponds till the moment of transportation is completed. Spent fuel transportation to the place of long-term storage is suggested to conduct in transport packaging casks (TPC) by railway to Kurchatov station where casks will be reloaded for transportation by auto-trailers. For the second Project stage the works have to be carried out on development of the following features: TPC design, technological process of transportation, design of storage facility and both nuclear fuel loading and reloading platforms. This part of this stage is yet completed and main project and technical solution are reported (TPC based on the one pack metal cask, technological process of TPC handling, Silo-type storage facility. As one of the option the TPC is reported based on heavy metal-concrete cask and indented for spent fuel transportation and storage (up to seven canisters with SFAs). Advantages and disadvantages of these TPC are reported compared to that of

  4. Spent fuel management strategy in Italy

    International Nuclear Information System (INIS)

    As a consequence of a national referendum in 1987, the Italian Government decided to close definitively all operating NPPs in Italy. Plans for decommissioning of the NPPs and disposal of the spent fuel had to be reviewed and the strategies revisited. The majority of spent fuel was by large that generated by ENEL NPPs, which decided to proceed with the interim storage of the spent fuel (< 250 t/HM) not covered by reprocessing contracts. ENEL finally decided to follow the strategy of interim dry storage in metallic casks on the plant sites, which could ensure a timely removal of the fuel from the to be decommissioned plant pools, in compliance with decommissioning programmes, independently from the availability of a centralized interim storage site. Therefore, the casks will be stored provisionally on Trino and Caorso sites, then they will be transported to the centralized interim facility, as soon as it will be made available by the Government. Current planning foresees that the Trino spent fuel pool shall be emptied by the end of 2002 and the Caorso pool at the end of 2004. An international bidding phase is currently underway. A smaller residual quantity of spent fuel is also currently owned in Italy by ENEA, the National Agency, responsible also for the nuclear research. Also ENEA has a programme of storing its spent fuel in dry metallic casks with the aim of transporting them to the national storage site as soon as it will be available. ENEL's Technical Specifications for the casks are stringent, but in line with other European installations of the same type, taking into account also recent US NRC regulatory documents, in particular on protection against aircraft crash. Design margins to accommodate site characteristics not currently identified (the European Utility Requirements reference site parameters have been used), have been introduced. Some important issues are identified, such as: definition and identification of failed fuel elements and/or pins, specific

  5. Expected very-near-field thermal environments for advanced spent-fuel and defense high-level waste packages

    International Nuclear Information System (INIS)

    The very-near-field thermal environments expected in a nuclear waste repository in a salt formation have been evaluated for the Westinghouse Form I advanced waste package concepts. The repository descriptions used to supplement the waste package designs in these analyses are realistic and take into account design constraints to assure conservatism. As a result, areal loadings are well below the acceptable values established for salt repositories. Predicted temperatures are generally well below any temperature limits which have been discussed for waste packages in a salt formation. These low temperatures result from the conservative repository designs. Investigations are also made of the sensitivity of these temperatures to areal loading, canister separation, and other design features

  6. The psychosocial consequences of spent fuel disposal

    International Nuclear Information System (INIS)

    In this report the potential psychosocial consequences of spent fuel disposal to inhabitants of a community are assessed on the basis of earlier research. In studying the situation, different interpretations and meanings given to nuclear power are considered. First, spent fuel disposal is studied as fear-arousing and consequently stressful situation. Psychosomatic effects of stress and coping strategies used by an individual are presented. Stress as a collective phenomenon and coping mechanisms available for a community are also assessed. Stress reactions caused by natural disasters and technological disasters are compared. Consequences of nuclear power plant accidents are reviewed, e.g. research done on the accident at Three Mile Island power plant. Reasons for the disorganising effect on a community caused by a technological disaster are compared to the altruistic community often seen after natural disasters. The potential reactions that a spent fuel disposal plant can arouse in inhabitants are evaluated. Both short-term and long-term reactions are evaluated as well as reactions under normal functioning, after an incident and as a consequence of an accident. Finally an evaluation of how the decision-making system and citizens' opportunity to influence the decision-making affect the experience of threat is expressed. As a conclusion we see that spent fuel disposal can arouse fear and stress in people. However, the level of the stress is probably low. The stress is at strongest at the time of the starting of the spent fuel disposal plant. With time people get used to the presence of the plant and the threat experienced gets smaller. (orig.)

  7. Numerical Estimation of the Spent Fuel Ratio

    Energy Technology Data Exchange (ETDEWEB)

    Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Durbin, Samuel [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wilke, Jason [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Margraf, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dunn, T. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-01-01

    Sabotage of spent nuclear fuel casks remains a concern nearly forty years after attacks against shipment casks were first analyzed and has a renewed relevance in the post-9/11 environment. A limited number of full-scale tests and supporting efforts using surrogate materials, typically depleted uranium dioxide (DUO 2 ), have been conducted in the interim to more definitively determine the source term from these postulated events. However, the validity of these large- scale results remain in question due to the lack of a defensible spent fuel ratio (SFR), defined as the amount of respirable aerosol generated by an attack on a mass of spent fuel compared to that of an otherwise identical surrogate. Previous attempts to define the SFR in the 1980's have resulted in estimates ranging from 0.42 to 12 and include suboptimal experimental techniques and data comparisons. Because of the large uncertainty surrounding the SFR, estimates of releases from security-related events may be unnecessarily conservative. Credible arguments exist that the SFR does not exceed a value of unity. A defensible determination of the SFR in this lower range would greatly reduce the calculated risk associated with the transport and storage of spent nuclear fuel in dry cask systems. In the present work, the shock physics codes CTH and ALE3D were used to simulate spent nuclear fuel (SNF) and DUO 2 targets impacted by a high-velocity jet at an ambient temperature condition. These preliminary results are used to illustrate an approach to estimate the respirable release fraction for each type of material and ultimately, an estimate of the SFR. This page intentionally blank

  8. The Spent Fuel Management in Finland and Modifications of Spent Fuel Storages

    International Nuclear Information System (INIS)

    The objective of this presentation is to share the Finnish regulator's (STUK) experiences on regulatory oversight of the enlargement of a spent fuel interim storage. An overview of the current situation of spent fuel management in Finland will also be given. In addition, the planned modifications and requirements set for spent fuel storages due to the Fukushima accident are discussed. In Finland, there are four operating reactors, one under construction and two reactors that have a Council of State's Decision-in-Principle to proceed with the planning and licensing of a new reactor. In Olkiluoto, the two operating ASEA-Atom BWR units and the Areva EPR under construction have a shared interim storage for the spent fuel. The storage was designed and constructed in 1980's. The option for enlarging the storage was foreseen in the original design. Considering three operating units to produce their spent fuel and the final disposal to begin in 2022, extra space in the spent fuel storage is estimated to be needed in around 2014. The operator decided to double the number of the spent fuel pools of the storage and the construction began in 2010. The capacity of the enlarged spent fuel storage is considered to be sufficient for the three Olkiluoto units. The enlargement of the interim storage was included in Olkiluoto NPP 1 and 2 operating license. The licensing of the enlargement was conducted as a major plant modification. The operator needed the approval from STUK to conduct the enlargement. Prior to the construction of this modification, the operator was required to submit the similar documentation as needed for applying for the construction license of a nuclear facility. When conducting changes in an old nuclear facility, the new safety requirements have to be followed. The major challenge in the designing the enlargement of the spent fuel storage was to modify it to withstand a large airplane crash. The operator chose to cover the pools with protecting slabs and also to

  9. Reprocessing of spent nuclear fuel

    International Nuclear Information System (INIS)

    This volume contains the following reports: Experimental facility for testing and development of pulsed columns and auxiliary devices; Chemical-technology study of the modified 'Purex' process; Chemical and radiometric control analyses; Chromatographic separation of rare earth elements on paper treated by di-n butylphosphate; Preliminary study of some organic nitrogen extracts significant in fuel reprocessing

  10. Stressmeter placement at spent fuel test in climax granite

    International Nuclear Information System (INIS)

    Vibrating wire stressmeters were installed in the Spent Fuel Facility at the Nevada Test Site. These stressmeters will measure the changes in in situ stress during the five-year spent fuel test. Before installation, laboratory tests were conducted to study reproducibility of placement and to develop a program hopefully to reduce corrosion of the stressmeters while in place at the Spent Fuel Facility. These laboratory tests are discussed along with the installation of the stressmeters at the Spent Fuel Facility

  11. Safety of WWER spent fuel storage, IAEA guides, national practices

    International Nuclear Information System (INIS)

    In this lecture are presented: IAEA safety programme; IAEA safety related documents until 1996; safety guides and technical reports published bu IAEA; IAEA safety related documents structure after 1996; IAEA documents on safe storage of spent fuel; Storage of WWER spent fuel and national aspects of spent fuel storage regulation. In the annex the excerpts from the Safety guide on the design of spent fuel storage facilities (IAEA Safety series No. 116) are included

  12. LWR spent fuel management in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    The spent fuel management strategy in the Federal Republic of Germany is based alternatively on interim storage and subsequent reprocessing of spent fuel, or on extended storage and direct disposal of spent fuel. By economic and strategic reasons the spent fuel burn-up is presently between 40 and 50 GWd/tHM and will target 55 GWd/tHM. (author). 7 figs

  13. Burnup credit implementation in spent fuel management

    International Nuclear Information System (INIS)

    The criticality safety analysis of spent fuel management systems has traditionally assumed that the fuel is fresh. This results in significant conservatism in the calculated value of the system's reactivity. The concept of allowing reactivity credit for spent fuel offers economic incentives. Burnup Credit (BUC) could reduce mass limitation during dissolution of highly enriched PWR assemblies at the La Hague reprocessing plant. Furthermore, accounting for burnup credit enables the operator to avoid the use of Gd soluble poison in the dissolver for MOX assemblies. Analyses performed by DOE and its contractors have indicated that using BUC to maximize spent fuel transportation cask capacities is a justifiable concept that would result in public risk benefits and cost savings while fully maintaining criticality safety margins. In order to allow for Fission Products and Actinides in Criticality-Safety analyses, an extensive BUC experimental programme has been developed in France in the framework of the CEA-COGEMA collaboration. The use of burnup credit implies a verification of the fuel burnup before loading for transport, storage, disposal, or reprocessing each assembly, to make sure that the burnup level achieved complies with the criteria established. Independent measurement systems, e.g. gamma spectrum detection systems, are needed to perform a true independent measurement of assembly burnup, without reliance on reactor records, using the gamma emission signatures fission products (mainly Cesium isotopes). (author)

  14. Spent fuel assembly source term parameters

    International Nuclear Information System (INIS)

    Containment of cask contents by a transport cask is a function of the cask body, one or more closure lids, and various bolting hardware, and seals associated with the cavity closure and other containment penetrations. In addition, characteristics of cask contents that impede the ability of radionuclides to move from an origin to the external environment also provide containment. In essence, multiple release barriers exist in series in transport casks, and the magnitude of the releasable activity in the cask is considerably lower than the total activity of its contents. A source term approach accounts for the magnitude of the releasable activity available in the cask by assessing the degree of barrier resistance to release provided by material characteristics and inherent barriers that impede the release of radioactive contents. Standardized methodologies for defining the spent-fuel transport packages with specified regulations have recently been developed. An essential part of applying the source term methodology involves characterizing the response of the spent fuel under regulatory conditions of transport. Thermal and structural models of the cask and fuel are analyzed and used to predict fuel rod failure probabilities. Input to these analyses and failure evaluations cover a wide range of geometrical and material properties. An important issue in the development of these models is the sensitivity of the radioactive source term generated during transport to individual parameters such as temperature and fluence level. This paper provides a summary of sensitivity analyses concentrating on the structural response and failure predictions of the spent fuel assemblies

  15. Characteristics of Loviisa NPP spent fuel

    International Nuclear Information System (INIS)

    The composition and radioactive characteristics of the spent fuel from Imatran Voima Oy's (IVO) Loviisa NPP (VVER-type PWR) have been estimated with the ORIGEN2 computer code (version 2.1) using the so-called PWR-UE cross section library. Four separate cases have been calculated. The main results of the calculations are the composition, activity, heat production, photon source and spectrum, and neutron source of the spent fuel as a function of cooling time. In the tables and figures of the report only the most important data of the large ORIGEN2 output files have been given. The ORIGEN2 results have also been compared with those calculated with the CASMO-HEX fuel assembly burnup program. (11 refs., 3 figs., 10 tabs.)

  16. Historical overview of domestic spent fuel shipments

    International Nuclear Information System (INIS)

    The purpose of this paper is to provide available historical data on most commercial and research reactor spent fuel shipments that have been completed in the United States between 1964 and 1989. This information includes data on the sources of spent fuel that has been shipped, the types of shipping casks used, the number of fuel assemblies that have been shipped, and the number of shipments that have been made. The data are updated periodically to keep abreast of changes. Information on shipments is provided for planning purposes; to support program decisions of the US Department of Energy's (DOE's) Office of Civilian Radioactive Waste Management (OCRWM); and to inform interested members of the public, federal, state, and local government, Indian tribes, and the transportation community. 5 refs., 7 figs., 2 tabs

  17. The E.D.F spent fuel

    International Nuclear Information System (INIS)

    In this article is studied the management of nuclear spent fuel. The loading of core reactor is detailed and the storage in pool is explained with the cooling time for fuel assemblies (between 1.5 years for UO2 and 2.5 years for MOX) and the defaults in pool alveoli. The fuel storage pool in nuclear power plants have to be managed in keeping the possibility to unload a complete reactor core. The right optimization goes through a high performance evacuation with the Aube storage plant and the reprocessing plant of Cogema la Hague. (N.C.)

  18. Spent Fuel Performance and Research in Germany

    International Nuclear Information System (INIS)

    For the long term management of spent fuel, dry storage in dual purpose casks and direct disposal remain the management policy in Germany. Due to the lacking progress in developing a repository for heat generating waste in deep geological formations the likely extension of the interim storage period beyond 40 years raises issues especially regarding degradation of fuel and the functional reliability of the cask closure system. OBJECTIVE: • Investigation of irradiated fuel rods exposed to mechanic stress; • Review of longterm performance of lid systems bolted to the cask body and tightened by Helicoflex® metallic gaskets

  19. Overview of Spent Fuel Management In China

    International Nuclear Information System (INIS)

    This paper briefly introduces the current reprocessing situation and challenges in the world, the policy and status of the nuclear energy development and SFM for the back end of the fuel cycle in China. Chinese government has already launched the nuclear energy medium- long-term development program, the opted policy of closed fuel cycle and the technical development strategy, the projected commercial reprocessing plant, The cold uranium test for nuclear power plant spent fuel reprocessing pilot plant is finished and the radioactive test is carried out in the early this year. The R&D program of reprocessing technology is emphasized. The challenges faced by China are described. (author)

  20. 77 FR 75065 - Rescinding Spent Fuel Pool Exclusion Regulations

    Science.gov (United States)

    2012-12-19

    ... consideration of spent fuel pool storage impacts from license renewal environmental review. The petition was... rescind the regulations excluding consideration of spent fuel storage impacts from license renewal... impacts of high-density pool storage of spent fuel as insignificant and thereby permit their...

  1. Spent fuel management system in Slovak Republic

    International Nuclear Information System (INIS)

    Slovakia is using nuclear energy since 1972, when the first Nuclear Power Plant A1 started its operation. Initially, spent nuclear fuel was shipped away to Russia but with gradual building of new units of VVER-440 type and social changes in 1989, it has become necessary to build the own storage facility. In 1987, the operation of the Interim Spent Fuel Storage (ISFS) was launched. The ISFS is the wet type storage facility, where the fuel of VVER-440 type is being stored in the pools of demineralized water. The original ISFS capacity was 5040 pc s of fuel assemblies (FA). Between 1997 and 1999, a reconstruction of the ISFS had been performed aimed at increasing the seismic resistance and increasing the storage capacity up to 14,112 pieces of FA. Slovakia has currently two VVER-440 units in the process of decommissioning, 4 units in operation and two units being under construction. In the future, a construction of new nuclear power plant at Bohunice site is being considered. At the current production of spent nuclear fuel, the storage capacity will be sufficient until 2024 and therefore the construction of a new dry storage type facility is being considered. After the interim storage of fuel, the last part of the back end of fuel cycle follows, either fuel reprocessing or its permanent disposal in a deep geological repository. Slovakia has not decided yet on a definitive alternative; but in 1996-2001, a project of site selection for a deep geological repository had already been executed. (author)

  2. Regeneration of ammonia borane spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sutton, Andrew David [Los Alamos National Laboratory; Davis, Benjamin L [Los Alamos National Laboratory; Gordon, John C [Los Alamos National Laboratory

    2009-01-01

    A necessary target in realizing a hydrogen (H{sub 2}) economy, especially for the transportation sector, is its storage for controlled delivery, presumably to an energy producing fuel cell. In this vein, the U.S. Department of Energy's Centers of Excellence (CoE) in Hydrogen Storage have pursued different methodologies, including metal hydrides, chemical hydrides, and sorbents, for the expressed purpose of supplanting gasoline's current > 300 mile driving range. Chemical H{sub 2} storage has been dominated by one appealing material, ammonia borane (H{sub 3}N-BH{sub 3}, AB), due to its high gravimetric capacity of H{sub 2} (19.6 wt %) and low molecular weight (30.7 g mol{sup -1}). In addition, AB has both hydridic and protic moieties, yielding a material from which H{sub 2} can be readily released in contrast to the loss of H{sub 2} from C{sub 2}H{sub 6} which is substantially endothermic. As such, a number of publications have described H{sub 2} release from amine boranes, yielding various rates depending on the method applied. The viability of any chemical H{sub 2} storage system is critically dependent on efficient recyclability, but reports on the latter subject are sparse, invoke the use of high energy reducing agents, and suffer from low yields. Our group is currently engaged in trying to find and fully demonstrate an energy efficient regeneration process for the spent fuel from H{sub 2} depleted AB with a minimum number of steps. Although spent fuel composition depends on the dehydrogenation method, we have focused our efforts on the spent fuel resulting from metal-based catalysis, which has thus far shown the most promise. Metal-based catalysts have produced the fastest rates for a single equivalent of H{sub 2} released from AB and up to 2.5 equiv. of H{sub 2} can be produced within 2 hours. While ongoing work is being carried out to tailor the composition of spent AB fuel, a method has been developed for regenerating the predominant product

  3. Evaluation of the Use of Synroc to Solidify the Cesium and Strontium Separations Product from Advanced Aqueous Reprocessing of Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Julia Tripp; Vince Maio

    2006-03-01

    This report is a literature evaluation on the Synroc process for determining the potential for application to solidification of the Cs/Sr strip product from advanced aqueous fuel separations activities.

  4. New developments in dry spent fuel storage

    International Nuclear Information System (INIS)

    As shown in various new examples, HABOG facility (Netherlands), CERNAVODA (Candu - Romania), KOZLODUY (WWER - Bulgaria), CHERNOBYL ( RMBK - Ukraine), MAYAK (Spent Fuel from submarine and Icebreakers - Russia), recent studies allow to confirm the flexibility and performances of the CASCAD system proposed by SGN, both in safety and operability, for the dry storage of main kinds of spent fuel. The main features are: A multiple containment barrier system: as required by international regulation, 2 independent barriers are provided (tight canister and storage pit); Passive cooling, while the Fuel Assemblies are stored in an inert atmosphere and under conditions of temperature preventing from degradation of rod cladding; Sub-criticality controlled by adequate arrangements in any conditions; Safe facility meeting ICPR 60 Requirements as well as all applicable regulations (including severe weather conditions and earthquake); Safe handling operations; Retrievability of the spent fuel either during storage period or at the end of planned storage period (100 years); Future Decommissioning of the facility facilitated through design optimisation; Construction and operating cost-effectiveness. (author)

  5. Development of spent fuel remote handling technology

    Energy Technology Data Exchange (ETDEWEB)

    Park, B. S.; Yoon, J. S.; Hong, H. D. (and others)

    2007-02-15

    In this research, the remote handling technology was developed for the ACP application. The ACP gives a possible solution to reduce the rapidly cumulative amount of spent fuels generated from the nuclear power plants in Korea. The remote technologies developed in this work are a slitting device, a voloxidizer, a modified telescopic servo manipulator and a digital mock-up. A slitting device was developed to declad the spent fuel rod-cuts and collect the spent fuel UO{sub 2} pellets. A voloxidizer was developed to convert the spent fuel UO{sub 2} pellets obtained from the slitting process in to U{sub 3}O{sub 8} powder. Experiments were performed to test the capabilities and remote operation of the developed slitting device and voloxidizer by using simulated rod-cuts and fuel in the ACP hot cell. A telescopic servo manipulator was redesigned and manufactured improving the structure of the prototype. This servo manipulator was installed in the ACP hot cell, and the target module for maintenance of the process equipment was selected. The optimal procedures for remote operation were made through the maintenance tests by using the servo manipulator. The ACP digital mockup in a virtual environment was established to secure a reliability and safety of remote operation and maintenance. The simulation for the remote operation and maintenance was implemented and the operability was analyzed. A digital mockup about the preliminary conceptual design of an enginnering-scale ACP was established, and an analysis about a scale of facility and remote handling was accomplished. The real-time diagnostic technique was developed to detect the possible fault accidents of the slitting device. An assessment of radiation effect for various sensors was also conducted in the radiation environment.

  6. Development of spent fuel remote handling technology

    International Nuclear Information System (INIS)

    In this research, the remote handling technology was developed for the ACP application. The ACP gives a possible solution to reduce the rapidly cumulative amount of spent fuels generated from the nuclear power plants in Korea. The remote technologies developed in this work are a slitting device, a voloxidizer, a modified telescopic servo manipulator and a digital mock-up. A slitting device was developed to declad the spent fuel rod-cuts and collect the spent fuel UO2 pellets. A voloxidizer was developed to convert the spent fuel UO2 pellets obtained from the slitting process in to U3O8 powder. Experiments were performed to test the capabilities and remote operation of the developed slitting device and voloxidizer by using simulated rod-cuts and fuel in the ACP hot cell. A telescopic servo manipulator was redesigned and manufactured improving the structure of the prototype. This servo manipulator was installed in the ACP hot cell, and the target module for maintenance of the process equipment was selected. The optimal procedures for remote operation were made through the maintenance tests by using the servo manipulator. The ACP digital mockup in a virtual environment was established to secure a reliability and safety of remote operation and maintenance. The simulation for the remote operation and maintenance was implemented and the operability was analyzed. A digital mockup about the preliminary conceptual design of an enginnering-scale ACP was established, and an analysis about a scale of facility and remote handling was accomplished. The real-time diagnostic technique was developed to detect the possible fault accidents of the slitting device. An assessment of radiation effect for various sensors was also conducted in the radiation environment

  7. An Indian perspective for transportation and storage of spent fuel

    International Nuclear Information System (INIS)

    Full text: The spent fuel discharged from the reactors are temporarily stored at the reactor pool. After a certain cooling time, the spent fuel is moved to the storage locations either on or off reactor site depending on the spent fuel management strategy. As India has opted for a closed fuel cycle for its nuclear energy development, reprocessing of the spent fuel, recycling of the reprocessed plutonium and uranium and disposal of the wastes from the reprocessing operations forms the spent fuel management strategy. Since the reprocessing operations are planned to match the nuclear energy programme, storage of the spent fuel in ponds are adopted prior to reprocessing. Transport of the spent fuel to the storage locations are carried out adhering to international and national guide lines. (author)

  8. Semiautomatic spent-fuel-handling machine

    International Nuclear Information System (INIS)

    The technology for the total automation of the entire fuel handling operation, has been in existence for several years. The simplest form, or first phase of modernization, is the semiautomatic fuel handling positioning system. Several of these types of platforms are in existence today, and recently CIMCORP/PaR systems completed a semiautomatic spent-fuel handling machine (SFHM) built for Calvert Cliffs, owned by Baltimore Gas and Electric. CIMCORP has provided a semiautomated spent nuclear fuel handling system consisting of the following: (1) newly designed refueling platform bridge and trolley; (2) CIMCORP CIMROC 4000 based automatic controls technology; (3) closed circuit TV surveillance of fuel grappling operations; and (4) direct replacement of the original system provided in 1971. All SFHM motions are driven under computer control, with fully automatic bridge and trolley traverse and manually activated hoisting and grappling. Position feedback for motion control and position indication is provided by resolvers. In operation, the technician selects machine destination on a touch screen and the control system automatically positions the bridge and trolley at the desired location. Future automation of fuel grappling and hoisting can be preformed with relatively few machine modifications

  9. Application of Spatial Data Modeling Systems, Geographical Information Systems (GIS), and Transportation Routing Optimization Methods for Evaluating Integrated Deployment of Interim Spent Fuel Storage Installations and Advanced Nuclear Plants

    International Nuclear Information System (INIS)

    The objective of this siting study work is to support DOE in evaluating integrated advanced nuclear plant and ISFSI deployment options in the future. This study looks at several nuclear power plant growth scenarios that consider the locations of existing and planned commercial nuclear power plants integrated with the establishment of consolidated interim spent fuel storage installations (ISFSIs). This research project is aimed at providing methodologies, information, and insights that inform the process for determining and optimizing candidate areas for new advanced nuclear power generation plants and consolidated ISFSIs to meet projected US electric power demands for the future.

  10. Application of Spatial Data Modeling Systems, Geographical Information Systems (GIS), and Transportation Routing Optimization Methods for Evaluating Integrated Deployment of Interim Spent Fuel Storage Installations and Advanced Nuclear Plants

    Energy Technology Data Exchange (ETDEWEB)

    Mays, Gary T [ORNL; Belles, Randy [ORNL; Cetiner, Sacit M [ORNL; Howard, Rob L [ORNL; Liu, Cheng [ORNL; Mueller, Don [ORNL; Omitaomu, Olufemi A [ORNL; Peterson, Steven K [ORNL; Scaglione, John M [ORNL

    2012-06-01

    The objective of this siting study work is to support DOE in evaluating integrated advanced nuclear plant and ISFSI deployment options in the future. This study looks at several nuclear power plant growth scenarios that consider the locations of existing and planned commercial nuclear power plants integrated with the establishment of consolidated interim spent fuel storage installations (ISFSIs). This research project is aimed at providing methodologies, information, and insights that inform the process for determining and optimizing candidate areas for new advanced nuclear power generation plants and consolidated ISFSIs to meet projected US electric power demands for the future.

