WorldWideScience

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)

    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

  4. 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.

  5. Advanced waste forms from spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ackerman, J.P.; McPheeters, C.C.

    1995-12-31

    More than one hundred spent nuclear fuel types, having an aggregate mass of more than 5000 metric tons (2700 metric tons of heavy metal), are stored by the United States Department of Energy. This paper proposes a method for converting this wide variety of fuel types into two waste forms for geologic disposal. The method is based on a molten salt electrorefining technique that was developed for conditioning the sodium-bonded, metallic fuel from the Experimental Breeder Reactor-II (EBR-II) for geologic disposal. The electrorefining method produces two stable, optionally actinide-free, high-level waste forms: an alloy formed from stainless steel, zirconium, and noble metal fission products, and a ceramic waste form containing the reactive metal fission products. Electrorefining and its accompanying head-end process are briefly described, and methods for isolating fission products and fabricating waste forms are discussed.

  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 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

  8. 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

  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. 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

  11. 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.

  12. 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

  13. 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

  14. 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.

  15. 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.

  16. 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

  17. 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

  18. 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.

  19. 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

  20. 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

  1. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Ro, Seung Gy; Byeon, Kee Hoh; Song, Dae Yong; Park, Seong Won; Shin, Young Jun

    1999-03-01

    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.

  3. 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

  4. 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)

  5. 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.

  6. 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

  7. 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.)

  8. 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

  9. 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

  10. 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)

  11. 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

  12. 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)

  13. 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

  14. 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

  15. 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

  16. 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)

  17. 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

  18. 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.

  19. 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)

  20. 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

  1. 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.

  2. 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

  3. 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.

  4. 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)

  5. 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)

  6. 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

  7. 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

  8. 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.

  9. 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.

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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

  15. Spent fuel characteristics & disposal considerations

    Energy Technology Data Exchange (ETDEWEB)

    Oversby, V.M.

    1996-06-01

    The fuel used in commercial nuclear power reactors is uranium, generally in the form of an oxide. The gas-cooled reactors developed in England use metallic uranium enclosed in a thin layer of Magnox. Since this fuel must be processed into a more stable form before disposal, we will not consider the characteristics of the Magnox spent fuel. The vast majority of the remaining power reactors in the world use uranium dioxide pellets in Zircaloy cladding as the fuel material. Reactors that are fueled with uranium dioxide generally use water as the moderator. If ordinary water is used, the reactors are called Light Water Reactors (LWR), while if water enriched in the deuterium isotope of hydrogen is used, the reactors are called Heavy Water reactors. The LWRs can be either pressurized reactors (PWR) or boiling water reactors (BWR). Both of these reactor types use uranium that has been enriched in the 235 isotope to about 3.5 to 4% total abundance. There may be minor differences in the details of the spent fuel characteristics for PWRs and BWRs, but for simplicity we will not consider these second-order effects. The Canadian designed reactor (CANDU) that is moderated by heavy water uses natural uranium without enrichment of the 235 isotope as the fuel. These reactors run at higher linear power density than LWRs and produce spent fuel with lower total burn-up than LWRs. Where these difference are important with respect to spent fuel management, we will discuss them. Otherwise, we will concentrate on spent fuel from LWRs.

  16. 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)

  17. 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

  18. 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)

  19. 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

  20. 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.

  1. 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)

  2. 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

  3. 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

  4. 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.

  5. 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.)

  6. 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)

  7. 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)

  8. 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)

  9. Developments in spent fuel storage

    Energy Technology Data Exchange (ETDEWEB)

    Stallings, R.A. [USDOE Office of Civilian Radioactive Waste Management, Washington, DC (United States)

    1995-04-01

    The author gives a brief review of the his efforts to negotiate a site for monitored retrieval storage (MRS) of spent fuels in 1994. His efforts were centered on finding a voluntary host for the MRS site. He found politician were not opposed but did not want to make it a campaign issue during 1994. The author and his office came to the conclusion that to find a site voluntarily, the project would have to be an economic opportunity for the region.

  10. 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

  11. Spent fuel canister docking station

    Energy Technology Data Exchange (ETDEWEB)

    Suikki, M. [Afore Oy, Turku (Finland)

    2006-01-15

    The working report for the spent fuel canister docking station presents a design for the operation and structure of the docking equipment located in the fuel handling cell for the spent fuel in the encapsulation plant. The report contains a description of the basic requirements for the docking station equipment and their implementation, the operation of the equipment, maintenance and a cost estimate. In the designing of the equipment all the problems related with the operation have been solved at the level of principle, nevertheless, detailed designing and the selection of final components have not yet been carried out. In case of defects and failures, solutions have been considered for postulated problems, and furthermore, the entire equipment was gone through by the means of systematic risk analysis (PFMEA). During the docking station designing we came across with needs to influence the structure of the actual disposal canister for spent nuclear fuel, too. Proposed changes for the structure of the steel lid fastening screw were included in the report. The report also contains a description of installation with the fuel handling cell structures. The purpose of the docking station for the fuel handling cell is to position and to seal the disposal canister for spent nuclear fuel into a penetration located on the cell floor and to provide suitable means for executing the loading of the disposal canister and the changing of atmosphere. The designed docking station consists of a docking ring, a covering hatch, a protective cone and an atmosphere-changing cap as well as the vacuum technology pertaining to the changing of atmosphere and the inert gas system. As far as the solutions are concerned, we have arrived at rather simple structures and most of the actuators of the system are situated outside of the actual fuel handling cell. When necessary, the equipment can also be used for the dismantling of a faulty disposal canister, cut from its upper end by machining. The

  12. 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.

  13. 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.

  14. 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)

  15. 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

  16. 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

  17. 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.

  18. 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

  19. 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.)

  20. 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.

  1. 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

  2. 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)

  3. 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

  4. 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

  5. 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.

  6. 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)

  7. 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)

  8. 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

  9. 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

  10. 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.