  11. Preliminary concepts: safeguards for spent light-water reactor fuels

    International Nuclear Information System (INIS)

    The technology available for safeguarding spent nuclear fuels from light-water power reactors is reviewed, and preliminary concepts for a spent-fuel safeguards system are presented. Essential elements of a spent-fuel safeguards system are infrequent on-site inspections, containment and surveillance systems to assure the integrity of stored fuel between inspections, and nondestructive measurements of the fuel assemblies. Key safeguards research and development activities necessary to implement such a system are identified. These activities include the development of tamper-indicating fuel-assembly identification systems and the design and development of nondestructive spent-fuel measurement systems

  12. Research reactor spent fuel management in Argentina

    International Nuclear Information System (INIS)

    The research reactor spent fuel (RRSF) management strategy will be presented as well as the interim storage experience. Currently, low-enriched uranium RRSF is in wet interim storage either at reactor site or away from reactor site in a centralized storage facility. High-enriched uranium RRSF from the centralized storage facility has been sent to the USA in the framework of the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program. The strategy for the management of the RRSF could implement the encapsulation for interim dry storage. As an alternative to encapsulation for dry storage some conditioning processes are being studied which include decladding, isotopic dilution, oxidation and immobilization. The immobilized material will be suitable for final disposal. (author)

  13. Cost analysis methodology of spent fuel storage

    International Nuclear Information System (INIS)

    The report deals with the cost analysis of interim spent fuel storage; however, it is not intended either to give a detailed cost analysis or to compare the costs of the different options. This report provides a methodology for calculating the costs of different options for interim storage of the spent fuel produced in the reactor cores. Different technical features and storage options (dry and wet, away from reactor and at reactor) are considered and the factors affecting all options defined. The major cost categories are analysed. Then the net present value of each option is calculated and the levelized cost determined. Finally, a sensitivity analysis is conducted taking into account the uncertainty in the different cost estimates. Examples of current storage practices in some countries are included in the Appendices, with description of the most relevant technical and economic aspects. 16 figs, 14 tabs

  14. Design of spent fuel storage facilities

    International Nuclear Information System (INIS)

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

  15. International safeguards for spent fuel storage

    International Nuclear Information System (INIS)

    This report analyzes the nonproliferation effectiveness and political and economic acceptability of prospective improvements in international safeguard techniques for LWR spent fuel storage. Although the applicability of item accounting considerably eases the safeguarding of stored spent fuel, the problem of verification is potentially serious. A number of simple gamma and neutron nondestructive assay techniques were found to offer considerable improvements, of a qualitative rather than quantitative nature, in verification-related data and information, and possess the major advantage of intruding very little on facility operations. A number of improved seals and monitors appear feasible as well, but improvements in the timeliness of detection will not occur unless the frequency of inspection is increased or a remote monitoring capability is established. Limitations on IAEA Safeguards resources and on the integration of results from material accounting and containment and surveillance remain problems

  16. Advanced fuels for fast reactors

    International Nuclear Information System (INIS)

    fuels originates from goals for achieving high burnup, operating at higher temperature, and the incorporation of the minor actinides (Np, Am, Cm) into the fuels. High burn-ups will allow uninterrupted reactor operations over longer periods of time and consequently, reduction of spent fuel volumes, and eventually a significant fuel cycle reduction cost. High burn-ups are however associated with physical limitations which are primary due to the swelling of the fuel and oxidation of cladding inner surface as well as the dimensional stability of core materials such as cladding and subassembly duct due to high fast neutron dose. Higher temperature operation also challenges the performance of cladding materials and hence advanced cladding materials are needed for high temperature operation. The irradiation performance database for (U,Pu)N mixed nitride (MN) fuels is substantially smaller than that for metal carbide (MC) fuels, and these fuels can be considered to be at an early stage of development relative to oxide and metal fuels. Compared to MC fuels, MN fuels exhibit less fuel swelling, lower fission gas release, however, the problem of the production of biologically hazardous 14C in nitride fuels fabricated using natural nitrogen poses a considerable concern for the nitride spent fuel waste management. Interest remains in nitride fuels due to the combination of high thermal conductivity and high melting point. The paper also addresses the technology readiness level (TRL) concept as applied to various fuel options. (author)

  17. Global Spent Fuel Logistics Systems Study (GSFLS). Volume 4. Pacific basin spent fuel logistics system

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-06-01

    This report summarizes the conceptual framework for a Pacific Basin Spent Fuel Logistics System (PBSFLS); and preliminarily describes programatic steps that might be taken to implement such a system. The PBSFLS concept is described in terms of its technical and institutional components. The preferred PBSFLS concept embodies the rationale of emplacing a fuel cycle system which can adjust to the technical and institutional non-proliferation solutions as they are developed and accepted by nations. The concept is structured on the basis of initially implementing a regional spent fuel storage and transportation system which can technically and institutionally accommodate downstream needs for energy recovery and long-term waste management solutions.

  18. Global Spent Fuel Logistics Systems Study (GSFLS). Volume 4. Pacific basin spent fuel logistics system

    International Nuclear Information System (INIS)

    This report summarizes the conceptual framework for a Pacific Basin Spent Fuel Logistics System (PBSFLS); and preliminarily describes programatic steps that might be taken to implement such a system. The PBSFLS concept is described in terms of its technical and institutional components. The preferred PBSFLS concept embodies the rationale of emplacing a fuel cycle system which can adjust to the technical and institutional non-proliferation solutions as they are developed and accepted by nations. The concept is structured on the basis of initially implementing a regional spent fuel storage and transportation system which can technically and institutionally accommodate downstream needs for energy recovery and long-term waste management solutions

  19. Wet spent fuel interim storage facility

    International Nuclear Information System (INIS)

    The article deals with the Spent Fuel Complementary Storage Unit, which was designed for the Almirante Alvaro Alberto Nuclear Power Station situated near Rio de Janeiro. The aim of the article is to present the technical solution of complementary storage. The design deals with different reactor technologies made by Areva and Westinghouse. The article also deals with the technically interesting solution of the storage tank heat removal and its dimensioning. (author)

  20. Safety assessment for spent fuel storage facilities

    International Nuclear Information System (INIS)

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

  1. Development of spent fuel examination technology

    International Nuclear Information System (INIS)

    The KMP and MBA was determined on the basis of the design of the containment and surveillance for the ACP facility. It was indicated from the safeguard ability evaluation that the pilot-scaled ACP facility is safeguardable. The prototype ASNC is composed of 20 Cd plate-attached He-3 tubes embedded in a high density PE cylinder containing a cavity of 10 cm in radius and a lead shield in its inside respectively and having a carbon neutron reflector on the top and bottom of the cylinder. The average and standard error of the neutron detection efficiency at 20 cm height are 19 % and 2 %, respectively, which satisfies the design requirements of 14 % efficiency and 2 % error. In the results of the integration analysis for the dry storage of the conditioned spent fuel, the vault type was found to be more profitable than the metal cask type, so the vault type was selected with the optimum storage system for the conditioned spent fuel. For a study of the efficient storage, the exponential decay constant for the PWR spent fuel is measured using the exponential experiment system installed in the KAERI PIEF. The neutron effective multiplication factor which is determined on the basis of the constant is shown to be roughly consistent with the calculated keff with the MCNP code. U-Nb binary alloy shows its oxidation rates one order of a magnitude less than that of the unalloyed uranium. The results from this study are applicable to the optimum stability method of the conditioned spent fuel

  2. Management and disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    The programme consists of the long-term and short-term programme, the continued bedrock investigations, the underground research laboratory, the decision-making procedure in the site selection process and information questions during the site selection process. The National Board for Spent Nuclear Fuel hereby subunits both the SKB's R and D Programme 86 and the Board's statement concerning the programme. Decisions in the matter have been made by the Board's executive committee. (DG)

  3. Treatment strategies for spent nuclear fuel

    International Nuclear Information System (INIS)

    Full text: Spent nuclear fuel is one of the big hazards of our time. The increasing demand for energy in the fast growing countries, mainly in Asia shows that nuclear power is not a passed technology belonging to history. Nuclear power is still our future if we are to be able to produce energy in a relatively cheap and environmentally friendly manner. However, everything has a drawback. In this case there are manly two: the mining of uranium ore and how to deal with the spent nuclear fuel. Mining can nowadays be made with a minimum of environmental impact and uranium mining is not more dangerous that normal coal mining. Probably even less so since the control and regulations are rather strict. Nuclear waste on the other hand may pose a threat to humanity for hundreds of thousands of years. There are mainly two strategies how to deal with it at present. Either the spent fuel is treated as waste and buried deep in the bedrock. This is planned in, e.g. Sweden and Finland. The other option is to use the uranium and plutonium in the waste for continuous energy production while the other actinides as well as the fission and corrosion products are vitrified and stored in the bedrock. Recently an 'add on' has been planned for the reprocessing countries and that is the so called transmutation option. Using this technique, not only the long lived elements in the spent fuel can be burned for energy production but the waste may be considered safe after less than 100 years. Even this is a very long time but compared to the original 100 000 years it is a time that may be possible to understand. (authors)

  4. Retrievability of spent nuclear fuel canisters

    International Nuclear Information System (INIS)

    As a part of the designing process of the Finnish spent nuclear fuel repository, a preliminary study has been carried out to investigate how the canisters could technically be retrieved to the ground surface. Possibility of retrieving a canister has been investigated in different phases of the disposal project. Retrievability has not been a design goal for the spent fuel repository. However, design of the repository includes some features that may ease the retrieval of canisters in the future. Spent fuel elements are packaged in massive copper-iron canisters, which are mechanically strong and long-lived. The repository consists of excavated tunnels in hard rock which are supposed to be very long-lived making the removal of the tunnel backfilling technically possible also in the future. As long as the bentonite buffer has not been installed the canister can be returned to the ground surface using the same equipment as was used when the canister was brought down to the repository and lowered into the hole. In the encapsulation station the spent fuel elements can be packaged in the other canister or in the transport cask. After a deposition tunnel has been backfilled and closed, the retrieval consists of tearing down the concrete structure at the entry of the deposition tunnel, removal of the tunnel backfilling, removal of the bentonite from the disposal hole and lifting up of the canister. Various methods, e.g., flushing the bentonite with saline solutions, can be used to detach the canister from a hole with fully saturated bentonite. Recovery will be technically possible also after closing of the disposal facility. Backfilling of the shafts and tunnels will be removed and additional new structures and systems will have to be built in the repository. After that canisters can be transported to the ground surface as described above. In addition, handling of the canisters at the ground surface will require additional facilities. Canisters can be packaged in the

  5. Scientific and technical aspects of creating spent nuclear fuel shipping and storage equipment

    International Nuclear Information System (INIS)

    Details of advanced capacity shipping packaging set development which allows shipping and storage of heat-producing spent fuel assemblies for pressurized water reactor VVER-1000 type are considered. Analogs and achieved advantages of TUK-146 construction are mentioned

  6. A present status for dry storage of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bang, K. S.; Lee, J. C.; Park, H. Y.; Seo, K. S

    2003-04-01

    National policy for management of a spent nuclear fuel does not establish in Korea yet. A storage capacity of a storage pool that is to store the spent nuclear fuel will be exceeded an amount of accumulation from the first Woljin nuclear power plant in 2007. Therefore it is necessary that dry storage facility is secured to store safely the spent nuclear fuel on site of the nuclear power plant until national policy for a back-end spent nuclear fuel cycle is established. In order to store safely spent nuclear fuel, it is important that the present status and technology on dry storage of spent nuclear fuel is looked over. Therefore, the present status on dry storage of spent nuclear fuel was analyzed so as to develop dry storage system and choose a proper dry storage method domestic.

  7. Current status of spent fuel management in China

    International Nuclear Information System (INIS)

    The administration organization and the laws and regulations on spent fuel management in China are described at first. In the year of 2000, CNN C and GNPJVC signed a contract of take-over mode for spent fuel, in which CNNC will take-over the spent fuel annually since 2003. In the year of 2001, CNNC and QNPC signed another contract of take-over mode for spent fuel. Transportation by road is the realistic selections of the transportation of the spent fuel from the nuclear power stations to LNFC. As for spent fuel management, the construction of CNNC's reprocessing pilot plant is going on and R and D of spent fuel management are launching continuously. (author)

  8. Recent developments - US spent fuel disposition

    International Nuclear Information System (INIS)

    One of a US utility's major risk factors in continuing to operate a nuclear plant is managing discharged spent fuel. The US Department of Energy (DOE) signed contracts with utilities guaranteeing government acceptance of spent fuel by 1988. However, on December 17, 1992, DOE Secretary Watkins wrote to Sen. J. Bennett Johnston (D-LA), Chairman of the Senate Energy Committee, indicating a reassessment of DOE's programs, the results of which will be presented to Congress in January 1993. He indicated the Department may not be able to meet the 1988 date, because of difficulty in finding a site for the Monitored Retrievable Storage facility. Watkins indicated that DOE has investigated an interim solution and decided to expedite a program to certify a multi-purpose standardized cask system for spent fuel receipt, storage, transport, and disposal. To meet the expectations of US utilities, DOE is considering a plan to use federal sites for interim storage of the casks. Secretary Watkins recommended the waste program be taken off-budget and put in a revolving fund established to ensure that money already collected from utilities will be available to meet the schedule for completion of the repository

  9. Spent Fuel Dry Storage Cask Thermal Test

    International Nuclear Information System (INIS)

    Most nuclear power plants maintain their spent fuel discharged at a reactor in wet storage pools. However, after several years of use, many pools are filled to capacity. Therefore, finding a sufficient capacity for storage is essential because of the continued delays in obtaining a safe, interim storage facility if nuclear power plants are to be allowed to continue to operate. Dry storage cask will be one solution for solving an interim storage problem. The dry storage cask consists of two separate components: an over-pack, and a canister. The structure strength part of the over-pack is made of carbon steel, and the inner cavity of the structure strength part is filled with concrete, which accomplishes the role as a radiation shield. The outer diameter of the dry storage cask is 3,550 mm and the its overall height is 5,885 mm. It weighs approximately 135 tons. The dry storage cask accommodates 24 PWR spent fuel assemblies with a burn-up of 55,000 MWD/MTU and a cooling time of 7 years. The decay heat from the 24 PWR spent fuel assemblies is 25.2 kW This paper discusses the experimental approach used to evaluate the heat transfer characteristics of the dry storage cask

  10. Considerations for the transportation of spent fuel

    International Nuclear Information System (INIS)

    In our society today the transportation of radioactive materials, and most particularly spent reactor fuel, is surrounded by considerable emotion and a wealth of information, good and bad. The transportation of these materials is viewed as unique and distinct from the transportation of other hazardous materials and as a particularly vulnerable component of the nuclear power activities of this nation. Added to this is the concept, widely held, that almost everyone is an expert on the transportation of radioactive materials. One significant contribution to this level of emotion is the notion that all roads (rail and highway), on which these goods will be transported, somehow traverse everyone's backyard. The issue of the transportation of spent fuel has thus become a political battleground. Perhaps this should not be surprising since it has all of the right characteristics for such politicization in that it is pervasive, emotional, and visible. In order that those involved in the discussion of this activity might be able to reach some rational conclusions, this paper offers some background information which might be useful to a broad range of individuals in developing their own perspectives. The intent is to address the safety of transporting spent fuel from a technical standpoint without the emotional content which is frequently a part of this argument

  11. Spent Fuel Working Group Report. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    O`Toole, T.

    1993-11-01

    The Department of Energy is storing large amounts of spent nuclear fuel and other reactor irradiated nuclear materials (herein referred to as RINM). In the past, the Department reprocessed RINM to recover plutonium, tritium, and other isotopes. However, the Department has ceased or is phasing out reprocessing operations. As a consequence, Department facilities designed, constructed, and operated to store RINM for relatively short periods of time now store RINM, pending decisions on the disposition of these materials. The extended use of the facilities, combined with their known degradation and that of their stored materials, has led to uncertainties about safety. To ensure that extended storage is safe (i.e., that protection exists for workers, the public, and the environment), the conditions of these storage facilities had to be assessed. The compelling need for such an assessment led to the Secretary`s initiative on spent fuel, which is the subject of this report. This report comprises three volumes: Volume I; Summary Results of the Spent Fuel Working Group Evaluation; Volume II, Working Group Assessment Team Reports and Protocol; Volume III; Operating Contractor Site Team Reports. This volume presents the overall results of the Working Group`s Evaluation. The group assessed 66 facilities spread across 11 sites. It identified: (1) facilities that should be considered for priority attention. (2) programmatic issues to be considered in decision making about interim storage plans and (3) specific vulnerabilities for some of these facilities.

  12. Remote technologies for handling spent fuel

    International Nuclear Information System (INIS)

    The nuclear programme in India involves building and operating power and research reactors, production and use of isotopes, fabrication of reactor fuel, reprocessing of irradiated fuel, recovery of plutonium and uranium-233, fabrication of fuel containing plutonium-239, uranium-233, post-irradiation examination of fuel and hardware and handling solid and liquid radioactive wastes. Fuel that could be termed 'spent' in thermal reactors is a source for second generation fuel (plutonium and uranium-233). Therefore, it is only logical to extend remote techniques beyond handling fuel from thermal reactors to fuel from fast reactors, post-irradiation examination etc. Fabrication of fuel containing plutonium and uranium-233 poses challenges in view of restriction on human exposure to radiation. Hence, automation will serve as a step towards remotisation. Automated systems, both rigid and flexible (using robots) need to be developed and implemented. Accounting of fissile material handled by robots in local area networks with appropriate access codes will be possible. While dealing with all these activities, it is essential to pay attention to maintenance and repair of the facilities. Remote techniques are essential here. There are a number of commonalities in these requirements and so development of modularized subsystems, and integration of different configurations should receive attention. On a long-term basis, activities like decontamination, decommissioning of facilities and handling of waste generated have to be addressed. While robotized remote systems have to be designed for existing facilities, future designs of facilities should take into account total operation with robotic remote systems. (author)

  13. Specialists' meeting on gas-cooled reactor fuel development and spent fuel treatment

    International Nuclear Information System (INIS)

    Topics covered during the 'Specialists' meeting on gas-cooled reactor fuel development and spent fuel treatment' were as follows: Selection of constructions and materials, fuel element development concepts; Fabrication of spherical coated fuel particles and fuel element on their base; investigation of fuel properties; Spent fuel treatment and storage; Head-end processing of HTGR fuel elements; investigation of HTGR fuel regeneration process; applicability of gas-fluorine technology of regeneration of spent HTGR fuel elements

  14. Status and trends in spent fuel reprocessing

    International Nuclear Information System (INIS)

    The management of spent fuel arising from nuclear power production is a crucial issue for the sustainable development of nuclear energy. The IAEA has issued several publications in the past that provide technical information on the global status and trends in spent fuel reprocessing and associated topics, and one reason for this present publication is to provide an update of this information which has mostly focused on the conventional technology applied in the industry. However, the scope of this publication has been significantly expanded in an attempt to make it more comprehensive and by including a section on emerging technologies applicable to future innovative nuclear systems, as are being addressed in such international initiatives as INPRO, Gen IV and MICANET. In an effort to be informative, this publication attempts to provide a state-of-the-art review of these technologies, and to identify major issues associated with reprocessing as an option for spent fuel management. It does not, however, provide any detailed information on some of the related issues such as safety or safeguards, which are addressed in other relevant publications. This report provides an overview of the status of reprocessing technology and its future prospects in terms of various criteria in Section 2. Section 3 provides a review of emerging technologies which have been attracting the interest of Member States, especially in the international initiatives for future development of innovative nuclear systems. A historical review of IAEA activities associated with spent fuel reprocessing, traceable back to the mid-1970s, is provided in Section 4, and conclusions in Section 5. A list of references is provided at the end the main text for readers interested in further information on the related topics. Annex I summarizes the current status of reprocessing facilities around the world, including the civil operational statistics of Purex-based plants, progress with decommissioning and

  15. Status of research reactor spent fuel world-wide

    International Nuclear Information System (INIS)

    Results compiled in the research reactor spent fuel database are used to assess the status of research reactor spent fuel world-wide. Fuel assemblies, their types, enrichment, origin of enrichment and geological distribution among the industrialised and developed countries of the world are discussed. Fuel management practices in wet and dry storage facilities and the concerns of reactor operators about long-term storage of their spent fuel are presented and some of the activities carried out by the International Atomic Energy Agency to address the issues associated with research reactor spent fuel are outlined. (author)

  16. Options for treatment of legacy and advanced nuclear fuels

    OpenAIRE

    Maher, Christopher John

    2014-01-01

    The treatment of advanced nuclear fuels is relevant to the stabilisation of legacy spent fuels or nuclear materials and fuels from future nuclear reactors. Historically, spent fuel reprocessing has been driven to recover uranium and plutonium for reuse. Future fuel cycles may also recover the minor actinides neptunium, americium and perhaps curium. These actinides would be fabricated into new reactor fuel to produce energy and for transmutation of the minor actinides. This has the potential t...

  17. Advanced Fuel Cycle Cost Basis

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert

    2007-04-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 26 cost modules—24 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, and high-level waste.

  18. Advanced Fuel Cycle Cost Basis

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert; E. Schneider

    2009-12-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 25 cost modules—23 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste.

  19. Advanced Fuel Cycle Cost Basis

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert; E. Schneider

    2008-03-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 25 cost modules—23 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste.

  20. New Methods for Evaluation of Spent Fuel Condition during Long-Term Storage in Slovakia

    OpenAIRE

    V. Kršjak; M. Mikloš

    2009-01-01

    Experiences with an advanced spent nuclear fuel management in Slovakia are presented in this paper. The evaluation and monitoring procedures are based on practices at the Slovak wet interim spent fuel storage facility in NPP Jaslovské Bohunice. Since 1999, leak testing of WWER-440 fuel assemblies are provided by special leak tightness detection system "Sipping in pool" delivered by Framatomeanp with external heating for the precise defects determination. In 2006, a new inspection stand "SVYP-...

  1. Nondestructive verification and assay systems for spent fuels

    International Nuclear Information System (INIS)

    This is an interim report of a study concerning the potential application of nondestructive measurements on irradiated light-water-reactor (LWR) fuels at spent-fuel storage facilities. It describes nondestructive measurement techniques and instruments that can provide useful data for more effective in-plant nuclear materials management, better safeguards and criticality safety, and more efficient storage of spent LWR fuel. In particular, several nondestructive measurement devices are already available so that utilities can implement new fuel-management and storage technologies for better use of existing spent-fuel storage capacity. The design of an engineered prototype in-plant spent-fuel measurement system is approx. 80% complete. This system would support improved spent-fuel storage and also efficient fissile recovery if spent-fuel reprocessing becomes a reality

  2. TRIGA spent fuel bundles safe storage

    Energy Technology Data Exchange (ETDEWEB)

    Negut, G.; Covaci, St. [Institute for Nuclear Research, Research Reactor Dept., Pitesti (Romania); Prisecaru, I.; Dupleac, D. [Bucharest Univ. Politehnica, Power and Nuclear Engineering Dept., Bucharest (Romania)

    2007-07-01

    TRIGA-SSR is a steady state research and material test reactor that has been in operation since 1980. The original TRIGA fuel was HEU (highly enriched uranium) with a U{sup 235} enrichment of 93 per cent. Almost all TRIGA HEU fuel bundles are now burned-up. Part of the spent fuel was loaded and transferred to US, in a Romania - DOE arrangement. The rest of the TRIGA fuel bundles have to be temporarily stored in the TRIGA facility. As the storage conditions had to be established with caution, neutron and thermal hydraulic evaluations of the storage conditions were required. Some criticality evaluations were made based on the SAR (Safety Analysis Report) data. Fuel constant axial temperature approximation effect is usual for criticality computations. TRIGA-SSR fuel bundle geometry and materials model for SCALE5-CSAS module allows the introduction of a fuel temperature dependency for the entire fuel active height, using different materials for each fuel bundle region. Previous RELAP5 thermal hydraulic computations for an axial and radial power distribution in the TRIGA fuel pin were done. Fuel constant temperature approximation overestimates pin factors for every core operating at high temperatures. From the thermal hydraulic point of view the worst condition of the storage grid occurs when the transfer channel is accidentally emptied of water from the pool, or the bundle is handled accidentally to remain in air. All the residual heat from the bundles has to be removed without fuel overheating and clad failure. RELAP5 computer code for residual heat removal was used in the assessment of residual heat removal. We made a couple of evaluations of TRIGA bundle clad temperatures in air cooling conditions, with different residual heat levels. The criticality computations have shown that the spent TRIGA fuel bundles storage grid is strongly sub-critical with k(eff) = 0.5951. So, there is no danger for a criticality accident for this storage grid type. The assessment is done

  3. TRIGA spent fuel bundles safe storage

    International Nuclear Information System (INIS)

    TRIGA-SSR is a steady state research and material test reactor that has been in operation since 1980. The original TRIGA fuel was HEU (highly enriched uranium) with a U235 enrichment of 93 per cent. Almost all TRIGA HEU fuel bundles are now burned-up. Part of the spent fuel was loaded and transferred to US, in a Romania - DOE arrangement. The rest of the TRIGA fuel bundles have to be temporarily stored in the TRIGA facility. As the storage conditions had to be established with caution, neutron and thermal hydraulic evaluations of the storage conditions were required. Some criticality evaluations were made based on the SAR (Safety Analysis Report) data. Fuel constant axial temperature approximation effect is usual for criticality computations. TRIGA-SSR fuel bundle geometry and materials model for SCALE5-CSAS module allows the introduction of a fuel temperature dependency for the entire fuel active height, using different materials for each fuel bundle region. Previous RELAP5 thermal hydraulic computations for an axial and radial power distribution in the TRIGA fuel pin were done. Fuel constant temperature approximation overestimates pin factors for every core operating at high temperatures. From the thermal hydraulic point of view the worst condition of the storage grid occurs when the transfer channel is accidentally emptied of water from the pool, or the bundle is handled accidentally to remain in air. All the residual heat from the bundles has to be removed without fuel overheating and clad failure. RELAP5 computer code for residual heat removal was used in the assessment of residual heat removal. We made a couple of evaluations of TRIGA bundle clad temperatures in air cooling conditions, with different residual heat levels. The criticality computations have shown that the spent TRIGA fuel bundles storage grid is strongly sub-critical with k(eff) = 0.5951. So, there is no danger for a criticality accident for this storage grid type. The assessment is done for

  4. Radionuclide release from research reactor spent fuel

    International Nuclear Information System (INIS)

    Numerous investigations with respect to LWR fuel under non oxidizing repository relevant conditions were performed. The results obtained indicate slow corrosion rates for the UO2 fuel matrix. Special fuel-types (mostly dispersed fuels, high enriched in 235U, cladded with aluminium) are used in German research reactors, whereas in German nuclear power plants, UO2-fuel (LWR fuel, enrichment in 235U up to 5%, zircaloy as cladding) is used. Irradiated research reactor fuels contribute less than 1% to the total waste volume. In Germany, the state is responsible for fuel operation and for fuel back-end options. The institute for energy research (IEF-6) at the Research Center Juelich performs investigation with irradiated research reactor spent fuels under repository relevant conditions. In the study, the corrosion of research reactor spent fuel has been investigated in MgCl2-rich salt brine and the radionuclide release fractions have been determined. Leaching experiments in brine with two different research reactor fuel-types were performed in a hot cell facility in order to determine the corrosion behaviour and the radionuclide release fractions. The corrosion of two dispersed research reactor fuel-types (UAlx-Al and U3Si2-Al) was studied in 400 mL MgCl2-rich salt brine in the presence of Fe2+ under static and initially anoxic conditions. Within these experimental parameters, both fuel types corroded in the experimental time period of 3.5 years completely, and secondary alteration phases were formed. After complete corrosion of the used research reactor fuel samples, the inventories of Cs and Sr were quantitatively detected in solution. Solution concentrations of Am and Eu were lower than the solubility of Am(OH)3(s) and Eu(OH)3(s) solid phases respectively, and may be controlled by sorption processes. Pu concentrations may be controlled by Pu(IV) polymer species, but the presence of Pu(V) and Pu(IV) oxyhydroxides species due to radiolytic effects cannot completely be

  5. State-of-the-art report of spent fuel management technology

    International Nuclear Information System (INIS)

    Essential technologies for a long-term management of domestic nuclear fuel have been described in this report. The technologies of interest are advanced processes for spent fuel management, spent fuel examination technology, evaluation of radiation effect on equipment, chemical characterization of spent fuel, and hot cell-related technology state of the art for the above-mentioned technologies has been reviewed and analyzed in detail. As a result, a future R and D direction that seems to be appropriate for us is drawn up in due consideration of in- and out-circumstances encountered with. (author). 304 refs., 28 tabs., 43 figs

  6. Operator declaration verification technique for spent fuel at reprocessing facilities

    International Nuclear Information System (INIS)

    A verification technique for use at reprocessing facilities, which integrates existing technologies to strengthen safeguards through the use of environmental monitoring, has been developed at Los Alamos National Laboratory. This technique involves the measurement of isotopic ratios of stable noble fission gases from on-stack emissions during reprocessing of spent fuel using high-precision mass spectrometry. These results are then compared to a database of calculated isotopic ratios using a data analysis method to determine specific fuel parameters (e.g., burnup, fuel type, reactor type, etc.). These inferred parameters can be used to verify operator declarations. The integrated system (mass spectrometry, reactor modeling, and data analysis) has been validated using on-stack measurements during reprocessing of fuel from a US production reactor. These measurements led to an inferred burnup that matched the declared burnup to within 3.9%, suggesting that the current system is sufficient for most safeguards applications. Partial system validation using gas samples from literature measurements of power reactor fuel has been reported elsewhere. This has shown that the technique developed here may have some difficulty distinguishing pressurized water reactor (PWR) from boiling water reactor (BWR) fuel; however, it consistently can distinguish light water reactor (either PWR or BWR) fuels from other reactor fuel types. Future validations will include advanced power reactor fuels (such as breeder reactor fuels) and research reactor fuels as samples become available

  7. The risks of the Taiwan research reactor spent fuel project

    International Nuclear Information System (INIS)

    The proposed action is to transport up to 118 spent fuel rods, to include canned spent fuel rod particulates immobilized on filters, from a research reactor in Taiwan by sea to Hampton Roads, Virginia, and then overland by truck to the Receiving Basin for Offsite Fuels at the Savannah River Site (SRS). At SRS, the spent fuel will be reprocessed to recover uranium and plutonium. 55 refs., 8 tabs

  8. Spent Nuclear Fuel Vibration Integrity Study

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [ORNL; Wang, Hong [ORNL; Jiang, Hao [ORNL; Yan, Yong [ORNL; Bevard, Bruce Balkcom [ORNL

    2016-01-01

    The objective of this research is to collect dynamic experimental data on spent nuclear fuel (SNF) under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT), the hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL). The collected CIRFT data will be utilized to support ongoing spent fuel modeling activities, and support SNF transportation related licensing issues. Recent testing to understand the effects of hydride reorientation on SNF vibration integrity is also being evaluated. CIRFT results have provided insight into the fuel/clad system response to transportation related loads. The major findings of CIRFT on the HBU SNF are as follows: SNF system interface bonding plays an important role in SNF vibration performance, Fuel structure contributes to the SNF system stiffness, There are significant variations in stress and curvature of SNF systems during vibration cycles resulting from segment pellets and clad interaction, and SNF failure initiates at the pellet-pellet interface region and appears to be spontaneous. Because of the non-homogeneous composite structure of the SNF system, finite element analyses (FEA) are needed to translate the global moment-curvature measurement into local stress-strain profiles. The detailed mechanisms of the pellet-pellet and pellet-clad interactions and the stress concentration effects at the pellet-pellet interface cannot be readily obtained directly from a CIRFT system measurement. Therefore, detailed FEA is used to understand the global test response, and that data will also be presented.

  9. Decay heat of fast reactor spent fuel

    International Nuclear Information System (INIS)

    Decay heat from JOYO Mk-II spent fuel subassemblies was measured to obtain experimental data and to improve the accuracy of related calculations. The measurement was taken in the JOYO spent fuel storage pond. The fuel burn-up was approximately 66 GWd/t and the cooling time was between 40 and 385 days. The decay heat was calculated with the ORIGEN2 code using the JENDL-3.2 cross section library and the JNDC-V2 decay data and fission yield data library. The fuel power used as an input to ORIGEN2 was calculated by the MAGI core management code system. The ratios between calculated and experimental values were between 0.94 and 0.89 and decreased with a longer cooling time. This systematic discrepancy is not fully understood, but the change with cooling time appears to be due to the actinide decay heat uncertainty. This indicated that cross sections of actinides are important to evaluate decay heat accurately. (author)

  10. Spent Nuclear Fuel Project operational staffing plan

    International Nuclear Information System (INIS)

    Using the Spent Nuclear Fuel (SNF) Project's current process flow concepts and knowledge from cognizant engineering and operational personnel, an initial assessment of the SNF Project radiological exposure and resource requirements was completed. A small project team completed a step by step analysis of fuel movement in the K Basins to the new interim storage location, the Canister Storage Building (CSB). This analysis looked at fuel retrieval, conditioning of the fuel, and transportation of the fuel. This plan describes the staffing structure for fuel processing, fuel movement, and the maintenance and operation (M ampersand O) staffing requirements of the facilities. This initial draft does not identify the support function resources required for M ampersand O, i.e., administrative and engineering (technical support). These will be included in future revisions to the plan. This plan looks at the resource requirements for the SNF subprojects, specifically, the operations of the facilities, balances resources where applicable, rotates crews where applicable, and attempts to use individuals in multi-task assignments. This plan does not apply to the construction phase of planned projects that affect staffing levels of K Basins

  11. Spent nuclear fuel management : Trends and back end scenarios

    International Nuclear Information System (INIS)

    All kinds of radioactive wastes including spent nuclear fuel have to be managed with special care to ensure public and environmental safety. Therefore any country including the nuclear energy option in its energy policy has to provide an appropriate program for the safe treatment and final disposal of spent nuclear fuel, which is an unavoidable product of nuclear power plants. Temporary storage is the essential step of all alternatives of spent fuel management and inevitable final step is the geological disposal. Therefore any national policy regarding spent fuel management must include short term and long term planning for the safe storage of spent nuclear fuel including final disposal in geological repository. Available alternatives for the spent fuel management are (1) the closed fuel cycle, (2) the once-through fuel cycle, (3) deferral of a final decision. The resources that are available to the country concerned will be the limiting conditions for establishing a national policy and the choice of the one of the above stated alternatives. Nuclear infrastructure, personnel training availability and relative cost of spent fuel alternatives are the main resources to be considered. In this study, national policy considerations regarding spent nuclear fuel management will be discussed and a frame of the spent fuel management strategy for Turkey will be proposed for back-end of fuel cycle in nuclear scenario

  12. Path forward of spent fuel management in the world

    International Nuclear Information System (INIS)

    Following topics among future trend of spent fuel management in the world will be reported. 1. Transportation of Spent Fuel after Storage: Safety and conformity to the transport regulations after long-term storage has been discussed in IAEA. Japan has proposed a holistic approach that will enable the transport after storage using record of spent fuel management. 2. Very Long-Term Storage of Spent Fuel: In USA, the Yucca mountain project for disposal was cancelled and an extended storage of spent fuel as long as 300 years is being proposed. The IAEA considered the very long-term storage of spent fuel is worldwide trend, and started to create its guideline for member states. 3. Potential Interface Issues in Spent Fuel Management: Effort to identify potential interface issues in spent fuel management including the above-mentioned 'Transportation of Spent Fuel after Storage' and 'Very Long-Term Storage of Spent Fuel' is being made in IAEA in order to realize safe and secured spent fuel management. 4. Lessons Learned on Spent Fuel Storage: The IAEA is collecting information on lessons learned from operations of wet and dry storage of spent fuel storage from the member states, including the Fukushima nuclear disaster in order to publish technical reports to share the lessons among the member states. 5. Safety and Security of Transport of Radioactive Material Transport: There was an international conference on safety and security of radioactive material transport to celebrate the 50th anniversary of the Transport Regulation of IAEA. The main topics of the conference was to harmonize and integrate the safety and security with proposal of path forward in the next 50 years to IAEA. (author)

  13. TRIGA spent fuel end-cap cut-out device

    International Nuclear Information System (INIS)

    During the TRIGA reactor operation most of the fuel is now burned-up. According to the Romania - USA agreement the spent fuel have to be returned to US. The spent fuel packaging is a difficult operation due to the high radioactivity and also to the dimensions which is to match the spent fuel casks sizes. The paper presents the problems met during the spent fuel packaging in the casks, the approach adopted, and the design of the spent fuel end-cap cut-out device. This device was designed and fabricated at INR Pitesti. It was used for end-cap cut-out in order to fit the spent fuel casks provided by US Idaho Falls partner. (author)

  14. Some factors to consider in handling and storing spent fuel

    International Nuclear Information System (INIS)

    This report includes information from various studies performed under the Wet Storage Task of the Behavior of Spent Fuel in Storage Project of the Commercial Spent Fuel Management (CSFM) Program at Pacific Northwest Laboratory. Wet storage experience has been summarized earlier in several other reports. This report summarizes pertinent items noted during FY 1985 concerning recent developments in the handling and storage of spent fuel and associated considerations. The subjects discussed include recent publications, findings, and developments associated with: (1) storage of water reactor spent fuel in water pools, (2) extended-burnup fuel, (3) fuel assembly reconstitution and reinsertion, (4) rod consolidation, (5) variations in the US Nuclear Regulatory Commission's definition of failed fuel, (6) detection of failed fuel rods, and (7) extended integrity of spent fuel. A list of pertinent publications is included

  15. Seismic qualification of spent fuel storage stacks

    International Nuclear Information System (INIS)

    CANDU reactors designed in Canada are built and operated worldwide for producing electricity economically. The operation of CANDU reactors yields spent fuel bundles that are removed from the reactor core by means of remote mechanisms. The spent fuel bundles are transferred to a Spent Fuel Bay (SFB) for underwater cooling and long term storage. Spent fuel bundles are loaded onto stainless steel trays. A number of such trays are stacked vertically and stored on the floor of the SFB. It is necessary that the storage stacks maintain their structural integrity and stability under a severe design earthquake. This paper presents the methods and process used for seismic qualification of the storage stacks by analysis. The finite element models of the storage stack are developed to represent the behavior of the structure. The models are created for each individual tray and then restructured by using the sub-structuring technique. In this process, the stiffness matrix of a tray is condensed into a number of key points that include the points of contact between trays. In a storage stack the trays are pressed together by their own deadweight and by the weight of the fuel bundles and locked in place against each other at a number of contact points having no structural continuity. It can generally be assumed that the trays are in contact with each other. However the effectiveness of contact between the trays during seismic motion is uncertain due to erratic deflection, manufacturing irregularities and field conditions. Consequently a sensitivity study is carried out to assess the effect of lack of continuity at some of the contact points. The stack models are grouped into a number of multiple stack arrangements with the safeguard covers. They are grouped into different configurations and analyzed. Since the stacks are submerged in water, hydrodynamic effects are considered in the seismic model to more accurately predict the seismic behavior. A time-history analysis was then

  16. A complete NUHOMS registered solution for storage and transport of high burnup spent fuel

    International Nuclear Information System (INIS)

    The discharge burnups of spent fuel from nuclear power plants keep increasing with plants discharging or planning to discharge fuel with burnups in excess of 60,000 MWD/MTU. Due to limited capacity of spent fuel pools, transfer of older cooler spent fuel from fuel pool to dry storage, and very limited options for transport of spent fuel, there is a critical need for dry storage of high burnup, higher heat load spent fuel so that plants could maintain their full core offload reserve capability. A typical NUHOMS registered solution for dry spent fuel storage is shown in the Figure 1. Transnuclear, Inc. offers two advanced NUHOMS registered solutions for the storage and transportation of high burnup fuel. One includes the NUHOMS registered 24PTH system for plants with 90.7 Metric Ton (MT) crane capacity; the other offers the higher capacity NUHOMS registered 32PTH system for higher crane capacity. These systems include NUHOMS registered - 24PTH and -32PTH Transportable Canisters stored in a concrete storage overpack (HSM-H). These canisters are designed to meet all the requirements of both storage and transport regulations. They are designed to be transported off-site either directly from the spent fuel pool or from the storage overpack in a suitable transport cask

  17. Spent Fuel Reprocessing: More Value for Money Spent in a Geological Repository?

    International Nuclear Information System (INIS)

    Today, each utility or country operating nuclear power plants can select between two long-term spent fuel management policies: either, spent fuel is considered as waste to dispose of through direct disposal or, spent fuel is considered a resource of valuable material through reprocessing-recycling. Reading and listening to what is said in the nuclear community, we understand that most people consider that the choice of policy is, actually, a choice among two technical paths to handle spent fuel: direct disposal versus reprocessing. This very simple situation has been recently challenged by analysis coming from countries where both policies are on survey. For example, ONDRAF of Belgium published an interesting study showing that, economically speaking for final disposal, it is worth treating spent fuel rather than dispose of it as a whole, even if there is no possibility to recycle the valuable part of it. So, the question is raised: is there such a one-to-one link between long term spent fuel management political option and industrial option? The purpose of the presentation is to discuss the potential advantages and drawbacks of spent fuel treatment as an implementation of the policy that considers spent fuel as waste to dispose of. Based on technical considerations and industrial experience, we will study qualitatively, and quantitatively when possible, the different answers proposed by treatment to the main concerns of spent-fuel-as-a-whole geological disposal

  18. Updating the Regulatory Framework for Spent Nuclear Fuel Reprocessing

    International Nuclear Information System (INIS)

    There is renewed domestic interest in establishing spent nuclear fuel recycle in the U.S. after about a 30 year hiatus. Introduction of safe, proliferation-resistant, and economical civilian nuclear fuel cycles, especially the reprocessing step, in the U.S. poses numerous technical, social, and regulatory challenges. Initially, fuel recycle activities are expected to focus on light water reactor fuels, but it is anticipated that recycle of fuel from advanced reactors such as liquid-metal-cooled reactors and gas-cooled reactors will follow. Proposed reprocessing technologies include processes for removing heat-producing and high-risk fission products and actinides from waste streams prior to disposal. Proposed reprocessing processes and operations raise a range of issues identified in this paper that would require new and revised regulations to effectively and efficiently ensure their safety. The NRC prepared a report (NUREG-1909) documenting the background, status, and potential future issues concerning recycle of spent nuclear fuel that is summarized in this paper. In response to the issues, the NRC Commissioners, and other stakeholders, the NRC staff has conducted two analyses to identify and prioritize regulatory gaps for spent fuel reprocessing facilities and held public meetings to obtain stakeholder input. The NRC staff is now working on a revised regulatory framework for reprocessing facilities with a goal of completing the revisions by FY 2012. This paper summarizes the contents of NUREG-1909 and the activities of the NRC staff to update the regulatory framework in to address the issues that have been identified. (author)

  19. Systems impacts of spent fuel disassembly alternatives

    International Nuclear Information System (INIS)

    Three studies were completed to evaluate four alternatives to the disposal of intact spent fuel assemblies in a geologic repository. A preferred spent fuel waste form for disposal was recommended on consideration of (1) package design and fuel/package interaction, (2) long-term, in-repository performance of the waste form, and (3) overall process performance and costs for packaging, handling, and emplacement. The four basic alternative waste forms considered were (1) end fitting removal, (2) fission gas venting, (3) disassembly and close packing, and (4) shearing/immobilization. None of the findings ruled out any alternative on the basis of waste package considerations or long-term performance of the waste form. The third alternative offers flexibility in loading that may prove attractive in the various geologic media under consideration, greatly reduces the number of packages, and has the lowest unit cost. These studies were completed in October, 1981. Since then Westinghouse Electric Corporation and the Office of Nuclear Waste Isolation have completed studies in related fields. This report is now being published to provide publicly the background material that is contained within. 47 references, 28 figures, 31 tables

  20. Systems impacts of spent fuel disassembly alternatives

    Energy Technology Data Exchange (ETDEWEB)

    1984-07-01

    Three studies were completed to evaluate four alternatives to the disposal of intact spent fuel assemblies in a geologic repository. A preferred spent fuel waste form for disposal was recommended on consideration of (1) package design and fuel/package interaction, (2) long-term, in-repository performance of the waste form, and (3) overall process performance and costs for packaging, handling, and emplacement. The four basic alternative waste forms considered were (1) end fitting removal, (2) fission gas venting, (3) disassembly and close packing, and (4) shearing/immobilization. None of the findings ruled out any alternative on the basis of waste package considerations or long-term performance of the waste form. The third alternative offers flexibility in loading that may prove attractive in the various geologic media under consideration, greatly reduces the number of packages, and has the lowest unit cost. These studies were completed in October, 1981. Since then Westinghouse Electric Corporation and the Office of Nuclear Waste Isolation have completed studies in related fields. This report is now being published to provide publicly the background material that is contained within. 47 references, 28 figures, 31 tables.

  1. Management and disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    The National Board for Spent Nuclear Fuel, in submitting its statement of comment to the Government on the Swedish Nuclear Fuel and Waste Management Company's (Svensk Kaernbraenslehantering AB, SKB) research programme, R and D Programme 86, has also put forward recommendations on the decision-making procedure and on the question of public information during the site selection process. In summary the Board proposes: * that the Government instruct the National Board for Spent Nuclear Fuel to issue certain directives concerning additions to and changes in R and D Programme 86, * that the Board's views on the decision-making procedure in the site selection process be taken into account in the Government's review of the so-called municipal veto in accordance with Chapter 4, Section 3 of the Act (1987:12) on the conservation of natural resources etc., NRL, * that the Board's views on the decision-making procedure and information questions during the site selection process serve as a basis for the continued work. Three appendices are added to the report: 1. Swedish review statements (SV), 2. International Reviews, 3. Report from the site selection group (SV)

  2. Spent fuel storage options: a critical appraisal

    International Nuclear Information System (INIS)

    The delayed decisions on nuclear fuel reprocessing strategies in the USA and other countries have forced the development of new long-term irradiated fuel storage techniques, to allow a larger volume of fuel to be held on the nuclear station site after removal from the reactor. The nuclear power industry has responded to the challenge by developing several viable options for long-term onsite storage, which can be employed individually or in tandem. They are: densification of storage in the existing spent fuel pool; building another fuel pool facility at the plant site; onsite cask park, and on site vault clusters. Desirable attributes of a storage option are: Safety: minimise the number of fuel handling steps; Economy: minimise total installed, and O and M cost; Security: protection from anti-nuclear protesters; Site adaptability: available site space, earthquake characteristics of the region and so on; Non-intrusiveness: minimise required modifications to existing plant systems; Modularisation: afford the option to adapt a modular approach for staged capital outlays; and Maturity: extent of industry experience with the technology. A critical appraisal is made of each of the four aforementioned storage options in the light of these criteria. (2 figures, 1 table, 4 references) (Author)

  3. Calibration of spent fuel measurement assembly

    International Nuclear Information System (INIS)

    The LVR-15 research reactor (Czech Republic) had been converted from the highly enriched IRT-2M to the low enriched IRT-4M fuel. For the possibility of the independent pre-transport evaluation of IRT-2M burnup, a spectrometric system was developed. This spectrometric system consists of the fuel holder, the collimator and the portable Canberra Big MAC HPGe (High Purity Germanium) detector. In order to have well reproducible and reliable experimental data for modeling of the measurement system, calibration with the 110mAg isotope with known activity was performed. This isotope was chosen for having energies similar to isotopes measured in fuel assemblies. The 110mAg isotope was prepared by irradiating of the silver foil in LVR-15 research reactor; its activity was evaluated in the LVR-15's spectrometric laboratory. From the measured data, an efficiency curve of the spectrometric system has been determined. The experimental data were compared to the calculation results with the MCNPX model of the spectrometric system. - Highlights: • Calibration of research reactor spent fuel measurement assembly. • On-site prepared 110mAg isotope used for the measurement. • Calculated self-shielding factor for the IRT-2M fuel. • Applicable to other research reactor fuel geometries

  4. Damage in spent nuclear fuel defined by properties and requirements

    International Nuclear Information System (INIS)

    The Nuclear Regulatory Commission's (NRC's) Spent Fuel Program Office (SFPO) has provided guidance in defining damaged fuel in Interim Staff Guidance, ISG-1. This guidance is similar to that developed by the American National Standards Institute (ANSI). Neither of these documents gives the logic behind its definition of damaged fuel. The paper discusses the requirements placed on spent fuel for dry interim storage and transport and the ways in which service requirements drive the definition of damage for spent fuel. Examples are given to illustrate the methodology, which focuses on defining damaged fuel based on the properties that the fuel must exhibit to meet the requirements of storage and/or transport. (author)

  5. Storage and Management of Spent Fuel of Fast Reactors in India

    International Nuclear Information System (INIS)

    Deploying fast breeder reactors on a commercial scale is vital for India in order to utilise the vast thorium reserves and to meet the long term energy needs of the country. As a first step, a 40 MWt, 13.5 MWe Fast Breeder Test Reactor (FBTR) was constructed and has been in operation since 1985. FBTR utilises mixed carbide fuel. The spent fuel sub assembly of FBTR is stored in an argon inverted container and cooled by air. India took the construction of indigenously designed 500 MWe Prototype Fast Breeder Reactor (PFBR) in September 2003. The reactor is in advanced stage of construction. It is planned to construct six more FBRs of 500 MWe each by the year 2023. All the 500 MWe fast reactors utilise oxide fuel, the most proven fuel in the case of the fast reactors. The spent fuel subassembly after internal storage for one campaign of 8 calendar months is washed to remove sodium and stored in water in a spent fuel storage lined with SS 304L liner and with a provision of leak monitoring of the liner welds. The failed subassembly is stored in a leak tight container before putting in the spent fuel storage. The road to getting faster growth in Indian indigenous nuclear power programme is through use of metallic fuel in fast breeder reactors. The metallic fuel design gives high breeding ratio and thus shorter doubling time. Both sodium and mechanical bonded metallic fuel design are under development. The growth of fast reactors is sensitive to the time period the spent fuel is out of the reactor i.e. from the time spent fuel is out of the reactor, cooled, reprocessed, refabricated and put back into the reactor. Storage of spent fuel subassemblies in sodium and co-located pyro-reprocessing facility is planned for the metallic fuel reactors. The paper describes in brief the system for storage and management of spent fuel of FBTR, PFBR and metallic fuel fast reactors. (author)

  6. Multinational approaches relevant to spent fuel management

    International Nuclear Information System (INIS)

    The storage of spent fuel is a suitable candidate for a multilateral approach, primarily at the regional level. Small countries with only a few nuclear power plants would benefit economically from large joint facilities. The storage of special nuclear materials in a few safe and secure facilities would also enhance safeguards and physical protection. However, the final disposal of spent fuel and high level radioactive waste is the best candidate for a multilateral approach. It would offer major economic benefits and substantial non-proliferation benefits in spite of the legal, political and public acceptance challenges to be expected in most countries. The transfer of nuclear waste from the exporting country to the host country of an interim storage facility or of a final repository would be done under bilateral or multilateral agreements at the commercial and governmental levels, in accordance with the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management. Bilateral or international oversight of joint facilities should be arranged, as needed, to achieve the confidence of the partners as to the safety and physical security of the proposed facility. Such monitoring should cover the adequacy of the technical design, its safety features, its environmental impact, the physical security of nuclear materials and possibly the financial management of the joint venture. After the initial choice of bilateral arrangements, some kind of international monitoring may become appropriate. Various organizations could fulfil such a function, in particular, the IAEA. Such monitoring would have nothing to do with nuclear safeguards; repository monitoring would be a parallel but independent activity of the IAEA. (author)

  7. Overview of the United States spent nuclear fuel program

    International Nuclear Information System (INIS)

    As a result of the end of the Cold War, the mission of the US Department of Energy (DOE) has shifted from an emphasis on nuclear weapons development and production to an emphasis on the safe management and disposal of excess nuclear materials including spent nuclear fuel from both production and research reactors. Within the US, there are two groups managing spent nuclear fuel. Commercial nuclear power plants are managing their spent nuclear fuel at the individual reactor sites until the planned repository is opened. All other spent nuclear fuel, including research reactors, university reactors, naval reactors, and legacy material from the Cold War is managed by DOE. DOE's mission is to safely and efficiently manage its spent nuclear fuel and prepare it for disposal. This mission involves correcting existing vulnerabilities in spent fuel storage; moving spent fuel from wet basins to dry storage; processing at-risk spent fuel; and preparing spent fuel in road-ready condition for repository disposal. Most of DOE's spent nuclear fuel is stored in underwater basins (wet storage). Many of these basins are outdated, and spent fuel is to be removed and transferred to more modern basins or to new dry storage facilities. In 1995, DOE completed a complex-wide environmental impact analysis that resulted in spent fuel being sent to one of three principal DOE sites for interim storage (up to 40 years) prior to shipment to a repository. This regionalization by fuel type will allow for economies of scale yet minimize unnecessary transportation. This paper discusses the national SNF program, ultimate disposition of SNF, and the technical challenges that have yet to be resolved, namely, release rate testing, non-destructive assay, alternative treatments, drying, and chemical reactivity

  8. DOE SPENT NUCLEAR FUEL DISPOSAL CONTAINER

    International Nuclear Information System (INIS)

    The DOE Spent Nuclear Fuel Disposal Container (SNF DC) supports the confinement and isolation of waste within the Engineered Barrier System of the Mined Geologic Disposal System (MGDS). Disposal containers are loaded and sealed in the surface waste handling facilities, transferred to the underground through the access mains, and emplaced in emplacement drifts. The DOE Spent Nuclear Fuel Disposal Container provides long term confinement of DOE SNF waste, and withstands the loading, transfer, emplacement, and retrieval loads and environments. The DOE SNF Disposal Containers provide containment of waste for a designated period of time, and limit radionuclide release thereafter. The disposal containers maintain the waste in a designated configuration, withstand maximum handling and rockfall loads, limit the individual waste canister temperatures after emplacement. The disposal containers also limit the introduction of moderator into the disposal container during the criticality control period, resist corrosion in the expected repository environment, and provide complete or limited containment of waste in the event of an accident. Multiple disposal container designs may be needed to accommodate the expected range of DOE Spent Nuclear Fuel. The disposal container will include outer and inner barrier walls and outer and inner barrier lids. Exterior labels will identify the disposal container and contents. Differing metal barriers will support the design philosophy of defense in depth. The use of materials with different failure mechanisms prevents a single mode failure from breaching the waste package. The corrosion-resistant inner barrier and inner barrier lid will be constructed of a high-nickel alloy and the corrosion-allowance outer barrier and outer barrier lid will be made of carbon steel. The DOE Spent Nuclear Fuel Disposal Containers interface with the emplacement drift environment by transferring heat from the waste to the external environment and by protecting

  9. DOE SPENT NUCLEAR FUEL DISPOSAL CONTAINER

    Energy Technology Data Exchange (ETDEWEB)

    F. Habashi

    1998-06-26

    The DOE Spent Nuclear Fuel Disposal Container (SNF DC) supports the confinement and isolation of waste within the Engineered Barrier System of the Mined Geologic Disposal System (MGDS). Disposal containers are loaded and sealed in the surface waste handling facilities, transferred to the underground through the access mains, and emplaced in emplacement drifts. The DOE Spent Nuclear Fuel Disposal Container provides long term confinement of DOE SNF waste, and withstands the loading, transfer, emplacement, and retrieval loads and environments. The DOE SNF Disposal Containers provide containment of waste for a designated period of time, and limit radionuclide release thereafter. The disposal containers maintain the waste in a designated configuration, withstand maximum handling and rockfall loads, limit the individual waste canister temperatures after emplacement. The disposal containers also limit the introduction of moderator into the disposal container during the criticality control period, resist corrosion in the expected repository environment, and provide complete or limited containment of waste in the event of an accident. Multiple disposal container designs may be needed to accommodate the expected range of DOE Spent Nuclear Fuel. The disposal container will include outer and inner barrier walls and outer and inner barrier lids. Exterior labels will identify the disposal container and contents. Differing metal barriers will support the design philosophy of defense in depth. The use of materials with different failure mechanisms prevents a single mode failure from breaching the waste package. The corrosion-resistant inner barrier and inner barrier lid will be constructed of a high-nickel alloy and the corrosion-allowance outer barrier and outer barrier lid will be made of carbon steel. The DOE Spent Nuclear Fuel Disposal Containers interface with the emplacement drift environment by transferring heat from the waste to the external environment and by protecting

  10. Surrogate Spent Nuclear Fuel Vibration Integrity Investigation

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [ORNL; Wang, Hong [ORNL; Bevard, Bruce Balkcom [ORNL; Howard, Rob L [ORNL

    2014-01-01

    Transportation packages for spent nuclear fuel (SNF) must meet safety requirements under normal and accident conditions as specified by federal regulations. During transportation, SNF experiences unique conditions and challenges to cladding integrity due to the vibrational and impact loading encountered during road or rail shipment. ORNL has been developing testing capabilities that can be used to improve our understanding of the impacts of vibration loading on SNF integrity, especially for high burn-up SNF in normal transportation operation conditions. This information can be used to meet nuclear industry and U.S. Nuclear Regulatory Commission needs in the area of safety of SNF storage and transportation operations.