  11. 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

  12. 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

  13. 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

  14. Inspection of state of spent fuel elements stored in RA reactor spent fuel storage pool

    Energy Technology Data Exchange (ETDEWEB)

    Aden, V.G.; Bulkin, S.Yu.; Sokolov, A.V. [Research and Development Institute of Power Engineering, Moscow (Russian Federation); Matausek, M.V.; Vukadin, Z. [VINCA Institute of Nuclear Science, Belgrade (Yugoslavia)

    1999-07-01

    About five thousand spent fuel elements from RA reactor have been stored for over 30 years in sealed aluminum barrels in the spent fuel storage pool. This way of storage does not provide complete information about the state of spent fuel elements or the medium inside the barrels, like pressure or radioactivity. The technology has recently been developed and the equipment has been manufactured to inspect the state of the spent fuel and to reduce eventual internal pressure inside the aluminum barrels. Based on the results of this inspection, a procedure will be proposed for transferring spent fuel to a more reliable storage facility. (author)

  15. 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

  16. 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

  17. 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)

  18. The cost of spent fuel storage

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J. R.; Palacios H, J. C.; Badillo, V.; Alonso, G., E-mail: ramon.ramirez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2011-11-15

    Spent fuel is one of the most important issues in the nuclear industry, currently spent fuel management is been cause of great amount of research, investments, constructing repositories or constructing the necessary facilities to reprocess the fuel, and later to recycle the plutonium recovered in thermal reactors. What is the best solution?, or What is the best technology for an specific solution? Many countries have deferred the decision on selecting an option, while others works actively constructing repositories and others implementing the reprocessing facilities to recycle the plutonium obtained from nuclear spent fuel. In Mexico the nuclear power is limited to two reactors BWR type and medium size. So the nuclear spent fuel discharged has been accommodated at reactor's spent fuel pools. Originally these pools have enough capacity to accommodate spent fuel for the 40 years of designed plant operation. However currently, the plants are under a process for extended power up-rate to 20% of original power and also there are plans to extended operational life for 20 more years. Under these conditions there will not be enough room for spent fuel in the pools. (Author)

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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)

  5. 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

  6. 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

  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. 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

  10. 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.

  11. 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)

  12. 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.

  13. 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)

  14. 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)

  15. 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.

  16. 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

  17. 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...

  18. 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...

  19. 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

  20. 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)

  1. 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)

  2. 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

  3. 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......

  4. 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)

  5. 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

  6. Spent fuel management of Jose Cabrera NPP

    Energy Technology Data Exchange (ETDEWEB)

    Blanco Zurro, J.E.; Garcia Costilla, M. [Area de Generacion - Unidad Nuclear, Gas Natural Fenosa, Avda. de San Luis, 77, 28033 Madrid (Spain); Lavara Sanz, A. [Division Nuclear, SOCOIN, P. del Club Deportivo, 1 - Edificio 5, 28223 Pozuelo de Alarcon, Madrid (Spain); Martinez Abad, J.E. [Departamento de Residuos de Alta Actividad, ENRESA, C/ Emilio Vargas, 7, 28043 Madrid (Spain)

    2010-07-01

    The definitive shutdown of Jose Cabrera Nuclear Power Plant took place on 30. of April 2006. From this moment, cooperation agreements between ENRESA and GAS NATURAL FENOSA were established to reach, among others objectives, its decommissioning, 3 years after the shutdown of the reactor. In order to accomplish the Spanish nuclear regulation, a spent fuel management plan was developed. This plan determined that the fuel assemblies placed in the spent fuel pool would be managed by means of their storage in an interim installation. For this reason, an Independent Spent Fuel Storage Installation (ISFSI) was built at plant site, pioneer in Spain by its characteristics of design. Different administrative authorizations from the point of view of nuclear safety as well as from the environmental were required for ISFSI licensing process. The transference and storage of spent fuel was carried out using the HI-STORM 100Z Dry Storage System, developed by HOLTEC INTERNATIONAL. This system, designed for the spent fuel storage in casks, supports abnormal and very hard accident conditions. The system has three main components: Storage Cask (HI-STORM), Transfer Cask (HI-TRAC) and Multipurpose Canister (MPC). In addition to this, the system has a specific Transport Cask (HI-STAR) for the future transport out of the Plant. More than 30 Design Modifications to the system and plant were implemented to solve structural problems and to include safety and ALARA improvements. The transfer of the spent fuel and its emplacement in the ISFSI began on January 2009 and finished on September of that year allowing starting the decommissioning process, three years and a half after Jose Cabrera NPP shutdown. (authors)

  7. 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

  8. 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.

  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 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.

  12. 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)

  13. 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

  14. 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

  15. 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

  16. 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

  17. 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.

  18. 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.

  19. 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)

  20. 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

  1. 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

  2. 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

  3. Direct reuse of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Mohamed, Nader M.A., E-mail: mnader73@yahoo.com

    2014-10-15

    Highlights: • A new design for the PWR assemblies for direct use of spent fuel was proposed. • The PWR spent fuel will be transferred directly (after a certain cooling time) to CANDU reactors. • The proposed assembly has four zircaloy-4 tubes contains a number of CANDU fuel bundles (7 or 8 bundles per tube) stacked end to end. • MCNPX is used for the calculations that showed that the burnup can be increased by about 25%. • Acceptable linear heat generation rate in hot rods and improved Pu proliferation resistance. - Abstract: In this paper we proposed a new design for the PWR fuel assembly for direct use of the PWR spent fuel without processing. The PWR spent fuel will be transferred directly (after a certain cooling time) to CANDU reactors which preferably built in the same site to avoid the problem of transportations. The proposed assembly has four zircaloy-4 tubes contains a number of CANDU fuel bundles (7 or 8 bundles per tube) stacked end to end. Each tube has the same inner diameter of that of CANDU pressure tube. The spaces between the tubes contain low enriched UO{sub 2} fuel rods and guide tubes. MCNPX code is used for the simulation and calculation of the burnup of the proposed assembly. The bundles after the discharge from the PWR with their materials inventories are burned in a CANDU cell after a certain decay time. The results were compared with reference results and the impact of this new design on the uranium utilization improvement and on the proliferation resistance of plutonium is discussed. The effect of this new design on the power peaking, moderator temperature coefficient of reactivity and CANDU coolant void reactivity are discussed as well.

  4. Spent nuclear fuel project integrated schedule plan

    Energy Technology Data Exchange (ETDEWEB)

    Squires, K.G.

    1995-03-06

    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.

  5. 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)

  6. 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

  7. 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

  8. 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

  9. 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)

  10. 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)

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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)

  17. 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

  18. 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.)