  11. Spent fuel transport: A continuous improvement

    International Nuclear Information System (INIS)

    Full text: Since the 70's, approximately 30,000 tons of spent fuel has been safely transported to COGEMA-La Hague plant from French Nuclear Power Plants (NPP) and foreign nuclear operators involving more than 5,500 shipments. COGEMA LOGISTICS made this possible by the continuous development of adapted transport casks duly licensed according to the regulations in force, the procurement of dedicated transport equipment such as wagons, trucks and ships, and an efficient transport organization providing a comprehensive door-to-door service. New markets are under development implying new routes and organization. This paper is aimed at presenting our approach to meet the future challenges. (author)

  12. Spent fuel management: Semi-dry storage

    International Nuclear Information System (INIS)

    To change the temporary underwater storage mode of nuclear spent fuel (NSF) from wet to semi-dry as a means of slowing down or even stopping corrosion of the cladding and thereby to ensure safe storage conditions for further temporary storage, experts of the Atomic Energy Research Institute developed a canning technology and automatic canning equipment. This equipment was commissioned at the AFR pond of the Budapest Research Reactor and the regulatory licence for NSF encapsulation was granted in March 2002. The technology uses a tube-type capsule made of an aluminium alloy with a wall thickness of 3 mm. The capsule is capable of accommodating one EK-lO, one triple VVR type assembly or three single VVR assemblies. Encapsulation utilizes a closed technology in which the capsule undergoes a powerful drying procedure (heated by an eddy current), is back-filled with dry nitrogen gas and then closed by a shrink-fit capsule head and subsequently welded. The program-controlled equipment of the canning machine is housed in a compact steel container. It includes a device for cropping the fuel legs and provides a means of reopening encapsulated NSF, if the need arises. The cycle time of the whole canning procedure is about 120 minutes allowing about 4 to 5 cans to be sealed per working day. The canning has been running since May 2002 and was divided into two phases. In Phase 1 the goal was to encapsulate all NSF assemblies irradiated before reactor upgrading (1986), while Phase 2 is a regular canning operation to deal with fuel that has decayed for about 10 years. Phase 1 was successfully completed in August 2004 by which time 82 EK-10, 228 single and 184 triple VVR fuel assemblies (altogether 342 cans) had been encapsulated. The canning machine and its accessories, including the cropping device, form a compact and mobile technology that ensures an almost completely automatic, safe and reliable encapsulation of spent fuel. (author)

  13. Investigation of the condition of spent-fuel pool components

    Energy Technology Data Exchange (ETDEWEB)

    Kustas, F.M.; Bates, S.O.; Opitz, B.E.; Johnson, A.B. Jr.; Perez, J.M. Jr.; Farnsworth, R.K.

    1981-09-01

    It is currently projected that spent nuclear fuel, which is discharged from the reactor and then stored in water pools, may remain in those pools for several decades. Other studies have addressed the expected integrity of the spent fuel during extended water storage; this study assesses the integrity of metallic spent fuel pool components. Results from metallurgical examinations of specimens taken from stainless steel and aluminum components exposed in spent fuel pools are presented. Licensee Event Reports (LERs) relating to problems with spent fuel components were assessed and are summarized to define the types of operational problems that have occurred. The major conclusions of this study are: aluminum and stainless steel spent fuel pool components have a good history of performance in both deionized and borated water pools. Although some operational problems involving pool components have occurred, these problems have had minimal impacts.

  14. Investigation of the condition of spent-fuel pool components

    International Nuclear Information System (INIS)

    It is currently projected that spent nuclear fuel, which is discharged from the reactor and then stored in water pools, may remain in those pools for several decades. Other studies have addressed the expected integrity of the spent fuel during extended water storage; this study assesses the integrity of metallic spent fuel pool components. Results from metallurgical examinations of specimens taken from stainless steel and aluminum components exposed in spent fuel pools are presented. Licensee Event Reports (LERs) relating to problems with spent fuel components were assessed and are summarized to define the types of operational problems that have occurred. The major conclusions of this study are: aluminum and stainless steel spent fuel pool components have a good history of performance in both deionized and borated water pools. Although some operational problems involving pool components have occurred, these problems have had minimal impacts

  15. Data management for spent fuel from power reactors in Argentina

    International Nuclear Information System (INIS)

    Updated data for the spent fuel management from the operating power reactors - as of December 31st, 2004 - as well as research reactors - as of March 1st, 2003 - in Argentina are presented. Data for the power reactor spent fuel are received from the nuclear power plant operator (Nucleoelectrica Argentina S.A.) twice a year for the cumulative spent fuel arising up to June 30th and December 31st. Data for the research reactor spent fuel are collected once a year from CNEA operators. At the time being, such data are not managed in a database management system but some of them are handled with a spreadsheet program in order to get total, average, lower and higher values. These values are being used to built the input of codes for calculating the composition, activity and thermal power for the spent fuel as a whole as well as the mass, activity and thermal power for spent fuel elements or nuclides. (author)

  16. Spent fuel management fee methodology and computer code user's manual

    International Nuclear Information System (INIS)

    The methodology and computer model described here were developed to analyze the cash flows for the federal government taking title to and managing spent nuclear fuel. The methodology has been used by the US Department of Energy (DOE) to estimate the spent fuel disposal fee that will provide full cost recovery. Although the methodology was designed to analyze interim storage followed by spent fuel disposal, it could be used to calculate a fee for reprocessing spent fuel and disposing of the waste. The methodology consists of two phases. The first phase estimates government expenditures for spent fuel management. The second phase determines the fees that will result in revenues such that the government attains full cost recovery assuming various revenue collection philosophies. These two phases are discussed in detail in subsequent sections of this report. Each of the two phases constitute a computer module, called SPADE (SPent fuel Analysis and Disposal Economics) and FEAN (FEe ANalysis), respectively

  17. Advanced cement solidification technique for spent resins

    International Nuclear Information System (INIS)

    In the past 40 years, the nuclear facilities of China Institute of Atomic Energy (CIAE) produced an amount of radioactive organic resins, a kind of problematic stream in nuclear industry. As these facilities were stepping into decommissioning, the treatment of the spent organic resins was put on the agenda. The various routes for spent resin treatment such as incineration, advanced oxidation, cement immobilization, etc, were considered. Each method has its advantages and disadvantages when applied in the treatment of spent resins. Since the quantities of the spent organic resins were relatively small and an experience with variety of cementation processes existed in CIAE, predominately for immobilization of the evaporated concentrates, the option of direct encapsulation of the spent organic resins into cementitious materials was adopted in 2003, as a preferred method from the point of view of saving the on the cost of the disposal. In order to realize the end goal, the main work consisted of: the survey of the source terms; cementitious material formula investigation; and the process development. This work, which was undertaken in the following years, is addressed as follows. Source terms of the spent resins in CIAE were to be made clear firstly. The results showed that a total of 24-29 m3 of spent resins was generated and accumulated in the past 40 years. Spent resin arose from two research reactors (heavy water reactor and light water reactor), and from the waste management plant. The amount of the spent resins from the heavy water reactor was 1m3 or so, but its radioactive concentration was high to ∼108-∼109Bq/m3. Two kinds of cements, ASC and OPC cement were selected next, as the solidifying matrix to be investigated. A mixture surface response approach was employed to design experiment and interpret data. In comparison, ASC was superior to OPC cement and it displayed preferable performances to encapsulate spent resins. The optimum formulation is:1) resin

  18. The status of spent fuel treatment in the United Kingdom

    International Nuclear Information System (INIS)

    Nuclear power has been used to generate electricity in the UK since the 1950s. Since that time a number of reactor and fuel types have been developed and are currently in use, requiring different spent fuel treatment routes. This paper reviews the spent fuel treatment technology along with the associated waste management and recycle facilities currently in use in the UK. (author)

  19. USA: energy policy and spent fuel and waste management

    International Nuclear Information System (INIS)

    The new US administration under President Bush has shifted political weights in the country's energy policy. The policy pursued by the Clinton administration, which had been focused strongly on energy efficiency and environmental protection, will be revoked in a number of points, and the focus instead will now be on economics and continuity of supply, also against the backdrop of the current power supply crisis in California. However, it is more likely that fossil-fired generating capacity will be expanded or added than new nuclear generating capacity. As far as the policy of managing radioactive waste is concerned, no fast and fundamental changes are expected. Low-level waste arising in medicine, research, industry, and nuclear power plants will be stored in a number of shallow ground burial facilities also involving more than one federal state. The Yucca Mountain repository project will be advanced with a higher budget, and WIPP (Waste Isolation Plant) in the state of New Mexico has been in operation since 1998. Plans for the management of spent fuel elements include interim stores called ISFSIs (Independent Spent Fuel Storage Installations) both near and independent of nuclear power sites. Nineteen sites have been licensed, another eighteen are ready to be licensed. In addition, also international spent fuel and nuclear waste management approaches are being discussed in the United States which, inter alia, are meant to offer comprehensive solutions to countries running only a small number of nuclear power plants. (orig.)

  20. Storage and Management of Spent Fuel in Hanaro Research Reactor

    International Nuclear Information System (INIS)

    HANARO, the only operating research reactor in the Republic of Korea, generates 5 spent fuel assemblies after 28 day operation at 30MWth. After a visual inspection, the discharged fuel is stored in the spent fuel storage pool. As of August 2009, the spent fuel pool accumulates 326 assemblies, not including 299 spent fuel rods from 2 TRIGA reactors which operated before HANARO, and which were sent back to the USA in June 1998, as decommissioning projects of the reactors are underway. The spent fuel pool of HANARO can store 1032 spent fuels, equivalent to 20 years of operation of the reactor, plus test fuels, which have been locally irradiated. The spent fuels are stored in modules supported by racks and modules can be piled up in 3 layers. The storage pool is a heavy concrete structure with an internal stainless steel lining. Demineralized water is used for radiation shielding and cooling purposes. An independent cooling and purification system maintains the water temperature below 40oC, electric conductivity below 5 μS/cm and pH between 5.5 and 6.5. The pool is equipped with a radiation monitoring system and an IAEA camera is used for surveillance. To accommodate our future needs we consider three options: To use the space previously available to accommodate the fuels from the shut down TRIGA type reactors; to expand the storage capacity, by changing the design of the storage module, or to return the spent fuels to the USA, taking advantage of the take-back programme. Another option under consideration is to store research reactor spent fuels together with the spent fuels from nuclear power plants (NPPs). This issue is under discussion and a new national policy for spent fuel management is expected to be defined soon, in a timely manner through national consensus by public consultation. (author)

  1. Spent nuclear fuel project - criteria document spent nuclear fuel final safety analysis report

    International Nuclear Information System (INIS)

    The criteria document provides the criteria and planning guidance for developing the Spent Nuclear Fuel (SNF) Final Safety Analysis Report (FSAR). This FSAR will support the US Department of Energy, Richland Operations Office decision to authorize the procurement, installation, installation acceptance testing, startup, and operation of the SNF Project facilities (K Basins, Cold Vacuum Drying Facility, and Canister Storage Building)

  2. HTGR Spent Fuel Treatment Program. HTGR Spent Fuel Treatment Development Program Plan

    International Nuclear Information System (INIS)

    The spent fuel treatment (SFT) program plan addresses spent fuel volume reduction, packaging, storage, transportation, fuel recovery, and disposal to meet the needs of the HTGR Lead Plant and follow-on plants. In the near term, fuel refabrication will be addressed by following developments in fresh fuel fabrication and will be developed in the long term as decisions on the alternatives dictate. The formulation of this revised program plan considered the implications of the Nuclear Waste Policy Act of 1982 (NWPA) which, for the first time, established a definitive national policy for management and disposal of nuclear wastes. Although the primary intent of the program is to address technical issues, the divergence between commercial and government interests, which arises as a result of certain provisions of the NWPA, must be addressed in the economic assessment of technically feasible alternative paths in the management of spent HTGR fuel and waste. This new SFT program plan also incorporates a significant cooperative research and development program between the United States and the Federal Republic of Germany. The major objective of this international program is to reduce costs by avoiding duplicate efforts

  3. Investigations into the behavior of high-burnup spent fuel and its further uses. Initial literature research

    International Nuclear Information System (INIS)

    The core of the text consists of 2 chapters: Description of advanced fuel cycles aimed at a further use of spent nuclear fuel (principles of scenarios of advanced fuel cycles (AFC); Mass-balance schemes of AFC scenarios; Technical and economic characteristics of AFC projects; A brief overview of technological processes of spent nuclear fuel processing; Transmutation; New aspects of radioactive waste management; Overview of nuclear fuel types for the advanced fuel cycle), and Strategy of the approach to advanced fuel cycles in the US. The Annex is devoted to advanced separation processes for application to nuclear fuel reprocessing (PUREX process; Extraction of the sum of trivalent lanthanides and actinides - TRUEX, DIAMEX, TRPO, DIDPA; Separation of trivalent actinides from trivalent lanthanides - SANEX, TALSPEAK; Separation of Am(III) from Cm(III); Separation of Cs and Sr; Dicarbollides; Complete spent fuel reprocessing processes - UREX+, ARTIST). (P.A.)

  4. Potential Interface Issues in Spent Fuel Management

    International Nuclear Information System (INIS)

    This publication is an output of a series of meetings to identify and evaluate issues and opportunities associated with interfaces in the back end of the fuel cycle (BEFC) and to describe effective management approaches based on the experience of Member States. During the meetings, participants from Member States and other international organizations shared and evaluated the main interfaces and potential interface issues among the spent fuel storage, transport, reprocessing and disposal of the BEFC, and also reviewed the national approaches to addressing these issues. The aim of this publication is to provide an approach to identify the interfaces in the BEFC as well as the potential issues that should be addressed. It also aims at responding to the solutions Member States most often find to be effective and, in some cases, were adjusted or revisited to reach the fixed target. Most of the interfaces and issues are country specific, as evidenced by the variety and diversity of examples provided in this publication

  5. Spent fuel disposal in deep geological repositories

    International Nuclear Information System (INIS)

    The technical reasons warranting a postponement of spent fuel disposal are explained, and the time schedule is given of activities associated with the design, construction, and start-up of a deep geological repository, stretching as far as 2070. The status of preparatory activities implemented so far in relation to the repository is summarized. The design is under development but its future is dependent on the Atomic Act which is being prepared by the Czech Government. The problem of liabilities associated with the fuel cycle back end is discussed. A brief overview of deep geological disposal plans in Finland, France, Germany and Sweden is presented. The estimated cost of ultimate radioactive waste disposal is given. (J.B.). 3 tabs., 1 fig

  6. Recriticality risk in PWR spent fuel pools

    International Nuclear Information System (INIS)

    In this paper we investigated the situation in a PWR Spent Fuel Pool (SFP) following a long-term loss of power / loss of cooling accident. In the SFP there is a large amount of water with soluble boron between the fuel assemblies. There may be a problem from the point of view of criticality safety if the water of the SFP starts to boil and evaporate. A thermal-hydraulic analysis was performed using a simplified model of the SFP. The thermal-hydraulic analysis shows that in all cases a chaotic boiling phenomenon develops. This indicates that even if there is an issue of (near-)criticality, it will have a very intermittent nature. The multiplication factor of the SFP was evaluated with a Monte Carlo calculation. The neutronic analysis was performed for several representative cooling situations. In all cases, the system remains (deeply) subcritical. (author)

  7. Spent fuel management: reprocessing or storage

    International Nuclear Information System (INIS)

    A review of the spent fuel management concepts generally adopted in several countries is presented, including an analysis of the brazilian situation. The alternatives are the reprocessing, the interim storage and the final disposal in a repository after appropriate conditioning. The commercial operating reprocessing facilities in the Western World are located in France and in the United Kingdom. In the USA the anti-reprocessing policy from 1977 changed in 1981, when the Government supported the resumption of commercial reprocessing and designated the private sector as responsible for providing these services. Small scale facilities are operating in India, Italy, Japan and West Germany. Pilot plant for LWR fuel are being planned by Spain, Pakistan and Argentina. (Author)

  8. Conceptual development of a test facility for research in spent fuel management

    International Nuclear Information System (INIS)

    Spent fuel management is an important issue for nuclear power programs, requiring careful planning and implementation. With the wait-and-see policy on spent fuel management in Korea, research efforts at KAERI are directed to developing advanced technologies for the safer and more efficient management of the accumulating spent fuels. In support of these research perspectives, a test facility on a pilot scale is being developed with provisions for integral demonstration of the multitude of technical functions required for advanced spent fuel management. The facility, named SMATER, is to be capable of handling full size assemblies of spent PWR fuel (as well as CANDU fuel) with a maximum capacity of 20 MTU/year (about 48 assemblies of PWR type). Major functions of the facility are consolidation of spent PWR fuel assembly into a half volume package and optionally, the 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 DUPIC fuel for reuse in CANDU reactors in the latter case. 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 material is configured in such a way as to maximize space utilization and ergonometrics, and to minimize costs within the given functional and operational requirements. (author)

  9. Transportation and storage of foreign spent power reactor fuel

    International Nuclear Information System (INIS)

    This report describes the generic actions to be taken by the Department of Energy, in cooperation with other US government agencies, foreign governments, and international organizations, in support of the implementation of Administration policies with respect to the following international spent fuel management activities: bilateral cooperation related to expansion of foreign national storage capacities; multilateral and international cooperation related to development of multinational and international spent fuel storage regimes; fee-based transfer of foreign spent power reactor fuel to the US for storage; and emergency transfer of foreign spent power reactor fuel to the US for storage

  10. Legal questions concerning the termination of spent fuel element reprocessing

    International Nuclear Information System (INIS)

    The thesis on legal aspects of the terminated spent fuel reprocessing in Germany is based on the legislation, jurisdiction and literature until January 2004. The five chapters cover the following topics: description of the problem; reprocessing of spent fuel elements in foreign countries - practical and legal aspects; operators' responsibilities according to the atomic law with respect to the reprocessing of Geman spent fuel elements in foreign countries; compatibility of the prohibition of Geman spent fuel element reprocessing in foreign countries with international law, European law and German constitutional law; results of the evaluation

  11. Calculation study of TNPS spent fuel pool using burnup credit

    International Nuclear Information System (INIS)

    Exampled by the spent fuel pool of TNPS which is consist of 2 × 5 fuel storage racks, the spent fuel high-density storage based on burnup credit (BUC) and related criticality safety issues were studied. The MONK9A code was used to analyze keff, of different enrichment fuels at different burnups. A reference loading curve was proposed in accordance with the system keff's changing with the burnup of different initially enriched nuclear fuels. The capacity of the spent fuel pool increases by 31% compared with the one that does not consider BUC. (authors)

  12. Buckling analysis of spent fuel basket

    International Nuclear Information System (INIS)

    The basket for a spent fuel shipping cask is subjected to compressive stresses that may cause global instability of the basket assemblies or local buckling of the individual members. Adopting the common buckling design practice in which the stability capacity of the entire structure is based on the performance of the individual members of the assemblies, the typical spent fuel basket, which is composed of plates and tubular structural members, can be idealized as an assemblage of columns, beam-columns and plates. This report presents the flexural buckling formulas for five load cases that are common in the basket buckling analysis: column under axial loads, column under axial and bending loads, plate under uniaxial loads, plate under biaxial loadings, and plate under biaxial loads and lateral pressure. The acceptance criteria from the ASME Boiler and Pressure Vessel Code are used to determine the adequacy of the basket components. Special acceptance criteria are proposed to address the unique material characteristics of austenitic stainless steel, a material which is frequently used in the basket assemblies

  13. Antineutrino monitoring of spent nuclear fuel

    CERN Document Server

    Brdar, Vedran; Kopp, Joachim

    2016-01-01

    Military and civilian applications of nuclear energy have left a significant amount of spent nuclear fuel over the past 70 years. Currently, in many countries world wide, the use of nuclear energy is on the rise. Therefore, the management of highly radioactive nuclear waste is a pressing issue. In this letter, we explore antineutrino detectors as a tool for monitoring and safeguarding nuclear waste material. We compute the flux and spectrum of antineutrinos emitted by spent nuclear fuel elements as a function of time, and we illustrate the usefulness of antineutrino detectors in several benchmark scenarios. In particular, we demonstrate how a measurement of the antineutrino flux can help to re-verify the contents of a dry storage cask in case the monitoring chain by conventional means gets disrupted. We then comment on the usefulness of antineutrino detectors at long-term storage facilities such as Yucca mountain. Finally, we put forward antineutrino detection as a tool in locating underground "hot spots" in ...

  14. Japan's spent fuel and plutonium management challenge

    International Nuclear Information System (INIS)

    Japan's commitment to plutonium recycling has been explicitly stated in its long-term program since 1956. Despite the clear cost disadvantage compared with direct disposal or storage of spent fuel, the Rokkasho reprocessing plant started active testing in 2006. Japan's cumulative consumption of plutonium has been only 5 tons to date and its future consumption rate is still uncertain. But once the Rokkasho reprocessing plant starts its full operation, Japan will separate about 8 tons of plutonium annually. Our analysis shows that, with optimum use of available at-reactor and away-from-reactor storage capacity, there would be no need for reprocessing until the mid-2020s. With an additional 30,000 tons of away-from-reactor (AFR) spent-fuel storage capacity reprocessing could be avoided until 2050. Deferring operation of the Rokkasho plant, at least until the plutonium stockpile had been worked down to the minimum required level, would also minimize international concern about Japan's plutonium stockpile. The authors are happy to acknowledge Frank von Hippel, Harold Feiveson, Jungming Kang, Zia Mian, M.V. Ramana, and other IPFM members, as well as the generous grant from the MacArthur Foundation for helping make this research possible.

  15. Buckling analysis of spent fuel basket

    Energy Technology Data Exchange (ETDEWEB)

    Lee, A.S.; Bumpas, S.E. [Lawrence Livermore National Lab., CA (United States)

    1995-05-01

    The basket for a spent fuel shipping cask is subjected to compressive stresses that may cause global instability of the basket assemblies or local buckling of the individual members. Adopting the common buckling design practice in which the stability capacity of the entire structure is based on the performance of the individual members of the assemblies, the typical spent fuel basket, which is composed of plates and tubular structural members, can be idealized as an assemblage of columns, beam-columns and plates. This report presents the flexural buckling formulas for five load cases that are common in the basket buckling analysis: column under axial loads, column under axial and bending loads, plate under uniaxial loads, plate under biaxial loadings, and plate under biaxial loads and lateral pressure. The acceptance criteria from the ASME Boiler and Pressure Vessel Code are used to determine the adequacy of the basket components. Special acceptance criteria are proposed to address the unique material characteristics of austenitic stainless steel, a material which is frequently used in the basket assemblies.

  16. Spent Fuel Management in Switzerland. Annex XIII

    International Nuclear Information System (INIS)

    In Switzerland, five nuclear power reactors (3 PWR, 2 BWR) at four sites (Beznau, Mühleberg, Gösgen and Leibstadt) are in operation, with a capacity of about 3200 MWe. A total amount of about 4200 tHM of spent fuel (SF) is expected, conservatively assuming 60 years of operation for each nuclear power plant (NPP). The management of SF and radioactive waste is governed by the federal legislation on nuclear energy. This legislation consists mainly of the following acts and ordinances: Nuclear Energy Act (2003), Nuclear Energy Ordinance (2004) and Ordinance on the Decommissioning and Waste Management Funds (2007, in force since 1st February 2008). Further requirements are detailed in regulatory guidelines. In the year 2000 Switzerland has ratified the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management and has, thus, committed itself to the fulfilment of respective obligations. Handling of SF may only take place in nuclear facilities. The Federal Council has appointed the Swiss Federal Nuclear Safety Inspectorate (ENSI) as the supervisory authority (regulatory body) for nuclear safety, physical protection, and radiation protection for nuclear facilities as well as the Swiss Federal Office of Energy (SFOE) as the supervisory authority for safeguards. ENSI also supervises the preparatory activities for the disposal of radioactive waste and the transport of radioactive material from and to nuclear facilities. In addition, ENSI is the competent authority with respect to the safe transport of dangerous goods of Class 7 in Switzerland

  17. Arrival condition of spent fuel after storage, handling, and transportation

    International Nuclear Information System (INIS)

    This report presents the results of a study conducted to determine the probable arrival condition of spent light-water reactor (LWR) fuel after handling and interim storage in spent fuel storage pools and subsequent handling and accident-free transport operations under normal or slightly abnormal conditions. The objective of this study was to provide information on the expected condition of spent LWR fuel upon arrival at interim storage or fuel reprocessing facilities or at disposal facilities if the fuel is declared a waste. Results of a literature survey and data evaluation effort are discussed. Preliminary threshold limits for storing, handling, and transporting unconsolidated spent LWR fuel are presented. The difficulty in trying to anticipate the amount of corrosion products (crud) that may be on spent fuel in future shipments is also discussed, and potential areas for future work are listed. 95 references, 3 figures, 17 tables

  18. Management of LEU aluminum-clad spent fuel in Argentina

    International Nuclear Information System (INIS)

    Full text: At present, the research and production reactor RA-3 is the only one in Argentina that generates aluminum-clad spent fuel. It was converted to reduced enrichment uranium in 1989 using LEU U3O8 fuel elements, which were developed in CNEA in the frame of the Program on Reduced Enrichment for Research and Test Reactors. The management strategy for the RA-3 spent fuel is presented in regard to interim wet and dry storage and the treatment prior to disposal. Particularly, different alternatives are discussed in relation to the processes being considered for the treatment of the spent fuel. In principle, these processes could be adjusted for spent fuels containing different fuel materials, e.g. U3O8, U3Si2 or U-Mo. A brief description of the available facilities for the spent fuel treatment is presented. (author)

  19. Technical bases for interim storage of spent nuclear fuel

    International Nuclear Information System (INIS)

    The experience base for water storage of spent nuclear fuel has evolved since 1943. The technology base includes licensing documentation, standards, technology studies, pool operator experience, and documentation from public hearings. That base reflects a technology which is largely successful and mundane. It projects probable satisfactory water storage of spent water reactor fuel for several decades. Interim dry storage of spent water reactor fuel is not yet licensed in the US, but a data base and documentation have developed. There do not appear to be technological barriers to interim dry storage, based on demonstrations with irradiated fuel. Water storage will continue to be a part of spent fuel management at reactors. Whether dry storage becomes a prominent interim fuel management option depends on licensing and economic considerations. National policies will strongly influence how long the spent fuel remains in interim storage and what its final disposition will be

  20. Management of spent fuel in the Slovak Republic

    International Nuclear Information System (INIS)

    Full text: The skills in handling spent fuel have been collected in Slovakia for more than 30 years. During this time period a well established spent fuel management system was created. The Slovak Government established the basic policy of spent fuel management in several resolutions. In 2000 the Slovak Government adopted the power policy of the Slovak Republic that is also related to the concept of fuel cycle back-end. In 2001, the Slovak Government in his Resolution No. 5/2001 accepted 'The proposal on the schedule of economical and material solution of the spent fuel management and decommissioning process of nuclear facilities' and decided to submit the 'Policy of decommissioning of nuclear facilities and management of spent fuel evaluated according to the act on environmental impact assessment' for a discussion on governmental level by the end of 2007. The state supervision on nuclear safety of spent fuel management is performed by the UJD. The legislative framework in the Slovak Republic is based on acts and regulations. Acts are at the highest legislative level. Based on general requirements described in the acts, the regulations describe more detailed requirements. Several guides were issued by UJD. Unlike the acts and regulations, guides are not binding for operators. Act No. 541/2004 Coll. on Peaceful Use of Nuclear Energy is the main legislative norm. In Slovakia there are six nuclear power units in operation. These units generate about 500 spent fuel assemblies (approximately 60 ton of heavy metal) per year. Temporary storage of the spent fuel after its unloading from the reactor core is carried out at the at-reactor spent fuel storage pools. The spent fuel is stored in a rack and cooled by boronated water. After at least 2.5 years of storage in the at-reactor pools, the spent fuel is removed to the Interim Spent Fuel Storage Facility (ISFSF). ISFSF was commissioned in 1988. During 1997-2000, it was subject to a reconstruction and seismic upgrade. The

  1. Survey of wet and dry spent fuel storage

    International Nuclear Information System (INIS)

    Spent fuel storage is one of the important stages in the nuclear fuel cycle and stands among the most vital challenges for countries operating nuclear power plants. Continuous attention is being given by the IAEA to the collection, analysis and exchange of information on spent fuel management. Its role in this area is to provide a forum for exchanging information and for coordinating and encouraging closer co-operation among Member States. Spent fuel management is recognized as a high priority IAEA activity. In 1997, the annual spent fuel arising from all types of power reactors worldwide amounted to about 10,500 tonnes heavy metal (t HM). The total amount of spent fuel accumulated worldwide at the end of 1997 was about 200,000 t HM of which about 130,000 t HM of spent fuel is presently being stored in at-reactor (AR) or away-from-reactor (AFR) storage facilities awaiting either reprocessing or final disposal and 70,000 t HM has been reprocessed. Projections indicate that the cumulative amount generated by 2010 may surpass 340,000 t HM and by the year 2015 395,000 t HM. Part of the spent fuel will be reprocessed and some countries took the option to dispose their spent fuel in a repository. Most countries with nuclear programmes are using the deferral of a decision approach, a 'wait and see' strategy with interim storage, which provides the ability to monitor the storage continuously and to retrieve the spent fuel later for either direct disposal or reprocessing. Some countries use different approaches for different types of fuel. Today the worldwide reprocessing capacity is only a fraction of the total spent fuel arising and since no final repository has yet been constructed, there will be an increasing demand for interim storage. The present survey contains information on the basic storage technologies and facility types, experience with wet and dry storage of spent fuel and international experience in spent fuel transport. The main aim is to provide spent fuel

  2. CFD Simulation of Spent Fuel in a Dry Storage System

    International Nuclear Information System (INIS)

    The spent fuel pool is expected to be full in few years. It is a serious problem one should not ignore. The dry storage type is considered as the interim storage system in Korea. The system stores spent fuel in a storage canister filled with an inert gas and the canister is cooled by a natural convection system using air or helium, radiation, and conduction. The spent fuel is heated by decay heat. The spent fuel is allowed to cool under a limiting temperature to avoid a fuel failure. Recently, the thermal hydraulic characteristics for a single bundle of the spent fuel were investigated through a CFD simulation. It would be of great interest to investigate the maximum fuel temperature in a dry storage system. The present paper deals with the thermal hydraulic characteristics of spent fuel for a dry storage system using the CFD method. A 3-D thermal flow simulation was carried out to predict the temperature of spent fuel. A dry storage system composed of 32 fuel bundles was modeled. The inlet temperature of the outer bundle is higher and that of inner bundle, however, is higher at the outlet. In a single fuel assembly, a center temperature of the fuel assembly was higher than elsewhere

  3. CFD Simulation of Spent Fuel in a Dry Storage System

    Energy Technology Data Exchange (ETDEWEB)

    Kwack, Young Kyun; In, Wang Kee; Shin, Chang Hwan; Chun, Tae Hyun; Kook, Dong Hak [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    The spent fuel pool is expected to be full in few years. It is a serious problem one should not ignore. The dry storage type is considered as the interim storage system in Korea. The system stores spent fuel in a storage canister filled with an inert gas and the canister is cooled by a natural convection system using air or helium, radiation, and conduction. The spent fuel is heated by decay heat. The spent fuel is allowed to cool under a limiting temperature to avoid a fuel failure. Recently, the thermal hydraulic characteristics for a single bundle of the spent fuel were investigated through a CFD simulation. It would be of great interest to investigate the maximum fuel temperature in a dry storage system. The present paper deals with the thermal hydraulic characteristics of spent fuel for a dry storage system using the CFD method. A 3-D thermal flow simulation was carried out to predict the temperature of spent fuel. A dry storage system composed of 32 fuel bundles was modeled. The inlet temperature of the outer bundle is higher and that of inner bundle, however, is higher at the outlet. In a single fuel assembly, a center temperature of the fuel assembly was higher than elsewhere.

  4. Introduction of new flasks for high burnup spent fuel

    International Nuclear Information System (INIS)

    New flasks have been designed to transport the high burnup spent fuels now becoming available from the world's nuclear power stations. Two versions have been designed: Excellox 6 for 5 metre PWR fuels and Excellox 7 with increased neutron shielding for 4.5 metre PWR and BWR fuels arising in Japan. The designs of these flasks have been finalised; Excellox 6 has been approved and validated as a Type B(U)F package and the first two have been manufactured and are now in routine service, with a third at an advanced stage of manufacture. The Excellox 7 design is ready for manufacture when service requirements for it have been settled. An account is given of the final adjustments to the design in the course of manufacture, the main steps and tests in the manufacturing process and the commissioning tests at the reprocessing and reactor sites. The entry of the flasks into service is reviewed. (author)

  5. TVO's new encapsulation method for spent nuclear fuel

    International Nuclear Information System (INIS)

    Teollisuuden Voima Oy has developed a new encapsulation method for spent nuclear fuel s.c. cold process. Instead of casting (400 deg C) molten lead the new canister is filled with cold granulated material like quartz sand, lead shots or glassa beads. The new canister concept ACPC (Advanced Cold Process Canister) consists of the outer oxygen free copper canister and of the inner steel canister. The function of the steel canister is merely to give mechanical strength. The copper canister acts as corrosion protection guaranteeing practically lifetime of millions of years for the ACPC concept

  6. The State of the WWER Nuclear Spent Fuel Management in Ukraine and Trends on the Optimal Choice of Spent Fuel Management Strategy

    International Nuclear Information System (INIS)

    The paper gives an overview of the operational state of the away from reactor interim spent fuel storage facility at the Zaporizhzhya NPP site, commissioned according to the dry storage cask technology that is developed by the US company Sierra Nuclear Corporation. The trends of the SFM at the WWER nuclear units were noted, which are caused by the advanced nuclear fuel implementation in order to improve the nuclear fuel utilization. (author)

  7. APPLICATIONS OF CURRENT TECHNOLOGY FOR CONTINUOUS MONITORING OF SPENT FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Drayer, R.

    2013-06-09

    Advancements in technology have opened many opportunities to improve upon the current infrastructure surrounding the nuclear fuel cycle. Embedded devices, very small sensors, and wireless technology can be applied to Security, Safety, and Nonproliferation of Spent Nuclear Fuel. Security, separate of current video monitoring systems, can be improved by integrating current wireless technology with a variety of sensors including motion detection, altimeter, accelerometer, and a tagging system. By continually monitoring these sensors, thresholds can be set to sense deviations from nominal values. Then alarms or notifications can be activated as needed. Safety can be improved in several ways. First, human exposure to ionizing radiation can be reduced by using a wireless sensor package on each spent fuel cask to monitor radiation, temperature, humidity, etc. Since the sensor data is monitored remotely operator stay-time is decreased and distance from the spent fuel increased, so the overall radiation exposure is reduced as compared to visual inspections. The second improvement is the ability to monitor continuously rather than periodically. If changes occur to the material, alarm thresholds could be set and notifications made to provide advanced notice of negative data trends. These sensor packages could also record data to be used for scientific evaluation and studies to improve transportation and storage safety. Nonproliferation can be improved for spent fuel transportation and storage by designing an integrated tag that uses current infrastructure for reporting and in an event; tracking can be accomplished using the Iridium satellite system. This technology is similar to GPS but with higher signal strength and penetration power, but lower accuracy. A sensor package can integrate all or some of the above depending on the transportation and storage requirements and regulations. A sensor package can be developed using off the shelf technology and applying it to each

  8. Applications of current technology for continuous monitoring of spent fuel

    International Nuclear Information System (INIS)

    Advancements in technology have opened many opportunities to improve upon the current infrastructure surrounding the nuclear fuel cycle. Embedded devices, very small sensors, and wireless technology can be applied to Security, Safety, and Nonproliferation of Spent Nuclear Fuel. Security, separate of current video monitoring systems, can be improved by integrating current wireless technology with a variety of sensors including motion detection, altimeter, accelerometer, and a tagging system. By continually monitoring these sensors, thresholds can be set to sense deviations from nominal values. Then alarms or notifications can be activated as needed. Safety can be improved in several ways. First, human exposure to ionizing radiation can be reduced by using a wireless sensor package on each spent fuel cask to monitor radiation, temperature, humidity, etc. Since the sensor data is monitored remotely operator stay-time is decreased and distance from the spent fuel increased, so the overall radiation exposure is reduced as compared to visual inspections. The second improvement is the ability to monitor continuously rather than periodically. If changes occur to the material, alarm thresholds could be set and notifications made to provide advanced notice of negative data trends. These sensor packages could also record data to be used for scientific evaluation and studies to improve transportation and storage safety. Nonproliferation can be improved for spent fuel transportation and storage by designing an integrated tag that uses current infrastructure for reporting and in an event; tracking can be accomplished using the Iridium satellite system. This technology is similar to GPS but with higher signal strength and penetration power, but lower accuracy. A sensor package can integrate all or some of the above depending on the transportation and storage requirements and regulations. A sensor package can be developed using off the shelf technology and applying it to each

  9. Visual inspection of underwater spent nuclear fuel

    International Nuclear Information System (INIS)

    There are many challenges associated with visual inspection of spent nuclear fuel stored in water. Two of the biggest challenges are high radiation fields and the old adaga ''water and electricity don't mix''. Two and one half years ago underwater inspections in nuclear fuel storage facilities were started at the Idaho National Engineering Laboratory. Systems have been operated around the clock for several months at a time. Camera systems have been exposed to radiation fields in excess of 10,000 grey per hour and have cumulative doses of several thousand grey. The video systems are crucial for fuel identification, repackaging, mechanical fastener verification, nondestructive examination probe placement and to examine the amount of corrosion to fuel cans and storage racks. Several camera systems fabricated from commercially available components are utilized in the underwater storage facilities. These include: the Rod camera, for steady pictures 7 meters deep in the water through a 1 centimeter crack in the floor; the Puppet camera, for close shots in buckets or rack ports; the Video Probe, for inspecting fuel cans in their storage position; and the Phantom II, submersible vehicle for visual information and parts retrieval. Waterproofing systems for filed deployment provides many learning opportunities. O-ring placement, pressurized housings, hermetically sealed connectors and silica gel contribute to successful visual inspections in water. Radiation effects were as expected, browning of the camera lens and fiber optics as well as the noise seen in the picture due to radiation bombarding the electronics inside the solid state camera. The waterproof housing provided excellent shielding for the camera system. The camera's orientation to the fuel and the amount of lighting also play an important part in reducing radiological degradation of the video picture. (author). 6 figs

  10. Storage of Spent Nuclear Fuel. Specific Safety Guide

    International Nuclear Information System (INIS)

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

  11. Spent fuel treatment options and application - An Indian perspective

    International Nuclear Information System (INIS)

    Reactors. Reprocessing facilities are being augmented to meet the fuel requirements of the second stage. Process developments are in progress to meet the reprocessing challenges of the fast reactor fuel cycle. For India, the building up of fissile material inventory at a fast pace is a prerequisite for the early introduction of thorium in the power programme. To meet the challenges of a thorium fuel cycle, reprocessing of irradiated thorium for 233U recovery was initiated almost from the beginning. After pilot scale studies in the early seventies, an engineering facility started operating in this domain for the recovery of 233U from thoria irradiated in research reactors using Thorex process. An engineering facility for the recovery of 233U from thoria rods irradiated in PHWRs during the initial flux flattening is expected to be in operation soon. As a predecessor to the third stage reactor for utilisation of thorium, an Advanced Heavy Water Reactor(AHWR) employing both (Th-Pu)O2 and (Th-233U)O2 fuels in one cluster is being inducted into the nuclear power programme. Reprocessing of the spent fuel from this reactor will pose many challenges and will necessitate the development of a flow sheet for the recovery of three components viz. U, Th and Pu. Currently fuel cycle studies are initiated to develop suitable flow sheets to close the AHWR fuel cycle. (author)

  12. Status and prospects for spent fuel management in China

    International Nuclear Information System (INIS)

    Full text: The main emphasis of China's nuclear industry has been shifted to peaceful uses of nuclear energy since the 1980s. Up to now, 7 units in 5 NPPs have been constructed and 4 units in 3 NPPs are under construction. The present total installed nuclear capacity is about 5600 MWe. Data of NPPs in Taiwan province is left open for the time being. With the development of the nuclear power, the amount of spent fuel discharged from NPPs is increasing rapidly in China. Chinese government pays its attention to the security of spent fuel, and carries out the policy of nuclear non-proliferation and guarantees the safety to the public and environment all the time. I. Current status of the commercial spent fuel storage. Two commercial nuclear power plants (3 units) are in operation in China, and have discharged spent fuel for several nuclear fuel cycles. All the discharged spent fuel store in the at-reactor (AR) storage pools with wet storage. By the end of December 2002, there are 232 spent fuel assemblies in the pool of Qinshan phase I (CN-1 unit), while 824 assemblies in the pools of NPP Daya Bay (CN-2, CN-3 unit). The accumulated amount of spent fuel generated by NPP Qinshan and NPP Daya Bay is about 445 t HM. For the limitation of the storage pools capacity, NPP Daya Bay is planning to transport the spent fuel stored at-reactor pools to northwest area for wet storage since 2003. The Everclean Environmental Engineering Corporation (EEEC) will be in charge of this transportation using NAC-STC dual-purpose casks that can be used for both shipping and storage. It is worth while to note that CANDU reactor in Qinshan phase III has different fuel structure and different discharged ways of spent fuel compared with other reactors we have had ever before in China. We will still pay attention to this problem. II. The amount of NPP spent fuel in the near future. Currently, there are approximately 60 t HM spent fuel discharged from NPPs every year in China. The NPPs under

  13. Method and device for cleaning spent fuel assembly

    International Nuclear Information System (INIS)

    A spent fuel assembly is immersed in a liquid metal in a pot disposed below a cleaning vessel which is under the floor of an argon gas cell, and the liquid metal in the pot is heated by a heater disposed at the periphery of the cleaning vessel, and the spent fuel assembly is preheated by the heated liquid metal. Then, in a state where the spent fuel is pulled up from the pot in the cleaning vessel, heating gases are blown to the fuel assembly from above, high temperature argon gases are blown to wash out the liquid metals deposited on the spent fuel assembly. In this way, the spent fuel assembly can be heated to a predetermined preheating temperature in a short period of time. Since the amount of the liquid metal to be recovered by a vapor trap is reduced, the capacity of a storage tank exclusively used for vapor trap can be reduced. (T.M.)

  14. Overview of spent fuel management options: technology and cost

    International Nuclear Information System (INIS)

    There are several spent fuel management options available to utilities beyond the traditional options of reracking and transshipment. Two different dry storage options have been licensed by NRC and it appears that NRC will issue a license amendment to allow for the storage of consolidated spent fuel in the very near future. The use of either of these spent fuel management options is highly dependent upon the individual requirements of the utility. There are several factors that must be considered when choosing a specific fuel management option. Among these are capital and operating costs, local political climate, structural strength of the spent fuel pool (rod consolidation), availability and cost of additional land (dry storage), and the amount of additional effort required, if any, to provide spent fuel in a form acceptable to DOE. Either of these spent fuel management options, dry storage or rod consolidation, is capable of meeting the needs of utilities for additional spent fuel storage until DOE begins accepting spent fuel for final disposal in 1998

  15. 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.

  16. Analytical methodology and facility description spent fuel policy

    International Nuclear Information System (INIS)

    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

  17. Final environmental impact statement: US Spent Fuel Policy. Storage of US spent power reactor fuel

    International Nuclear Information System (INIS)

    The activities associated with implementing or not implementing the proposed policy are similar for a given disposition facility startup date, and environmental impacts vary with the amount of fuel received, the number of Interim Spent Fuel Storage (ISFS) facilities required, the storage time, and to a lesser degree to the amount of spent fuel transported. The environmental impacts from all alternatives considered, either from implementing or not implementing the spent fuel storage policy, are small. The decreased resource consumptions and environmental impacts of alternatives that assume reactor discharge basin operation at less than full-core reserve must be balanced against the reduced flexibility in reactor operation and the possibility of forced shutdowns which could lead to the use of higher-cost substitute power or reduction of electrical power generation. Providing full-core reserve capacity is prudent and economical to avoid reactor outages due to inspections or emergency situations. The impacts for decentralized ISFSs providing full-core reserve are considered the same for either government or private facilities. Nevertheless, utilities have operated without full-core reserve rather than shut down. At-reactor storage increases environmental effects compared with ISFS basin storage because additional storage basins are constructed and operated. However, the impacts are relatively small compared with available resources and risks from natural radiation sources

  18. Studies and research concerning BNFP: LWR spent fuel storage

    International Nuclear Information System (INIS)

    This report describes potential spent fuel storage expansion programs using the Barnwell Nuclear Fuel Plant--Fuel Receiving and Storage Station (BNFP-FRSS) as a model. Three basic storage arrangements are evaluated with cost and schedule estimates being provided for each configuration. A general description of the existing facility is included with emphasis on the technical and equipment requirements which would be necessary to achieve increased spent fuel storage capacity at BNFP-FRSS

  19. Management of spent fuel from power reactors in Argentina

    International Nuclear Information System (INIS)

    A brief description of the two operative nuclear power plant and their fuel assemblies is given, as well as the fuel consumption and the expected quantities of spent fuel to be accumulated at the end of their lives. It is also described the legal framework, the organization and the current practice for the storage of the spent nuclear fuel in both nuclear power plants, as well as the management strategy for the future. (author)

  20. Monitoring instrumentation spent fuel management program. Final report

    Energy Technology Data Exchange (ETDEWEB)

    1979-01-01

    Preliminary monitoring system methodologies are identified as an input to the risk assessment of spent fuel management. Conceptual approaches to instrumentation for surveillance of canister position and orientation, vault deformation, spent fuel dissolution, temperature, and health physics conditions are presented. In future studies, the resolution, reliability, and uncertainty associated with these monitoring system methodologies will be evaluated.

  1. Monitoring instrumentation spent fuel management program. Final report

    International Nuclear Information System (INIS)

    Preliminary monitoring system methodologies are identified as an input to the risk assessment of spent fuel management. Conceptual approaches to instrumentation for surveillance of canister position and orientation, vault deformation, spent fuel dissolution, temperature, and health physics conditions are presented. In future studies, the resolution, reliability, and uncertainty associated with these monitoring system methodologies will be evaluated

  2. Radwaste management and spent fuel management in JAVYS

    International Nuclear Information System (INIS)

    In this work authors present radwaste management and spent fuel management in JAVYS, a.s. Processing of radioactive wastes (RAW) in the Bohunice Radioactive Waste Processing Center and surface storage of RAW in National RAW Repository as well as Interim Spent fuel storage in Jaslovske Bohunice are presented.

  3. Near-field chemistry of the spent nuclear fuel repository

    International Nuclear Information System (INIS)

    Factors affecting near-field chemistry of the spent nuclear fuel repository as well as the involved mutual interactions are described on the basis of literature. The most important processes in the near-field (spent-fuel, canister and bentonite) are presented. The related examples on near-field chemistry models shed light on the extensive problematics of near-field chemistry. (authors)

  4. Spent-fuel photon and neutron source spectra

    International Nuclear Information System (INIS)

    Computational activities at Oak Ridge National Laboratory have been performed to develop appropriate data and techniques for computing the photon and neutron source spectra of spent fuel. The methods reviewed here include both the determination of spent-fuel composition and the radiation source spectra associated with these isotopic inventories

  5. Breeder Spent Fuel Handling Program multipurpose cask design basis document

    International Nuclear Information System (INIS)

    The Breeder Spent Fuel Handling (BSFH) Program multipurpose cask Design Basis Document defines the performance requirements essential to the development of a legal weight truck cask to transport FFTF spent fuel from reactor to a reprocessing facility and the resultant High Level Waste (HLW) to a repository. 1 ref

  6. Improvement of shacking helical elevators used in spent fuel reprocessing

    International Nuclear Information System (INIS)

    For reprocessing cut spent fuel elements are introduced in a tank and raised gradually with an helical ramp by a back and forth motion around a vertical axis. Spent fuel is dissolved and hulls are recovered at the top of the ramp

  7. An approach to meeting the spent fuel standard

    Energy Technology Data Exchange (ETDEWEB)

    Makhijani, A. [Institute for Energy and Environmental Research, Takoma Park, MD (United States)

    1996-05-01

    The idea of the spent fuel standard is that there should be a high surface gamma radiation to prevent theft. For purposes of preventing theft, containers should be massive, and the plutonium should be difficult to extract. This report discusses issues associated with the spent fuel standard.

  8. Contributions to LWR spent fuel storage and transport

    International Nuclear Information System (INIS)

    The papers included in this document describe the aspects of spent LWR fuel storage and transport-behaviour of spent fuel during storage; use of compact storage packs; safety of storage; design of storage facilities AR and AFR; description of transport casks and transport procedures

  9. 77 FR 28406 - Spent Fuel Transportation Risk Assessment

    Science.gov (United States)

    2012-05-14

    ... COMMISSION Spent Fuel Transportation Risk Assessment AGENCY: Nuclear Regulatory Commission. ACTION: Draft... issuing for public comment a draft NUREG, NUREG-2125, ``Spent Fuel Transportation Risk Assessment (SFTRA... safety assurance. In that assessment, the measure of safety was the risk of radiation doses to the...

  10. Spent nuclear fuel project quality assurance program plan

    Energy Technology Data Exchange (ETDEWEB)

    Lacey, R.E.

    1997-05-09

    This main body of this document describes how the requirements of 10 CFR 830.120 are met by the Spent Nuclear Fuel Project through implementation of WHC-SP-1131. Appendix A describes how the requirements of DOE/RW-0333P are met by the Spent Nuclear Fuel Project through implementation of specific policies, manuals, and procedures.