  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. 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

  1. Report on interim storage of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    1993-04-01

    The report on interim storage of spent nuclear fuel discusses the technical, regulatory, and economic aspects of spent-fuel storage at nuclear reactors. The report is intended to provide legislators state officials and citizens in the Midwest with information on spent-fuel inventories, current and projected additional storage requirements, licensing, storage technologies, and actions taken by various utilities in the Midwest to augment their capacity to store spent nuclear fuel on site.

  2. 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

  3. 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...

  4. 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

  5. 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

  6. 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)

  7. 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

  8. 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.

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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)

  14. 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)

  15. 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

  16. 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.

  17. 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

  18. 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)

  19. 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

  20. 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.

  1. 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

  2. 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.

  3. 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.

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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.

  10. 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

  11. 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.

  12. 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

  13. 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-...

  14. 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

  15. 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)

  16. 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

  17. 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

  18. 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.

  19. Transportation capabilities study of DOE-owned spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Clark, G.L.; Johnson, R.A.; Smith, R.W. [Packaging Technology, Inc., Tacoma, WA (United States); Abbott, D.G.; Tyacke, M.J. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

    1994-10-01

    This study evaluates current capabilities for transporting spent nuclear fuel owned by the US Department of Energy. Currently licensed irradiated fuel shipping packages that have the potential for shipping the spent nuclear fuel are identified and then matched against the various spent nuclear fuel types. Also included are the results of a limited investigation into other certified packages and new packages currently under development. This study is intended to support top-level planning for the disposition of the Department of Energy`s spent nuclear fuel inventory.

  20. Georgia Tech Studies of Sub-Critical Advanced Burner Reactors with a D-T Fusion Tokamak Neutron Source for the Transmutation of Spent Nuclear Fuel

    Science.gov (United States)

    Stacey, W. M.

    2009-09-01

    The possibility that a tokamak D-T fusion neutron source, based on ITER physics and technology, could be used to drive sub-critical, fast-spectrum nuclear reactors fueled with the transuranics (TRU) in spent nuclear fuel discharged from conventional nuclear reactors has been investigated at Georgia Tech in a series of studies which are summarized in this paper. It is found that sub-critical operation of such fast transmutation reactors is advantageous in allowing longer fuel residence time, hence greater TRU burnup between fuel reprocessing stages, and in allowing higher TRU loading without compromising safety, relative to what could be achieved in a similar critical transmutation reactor. The required plasma and fusion technology operating parameter range of the fusion neutron source is generally within the anticipated operational range of ITER. The implications of these results for fusion development policy, if they hold up under more extensive and detailed analysis, is that a D-T fusion tokamak neutron source for a sub-critical transmutation reactor, built on the basis of the ITER operating experience, could possibly be a logical next step after ITER on the path to fusion electrical power reactors. At the same time, such an application would allow fusion to contribute to meeting the nation's energy needs at an earlier stage by helping to close the fission reactor nuclear fuel cycle.

  1. 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)

  2. 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

  3. 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)

  4. 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.

  5. 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

  6. 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

  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. 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 ...

  9. 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

  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. Transportation and storage of foreign spent power reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    1979-09-30

    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.

  12. 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

  13. 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)

  14. 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

  15. 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...

  16. 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

  17. Development of spent fuel remote handling technology - Kinematic analysis of bilateral arms for abnormal spent fuels

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kyu Won; Yoo, Ju Sang; Kim, Jong Yoon [Chungbuk National University, Chongju (Korea)

    2000-03-01

    In the project of 'Development of Spent Fuel Remote Handling Technology', Preprocessing technique, mechanism and teleoperation technique are being developed. One of the mechanisms is a device for disassembling of the spent fuel bundle. However, there may be abnormal fuel bar among the fuel bundle, In this case the unpacking task will be difficult and dangerous. So, in that case, a force reflected teleoperation manipulator is desirable. The system is composed of a anthropomorphic input device at control site, power manipulator at remote site and control system. In this research, the forward and inverse kinematic equations of input device and manipulators has been solved, respectively. In addition, the mapping algorithm is proposed and shown using computer simulation. The reaction force of the telemanipulator with the environmental object is reflected through control system. The reaction force is decomposed into joint torque of the input device based on the jacobian equation. The obtained theoretical relations are verified through computer simulation and they will be used effectively in the spent fuel remote handling technology. 6 refs., 26 figs., 7 tabs. (Author)

  18. 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

  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

    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.

  20. 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.

  1. 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.

  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. 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

  5. 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).

  6. 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

  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. 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

  10. 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)

  11. 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

  12. 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.

  13. 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

  14. 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

  15. 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.

  16. 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.

  17. 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

  18. 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...

  19. 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

  20. 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

  1. 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)

  2. 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.

  3. Storage of LWR spent fuel in air: Volume 1: Design and operation of a spent fuel oxidation test facility

    Energy Technology Data Exchange (ETDEWEB)

    Thornhill, C.K.; Campbell, T.K.; Thornhill, R.E.

    1988-12-01

    This report describes the design and operation and technical accomplishments of a spent-fuel oxidation test facility at the Pacific Northwest Laboratory. The objective of the experiments conducted in this facility was to develop a data base for determining spent-fuel dry storage temperature limits by characterizing the oxidation behavior of light-water reactor (LWR) spent fuels in air. These data are needed to support licensing of dry storage in air as an alternative to spent-fuel storage in water pools. They are to be used to develop and validate predictive models of spent-fuel behavior during dry air storage in an Independent Spent Fuel Storage Installation (ISFSI). The present licensed alternative to pool storage of spent fuel is dry storage in an inert gas environment, which is called inerted dry storage (IDS). Licensed air storage, however, would not require monitoring for maintenance of an inert-gas environment (which IDS requires) but does require the development of allowable temperature limits below which UO/sub 2/ oxidation in breached fuel rods would not become a problem. Scoping tests at PNL with nonirradiated UO/sub 2/ pellets and spent-fuel fragment specimens identified the need for a statistically designed test matrix with test temperatures bounding anticipated maximum acceptable air-storage temperatures. This facility was designed and operated to satisfy that need. 7 refs.

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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

  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. Spent fuel data base: commercial light water reactors. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hauf, M.J.; Kniazewycz, B.G.

    1979-12-01

    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.

  15. 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

  16. 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

  17. 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...