  11. Environmental phylosophy of the spent fuel and radioactive waste management

    International Nuclear Information System (INIS)

    The management of spent fuel and radioactive wastes differs considerably from the management of other types of waste. The main difference of spent fuel and radioactive waste management is the necessity of control and supervision over very long periods of time. Various aspects relating to this issue are addressed. (author)

  12. Characterization plan for Hanford spent nuclear fuel

    International Nuclear Information System (INIS)

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

  13. Nevada commercial spent nuclear fuel transportation experience

    International Nuclear Information System (INIS)

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

  14. Exponential experiments on PWR spent fuel assemblies

    International Nuclear Information System (INIS)

    An Exponential experiment system which is composed of a neutron detector, a signal analysis system and a neutron source, Cf-252 has been installed in order to experimentally determine the neutron effective multiplication factor for PWR spent fuel assembly. The axial background neutron flux is measured as a preliminary performance test. From the results, the spacer grid position is determined to be consistent with the design specifications within a 2.3% relative error. The induced fission neutron for four of the assemblies is also measured by scanning the neutron source, Cf-252 or the neutron detector. The exponential decay constants have been evaluated by the application of Poisson regression to the net induced fission neutron counts. It was revealed that the average exponential decay constants for the C15, J14, G23 and J44 assemblies were determined to be 0.130, 0.127, 0.125 and 0.121, respectively. (author)

  15. Study on spent fuel rejuvenation in PROMETHEUS fusion reactor

    International Nuclear Information System (INIS)

    This study presents the spent fuel rejuvenation potential of the PROMETHEUS-H fusion reactor. For this purpose, three different spent fuels were selected, i.e. (1) CANDU (2) PWR-UO2 and (3) PWR-MOX spent fuels. The spent fuel (volume fraction of 60%), spherically prepared and cladded with SiC (volume fraction of 10%), was located in the fuel zone (FZ) in the blanket of the modified PROMETHEUS-H fusion reactor. The FZ was cooled with high pressure helium gas (volume fraction of 30%) for the nuclear heat transfer. The neutronic calculations were performed by solving the Boltzmann transport equation with the help of the neutron transport code XSDRNPM-S/SCALE 4.3. The calculations of the time dependent atomic densities of the isotopes were performed for an operation period (OP) of up to 4 years with a 75% plant factor (η) under a first wall neutron load (P) of 4.7 MW/m2. The temporal variations of the atomic densities of the isotopes in the spent fuel composition and other physical parameters were calculated for a discrete time interval (Δt) of 1/12 year (one month) by using the interface program (code). In all investigated spent fuel cases, the tritium self sufficiency is maintained for the DT fusion driver along the OP. The CANDU spent fuel becomes usable in a conventional CANDU reactor after a regeneration time of ∼5.5 months. The CFFE value approaches 3.5% in the blanket fuelled with the PWR-UO2 and PWR-MOX spent fuels after 41 and 35 months, respectively. The plutonium component can never reach a nuclear weapon grade quality during the spent fuel rejuvenation. Consequently, the modified PROMETHEUS-H fusion reactor has high neutronic performance for the rejuvenation of the spent fuels

  16. Security Requirements for Spent Fuel Storage Systems

    International Nuclear Information System (INIS)

    The U.S. Nuclear Regulatory Commission (Commission or NRC) requires high assurance of adequate protection of public health and safety and the common defense and security for the storage of spent nuclear fuel. Following the terrorist attacks of September 11, 2001, the NRC has achieved this requisite high assurance for all independent spent fuel storage installations (ISFSIs) through a combination of existing security regulations and the issuance of new security orders to individual licensees. However, the NRC's current security regulations for ISFSIs are quite complex and pose challenges both to NRC staff and to the regulated industry. This regulatory complexity is due to multiple factors, including: two different types of ISFSI licenses (general licenses and specific licenses) and varying applicability of regulations based upon whether the ISFSI is collocated with an operating power reactor, collocated with a decommissioning power reactor, or is located away from any power reactors. The NRC's ISFSI security regulations were last comprehensively updated in the early 1990's. Moreover, the nature and characteristics of the threat environment have evolved significantly since that time. The Commission has directed the NRC staff to begin development of a risk-informed and performance-based update to the ISFSI security regulations which will enhance the ISFSI security regulations, while continuing to ensure the common defense and security and public health and safety are adequately protected under the current threat environment. The NRC staff is developing the technical bases supporting this ISFSI security rulemaking. The NRC's specific goals for this rulemaking are to update ISFSI security requirements to apply consistently to both types of ISFSI licensees, to improve the clarity of NRC regulations, to generically incorporate the provisions of the post-9/11 security orders, and to incorporate the Commission's direction on several specific policy issues. The Commission

  17. International conference on storage of spent fuel from power reactors. Book of extended synopses

    International Nuclear Information System (INIS)

    The management of spent nuclear fuel is a key aspect characterizing the use of nuclear power around the world. At the international level, there is an ongoing debate focused on this issue. At the national level, spent fuel management often provokes public concern. Continuous attention is being given by the IAEA to the collection, analysis and exchange of information on spent fuel storage. Its role in this area is to: provide a forum for exchanging information; identify the key issues for long term storage; and co-ordinate and encourage closer co-operation among Member States in certain research and development activities that are of common interest. Meetings on this topic have been organized about once every four years since 1987. The objectives of the Conference were to: review recent advances in spent fuel storage technology; exchange information on the state of the art of and prospects for spent fuel storage; review and discuss the worldwide situation and the major factors influencing national policies in this field; exchange information on operating experience with wet and dry storage facilities; identify the most important directions for future national efforts and international co-operation in this area. The following subjects were covered in the topical sessions: National Programmes: the status and trends of spent fuel storage in Member States, spent fuel arising, amount of spent fuel stored, wet and dry storage capacities, storage facilities under construction and in planning and the national policy for the back end of the fuel cycle; Technologies: technological approaches for long term storage, new storage concepts, re-racking of fuel pools, spent fuel and material behaviour in long term storage; Experience and Licensing: experience in wet and dry storage, problems with materials in fuel pools, licensing practices for spent fuel storage facilities, license extension and re-licensing of existing facilities; R and D and Special Aspects: highly enriched fuel

  18. The TVO concept for direct disposal of spent fuel

    International Nuclear Information System (INIS)

    Teollisuuden Voima Oy (TVO) is responsible for the management of spent fuel produced by the Olkiluoto power plant. TVO's current programme of spent fuel management is based on the guidelines and time schedule set by the Finnish Government. TVO has studied a final disposal concept in which the spent fuel bundles are encapsulated in copper canisters and emplaced in Finnish bedrock. According to the plan the final repository for spent fuel will be in operation by 2020. TVO's updated technical plans for the disposal of spent fuel together with a performance analysis (TVO-92) will be submitted to the authorities by the end of 1992. The paper describes TVO's new encapsulation process, canister design and repository layout. (author). 5 refs, 6 figs

  19. Case histories of West Valley spent fuel shipments: Final report

    International Nuclear Information System (INIS)

    In 1983, NRC/FC initiated a study on institutional issues related to spent fuel shipments originating at the former spent fuel processing facility in West Valley, New York. FC staff viewed the shipment campaigns as a one-time opportunity to document the institutional issues that may arise with a substantial increase in spent fuel shipping activity. NRC subsequently contracted with the Aerospace Corporation for the West Valley Study. This report contains a detailed description of the events which took place prior to and during the spent fuel shipments. The report also contains a discussion of the shipment issues that arose, and presents general findings. Most of the institutional issues discussed in the report do not fall under NRC's transportation authority. The case histories provide a reference to agencies and other institutions that may be involved in future spent fuel shipping campaigns. 130 refs., 7 figs., 19 tabs

  20. Case histories of West Valley spent fuel shipments: Final report

    Energy Technology Data Exchange (ETDEWEB)

    1987-01-01

    In 1983, NRC/FC initiated a study on institutional issues related to spent fuel shipments originating at the former spent fuel processing facility in West Valley, New York. FC staff viewed the shipment campaigns as a one-time opportunity to document the institutional issues that may arise with a substantial increase in spent fuel shipping activity. NRC subsequently contracted with the Aerospace Corporation for the West Valley Study. This report contains a detailed description of the events which took place prior to and during the spent fuel shipments. The report also contains a discussion of the shipment issues that arose, and presents general findings. Most of the institutional issues discussed in the report do not fall under NRC's transportation authority. The case histories provide a reference to agencies and other institutions that may be involved in future spent fuel shipping campaigns. 130 refs., 7 figs., 19 tabs.

  1. IAEA spent fuel management programme - Past and present

    International Nuclear Information System (INIS)

    The Agency's programme continued to cover developments at all stages of the nuclear fuel cycle. Spent fuel management is one of the most important parts of the nuclear fuel cycle and it is a vital and common problem for all countries developing nuclear power. The Agency's role in this area is to provide a forum for exchanging information on spent fuel management, including technical, environmental, economical and safety aspects of transport, storage and reprocessing of spent fuel, and to encourage closer co-operation among Member States in certain research and development activities that are of common interest. An increased emphasis was placed on nuclear safety, economics, development concepts for fuel storage etc., and on the integrated or multidisciplinary rather than single technique approached. The Agency's activities were performed in the field of spent fuel management are reviewed in this paper

  2. Try out of technology for dry container storage of spent nuclear fuel of research reactors

    International Nuclear Information System (INIS)

    Full text: At present a great attention is paid to the technology for fuel storage. As for uranium-aluminum fuel applied in the majority of research reactors, recommendations were elaborated on the basis of multiple investigations regarding the water-chemical conditions of wet storage as well as criteria and mechanisms of fuel degradation evaluation during its storage. Technology of dry storage of such fuel is recommended. Dry storage of spent nuclear fuel is an alternative to the wet one but it does not exclude the preliminary cooling of fuel in water so as to decrease the radioactivity level and heat release. At SSC RF RIAR a 'Program of experimental testing of technology for dry container storage of spent nuclear fuel of the research reactors using transport packaging 108/1' is implemented. The aim of the Program is to try out the technology for dry storage of spent fuel assemblies of the research reactors in the advance surveillance mode; determine the limiting admissible period and conditions of dry and wet storage; try out of the monitoring technology of spent assembly conditions; investigate various conditions of the spent fuel assembly storage (temperature and inner medium); determine periodically the spent fuel assembly conditions after stage-by-stage testing; evaluate the container structural material conditions; try out the product drying technology before dry storage; control the cladding integrity state and perform the physical inspection of spent fuel assemblies before dry storage. Only standard spent fuel assemblies of the MIR reactor after their operation and dry storage within the time period from 4 to 40 years are the objects of the first investigation stage. The temperature in the container does not exceed 180 deg. C due to the spent fuel assembly heat release. Air is used as medium. The examinations are performed at the bench located in the open railway platform. To evaluate the external conditions of the assembly and perform the cladding

  3. Spent fuel management: Current status and prospects 1993

    International Nuclear Information System (INIS)

    Spent fuel management has always been one of the most important stages in the nuclear fuel cycle and it is still one of the most vital problems common to all countries with nuclear reactors. It begins with the discharge of spent fuel from a power or a research reactor and ends with its ultimate disposition, either by direct disposal or by reprocessing of the spent fuel. Two options exist at present - an open, once-through cycle with direct disposal of the spent fuel and a closed cycle with reprocessing of the spent fuel and recycling of plutonium and uranium in new mixed oxide fuels. The selection of a spent fuel strategy is a complex procedure in which many factors have to be weighed, including political, economic and safeguards issues as well as protection of the environment. Continuous attention is being given by the IAEA to the collection, analysis and exchange of information on spent fuel management. Its role in this area is to provide a forum for the exchange of information and to co-ordinate and to encourage closer co-operation among Member States in certain research an development activities that are of common interest. Refs, figs and tabs

  4. International safeguards aspects of spent fuel in permanent geological repositories

    International Nuclear Information System (INIS)

    The practice of not reprocessing spent fuel (the once-through cycle) poses one of the requirements of spent fuel management. States which decide not to reprocess spent fuel for recovery of the contained plutonium intend to dispose of the fuel in a geologic repository after appropriate conditioning. Storage facilities at reactors and away-from-reactor facilities will be required for storing and cooling the fuel until suitable repositories are available. In several states, reprocessing of spent fuel is neither envisaged nor considered to be economical. Recent developments make the disposal of spent fuel in geologic repositories more attractive than previously believed, thus introducing new challenges to safeguards. The nuclear community has expressed concern about the pressing need to address issues of long-term safeguards for the disposal of spent fuel in geologic repositories. According to the authors, the IAEA must develop safeguards requirements and methodology for geologic disposal facilities for spent fuel and formulate a safeguards policy before such facilities enter into operation

  5. Pyrochemical processes for LWR spent fuel

    International Nuclear Information System (INIS)

    Pyrochemical processes are under development at Argonne National Laboratory for recovery of transuranium (TRU) elements from light water reactor spent fuel. The recovered TRU elements will be used as fuel in the integral fast reactor (IFR). Burning these long-lived isotopes for electrical power generation has the additional benefit of reducing the burden on the geological repository for their long-term containment. The goals for the process include: greater than 99.90% recovery of TRU elements, a metal product that is compatible with the IFR fuel cycle, retention of some fission products in the TRU product to enhance proliferation resistance, and a simple process that is economically attractive. The TRU product will be inserted into the IFR fuel cycle for fission product decontamination and incorporation into the fuel. Based on research and development at ANL in the 1960 and 1970 and a comparison of known processes for separation of TRU elements from uranium fuel, three conceptual processes were identified that seem to offer high potential for achieving the desired goals. All three conceptual processes include a reduction step to convert the oxide fuel to metal, an electrochemical step to recover the reductant (calcium) from its oxide, a TRU extraction step to separate TRU elements from the bulk uranium, and a retort step to recover the TRU product from a solvent metal. The candidate processes differ primarily in the methods used to separate the TRU elements from uranium. The salt transport process effects this separation by molten salt extraction; the magnesium extraction process uses the differential solubility of TRU elements in magnesium relative to that of uranium; and the zinc-magnesium process uses phase separation to recover TRU elements, which are soluble in a Zn-Mg alloy. The chemical feasibility of each step of the three concepts has been demonstrated in small-scale experiments. Candidate containment materials have been selected and tested at the

  6. Spent fuel management: Current status and prospects 1990

    International Nuclear Information System (INIS)

    Spent Fuel Management has always been one of the most important steps in the Nuclear Fuel Cycle and it still is one of the most vital and common problem for all countries. Projections for spent fuel arisings by the year 2010 range between 400,000 and 450,000 t of spent nuclear fuel. It is recognized that this fuel will either be stored and later disposed of in a deep geological repository (once-through fuel cycle) or stored and then reprocessed (closed fuel cycle). While some countries have concluded which choice they will make, others are applying the ''wait and see'' attitude. This continues to place great emphasis on short and long term storage technologies since much of the spent fuel will remain in storage in the next 20 years. The nuclear community recognizes the importance that design, technological, economic and material problems in spent fuel storage concepts and continues to encourage the international cooperation in such areas. This past year several nations have made decisions which impact on the projected storage volume (the Federal Republic of Germany has cancelled their reprocessing plant) and plan to contract the reprocessing with other nations. Argentina has delayed its reprocessing efforts. At the same time, while there are plans for recycle of plutonium in thermal reactors, the plans for its use in fast reactors have been delayed. These unforeseen changes reflect the constantly changing nature of the back-end of the fuel cycle and reinforce the importance of cooperation in these activities. The main objective of the Advisory Group on Spent Fuel Management is to review the world-wide situation in spent fuel management, to define the most important directions of national efforts and international cooperation, to exchange information on the present status and progress in performing the back-end of the nuclear fuel cycle and to elaborate recommendations for future Agency programmes in the field of spent fuel management. Refs, figs and tabs

  7. Waste management planned for the advanced fuel cycle facility

    International Nuclear Information System (INIS)

    The U.S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP) program has been proposed to develop and employ advanced technologies to increase the proliferation resistance of spent nuclear fuels, recover and reuse nuclear fuel resources, and reduce the amount of wastes requiring permanent geological disposal. In the initial GNEP fuel cycle concept, spent nuclear fuel is to be reprocessed to separate re-usable transuranic elements and uranium from waste fission products, for fabricating new fuel for fast reactors. The separated wastes would be converted to robust waste forms for disposal. The Advanced Fuel Cycle Facility (AFCF) is proposed by DOE for developing and demonstrating spent nuclear fuel recycling technologies and systems. The AFCF will include capabilities for receiving and reprocessing spent fuel and fabricating new nuclear fuel from the reprocessed spent fuel. Reprocessing and fuel fabrication activities will generate a variety of radioactive and mixed waste streams. Some of these waste streams are unique and unprecedented. The GNEP vision challenges traditional U.S. radioactive waste policies and regulations. Product and waste streams have been identified during conceptual design. Waste treatment technologies have been proposed based on the characteristics of the waste streams and the expected requirements for the final waste forms. Results of AFCF operations will advance new technologies that will contribute to safe and economical commercial spent fuel reprocessing facilities needed to meet the GNEP vision. As conceptual design work and research and design continues, the waste management strategies for the AFCF are expected to also evolve. (authors)

  8. 75 FR 36449 - Yankee Atomic Electric Co.; Yankee Atomic Independent Spent Fuel Storage Installation; Issuance...

    Science.gov (United States)

    2010-06-25

    ... COMMISSION Yankee Atomic Electric Co.; Yankee Atomic Independent Spent Fuel Storage Installation; Issuance of... impact. FOR FURTHER INFORMATION CONTACT: John Goshen, Project Manager, Division of Spent Fuel Storage and... an independent spent fuel storage installation (ISFSI) associated with the decommissioned...

  9. 75 FR 9452 - Solicitation of Topics for Discussion at a Spent Fuel Storage and Transportation Licensing...

    Science.gov (United States)

    2010-03-02

    ... COMMISSION Solicitation of Topics for Discussion at a Spent Fuel Storage and Transportation Licensing... Spent Fuel Storage and Transportation Licensing Conference. SUMMARY: The U.S. Nuclear Regulatory... entitled, ``Spent Fuel Storage and Transportation Licensing Conference.'' The purpose of the...

  10. Spent fuel management options and nuclear fuel supplies in Germany

    International Nuclear Information System (INIS)

    The spent fuel management pathway adopted has a direct bearing on the supply of nuclear fuel. Compared to direct disposal, reprocessing is able to reduce the consumption of uranium, thus making nuclear power a quasi-indigenous source of power. The breeder technology was developed to make use of as many fuel constituents of natural uranium as possible, especially Pu-239. When used in mixed oxide fuel assemblies, plutonium can be burnt even in light water reactors. On the basis of three different scenarios for the development of the installed nuclear generating capacity, the annual uranium requirement up to 2030 is simulated in a computer model. The parameters influencing the calculation are the time, final storage, reprocessing, the use of mixed oxide fuel, and a higher fuel burnup. The service life of a nuclear power plant is assumed to be 35 years throughout. All steps of the nuclear fuel cycle are modeled, from purchasing the natural uranium to final storage. In each of the three scenarios, the model calculations arrive at clearly lower prices of natural uranium, of approx. US Dollar 65/kg of U, than actually prevailed in the second half of the seventies, i.e. more than US Dollar 190/kg of U. (orig.)

  11. Status of spent fuels in Japanese research reactors

    International Nuclear Information System (INIS)

    There are now eleven research and test reactors in operation in Japan. Spent fuel issues might cause problems at the JRR-3M and JMTR reactors in the near future. Increasing the number of spent fuel racks at these reactors is now under consideration because the existing capacity is almost filled. The commissioning of extra racks will allow space for the normal discharge of fuel from these reactors for several more years. The current management of spent fuel from the eleven operational reactors is suitable to meet their needs. (author). 3 tabs

  12. Spent fuel data base: commercial light water reactors

    International Nuclear Information System (INIS)

    As a consequence of this country's non-proliferation policy, the reprocessing of spent nuclear fuel has been delayed indefinitely. This has resulted in spent light water reactor (LWR) fuel being considered as a potential waste form for disposal. Since the Nuclear Regulatory Commission (NRC) is currently developing methodologies for use in the regulation of the management and disposal of high-level and transuranic wastes, a comprehensive data base describing LWR fuel technology must be compiled. This document provides that technology baseline and, as such, will support the development of those evaluation standards and criteria applicable to spent nuclear fuel

  13. Volatile fission product distributions in LWR spent fuel rods

    International Nuclear Information System (INIS)

    Results presented are from spent fuel characterizations being conducted by the Materials Characterization Center at Pacific Northwest Laboratory on a variety of spent commercial power reactor fuels designated as approved testing materials (ATMs). These ATMs have a variety of burnup levels and fission gas releases; they include fuel from both pressurized water and boiling water reactor designs. The purpose of this work is to provide a source of well-characterized spent fuel for testing in the U.S. Department of Energy Office of Civilian Radioactive Waste Management repository programs and, potentially, other programs

  14. MTR radiological database for SRS spent nuclear fuel facilities

    International Nuclear Information System (INIS)

    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

  15. Advanced fuel cycles in CANDU reactors

    International Nuclear Information System (INIS)

    This paper re-examines the rationale for advanced nuclear fuel cycles in general, and for CANDU advanced fuel cycles in particular. The traditional resource-related arguments for more uranium nuclear fuel cycles are currently clouded by record-low prices for uranium. However, the total known conventional uranium resources can support projected uranium requirements for only another 50 years or so, less if a major revival of the nuclear option occurs as part of the solution to the world's environmental problems. While the extent of the uranium resource in the earth's crust and oceans is very large, uncertainty in the availability and price of uranium is the prime resource-related motivation for advanced fuel cycles. There are other important reasons for pursuing advanced fuel cycles. The three R's of the environmental movement, reduce, recycle, reuse, can be achieved in nuclear energy production through the employment of advanced fuel cycles. The adoption of more uranium-conserving fuel cycles would reduce the amount of uranium which needs to be mined, and the environmental impact of that mining. Environmental concerns over the back end of the fuel cycle can be mitigated as well. Higher fuel burnup reduces the volume of spent fuels which needs to be disposed of. The transmutation of actinides and long-lived fission products into short-lived fission products would reduce the radiological hazard of the waste from thousands to hundreds of years. Recycling of uranium and/or plutonium in spent fuel reuses valuable fissile material, leaving only true waste to be disposed of. Advanced fuel cycles have an economical benefit as well, enabling a ceiling to be put on fuel cycle costs, which are

  16. A method for spent fuel cladding leaktightness assessment based on SCALE 4.4a calculations

    International Nuclear Information System (INIS)

    Experience with advanced spent nuclear fuel management in Slovakia is described. The evaluation and monitoring procedures are based on approaches practised by the Slovak wet interim spent fuel storage facility at the Jaslovske Bohunice NPP. Since 1999, leak testing of WWER-440 fuel assemblies has been provided by a special leak detection system known as 'Sipping in Pool'. In 2006 a new inspection stand called 'SVYP-440' for monitoring the spent nuclear fuel condition was deployed. This stand makes it possible to open WWER-440 fuel assemblies, seek for leaking fuel pins, examine them and store them under hermetic conditions. Optimal ways of spent fuel disposal and monitoring of nuclear fuel condition were designed. Applying an appropriate degree of conservativeness, the authors introduce a new factor for specifying the level of spent fuel leaktightness. A suitable combination of computer simulations (based on the SCALE 4.4a code) and Sipping in Pool measurements is used to introduce limiting values for the fuel cladding leaktightness factor. (orig.)

  17. Packaging of spent fuel elements into special containers

    International Nuclear Information System (INIS)

    This report contains detailed description of the procedure for packaging the spent fuel elements from the fuel channels into the special steel containers. The previously cooled fuel elements are packaged into containers by the existing crane and transported later into the spen fuel storage. Instructions for crane operation are included

  18. Methods for expanding the capacity of spent fuel storage facilities

    International Nuclear Information System (INIS)

    At the beginning of 1989 more than 55,000 metric tonnes of heavy metal (MTHM) of spent Light Water Reactor (LWR) and Heavy Water Reactor (HWR) fuel had been discharged worldwide from nuclear power plants. Only a small fraction of this fuel has been reprocessed. The majority of the spent fuel assemblies are currently held at-reactor (AR) or away-from-reactor (AFR) in storage awaiting either chemical processing or final disposal depending on the fuel concept chosen by individual countries. Studies made by NEA and IAEA have projected that annual spent fuel arising will reach about 10,000 t HM in the year 2000 and cumulative arising will be more than 200,000 t HM. Taking into account the large quantity of spent fuel discharged from NPP and that the first demonstrations of the direct disposal of spent fuel or HLW are expected only after the year 2020, long-term storage will be the primary option for management of spent fuel until well into the next century. There are several options to expand storage capacity: (1) to construct new away-from-reactor storage facilities, (2) to transport spent fuel from a full at-reactor pool to another site for storage in a pool that has sufficient space to accommodate it, (3) to expand the capacity of existing AR pools by using compact racks, double-tierce, rod consolidation and by increasing the dimensions of existing pools. The purpose of the meeting was: to exchange new information on the international level on the subject connected with the expansion of storage capacities for spent fuel; to elaborate the state-of-the-art of this problem; to define the most important areas for future activity; on the basis of the above information to give recommendations to potential users for selection and application of the most suitable methods for expanding spent fuel facilities taking into account the relevant country's conditions. Refs, figs and tabs

  19. Safety and Licensing of Spent Fuel Storage Facilities

    International Nuclear Information System (INIS)

    All operating nuclear power reactors in the United States (U.S.) are storing spent fuel on-site in spent fuel pools licensed by the U.S. Nuclear Regulatory Commission (NRC). Spent fuel pools at U.S. reactors were not designed to store the full amount of spent fuel generated during the lifetime of plant operation. Consequently, most utilities expanded their storage capacity by the use of high-density storage racks. Even with the high density racks, most utilities will need additional storage capacity. When it became apparent that nuclear power plants were going to need additional spent fuel storage space, the NRC amended its regulations in 1980 to provide nuclear power plants with alternate spent fuel storage in an independent spent fuel storage installation (ISFSI). NRC provides for a 20-year specific license with the option to renew the license for additional 20-year terms. In 1990, the NRC implemented the General License option to ease the burden on nuclear power plants that have a license to either operate or possess fuel. The general license for each storage cask terminates 20 years after the storage cask is first used by the licensee. The first storage cask using a general license was loaded in 1994. This paper discusses NRC experiences and its knowledge gained in licensing over the past 30 years and renewing the licenses for three ISFSIs and how this knowledge has driven the NRC to revise its guidance and thought processes for dry storage. (author)

  20. Thermal Analysis of CANDU Spent Fuel Bay Cooling System

    International Nuclear Information System (INIS)

    The spent fuel bay cooling and purification system for Wolsong Nuclear Power Plant (NPP) Units 2, 3 and 4 was designed to remove heat from the spent fuel bay generated by 10 years accumulation of spent fuel at an 80% capacity factor refueling rate plus an emergency discharge of one-half the core fuel inventory over a 20-day period for 25.5 .deg. C of the cooling sea water temperature. The heat load in the spent fuel bay depends on the capacity factor refueling rate and the amount of spent fuel accumulated at the spent fuel bay. An 80% capacity factor refueling rate was considered as a design condition, but the highest capacity factor refueling rate of 93.75% for Wolsong NPPs was calculated based on nine (9) years of operating experience from 2000 to 2008. For the abnormal operating condition, the operating temperature of spent fuel bay does not meet with the acceptance criterion of 49 .deg. C for the conditions of the capacity factor refueling rate of 93.75%. These operating modes are not recommended for the abnormal operating condition

  1. Thermal Analysis of CANDU Spent Fuel Bay Cooling System

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jeong Mann; Jang, Ho Cheol; Jang, Jin A.; Kim, Eun Kee [KEPCO Engineering and Construction Company, Daejeon (Korea, Republic of); Park, WanGyu [KHNP, Uljingun (Korea, Republic of)

    2015-05-15

    The spent fuel bay cooling and purification system for Wolsong Nuclear Power Plant (NPP) Units 2, 3 and 4 was designed to remove heat from the spent fuel bay generated by 10 years accumulation of spent fuel at an 80% capacity factor refueling rate plus an emergency discharge of one-half the core fuel inventory over a 20-day period for 25.5 .deg. C of the cooling sea water temperature. The heat load in the spent fuel bay depends on the capacity factor refueling rate and the amount of spent fuel accumulated at the spent fuel bay. An 80% capacity factor refueling rate was considered as a design condition, but the highest capacity factor refueling rate of 93.75% for Wolsong NPPs was calculated based on nine (9) years of operating experience from 2000 to 2008. For the abnormal operating condition, the operating temperature of spent fuel bay does not meet with the acceptance criterion of 49 .deg. C for the conditions of the capacity factor refueling rate of 93.75%. These operating modes are not recommended for the abnormal operating condition.