  18. 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)

  19. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-02-01

    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.

  1. 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

  2. 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.

  3. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    1978-05-01

    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.

  5. 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

  6. Development of the Combination Method of PWR Spent Fuel for DUPIC Fuel Preparation

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Ju Ho; Kim, S. K.; Jung, T. C.; June, T. H.; Lee, J. M.; Kim, I. S.; Park, C. S.; Kim, M. J.; An, J. I.; Park, S. H. [Kyung Hee University, Seoul (Korea, Republic of)

    1997-07-15

    Optimum finding method of PWR spent fuel was developed in application of DUPIC fuel composition to nuclear fuel production. In order to make the database of the PWR spent fuel for the optimum composition, composition data of the PWR spent fuels from Youngkwang unit 1 and 2, Kori unit 3 and 4 and Uljin unit 1 and 2 were collected, analyzed and stored. Artificial intelligent access was attempted in optimizing the composition, and the combination algorithm for PWR spent fuel was developed. In this work database of the composition data of the PWR spent fuels from Youngkwang unit 1 and 2, Kori unit 3 and 4 and Uljin unit 1 and 2 as well as their combination algorithm for PWR spent fuel were developed. The combination algorithm is to find the combination of the spent fuel assembly which is quite close to the requirement per unit mass of DUPIC fuel. The required data are total weight of the fuel, tolerance of the errors, importance of the elements and the discharge data. This combination algorithm enables to find the optimum PWR spent fuel assembly for DUPIC fuel with the database of the spent fuels according to the DUPIC fuel standards. The combination algorithm developed in this work can afford the technical support to fuel supply in preparing the DUPIC fuel, and make contribution in DUPIC fuel cycle technology. It can be directly used in DUPIC fuel cycle technology, and can be also used in the management of the spent fuels with respect to their compositions and ingredients as well as the nuclear safeguards. Composition of the PWR spent fuel in each assembly depends on the initial concentration, degree of combustion, specific power, and its location in the reactor core. They may be affected by the kind of fuel rod and its axial length. Therefore, analysis procedures in these regards should be established for the effective application of the results of this work. 6 refs., 12 tabs., 46 figs. (author)

  7. Development of the ACP safeguards neutron counter for PWR spent fuel rods

    Science.gov (United States)

    Lee, Tae-Hoon; Menlove, Howard O.; Lee, Sang-Yoon; Kim, Ho-Dong

    2008-04-01

    An advanced neutron multiplicity counter has been developed for measuring spent fuel in the Advanced spent fuel Conditioning Process (ACP) at the Korea Atomic Energy Research Institute (KAERI). The counter uses passive neutron multiplicity counting to measure the 244Cm content in spent fuel. The input to the ACP process is spent fuel from pressurized water reactors (PWRs), and the high intensity of the gamma-ray exposure from spent fuel requires a careful design of the counter to measure the neutrons without gamma-ray interference. The nuclear safeguards for the ACP facility requires the measurement of the spent fuel input to the process and the Cm/Pu ratio for the plutonium mass accounting. This paper describes the first neutron counter that has been used to measure the neutron multiplicity distribution from spent fuel rods. Using multiple samples of PWR spent fuel rod-cuts, the singles (S), doubles (D), and triples (T) rates of the neutron distribution for the 244Cm nuclide were measured and calibration curves were produced. MCNPX code simulations were also performed to obtain the three counting rates and to compare them with the measurement results. The neutron source term was evaluated by using the ORIGEN-ARP code. The results showed systematic difference of 21-24% in the calibration graphs between the measured and simulation results. A possible source of the difference is that the burnup codes have a 244Cm uncertainty greater than ±15% and it would be systematic for all of the calibration samples. The S/D and D/T ratios are almost constant with an increment of the 244Cm mass, and this indicates that the bias is in the 244Cm neutron source calculation using the ORIGEN-ARP source code. The graphs of S/D and D/T ratios show excellent agreement between measurement and MCNPX simulation results.

  8. 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

  9. 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

  10. 78 FR 16601 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System

    Science.gov (United States)

    2013-03-18

    ... National Technology Transfer and Advancement Act of 1995 (Pub. L. 104-113) requires that Federal agencies... spent fuel storage regulations by revising the NAC International, Inc. (NAC) Modular Advanced Generation... explains why the rule would be inappropriate, including challenges to the rule's underlying premise...

  11. 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

  12. 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

  13. 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

  14. 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.

  15. 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.

  16. 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

  17. 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)

  18. 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

  19. 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

  20. 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.

  1. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Swinhoe, Martyn Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Goodsell, Alison [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Ianakiev, Kiril Dimitrov [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Iliev, Metodi [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Desimone, David J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rael, Carlos D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Henzl, Vladimir [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Polk, Paul John [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-07-28

    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. 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 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.

  3. 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.

  4. 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

  5. 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”.

  6. Microbiology of spent nuclear fuel storage basins.

    Science.gov (United States)

    Santo Domingo, J W; Berry, C J; Summer, M; Fliermans, C B

    1998-12-01

    Microbiological studies of spent nuclear fuel storage basins at Savannah River Site (SRS) were performed as a preliminary step to elucidate the potential for microbial-influenced corrosion (MIC) in these facilities. Total direct counts and culturable counts performed during a 2-year period indicated microbial densities of 10(4) to 10(7) cells/ml in water samples and on submerged metal coupons collected from these basins. Bacterial communities present in the basin transformed between 15% and 89% of the compounds present in Biologtrade mark plates. Additionally, the presence of several biocorrosion-relevant microbial groups (i.e., sulfate-reducing bacteria and acid-producing bacteria) was detected with commercially available test kits. Scanning electron microscopy and X-ray spectra analysis of osmium tetroxide-stained coupons demonstrated the development of microbial biofilm communities on some metal coupons submerged for 3 weeks in storage basins. After 12 months, coupons were fully covered by biofilms, with some deterioration of the coupon surface evident at the microscopical level. These results suggest that, despite the oligotrophic and radiological environment of the SRS storage basins and the active water deionization treatments commonly applied to prevent electrochemical corrosion in these facilities, these conditions do not prevent microbial colonization and survival. Such microbial densities and wide diversity of carbon source utilization reflect the ability of the microbial populations to adapt to these environments. The presumptive presence of sulfate-reducing bacteria and acid-producing bacteria and the development of biofilms on submerged coupons indicated that an environment for MIC of metal components in the storage basins may occur. However, to date, there has been no indication or evidence of MIC in the basins. Basin chemistry control and corrosion surveillance programs instituted several years ago have substantially abated all corrosion mechanisms.

  7. Microbiology of spent nuclear fuel storage basins.

    Science.gov (United States)

    Santo Domingo, J W; Berry, C J; Summer, M; Fliermans, C B

    1998-12-01

    Microbiological studies of spent nuclear fuel storage basins at Savannah River Site (SRS) were performed as a preliminary step to elucidate the potential for microbial-influenced corrosion (MIC) in these facilities. Total direct counts and culturable counts performed during a 2-year period indicated microbial densities of 10(4) to 10(7) cells/ml in water samples and on submerged metal coupons collected from these basins. Bacterial communities present in the basin transformed between 15% and 89% of the compounds present in Biologtrade mark plates. Additionally, the presence of several biocorrosion-relevant microbial groups (i.e., sulfate-reducing bacteria and acid-producing bacteria) was detected with commercially available test kits. Scanning electron microscopy and X-ray spectra analysis of osmium tetroxide-stained coupons demonstrated the development of microbial biofilm communities on some metal coupons submerged for 3 weeks in storage basins. After 12 months, coupons were fully covered by biofilms, with some deterioration of the coupon surface evident at the microscopical level. These results suggest that, despite the oligotrophic and radiological environment of the SRS storage basins and the active water deionization treatments commonly applied to prevent electrochemical corrosion in these facilities, these conditions do not prevent microbial colonization and survival. Such microbial densities and wide diversity of carbon source utilization reflect the ability of the microbial populations to adapt to these environments. The presumptive presence of sulfate-reducing bacteria and acid-producing bacteria and the development of biofilms on submerged coupons indicated that an environment for MIC of metal components in the storage basins may occur. However, to date, there has been no indication or evidence of MIC in the basins. Basin chemistry control and corrosion surveillance programs instituted several years ago have substantially abated all corrosion mechanisms

  8. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Dana, W.P.

    1995-12-01

    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

  10. 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)

  11. 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...

  12. 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...

  13. Economic and Innovative Spent Fuel Pool Level Instrumentation

    Energy Technology Data Exchange (ETDEWEB)

    Legrand, R.; Scecina, J.; Koenig, W.

    2014-07-01

    At Fukushima, significant attention and resources were concentrated on providing makeup water to the Spent Fuel Pools (SFP). No level indication was available and the assumption was that the water level was dangerously low. As it later turned out, the applied resources could have been used more effectively elsewhere. In reaction, Nuclear Regulatory Authorities worldwide have issued orders to add rugged, seismically qualified level instrumentation to Spent Fuel Pools. (Author)

  14. 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

  15. Safety Aspects of Dry Spent Fuel Storage and Spent Fuel Management - 13559

    International Nuclear Information System (INIS)

    Dry storage systems are characterized by passive and inherent safety systems ensuring safety even in case of severe incidents or accidents. After the events of Fukushima, the advantages of such passively and inherently safe dry storage systems have become more and more obvious. As with the storage of all radioactive materials, the storage of spent nuclear fuel (SF) and high-level radioactive waste (HLW) must conform to safety requirements. Following safety aspects must be achieved throughout the storage period: - safe enclosure of radioactive materials, - safe removal of decay heat, - securing nuclear criticality safety, - avoidance of unnecessary radiation exposure. The implementation of these safety requirements can be achieved by dry storage of SF and HLW in casks as well as in other systems such as dry vault storage systems or spent fuel pools, where the latter is neither a dry nor a passive system. Furthermore, transport capability must be guaranteed during and after storage as well as limitation and control of radiation exposure. The safe enclosure of radioactive materials in dry storage casks can be achieved by a double-lid sealing system with surveillance of the sealing system. The safe removal of decay heat must be ensured by the design of the storage containers and the storage facility. The safe confinement of radioactive inventory has to be ensured by mechanical integrity of fuel assembly structures. This is guaranteed, e.g. by maintaining the mechanical integrity of the fuel rods or by additional safety measures for defective fuel rods. In order to ensure nuclear critically safety, possible effects of accidents have also to be taken into consideration. In case of dry storage it might be necessary to exclude the re-positioning of fissile material inside the container and/or neutron moderator exclusion might be taken into account. Unnecessary radiation exposure can be avoided by the cask or canister vault system itself. In Germany dry storage of SF in

  16. Reactor-specific spent fuel discharge projections, 1987-2020

    Energy Technology Data Exchange (ETDEWEB)

    Walling, R.C.; Heeb, C.M.; Purcell, W.L.

    1988-03-01

    The creation of five reactor-specific spent fuel data bases that contain information on the projected amounts of spent fuel to be discharged from U.S. commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water eactors (BWR), and one existing high temperature gas reactor (HTGR). The projections are based on individual reactor information supplied by the U.S. reactor owners. The basic information is adjusted to conform to Energy Information Administration (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: No New Orders (assumes increasing burnup), No New Orders with No Increased Burnup, Upper Reference (assumes increasing burnup), Upper Reference with No Increased Burnup, and Lower Reference (assumes increasing burnup). Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum at-reactor storage, and for storage requirements assuming maximum at-reactor storage plus intra-utility transshipment of spent fuel. 8 refs., 8 figs., 10 tabs.