  2. International status of dry storage of spent fuels

    International Nuclear Information System (INIS)

    Spent fuel from the world's nuclear power reactors, or the high-level radioactive wastes from reprocessing of the spent fuels, are planned to be disposed of in national deep geological repositories in the respective countries of origin. The plans for most countries with nuclear power call for spent fuel or high-level waste disposal to start between 2010 and about 2050. Although storage in water pools is the primary method for management of spent nuclear fuels for the first few years after discharge from the reactor, dry storage has been implemented in several countries and is being considered in others. Dry storage is generally planned for an interim period (from 10 to as long as 100 years) until the spent fuel is disposed of or until a final decision is made on reprocessing. Dry storage is also being used to supplement wet storage capacity at some nuclear power stations. This paper summarizes the world-wide status of dry spent fuel storage and information on the expected long-term integrity of the dry-stored spent fuel based on experience, particularly for Zircaloy-clad fuels. The paper also addresses briefly the dry storage of solidified high-level radioactive wastes. This paper is based on work carried out for the US Department of Energy (DOE) by the Pacific Northwest Laboratory

  3. Shippingport Spent Fuel Canister System Description

    Energy Technology Data Exchange (ETDEWEB)

    JOHNSON, D.M.

    2000-03-27

    In 1978 and 1979, a total of 72 blanket fuel assemblies (BFAs), irradiated during the operating cycles of the Shippingport Atomic Power Station's Pressurized Water Reactor (PWR) Core 2 from April 1965 to February 1974, were transferred to the Hanford Site and stored in underwater storage racks in Cell 2R at the 221-T Canyon (T-Plant). The initial objective was to recover the produced plutonium in the BFAs, but this never occurred and the fuel assemblies have remained within the water storage pool to the present time. The Shippingport Spent Fuel Canister (SSFC) is a confinement system that provides safe transport functions (in conjunction with the TN-WHC cask) and storage for the BFAs at the Canister Storage Building (CSB). The current plan is for these BFAs to be retrieved from wet storage and loaded into SSFCs for dry storage. The sealed SSFCs containing BFAs will be vacuum dried, internally backfilled with helium, and leak tested to provide suitable confinement for the BFAs during transport and storage. Following completion of the drying and inerting process, the SSFCs are to be delivered to the CSB for closure welding and long-term interim storage. The CSB will provide safe handling and dry storage for the SSFCs containing the BFAs. The purpose of this document is to describe the SSFC system and interface equipment, including the technical basis for the system, design descriptions, and operations requirements. It is intended that this document will be periodically updated as more equipment design and performance specification information becomes available.

  4. Shippingport Spent Fuel Canister System Description

    International Nuclear Information System (INIS)

    In 1978 and 1979, a total of 72 blanket fuel assemblies (BFAs), irradiated during the operating cycles of the Shippingport Atomic Power Station's Pressurized Water Reactor (PWR) Core 2 from April 1965 to February 1974, were transferred to the Hanford Site and stored in underwater storage racks in Cell 2R at the 221-T Canyon (T-Plant). The initial objective was to recover the produced plutonium in the BFAs, but this never occurred and the fuel assemblies have remained within the water storage pool to the present time. The Shippingport Spent Fuel Canister (SSFC) is a confinement system that provides safe transport functions (in conjunction with the TN-WHC cask) and storage for the BFAs at the Canister Storage Building (CSB). The current plan is for these BFAs to be retrieved from wet storage and loaded into SSFCs for dry storage. The sealed SSFCs containing BFAs will be vacuum dried, internally backfilled with helium, and leak tested to provide suitable confinement for the BFAs during transport and storage. Following completion of the drying and inerting process, the SSFCs are to be delivered to the CSB for closure welding and long-term interim storage. The CSB will provide safe handling and dry storage for the SSFCs containing the BFAs. The purpose of this document is to describe the SSFC system and interface equipment, including the technical basis for the system, design descriptions, and operations requirements. It is intended that this document will be periodically updated as more equipment design and performance specification information becomes available

  5. Thermal Cooling Limits of Sabotaged Spent Fuel Pools

    International Nuclear Information System (INIS)

    To develop the understanding and predictive measures of the post 'loss of water inventory' hazardous conditions as a result of the natural and/or terrorist acts to the spent fuel pool of a nuclear plant. This includes the thermal cooling limits to the spent fuel assembly (before the onset of the zircaloy ignition and combustion), and the ignition, combustion, and the subsequent propagation of zircaloy fire from one fuel assembly to others.

  6. Spent nuclear fuel discharges from U.S. reactors 1994

    International Nuclear Information System (INIS)

    Spent Nuclear Fuel Discharges from US Reactors 1994 provides current statistical data on fuel assemblies irradiated at commercial nuclear reactors operating in the US. This year's report provides data on the current inventories and storage capacities at these reactors. Detailed statistics on the data are presented in four chapters that highlight 1994 spent fuel discharges, storage capacities and inventories, canister and nonfuel component data, and assembly characteristics. Five appendices, a glossary, and bibliography are also included. 10 figs., 34 tabs

  7. Thermal Cooling Limits of Sbotaged Spent Fuel Pools

    Energy Technology Data Exchange (ETDEWEB)

    Dr. Thomas G. Hughes; Dr. Thomas F. Lin

    2010-09-10

    To develop the understanding and predictive measures of the post “loss of water inventory” hazardous conditions as a result of the natural and/or terrorist acts to the spent fuel pool of a nuclear plant. This includes the thermal cooling limits to the spent fuel assembly (before the onset of the zircaloy ignition and combustion), and the ignition, combustion, and the subsequent propagation of zircaloy fire from one fuel assembly to others

  8. Economics of National Waste Terminal Storage Spent Fuel Pricing Study

    International Nuclear Information System (INIS)

    The methodology for equitably pricing commercial nuclear spent fuel management is developed, and the results of four sample calculations are presented. The spent fuel management program analyzed places encapsulated spent fuel in bedded salt while maintaining long-term retrievability. System design was reasonable but not optimum. When required, privately-owned Away From Reactor (AFR) storage is provided and the spent fuel placed in AFR storage is eventually transported to final storage. Applicable Research and Development and Government Overhead are included. The cost of each component by year was estimated from the most recent applicable data source available. These costs were input to the pricing methodology to establish a one-time charge whose present value exactly recovered the present value of the expenditure flow. The four cases exercised were combinations of a high and a low quantity of spent fuel managed, with a single repository (venture) or a multiple repository (campaign) approach to system financial structure. The price for spent fuel management calculated ranged from 116 to 152 dollars (1978) per kilogram charged initially to the reactor. The effect of spent fuel receiving rate on price is illustrated by the fact that the extremes of price did not coincide with the cases having the extremes of undiscounted cost. These prices for spent fuel management are comparable in magnitude to other fuel cycle costs. The range of variation is small because of compensating effects, i.e., additional costs for high early deliveries (AFR and transportation) versus lower present value of future revenue for later delivery cases. The methodology contains numerous conservative assumptions, provisions for contingencies, and covers the complete set of spent fuel management expenses

  9. US Industry Spent Fuel Management (R&D and Operating Experience)

    International Nuclear Information System (INIS)

    In the USA, it is highly likely that commercial spent nuclear fuel will need to be stored at the reactor sites for many decades until centralized interim storage or final disposal becomes available. Given the economic and worker dose involved in transferring spent fuel from wet (spent fuel pools) to dry storage, it would be desirable to continue to store and eventually ship spent fuel in the original dry storage systems rather than incurring the additional economic and radiological penalties associated with repackaging. Currently available information on long term ageing have allowed several utilities to obtain license extensions for their existing dry storage installations for up to 60 years after initial fuel loading. Given the uncertainty about the eventual readiness of options for ultimate disposal, it may be necessary to store spent fuel at reactor sites for periods well beyond 60 years. In addition, risk information assessment of criticality risks during transportation, including improvements in criticality methods (burnup credit), are being developed to provide more rational bases for future improvements in operations and regulations. Objective: 1. To identify the needs for enhancing existing technical bases for extended storage and transportation of commercial spent nuclear fuel. 2. To initiate work on high-priority ageing management topics. 3. To advance the state of the art in assessing criticality safety for wet storage and transportation applications

  10. New Methods for Evaluation of Spent Fuel Condition during Long-Term Storage in Slovakia

    Directory of Open Access Journals (Sweden)

    M. Mikloš

    2009-01-01

    Full Text Available Experiences with an advanced spent nuclear fuel management in Slovakia are presented in this paper. The evaluation and monitoring procedures are based on practices at the Slovak wet interim spent fuel storage facility in NPP Jaslovské Bohunice. Since 1999, leak testing of WWER-440 fuel assemblies are provided by special leak tightness detection system “Sipping in pool” delivered by Framatomeanp with external heating for the precise defects determination. In 2006, a new inspection stand “SVYP-440” for monitoring of spent nuclear fuel condition was inserted. This stand has the possibility to open WWER-440 fuel assemblies and examine fuel elements. Optimal ways of spent fuel disposal and monitoring of nuclear fuel condition were designed. With appropriate approach of conservativeness, new factor for specifying spent fuel leak tightness is introduced in the paper. By using computer simulations (based on SCALE 4.4a code for fission products creation and measurements by system “Sipping in pool,” the limit values of leak tightness were established.

  11. Status of dry storage of spent nuclear fuel

    International Nuclear Information System (INIS)

    Spent-fuel storage has been identified as the key element of spent-fuel management. Over 45,000 t of spent water reactor fuel has been discharged worldwide, of which only ∼7% has been reprocessed. Estimates by the Organization for Economic Cooperation and Development and the International Atomic Energy Agency (IAEA) indicate that the amount of spent fuel being generated will increase significantly. About 200,000 t of heavy metal of spent fuel could be accumulated by the year 2000. Many countries are involved in the development of new ways, including dry storage, for handling and storing the spent fuel. These new technologies require new reviews and perhaps new approaches for domestic and international safeguards. The IAEA has been involved in surveying worldwide experience related to the storage of spent fuel. This paper summarizes the efforts of an international working group to survey the experience with dry storage and related innovations that might have an impact on safeguarding procedures in the future

  12. Spent Fuel Transportation Package Performance Study - Experimental Design Challenges

    International Nuclear Information System (INIS)

    Numerous studies of spent nuclear fuel transportation accident risks have been performed since the late seventies that considered shipping container design and performance. Based in part on these studies, NRC has concluded that the level of protection provided by spent nuclear fuel transportation package designs under accident conditions is adequate. [1] Furthermore, actual spent nuclear fuel transport experience showcase a safety record that is exceptional and unparalleled when compared to other hazardous materials transportation shipments. There has never been a known or suspected release of the radioactive contents from an NRC-certified spent nuclear fuel cask as a result of a transportation accident. In 1999 the United States Nuclear Regulatory Commission (NRC) initiated a study, the Package Performance Study, to demonstrate the performance of spent fuel and spent fuel packages during severe transportation accidents. NRC is not studying or testing its current regulations, a s the rigorous regulatory accident conditions specified in 10 CFR Part 71 are adequate to ensure safe packaging and use. As part of this study, NRC currently plans on using detailed modeling followed by experimental testing to increase public confidence in the safety of spent nuclear fuel shipments. One of the aspects of this confirmatory research study is the commitment to solicit and consider public comment during the scoping phase and experimental design planning phase of this research

  13. Loss of spent fuel pool cooling PRA: Model and results

    International Nuclear Information System (INIS)

    This letter report documents models for quantifying the likelihood of loss of spent fuel pool cooling; models for identifying post-boiling scenarios that lead to core damage; qualitative and quantitative results generated for a selected plant that account for plant design and operational practices; a comparison of these results and those generated from earlier studies; and a review of available data on spent fuel pool accidents. The results of this study show that for a representative two-unit boiling water reactor, the annual probability of spent fuel pool boiling is 5 x 10-5 and the annual probability of flooding associated with loss of spent fuel pool cooling scenarios is 1 x 10-3. Qualitative arguments are provided to show that the likelihood of core damage due to spent fuel pool boiling accidents is low for most US commercial nuclear power plants. It is also shown that, depending on the design characteristics of a given plant, the likelihood of either: (a) core damage due to spent fuel pool-associated flooding, or (b) spent fuel damage due to pool dryout, may not be negligible

  14. Preliminary design and analysis on nuclear fuel cycle for fission-fusion hybrid spent fuel burner

    International Nuclear Information System (INIS)

    A wet-processing-based fuel cycle and a dry-processing were designed for a fission-fusion hybrid spent fuel burner (FDS-SFB). Mass flow of SFB was preliminarily analyzed. The feasibility analysis of initial loaded fuel inventory, recycle fuel fabrication and spent fuel reprocessing were preliminarily evaluated. The results of mass flow of FDS-SFB demonstrated that the initial loaded fuel inventory, recycle fuel fabrication and spent fuel reprocessing of nuclear fuel cycle of FDS-SFB is preliminarily feasible. (authors)

  15. Remote technology in the spent fuel route in the UK

    International Nuclear Information System (INIS)

    Remote technologies employed in front end (commercial) reprocessing operations of metallic and oxide fuel at Sellafield in the UK are described. An overview of the transportation, fuel receiving and preparation facilities are given together with the remote technology developments employed to improve operations. It is concluded that the facilities and remote technology used within them are mature and based upon simple and robust principles. Remote operations and maintenance in these facilities is often easier than in many facilities downstream of the dissolution stage. Fuel design considerations for shearing and handling are described and it is concluded that advanced and higher burnup fuel can be accommodated by existing reprocessing and interim storage routes with current remote technologies. Two different storage systems are available from UK companies which use existing spent fuel handling technology/equipment. The pace of remote technology development is currently being set by the demands of other nuclear process areas such as decommissioning and plant clean out; these will spin-off into front end processes. (author)

  16. China's spent nuclear fuel management: Current practices and future strategies

    International Nuclear Information System (INIS)

    Although China's nuclear power industry is relatively young and the management of its spent nuclear fuel is not yet a concern, China's commitment to nuclear energy and its rapid pace of development require detailed analyses of its future spent fuel management policies. The purpose of this study is to provide an overview of China's fuel cycle program and its reprocessing policy, and to suggest strategies for managing its future fuel cycle program. The study is broken into four sections. The first reviews China's current nuclear fuel cycle program and facilities. The second discusses China's current spent fuel management methods and the storage capability of China's 13 operational nuclear power plants. The third estimates China's total accumulated spent fuel, its required spent fuel storage from present day until 2035, when China expects its first commercialized fast neutron reactors to be operational, and its likely demand for uranium resources. The fourth examines several spent fuel management scenarios for the present period up until 2035; the financial cost and proliferation risk of each scenario is evaluated. The study concludes that China can and should maintain a reprocessing operation to meet its R and D activities before its fast reactor program is further developed. - Highlights: → This study provides an overview of China's fuel cycle program and its reprocessing policy.→ This study suggests strategies for managing its future fuel cycle program.→ China will experience no pressure to lessen the burden of spent fuel storage in the next 30 years.→ China should maintain sufficient reprocessing operations to meet its demands for R and D activities.→ China should actively invest on R and D activities of both fuel cycling and fast reactor programs.

  17. Management of Spent Nuclear Fuel from Nuclear Power Plant Reactor

    International Nuclear Information System (INIS)

    Management of spent nuclear fuel from Nuclear Power Plant (NPP) reactor had been studied to anticipate program of NPP operation in Indonesia. In this paper the quantity of generated spent nuclear fuel (SNF) is predicted based on the national electrical demand, power grade and type of reactor. Data was estimated using Pressurized Water Reactor (PWR) NPP type 1.000 MWe and the SNF management overview base on the experiences of some countries that have NPP. There are four strategy nuclear fuel cycle which can be developed i.e: direct disposal, reprocessing, DUPlC (Direct Use of Spent PWR Fuel In Candu) and wait and see. There are four alternative for SNF management i.e : storage at the reactor building (AR), away from reactor (AFR) using wet centralized storage, dry centralized storage AFR and prepare for reprocessing facility. For the Indonesian case, centralized facility of the wet type is recommended for PWR or BWR spent fuel. (author)

  18. Application of ALARA principles to shipment of spent nuclear fuel

    International Nuclear Information System (INIS)

    The public exposure from spent fuel shipment is very low. In view of this low exposure and the perfect safety record for spent fuel shipment, existing systems can be considered satisfactory. On the other hand, occupational exposure reduction merits consideration and technology improvement to decrease dose should concentrate on this exposure. Practices that affect the age of spent fuel in shipment and the number of times the fuel must be shipped prior to disposal have the largest impact. A policy to encourage a 5-year spent fuel cooling period prior to shipment coupled with appropriate cask redesign to accommodate larger loads would be consistent with ALARA and economic principles. And finally, bypassing high population density areas will not in general reduce shipment dose

  19. Nuclear reactor spent fuel valuation: procedure, applications, and analysis

    International Nuclear Information System (INIS)

    A preliminary approach that values nuclear reactor generated spent fuel is developed and applied in this report. There is no intent to assess the merits of reprocessing and recycling but rather to outline a procedure that may provide a basis for international negotiations on nation-to-nation spent fuel transfer. The valuation procedure described estimates the net present discounted value (PDV) of the benefit incurred when the reprocessed plutonium and uranium contained in the spent fuel are recycled, less the PDV of the stream of costs associated with transporting, storing, and reprocessing the spent fuel and fabricating, storing, and safeguarding the new mixed oxide fuel (MOX). The parameters that affect the net PDV most strongly are the discount rate, yellowcake prices, reprocessing costs, timing of recycle or disposal, and, to a lesser extent, enrichment costs

  20. Behaviour of spent fuel assemblies during extended storage

    International Nuclear Information System (INIS)

    This report is the final report of the IAEA Co-ordinated Research Programme on Behaviour of Spent Fuel Assemblies During Extended Storage (BEFAST, Phase I, 1981-86). It contains the results on wet and dry spent fuel storage technologies obtained from 11 institutes (10 countries: Austria, Canada, Czechoslovakia, Finland, German Democratic Republic, Hungary, Japan, Sweden, USA and USSR) participating in the BEFAST CRP during the time period 1981-86. Names of participating institutes and chief investigators are given. The interim spent fuel storage has been recognized as an important independent step in the nuclear fuel cycle. Due to the delay in commercial reprocessing of spent fuel in some cases it should be stored up to 30-50 years or more before reprocessing or final disposal. This programme was evaluated by all its participants and observers as very important and helpful for the nuclear community and it was decided to continue it further (1986-91) as BEFAST, Phase II

  1. Long-term leaching of irradiated spent fuel

    International Nuclear Information System (INIS)

    Spent Light Water Reactor (LWR) fuel with burnups of 9, 28 and 54 MWd/kg U were leach tested at 250C in deionized water in a Paige apparatus. No discernible differences in leach rates were observed due to burnup. Additionally, the 28 MWd/kg U fuel was IAEA leach tested in five different leachants using the IAEA method. Deionized water gave the highest leach rates and a calcium chloride solution gave the lowest leach rates. An accelerated leaching period was observed during the Paige leach test of the 54 MWd/kg U spent fuel. Comparison between spent fuel and borosilicate waste glass leach rates was made. In sodium bicarbonate solution the leach rates are near equal and the glass becomes increasingly more durable with CaCl2 solution, followed by sodium chloride solution, WIPP B brine and deionized water where the glass is two to three orders of magnitude more leach resistant than spent fuel. 16 figures

  2. Existing Condition Analysis of Dry Spent Fuel Storage Technology

    Institute of Scientific and Technical Information of China (English)

    LI Ning; XU Lan; HAO Jian-sheng

    2016-01-01

    As in some domestic nuclear power plants, spent fuel pools near capacity, away-from-reactor type storage should be arranged to transfer spent fuel before the pool capacity is full and the plants can operate in safety. This study compares the features of wet and dry storage technology, analyzes the actualities of foreign dry storage facilities and then introduces structural characteristics of some foreign dry storage cask. Finally, a glance will be cast on the failure of away-from-reactor storage facilities of Pressurized Water Reactor(PWR)to meet the need of spent-fuel storage. Therefore, this study believes dry storage will be a feasible solution to the problem.

  3. Dry spent fuel storage in the 1990's

    International Nuclear Information System (INIS)

    In the US, for the decade of the 1990's, at-reactor-site dry spent fuel storage has become the predominant option outside of reactor spent fuel pools. This development has resulted from failure, in the 1980's, of a viable reprocessing option for commercial power reactors, and delay in geologic repository development to an operational date at or beyond the year 2010. Concurrently, throughout the 1980's, aggressive technical and regulatory efforts by the Federal Government, coordinated with nuclear industry, have led to successful evolution of dry spent fuel storage as a utility option

  4. Spent fuel test project, Climax granitic stock, Nevada Test Site

    International Nuclear Information System (INIS)

    The Spent Fuel Test-Climax (SFT-C) is a test of dry geologic storage of spent nuclear reactor fuel. The SFT-C is located at a depth of 420 m in the Climax granitic stock at the Nevada Test Site. Eleven canisters of spent commercial PWR fuel assemblies are to be stored for 3 to 5 years. Additional heat is supplied by electrical heaters, and more than 800 channels of technical information are being recorded. The measurements include rock temperature, rock displacement and stress, joint motion, and monitoring of the ventilation air volume, temperature, and dewpoint

  5. Transportation 2000. Spent fuel transportation trends in the new millenium

    International Nuclear Information System (INIS)

    The paper will provide a comparison of foreign research reactor spent fuel transportation today verses the assumptions used by the Department of Energy in the Environmental Impact Statement. In addition, it will suggest changes that are likely to occur in transportation logistics through the remainder of the U.S. spent fuel returns program. Cask availability, certification status, shipment strategy, cost issues, and public acceptance are among the topical areas that will be examined. Transportation requirements will be assessed in light of current participation in the returns program and the tendency for shipment plans to shift toward spent fuel return toward the end of the 13 year period of eligibility. (author)

  6. The united kingdom's changing requirements for spent fuel storage

    International Nuclear Information System (INIS)

    The UK is adopting an open fuel cycle, and is necessarily moving to a regime of long term storage of spent fuel, followed by geological disposal once a geological disposal facility (GDF) is available. The earliest GDF receipt date for legacy spent fuel is assumed to be 2075. The UK is set to embark on a programme of new nuclear build to maintain a nuclear energy contribution of 16 GW. Additionally, the UK are considering a significant expansion of nuclear energy in order to meet carbon reduction targets and it is plausible to foresee a scenario where up to 75 GW from nuclear power production could be deployed in the UK by the mid 21. century. Such an expansion, could lead to spent fuel storage and its disposal being a dominant issue for the UK Government, the utilities and the public. If the UK were to transition a closed fuel cycle, then spent fuel storage should become less onerous depending on the timescales. The UK has demonstrated a preference for wet storage of spent fuel on an interim basis. The UK has adopted an approach of centralised storage, but a 16 GW new build programme and any significant expansion of this may push the UK towards distributed spent fuel storage at a number of reactors station sites across the UK

  7. Licensing of spent fuel dry storage and consolidated rod storage

    International Nuclear Information System (INIS)

    The results of this study, performed by Pacific Northwest Laboratory (PNL) and sponsored by the US Department of Energy (DOE), respond to the nuclear industry's recommendation that a report be prepared that collects and describes the licensing issues (and their resolutions) that confront a new applicant requesting approval from the US Nuclear Regulatory Commission (NRC) for dry storage of spent fuel or for large-scale storage of consolidated spent fuel rods in pools. The issues are identified in comments, questions, and requests from the NRC during its review of applicants' submittals. Included in the report are discussions of (1) the 18 topical reports on cask and module designs for dry storage fuel that have been submitted to the NRC, (2) the three license applications for dry storage of spent fuel at independent spent fuel storage installations (ISFSIs) that have been submitted to the NRC, and (3) the three applications (one of which was later withdrawn) for large-scale storage of consolidated fuel rods in existing spent fuel storage pools at reactors that were submitted tot he NRC. For each of the applications submitted, examples of some of the issues (and suggestions for their resolutions) are described. The issues and their resolutions are also covered in detail in an example in each of the three subject areas: (1) the application for the CASTOR V/21 dry spent fuel storage cask, (2) the application for the ISFSI for dry storage of spent fuel at Surry, and (3) the application for full-scale wet storage of consolidated spent fuel at Millstone-2. The conclusions in the report include examples of major issues that applicants have encountered. Recommendations for future applicants to follow are listed. 401 refs., 26 tabs

  8. Investigation of potential utilization for light water reactor spent fuel and thorium fuel in ACR-700

    International Nuclear Information System (INIS)

    The potential utilization of light water reactor (LWR) spent fuel and thorium fuel is a very efficient way for solving spent fuel processing and nuclear fuel shortage problems. Four different mixed fuel types with LWR spent fuel and Th fuel were researched in ACR-700 using SCALE5.1 code system. Compared with the ACR-700 standard fuel, the fuel cycle model 1 and 4 can reach higher burnup. According to the research on the lattice physics of ACR-700 fuel bundle, it is indicated that LWR spent fuel after separation of the fission products is a very prospective fuel in ACR-700 and 232Th can be a preferable fertile isotope if sufficient booster fuel is supplied. (authors)

  9. Thermal hydraulic feasibility assessment for the Spent Nuclear Fuel Project

    International Nuclear Information System (INIS)

    A series of scoping analyses have been completed investigating the thermal-hydraulic performance and feasibility of the Spent Nuclear Fuel Project (SNFP) Integrated Process Strategy (IPS). The SNFP was established to develop engineered solutions for the expedited removal, stabilization, and storage of spent nuclear fuel from the K Basins at the U.S. Department of Energy's Hanford Site in Richland, Washington. The subject efforts focused on independently investigating, quantifying, and establishing the governing heat production and removal mechanisms for each of the IPS operations and configurations, obtaining preliminary results for comparison with and verification of other analyses, and providing technology-based recommendations for consideration and incorporation into the design bases for the SNFP. The goal was to develop a series fo thermal-hydraulic models that could respond to all process and safety-related issues that may arise pertaining to the SNFP. A series of sensitivity analyses were also performed to help identify those parameters that have the greatest impact on energy transfer and hence, temperature control. It is anticipated that the subject thermal-hydraulic models will form the basis for a series of advanced and more detailed models that will more accurately reflect the thermal performance of the IPS and alleviate the necessity for some of the more conservative assumptions and oversimplifications, as well as form the basis for the final process and safety analyses

  10. Transporting spent reactor fuel: allegations and responses

    International Nuclear Information System (INIS)

    A January 1982 monthly newsletter from the Council on Economic Priorities (CEP) was entirely devoted to the presentation of a broad-ranging series of allegations that the transportation of spent fuel in particular, and other high-level radioactive materials by inference is currently being conducted in this country in an unsafe manner. This newsletter preceded the release of a book authored by Marvin Resnikoff on the same subject by over a year. This book titled The Next Nuclear Gamble contained substantially the same allegations as the newsletter, although the book devoted space to a greatly increased number of specific examples. This paper reduces those allegations contained in the executive summary and the recommendations contained in the last chapter of the book to a manageable number by combining the many specific issues into a few topics. Each of these topics is then addressed. As such, this is an abbreviated analysis of The Next Nuclear Gamble and does not address much of the fine detail. In spite of that, it would be possible to address each of the details within the book on a similar basis. The intent of this document is to provide background information for those who are questioned on the validity of the allegations made by the CEp

  11. The management status of the spent fuel in HANARO(1995-2009)

    International Nuclear Information System (INIS)

    In HANARO, the spent fuels are stored in the spent fuel storage pool of the reactor hall. The capacity of the spent fuel storage pool was designed to store 600 bundles for 36 rods fuel, 432 bundles for 18 rods fuel, 315 rods for TRIGA reactor fuel and the fuels loaded in the reactor core. The spent fuel storage pool can store spent fuels discharged from the reactor core for 20 years normal operation. As for July 2009, the spent fuel 337 bundles are stored in the spent fuel storage pool. There are 217 bundles of 36 rods fuel and 120 bundles of 18 rods fuel. In this report, the information of the spent fuel about the loading date in the reactor core, discharged date, burnup, invisible inspection results and loading position in the spent fuel storage pool are described

  12. Development of information management system on LWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, B. D.; Lee, S. H.; Song, D. Y.; Jeon, I.; Park, S. J.; Seo, D. S. [KAERI, Taejon (Korea, Republic of)

    2002-10-01

    LWRs in Korea should manage all the information of spent fuel to implement the obligations under Korea-IAEA safeguards agreement and to perform the nuclear material accountancy work at the facility level. The information management system on LWR spent fuel was developed to manage all movement records from receipt to shipment of LWR fuels, and to get the necessary information such as nuclear fuel inventory lists and status, maps of fresh fuel storage, reactor and spent fuel pool, receipt and shipment records and so on. This information management system has a function to setup the system environments to cover the various kinds of storage types for all LWRs ; reactor, spent fuel pool and fresh fuel storage. The movements of nuclear fuel between the storages can be easily done by double click of the mouse to the destination. It also has a several error checking routines for maintaining the correct accounting data. Using this information management system of LWR spent fuel, facility operators can perform efficiently and effectively the safeguards related works including nuclear material accountancy at each facility.