  17. Operating Experience and Condition Assessment of Spent Fuel Dry Storage Silos and Spent Fuel Pool at Embalse NPP

    International Nuclear Information System (INIS)

    At Embalse Nuclear Power Plant (NPP), spent fuel removed from the reactor core is placed in a carbon steel basket before it is removed from the Plant spent fuel storage pool. Then, baskets are carried in a shielding container to a storage silo, where they remain until their final disposal. The silo system consists in a concrete cylinder of 2.80 m external diameter and 0.85 m thick, internally lined with a carbon steel cylinder of 9.5 mm thick. This structure is supported by a 0.60 m thick concrete slab. This work reviews the Condition Assessment of Embalse Spent Fuel Dry Storage Silos and was performed following the procedures implemented in the Embalse Refurbishment Project. A review of nondestructive and destructive methods is presented so as to assess the condition of concrete and carbon steel of this structure. Future tasks to be performed in the Spent Fuel Pool is presented. (author)

  18. Fabrication and Installation of Radiation Shielded Spent Fuel Fusion System

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soon Dal; Park, Yang Soon; Kim, Jong Goo; Ha, Yeong Keong; Song, Kyu Seok

    2010-02-15

    Most of the generated fission gases are retained in the fuel matrix in supersaturated state, thus alter the original physicochemical properties of the fuel. And some of them are released into free volume of a fuel rod and that cause internal pressure increase of a fuel rod. Furthermore, as extending fuel burnup, the data on fission gas generation(FGG) and fission gas release(FGR) are considered very important for fuel safety investigation. Consequently, it is required to establish an experimental facility for handling of highly radioactive sample and to develop an analytical technology for measurement of retained fission gas in a spent fuel. This report describes not only on the construction of a shielded glove box which can handle highly radioactive materials but also on the modifications and instrumentations of spent fuel fusion facilities and collection apparatuses of retained fission gas

  19. Spent nuclear fuel discharges from US reactors 1993

    Energy Technology Data Exchange (ETDEWEB)

    1995-02-01

    The Energy Information Administration (EIA) of the U.S. Department of Energy (DOE) administers the Nuclear Fuel Data Survey, Form RW-859. This form is used to collect data on fuel assemblies irradiated at commercial nuclear reactors operating in the United States, and the current inventories and storage capacities of those reactors. These data are important to the design and operation of the equipment and facilities that DOE will use for the future acceptance, transportation, and disposal of spent fuels. The data collected and presented identifies trends in burnup, enrichment, and spent nuclear fuel discharged form commercial light-water reactor as of December 31, 1993. The document covers not only spent nuclear fuel discharges; but also site capacities and inventories; canisters and nonfuel components; and assembly type characteristics.

  20. 77 FR 60482 - Yankee Atomic Electric Company; Yankee Rowe Independent Spent Fuel Storage Installation, Staff...

    Science.gov (United States)

    2012-10-03

    ... COMMISSION Yankee Atomic Electric Company; Yankee Rowe Independent Spent Fuel Storage Installation, Staff... for the storage of spent fuel in an Independent Spent Fuel Storage Installation (ISFSI) to persons... part 72 general license for storage of spent fuel and greater than Class C waste at the Yankee......

  1. 10 CFR 171.15 - Annual fees: Reactor licenses and independent spent fuel storage licenses.

    Science.gov (United States)

    2010-01-01

    ... only status and has spent fuel onsite, and each independent spent fuel storage 10 CFR part 72 licensee... onsite, and to each independent spent fuel storage 10 CFR part 72 licensee who does not hold a 10 CFR... NRC § 171.15 Annual fees: Reactor licenses and independent spent fuel storage......

  2. 10 CFR 72.230 - Procedures for spent fuel storage cask submittals.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Procedures for spent fuel storage cask submittals. 72.230... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Approval of Spent Fuel Storage Casks § 72.230 Procedures for spent fuel storage cask...

  3. 75 FR 33853 - Maine Yankee Atomic Power Company; Independent Spent Fuel Storage Installation; Issuance of...

    Science.gov (United States)

    2010-06-15

    ... COMMISSION Maine Yankee Atomic Power Company; Independent Spent Fuel Storage Installation; Issuance of... Manager, Division of Spent Fuel Storage and Transportation, ] Office of Nuclear Material Safety and..., to store spent nuclear fuel under a general license in an independent spent fuel storage...

  4. 77 FR 33005 - Connecticut Yankee Atomic Power Company; Haddam Neck Independent Spent Fuel Storage Installation...

    Science.gov (United States)

    2012-06-04

    ... COMMISSION Connecticut Yankee Atomic Power Company; Haddam Neck Independent Spent Fuel Storage Installation... license is issued for the storage of spent fuel in an Independent Spent Fuel Storage Installation (ISFSI... also holds a 10 CFR part 72 general license which allows storage of spent fuel and greater......

  5. International safeguards concerns of Spent Fuel Disposal Program

    International Nuclear Information System (INIS)

    The purpose of this paper is to stimulate discussions on the subjects of safeguarding large quantities of plutonium contained in spent fuels to be disposed of in geologic respositories. All the spent fuel disposal scenarios examined here pose a variety of safeguards problems, none of which are adequately addressed by the international safeguards community. The spent fuels from once-through fuel cycles in underground repositories would become an increasingly attractive target for diversion because of their plutonium content and decreasing radioactivity. Current design of the first geologic repository in the US will have the capacity to accommodate wastes equivalent to 70,000 Mt of uranium from commercial and defense fuel cycles. Of this, approximately 62,000 Mt uranium equivalent will be commerical spent fuel, containing over 500 Mt of plutonium. International safeguards commitments may require us to address the safeguards issues of disposing of such large quanities of plutonium in a geologic repository, which has the potential to become a plutonium mine in the future. This paper highlights several issues that should be addressed in the near term by US industries and the DOE before geologic repositories for spent fuels become a reality

  6. Spent fuel storage and waste management fuel cycle optimization using CAFCA

    International Nuclear Information System (INIS)

    Spent fuel storage modeling is at the intersection of nuclear fuel cycle system dynamics and waste management policy. A model that captures the economic parameters affecting used nuclear fuel storage location options, which complements fuel cycle economic assessment has been created using CAFCA (Code for Advanced Fuel Cycles Assessment) of MIT. Research has also expanded to the study on dependency of used nuclear fuel storage economics, environmental impact, and proliferation risk. Three options of local, regional, and national storage were studied. The preliminary product of this research is the creation of a system dynamics tool known as the Waste Management Module which provides an easy to use interface for education on fuel cycle waste management economic impacts. Storage options costs can be compared to literature values with simple variation available for sensitivity study. Additionally, a first of a kind optimization scheme for the nuclear fuel cycle analysis is proposed and the applications of such an optimization are discussed. The main tradeoff for fuel cycle optimization was found to be between economics and most of the other identified metrics. (authors)

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

    International Nuclear Information System (INIS)

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

  8. Studies on spent nuclear fuel evolution during storage

    Energy Technology Data Exchange (ETDEWEB)

    Rondinella, V.V.; Wiss, T.A.G.; Papaioannou, D.; Nasyrow, R. [European Commission Joint Research Centre, Karlsruhe (Germany). Inst. for Transuranium Elements

    2015-07-01

    Initially conceived to last only a few decades (40 years in Germany), extended storage periods have now to be considered for spent nuclear fuel due to the expanding timeline for the definition and implementation of the disposal in geologic repository. In some countries, extended storage may encompass a timeframe of the order of centuries. The safety assessment of extended storage requires predicting the behavior of the spent fuel assemblies and the package systems over a correspondingly long timescale, to ensure that the mechanical integrity and the required level of functionality of all components of the containment system are retained. Since no measurement of ''old'' fuel can cover the ageing time of interest, spent fuel characterization must be complemented by studies targeting specific mechanisms that may affect properties and behavior of spent fuel during extended storage. Tests conducted under accelerated ageing conditions and other relevant simulations are useful for this purpose. During storage, radioactive decay determines the overall conditions of spent fuel and generates heat that must be dissipated. Alpha-decay damage and helium accumulation are key processes affecting the evolution of properties and behavior of spent fuel. The radiation damage induced by a decay event during storage is significantly lower than that caused by a fission during in-pile operation: however, the duration of the storage is much longer and the temperature levels are different. Another factor potentially affecting the mechanical integrity of spent fuel rods during storage and handling / transportation is the behavior of hydrogen present in the cladding. At the Institute for Transuranium Elements, part of the Joint Research Centre of the European Commission, spent fuel alterations as a function of time and activity are monitored at different scales, from the microstructural level (defects and lattice parameter swelling) up to macroscopic properties such as

  9. Development of Pyrochemical Reprocessing of the Spent Nuclear Fuel and Prospects of Closed Fuel Cycle

    Directory of Open Access Journals (Sweden)

    R. Tulackova

    2007-01-01

    Full Text Available Molten-Salt Reactor (MSR is a design of an advanced reactor system from the GEN IV family working in thermal or epithermal neutron spectrum and using thorium or transuranium fuel in the form of molten fluorides. It is based on the experience with the development of the molten-salt reactor technology in the Oak-Ridge National Laboratory in the United States. The MSR fuel cycle with integrated reprocessing represents one of the potential ways both for significant decrease of total amount of radioactive wastes for final deposition and for utilization of nuclear energy for electricity and heat production as effectively as possible. There are two pyrochemical reprocessing techniques studied in NRI Rez plc which are considered to be applied both for reprocessing of already existing spent fuel and for preparation and „on-line“ reprocessing of MSR fuel: (i the Fluoride Volatility Method (FVM, which performs chemical conversion of spent thermal oxide fuel components into fluorides and their consequent separation by means of their different volatility, thermal stability and chemical affinity to various sorbents; and (ii electrochemical separation of the actinides (Ans and fission products (FP, represented mainly by lanthanides (Lns, from each other by electrolytic deposition method on solid cathode in molten fluoride media.

  10. Spent fuel in geologic repositories: Swedish aspects of safeguards

    International Nuclear Information System (INIS)

    When the safeguards system was defined and established, spent fuel was destined for reprocessing. The deposition without reprocessing represents a conceptual change for the back end of the fuel cycle in that nuclear material is not intended for further use in any nuclear application. Much effort has been devoted to finding a solution whereby the spent fuel can be left unattended forever. The disposal of spent fuel, and specifically the need to protect humans and the environment in the distant future, is given particular attention in all countries engaged in nuclear generation. The system considered for spent-fuel disposal in Sweden includes the encapsulation of the spent fuel elements in a corrosion-resistant tight canister and disposal of the canister at ∼500m depth in Swedish bedrock. The canister will be made of thick copper (for corrosion resistance) and will have an internal steel container (for mechanical strength). The encapsulation is planned to be performed in a new facility to be built adjacent to the central interim storage facility, CLAB

  11. Shielding calculations for spent CANDU fuel transport cask

    International Nuclear Information System (INIS)

    CANDU spent fuel discharged from the reactor core contains Pu, so, a special attention must be focussed into two directions: tracing for the fuel reactivity in order to prevent critical mass formation and personnel protection during the spent fuel manipulation. Shielding analyses, an essential component of the nuclear safety, take into account the difficulties occurred during the manipulation, transport and storage of spent fuel bundles, both for personnel protection and impact on the environment. The main objective here consists in estimations on radiation doses in order to reduce them under specified limit values. In order to perform the shielding calculations for the spent fuel transport cask three different codes were used: XSDOSE code and MORSE-SGC code, both incorporated in the SCALE4.4a system, and PELSHIE-3 code, respectively. As source of radiation one spent standard CANDU fuel bundle was used. All the geometrical and material data, related to the transport casks, were considered according to the shipping cask type B model, whose prototype has been realized and tested in the Institute for Nuclear Research Pitesti. The radial gamma dose rates estimated to the cask wall and in air, at different distances from the cask, are presented together with a comparison between the dose rates values obtained by all three recipes of shielding calculations. (authors)

  12. Shippingport Spent Fuel Canister (SSFC) Design Report Project W-518

    Energy Technology Data Exchange (ETDEWEB)

    JOHNSON, D.M.

    2000-01-27

    The SSFC Design Report Describes A spent fuel canister for Shippingport Core 2 blanket fuel assemblies. The design of the SSFC is a minor modification of the MCO. The modification is limited to the Shield Plug which remains unchanged with regard to interfaces with the canister shell. The performance characteristics remain those for the MCO, which bounds the payload of the SSFC.

  13. Spent fuel receipt and storage at the Morris Operation

    International Nuclear Information System (INIS)

    Operating and maintenance activities in an independent spent fuel storage facility are described, and current regulations governing such activities are summarized. This report is based on activities at General Electric's licensed storage facility located near Morris, Illinois, and includes photographs of cask and fuel handling equipment used during routine operations

  14. Separator assembly for use in spent nuclear fuel shipping cask

    Science.gov (United States)

    Bucholz, James A.

    1983-01-01

    A separator assembly for use in a spent nuclear fuel shipping cask has a honeycomb-type wall structure defining parallel cavities for holding nuclear fuel assemblies. Tubes formed of an effective neutron-absorbing material are embedded in the wall structure around each of the cavities and provide neutron flux traps when filled with water.