  13. Safeguards approach for spent fuel transfers to dry storage

    International Nuclear Information System (INIS)

    Full text: Transfers of spent fuel to dry storage, where the fuel is not easily accessible for verification, have become a common occurrence at many nuclear reactors. The International Atomic Energy Agency (IAEA) safeguards on such transfers currently consume a considerable amount of inspection effort and are forecast to increase further. Under traditional safeguards, spent fuel transfers have been inspected by continuous inspector presence or by unattended instruments. The IAEA has recently developed a new safeguards approach for spent fuel transfers to dry storage for States under integrated safeguards. For States with a comprehensive safeguards agreement and an additional protocol in force for which the Agency has completed sufficient activities and evaluation and found no indication of diversion of nuclear material placed under safeguards, and no indication of undeclared nuclear material or activities for the State as a whole, safeguards activities can be optimized in light of the added safeguards assurance provided by measures of the additional protocol. The paper introduces a new safeguards policy, which was developed for transfers of spent fuel to dry storage for States under integrated safeguards, and describes a safeguards approach for transfers of spent fuel to dry storage at common reactors. The policy provides basis for the use of unannounced inspections to confirm the operator declarations of spent fuel transfer activities. During transfers of spent fuel to placement in dry storage, continuity of knowledge will be maintained by an unannounced inspection programme or some combination with temporary containment/surveillance (C/S) measures. With the policy, it will no longer be necessary for IAEA inspectors to be physically present during all spent fuel transfers. Traditional Safeguards Practices. When spent fuel is discharged from a reactor to a spent fuel pond, it is verified by Agency inspectors or by unattended instruments. In traditional safeguards

  14. Development of Techniques for Spent Fuel Assay – Differential Dieaway Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Swinhoe, Martyn Thomas [Los Alamos National Laboratory; Goodsell, Alison [NEN-1: SAFEGUARDS SCIENCE AND TECHNO; Ianakiev, Kiril Dimitrov [Los Alamos National Laboratory; Iliev, Metodi [Los Alamos National Laboratory; Desimone, David J. [Los Alamos National Laboratory; Rael, Carlos D. [Los Alamos National Laboratory; Henzl, Vladimir [Los Alamos National Laboratory; Polk, Paul John [Los Alamos National Laboratory

    2016-08-24

    This report summarizes the work done under a DNDO R&D funded project on the development of the differential dieaway method to measure plutonium in spent fuel. There are large amounts of plutonium that are contained in spent fuel assemblies and currently there is no way to make quantitative non-destructive assay. This has led NA24 under the Next Generation Safeguards Initiative (NGSI) to establish a multi-year program to investigate, develop and implement measurement techniques for spent fuel [1, 2, 3]. The techniques which are being experimentally tested by the existing NGSI project do not include any pulsed neutron active techniques. The present work covers the active neutron differential dieaway technique and has advanced the state of knowledge of this technique as well as produced a design for a practical active neutron interrogation instrument for spent fuel. Monte Carlo results from the NGSI effort show that much higher accuracy (1-2%) for the Pu content in spent fuel assemblies can be obtained with active neutron interrogation techniques [4, 5] than passive techniques and this would allow their use for nuclear material accountancy independently of any information from the operator. The main purpose of this work was to develop an active neutron interrogation technique for spent nuclear fuel.

  15. Pyroprocessing of Light Water Reactor Spent Fuels Based on an Electrochemical Reduction Technology

    International Nuclear Information System (INIS)

    A concept of pyroprocessing light water reactor (LWR) spent fuels based on an electrochemical reduction technology is proposed, and the material balance of the processing of mixed oxide (MOX) or high-burnup uranium oxide (UO2) spent fuel is evaluated. Furthermore, a burnup analysis for metal fuel fast breeder reactors (FBRs) is conducted on low-decontamination materials recovered by pyroprocessing. In the case of processing MOX spent fuel (40 GWd/t), UO2 is separately collected for ∼60 wt% of the spent fuel in advance of the electrochemical reduction step, and the product recovered through the rare earth (RE) removal step, which has the composition uranium:plutonium:minor actinides:fission products (FPs) = 76.4:18.4:1.7:3.5, can be applied as an ingredient of FBR metal fuel without a further decontamination process. On the other hand, the electroreduced alloy of high-burnup UO2 spent fuel (48 GWd/t) requires further decontamination of residual FPs by an additional process such as electrorefining even if RE FPs are removed from the alloy because the recovered plutonium (Pu) is accompanied by almost the same amount of FPs in addition to RE. However, the amount of treated materials in the electrorefining step is reduced to ∼10 wt% of the total spent fuel owing to the prior UO2 recovery step. These results reveal that the application of electrochemical reduction technology to LWR spent oxide fuel is a promising concept for providing FBR metal fuel by a rationalized process

  16. Recycling of nuclear spent fuel with AIROX processing

    International Nuclear Information System (INIS)

    This report examines the concept of recycling light water reactor (LWR) fuel through use of a dry-processing technique known as the AIROX (Atomics International Reduction Oxidation) process. In this concept, the volatiles and the cladding from spent LWR fuel are separated from the fuel by the AIROX process. The fuel is then reenriched and made into new fuel pins with new cladding. The feasibility of the concept is studied from a technical and high level waste minimization perspective

  17. Spent fuel dry storage technology development: thermal evaluation of sealed storage cask containing spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Schmitten, P.F.; Wright, J.B.

    1980-08-01

    A PWR spent fuel assembly was encapsulated inside the E-MAD Hot Bay and placed in a instrumented above surface storage cell during December 1978 for thermal testing. Instrumentation provided to measure canister, liner and concrete temperatures consisted of thermocouples which were inserted into tubes on the outside of the canister and liner and in three radial positions in the concrete. Temperatures from the SSC test assembly have been recorded throughout the past 16 months. Canister and liner temperatures have reached their peak values of 200{sup 0}F and 140{sup 0}F, respectively. Computer predictions of the transient and steady-state temperatures show good agreement with the test data.

  18. Spent fuel dry storage technology development: thermal evaluation of sealed storage cask containing spent fuel

    International Nuclear Information System (INIS)

    A PWR spent fuel assembly was encapsulated inside the E-MAD Hot Bay and placed in a instrumented above surface storage cell during December 1978 for thermal testing. Instrumentation provided to measure canister, liner and concrete temperatures consisted of thermocouples which were inserted into tubes on the outside of the canister and liner and in three radial positions in the concrete. Temperatures from the SSC test assembly have been recorded throughout the past 16 months. Canister and liner temperatures have reached their peak values of 2000F and 1400F, respectively. Computer predictions of the transient and steady-state temperatures show good agreement with the test data

  19. Aspects of VVR-S spent fuel management

    International Nuclear Information System (INIS)

    The management of two fuel types is presented, for the period 1957-1997. The research reactor VVR-S used EK-10 and S-36 fuel supplied by the former Soviet Union. The status of the spent fuel and possible options for medium term storage are described. (author)

  20. Cask operation and maintenance for spent fuel storage

    International Nuclear Information System (INIS)

    Interim storage is an essential platform for any option to be chosen later as an endpoint for spent fuel management. In view of such a circumstance, the most imminent service required for the spent fuel management worldwide is to provide adequate storage for the future spent fuel inventory arising either from the continued operation of nuclear power plants or from the removal of spent fuel in preparation for plant decommissioning. While the bulk of the global inventory of spent fuel are still stored in AR pools, dry storage has become a prominent alternative especially for newly built AFR facilities, with more than 17,000 t HM already stored in dry storage facilities worldwide. Storage in cask under inert conditions has become the preferred option, given the advantages including passive cooling features and modular mode of capacity increase. In terms of economics, dry storage is particularly propitious for long-term storage in that operational costs are minimized by the passive cooling features. The trend toward dry storage, especially in cask type, is likely to continue with an implication that and the supply will closely follow the increasing demand for storage by incremental additions of casks to the effect of minimizing cost penalty of the idle capacities typical of pool facilities. A variety of storage systems have been developed to meet specific requirements of different reactor fuels and a large number of designs based on these generic technologies are now available for the spent fuel containers (horizontal, vertical etc) and storage facilities. Multi-purpose technologies (i.e. a single technology for storage, transportation and disposal) have also been studied. Recent concern on security measures for protection of spent fuel has prompted a consideration on the possibility of placing storage facility underground. The future evolution of requirements and technologies will bring important impacts on cask operation and maintenance for spent fuel storage

  1. Spent Nuclear Fuel (SNF) Project Product Specification

    International Nuclear Information System (INIS)

    The process for removal of Spent Nuclear Fuel (SNF) from the K Basins has been divided into major sub-systems. The Fuel Retrieval System (FRS) removes fuel from the existing storage canisters, cleans it, and places it into baskets. The multi-canister overpack (MCO) loading system places the baskets into an MCO that has been pre-loaded in a cask. The cask, containing a loaded MCO, is then transferred to the Cold Vacuum Drying (CVD) Facility. After drying at the CVD Facility, the cask, and MCO, are transferred to the Canister Storage Building (CSB), where the MCO is removed from the cask, staged, inspected, sealed (by welding), and stored until a suitable permanent disposal option is implemented. The purpose of this document is to specify the process related characteristics of an MCO at the interface between major process systems. The characteristics are derived from the primary technical documents that form the basis for safety analysis and design calculations. This document translates the calculation assumptions into implementation requirements and describes the method of verifying that the requirement is achieved. These requirements are used to define validation test requirements and describe requirements that influence multiple sub-project safety analysis reports. This product specification establishes limits and controls for each significant process parameter at interfaces between major sub-systems that potentially affect the overall safety and/or quality of the SNF packaged for processing, transport, and interim dry storage. The product specifications in this document cover the SNF packaged in MCOs to be transported throughout the SNF Project. The description of the product specifications are organized in the document as follows: Section 2.0--Summary listing of product specifications at each major sub-system interface. Section 3.0--Summary description providing guidance as to how specifications are complied with by equipment design or processing within a major

  2. Spent Nuclear Fuel (SNF) Project Product Specification

    Energy Technology Data Exchange (ETDEWEB)

    PAJUNEN, A.L.

    2000-12-07

    The process for removal of Spent Nuclear Fuel (SNF) from the K Basins has been divided into major sub-systems. The Fuel Retrieval System (FRS) removes fuel from the existing storage canisters, cleans it, and places it into baskets. The multi-canister overpack (MCO) loading system places the baskets into an MCO that has been pre-loaded in a cask. The cask, containing a loaded MCO, is then transferred to the Cold Vacuum Drying (CVD) Facility. After drying at the CVD Facility, the cask, and MCO, are transferred to the Canister Storage Building (CSB), where the MCO is removed from the cask, staged, inspected, sealed (by welding), and stored until a suitable permanent disposal option is implemented. The purpose of this document is to specify the process related characteristics of an MCO at the interface between major process systems. The characteristics are derived from the primary technical documents that form the basis for safety analysis and design calculations. This document translates the calculation assumptions into implementation requirements and describes the method of verifying that the requirement is achieved. These requirements are used to define validation test requirements and describe requirements that influence multiple sub-project safety analysis reports. This product specification establishes limits and controls for each significant process parameter at interfaces between major sub-systems that potentially affect the overall safety and/or quality of the SNF packaged for processing, transport, and interim dry storage. The product specifications in this document cover the SNF packaged in MCOs to be transported throughout the SNF Project. The description of the product specifications are organized in the document as follows: Section 2.0--Summary listing of product specifications at each major sub-system interface. Section 3.0--Summary description providing guidance as to how specifications are complied with by equipment design or processing within a major

  3. ICP-AES technique in spent nuclear fuel reprocessing plant

    International Nuclear Information System (INIS)

    This report describes the glove box adaptation of a High Resolution Atomic Emission Spectrometer (HR-AES) for the determination of trace metallic elements in products of Reprocessing plant and various raw materials used for nuclear spent fuel reprocessing

  4. Scientists warn of 'trillion-dollar' spent-fuel risk

    Science.gov (United States)

    Gwynne, Peter

    2016-07-01

    A study by two Princeton University physicists suggests that a major fire in the spent nuclear fuel stored on the sites of US nuclear reactors could “dwarf the horrific consequences of the Fukushima accident”.

  5. Technical considerations of on-site spent fuel storage

    International Nuclear Information System (INIS)

    As more time is spent on power generation, more nuclear plants will face the dilemma of finding spent fuel storage space. Presently, there are numerous options available to nuclear power plant owners. The most common options are wet storage, such as fuel consolidation in a spent fuel pool, and dry storage, such as vault and cask storage. Choosing the most suitable option for a particular power plant is not an easy task. The primary selection considerations are licensing and financial. To achieve the optimum licensing and financial goals, a thorough technical evaluation of plant design, environmental requirements, and safety significance are essential. The purpose of this presentation is to benefit other nuclear plant owners by sharing the knowledge gained in selecting and evaluating the on-site spent fuel storage plan implemented at Power Company's Palisades Plant

  6. Time history analysis method for spent fuel racks

    International Nuclear Information System (INIS)

    Background: Spent fuel racks are important facilities to store the spent fuel which are free standing in the spent fuel pool. The response of racks to seismic load is highly nonlinear and involves a complex combination of motions: sliding, impact, twisting and turning. Purpose: An analysis method should be built to accurately replicate these nonlinear responses. Methods: The whole pool multi-rack FEA model was developed and time history analysis was performed which contains the consideration of effect of sliding, impact and friction and the fluid structure interaction effect. Results: The analysis results such as displacement and force under seismic loads were obtained. Conclusion: The method can be used to the seismic analysis for spent fuel racks. (authors)

  7. Spent-fuel verification with the Los Alamos fork detector

    International Nuclear Information System (INIS)

    The Los Alamos fork detector for the verification of spent-fuel assemblies has generated precise, reproducible data. The data analyses have now evolved to the point of placing tight restrictions on a diverter's actions

  8. An economic analysis of spent fuel management and storage

    International Nuclear Information System (INIS)

    Spent fuel management is becoming a key issue not only in the countries that have already experienced years of nuclear operation but also in the Asian countries that started nuclear utilization rather lately. This paper summarizes the key aspects that essentially determine optimal conditions for desired spent fuel management strategies from the engineering-economic point of view, in both national and regional perspectives. The term 'desired' is intended to highlight positive and beneficial aspects of such strategies, namely mobile and timely exploitation of spent fuel storage. Among all, the economy of scale, the economy of scope, the learning-by-doing effect, and benefits of R and D are reviewed theoretically and empirically, and the paper overviews to what extent these factors are implemented in solving spent fuel management strategy optimization problem. (author)

  9. Interim storage of CANDU spent fuel and safety performance

    International Nuclear Information System (INIS)

    'Full text:' Pickering Waste Management Facility (PWMF) is operational since November 1995 and safely storing spent fuel from Pickering 8 CANDU reactors. To date, equivalent to 22 reactor-years worth of spent fuel have been loaded, processed and stored in Dry Storage Containers (DSC). One DSC contains spent fuel from approximately one reactor-month of full power operation. The design life for the storage containers is 50 years. A Nuclear Waste Management Organization (NWMO) has been formed to advise on the long-term Canadian strategy for management of spent fuel. This paper will present the DSC processing steps, radiological hazard magnitude experienced during the DSC loading and processing for interim storage. A brief description of environmental and occupational safety performance will be presented. (author)

  10. Summary of the transportation of spent fuel attitude survey

    International Nuclear Information System (INIS)

    The proposed repository at Yucca Mountain, Nevada will increase highway and railway transportation of spent fuel and high level nuclear wastes. The purpose of the survey was to determine the attitudes and differences in attitudes of important actors in the transportation of spent fuel. The three major areas of investigation were 1) perceived risks associated with the transportation of spent fuel, 2) confidence in the government and others responsible for transporting spent fuel, and 3) certain transportation requirements. Response was 34.3% of the original mailing and included: 193 safety personnel, 141 employees of the nuclear industry, 260 government employees, 34 native Americans, and 9 employees of environmental organizations. This paper summarizes overall and group attitudes and opinions for the three areas mentioned above. (author)

  11. A study on the safety of spent fuel management

    International Nuclear Information System (INIS)

    The types and probabilities of events which may occur during the process of reception, transfer and storage of spent fuels in an away-from-reactor (AFR) spent fuel storage facility were analyzed in order to calculate the amount of radioactive material released to operation area and atmosphere, and the basic model for predicting the radioactive source-term under normal and abnormal operations were developed. Also, oxidation and dissolution of U02 pellet was investigated to estimate the amount of radioactive materials released from spent fuel and the release characteristics of radionuclides from defected spent fuel rods was analyzed. Basic information using FIRAC code to analyze the ventilation system during fire accident was prepared and FIRIN was detached from FIRAC modified to simulate the compartment fire by personal computer. (Author)

  12. Advanced fuel cycles options for LWRs and IMF benchmark definition

    International Nuclear Information System (INIS)

    In the paper, different advanced nuclear fuel cycles including thorium-based fuel and inert-matrix fuel are examined under light water reactor conditions, especially VVER-440, and compared. Two investigated thorium based fuels include one solely plutonium-thorium based fuel and the second one plutonium-thorium based fuel with initial uranium content. Both of them are used to carry and burn or transmute plutonium created in the classical UOX cycle. The inert-matrix fuel consist of plutonium and minor actinides separated from spent UOX fuel fixed in Yttria-stabilised zirconia matrix. The article shows analysed fuel cycles and their short description. The conclusion is concentrated on the rate of Pu transmutation and Pu with minor actinides cumulating in the spent advanced thorium fuel and its comparison to UOX open fuel cycle. Definition of IMF benchmark based on presented scenario is given. (authors)

  13. Foreign materials in a deep repository for spent nuclear fuels

    International Nuclear Information System (INIS)

    The effects of foreign substances introduced into a spent-fuel repository are reviewed. Possible impacts on processes and barrier-functions are examined, and the following areas are identified: Corrosion of the spent-fuel canister through the presence of sulfur and substances that favor microbial growth; impacts on the bentonite properties through the presence of cations as calcium, potassium and iron; radionuclide transport through the presence of complex-formers and surface-active substances

  14. Public Opinion Surveys in Spent Nuclear Fuel Management

    OpenAIRE

    Vasilieva, E.

    2002-01-01

    Russia's plans to import foreign SNF for storage and reprocessing meet serious public opposition. As a start of taking into account public concerns, programs of public involvement can be designed and implemented. In the paper, approaches to decision-making on spent nuclear fuel management that differ in their commitment to public participation are discussed. The review of public opinion surveys in Russia that investigated public attitudes to spent fuel is given. Finally, the experience of sev...

  15. Alternative measuring approaches in gamma scanning on spent nuclear fuel

    OpenAIRE

    Sihm Kvenangen, Karen

    2007-01-01

    In the future, the demand for energy is predicted to grow and more countries plan to utilize nuclear energy as their source of electric energy. This gives rise to many important issues connected to nuclear energy, such as finding methods that can verify that the spent nuclear fuel has been handled safely and used in ordinary power producing cycles as stated by the operators. Gamma ray spectroscopy is one method used for identification and verification of spent nuclear fuel. In the specific ga...

  16. Site selection - location of the repository for spent nuclear fuel

    International Nuclear Information System (INIS)

    This document describes the localization work and SKB's choice of site for the repository. Furthermore, SKB's basis and rationale for the decisions taken during the work are reported. The document is Appendix PV of applications under the Nuclear Activities Act and the Environmental Code to both build and operate an encapsulation plant adjacent to the central interim storage facility for spent nuclear fuel in Oskarshamn, and to construct and operate a disposal facility for spent nuclear fuel at Forsmark in Oesthammar municipality

  17. Spent nuclear fuel Canister Storage Building CDR Review Committee report

    International Nuclear Information System (INIS)

    The Canister Storage Building (CSB) is a subproject under the Spent Nuclear Fuels Major System Acquisition. This subproject is necessary to design and construct a facility capable of providing dry storage of repackaged spent fuels received from K Basins. The CSB project completed a Conceptual Design Report (CDR) implementing current project requirements. A Design Review Committee was established to review the CDR. This document is the final report summarizing that review

  18. An overview on the nuclear spent fuel management in Romania

    International Nuclear Information System (INIS)

    The sources of radioactive waste in Romania are users of radiation and radioactive materials in industry (including nuclear electricity generation), medicine, agriculture and research and also the processing of materials that are naturally radioactive, such as uranium ores. The different types of radioactive waste are classified into four categories of waste: excepted waste, low level waste, medium level waste and high level waste. A spent fuel management sub-programme as a part of the Radioactive Waste Management programme was initiated by the former Romanian Electricity Company (RENEL) in 1992. Within the frame of R and D of the Radioactive Waste and Spent Fuel Management Programme, the topics cover investigations, studies and research to identify the sites and the conceptual designs for a Spent Fuel Interim Storage Facility (SFISF) and also a Spent Fuel Disposal Facility (SFDF). Changes in the organization of the nuclear activities of RENEL, involving both responsibilities and financing aspects, led to interruption of the programme. The programme includes study of the main methods and the existing technologies for the design, operation and safety of an interim storage facility (including transport aspects). It also includes analysis of details on the site selection for this facility and for a spent fuel final disposal facility. The achievement of the spent fuel interim storage facility is proceeding. The results from the studies performed in the last years will permit us to prepare the feasibility study next year and the documentation required by our regulatory body for starting the process to obtain a license for a SFISF at Cernavoda. A second phase is the assessment of a long term strategy to select and adopt a proven disposal technology for spent fuel, corresponding with a selected site. The status of the work performed in the frame of this programme and also the situation of the spent fuel from research reactors are presented. (author)

  19. Scientific reference on the long time evolution of spent fuels

    International Nuclear Information System (INIS)

    This report is published in the framework of the 1991 French law for the nuclear waste management. The state of the art reported here concerns the long term evolution of spent fuel in the various environmental conditions corresponding to dry storage and geological disposal: closed system, air and water saturated medium. This review is based on the results of the french PRECCI project (Research Program on Long term Evolution of Spent Nuclear Fuel) and on literature data. (authors)

  20. Attitudes of the knowledgeable toward the transportation of spent fuel

    International Nuclear Information System (INIS)

    The purpose of the paper is to discuss the recent data analysis of a mail-in survey which investigated the attitudes of people having some knowledge of the nuclear industry. The survey contained 74 items which solicited 4 major types of information. Thirty-seven items investigated respondents' perception of risk concerning spent fuel transportation, 12 examined respondents' confidence in certain governing agencies' abilities of managing the safe transportation of spent fuel, 19 examined respondents' attitudes toward certain special precautionary actions for spent fuel transport, and 6 requested demographic data. The 637 people who responded to the survey were divided into 5 mutually exclusive groups for analysis. These included: safety professionals, government employees, employees of the nuclear industry, employees of environmental organizations, and Native Americans. The safety, government, and Native American groups were divided into subgroups for further analysis. Overall and group responses concerned with special precautionary actions for the transportation of spent fuel were examined. Native American, safety, and government subgroup responses to items which examined respondents' perceived risks, confidence in certain governing agencies, and attitudes concerning special precautionary actions for the transportation of spent fuel were also investigated. Additionally, risk perceptions of spent fuel transport was compared between people residing no more than 5 miles from an interstate highway to those living no less than 16 miles from an interstate highway. The results of the analysis are divided into 4 sections: A. Sample Identification; B. Overall and Group Attitudes to Certain Special Precautionary Actions For Transporting Spent Fuel; C. Subgroup Attitudes toward the Transportation of Spent Fuel; D. Risk Perceptions of People Living No More than 5 Miles vs. People Living No Less than 16 Miles from and Interstate Highway. Generally conclusions are also provided