  15. Implementation of burnup credit in PWR spent fuel storage pools

    International Nuclear Information System (INIS)

    Implementation of burnup credit in spent fuel storage of LWR fuel at nuclear power plants is approved in Germany since the beginning of 2000. The burnup credit methods applied have to comply with the newly developed German criticality safety standard DIN 25471 passed in November 1999 and published in September 2000, cp. (orig.)

  16. Spent fuel receipt and storage at the Morris Operation

    Energy Technology Data Exchange (ETDEWEB)

    Astrom, K A; Eger, K J

    1978-06-01

    Operating and maintenance activities in an independent spent fuel storage facility are described, and current regulations governing such activities are summarized. This report is based on activities at General Electric's licensed storage facility located near Morris, Illinois, and includes photographs of cask and fuel handling equipment used during routine operations.

  17. The Sliding and Overturning Analysis of Spent Fuel Storage Rack Based on Dynamic Analysis Model

    OpenAIRE

    Liu, Yu; Lu, Daogang; Wang, Yuanpeng; LIU, HONGDA

    2016-01-01

    Spent fuel rack is the key equipment for the storage of spent fuel after refueling. In order to investigate the performance of the spent fuel rack under the earthquake, the phenomena including sliding, collision, and overturning of the spent fuel rack were studied. An FEM model of spent fuel rack is built to simulate the transient response under seismic loading regarding fluid-structure interaction by ANSYS. Based on D’Alambert’s principle, the equilibriums of force and momentum were establis...

  18. Lessons learned from a review of international approaches to spent fuel management

    OpenAIRE

    Hambley David; Laferrere Alice; Walters W. Steven; Hodgson Zara; Wickham Steven; Richardson Phillip

    2016-01-01

    Worldwide, a variety of approaches to the management of spent fuel have been adopted. A review of approaches adopted internationally was undertaken to inform decision making on spent fuel management in UK. The review surveyed spent fuel storage and disposal practices, standards, trends and recent developments in 16 countries and carried out more detailed studies into the evolution of spent fuel storage and disposal strategies in four countries. The review highlighted that: (1) spent fuel mana...

  19. Storage of spent fuel from power reactors. Proceedings of a symposium

    International Nuclear Information System (INIS)

    The symposium 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 an international cooperation in this area should take. Dominant message retrieved from the symposium are that the primary spent fuel management solution for the next decades will be interim storage, the duration of time of interim storage becomes longer than earlier anticipated and the storage facilities will have to be designed for receiving also 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 different countries. The changes in nuclear policy and philosophy across the world, and practical considerations, have made interim storage a real necessity in the nuclear power industry. This is being addressed adequately by utilities, vendors and regulators alike

  20. An Applied Study of Implementation of the Advanced Decommissioning Costing Methodology for Intermediate Storage Facility for Spent Fuel in Studsvik, Sweden with special emphasis to the application of the Omega code

    Energy Technology Data Exchange (ETDEWEB)

    Kristofova, Kristina; Vasko, Marek; Daniska, Vladimir; Ondra, Frantisek; Bezak, Peter [DECOM Slovakia, spol. s.r.o., J. Bottu 2, SK-917 01 Trnava (Slovakia); Lindskog, Staffan [Swedish Nuclear Power Inspectorate, Stockholm (Sweden)

    2007-01-15

    The presented study is focused on an analysis of decommissioning costs for the Intermediate Storage Facility for Spent Fuel (FA) facility in Studsvik prepared by SVAFO and a proposal of the advanced decommissioning costing methodology application. Therefore, this applied study concentrates particularly in the following areas: 1. Analysis of FA facility cost estimates prepared by SVAFO including description of FA facility in Studsvik, summarised input data, applied cost estimates methodology and summarised results from SVAFO study. 2. Discussion of results of the SVAFO analysis, proposals for enhanced cost estimating methodology and upgraded structure of inputs/outputs for decommissioning study for FA facility. 3. Review of costing methodologies with the special emphasis on the advanced costing methodology and cost calculation code OMEGA. 4. Discussion on implementation of the advanced costing methodology for FA facility in Studsvik together with: - identification of areas of implementation; - analyses of local decommissioning infrastructure; - adaptation of the data for the calculation database; - inventory database; and - implementation of the style of work with the computer code OMEGA.

  1. Development of code SFINEL (Spent fuel integrity evaluator)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yong Soo; Min, Chin Young; Ohk, Young Kil; Yang, Yong Sik; Kim, Dong Ju; Kim, Nam Ku [Hanyang University, Seoul (Korea)

    1999-01-01

    SFINEL code, an integrated computer program for predicting the spent fuel rod integrity based on burn-up history and major degradation mechanisms, has been developed through this project. This code can sufficiently simulate the power history of a fuel rod during the reactor operation and estimate the degree of deterioration of spent fuel cladding using the recently-developed models on the degradation mechanisms. SFINEL code has been thoroughly benchmarked against the collected in-pile data and operating experiences: deformation and rupture, and cladding oxidation, rod internal pressure creep, then comprehensive whole degradation process. (author). 75 refs., 51 figs., 5 tabs.

  2. Spent LWR fuel leach tests: Waste Isolation Safety Assessment program

    International Nuclear Information System (INIS)

    Spent light-water-reactor (LWR) fuels with burnups of 54.5, 28 and 9 MWd/kgU were leach-tested in deionized water at 250C. Fuel burnup has no apparent effect on the calculated leach rates based upon the behavior of 137Cs and 239+240Pu. A leach test of 54.5 MWd/kgU spent fuel in synthetic sea brine showed that the cesium-based leach rate is lower in sea brine than in deionized water. A rise in the leach rate was observed after approximately 600 d of cumulative leaching. During the rise, the leach rate for all the measured radionuclides become nearly equal. Evidence suggests that exposure of new surfaces to the leachant may cause the increase. As a result, experimental work to study leaching mechanisms of spent fuel has been initiated. 22 figures

  3. Handling of final storage of unreprocessed spent nuclear fuel

    International Nuclear Information System (INIS)

    In this report the various facilities incorporated in the proposed handling chain for spent fuel from the power stations to the final repository are discribed. Thus the geological conditions which are essential for a final repository is discussed as well as the buffer and canister materials and how they contribute towards a long-term isolation of the spent fuel. Furthermore one chapter deals with leaching of the deposited fuel in the event that the canister is penetrated as well as the transport mechanisms which determine the migration of the radioactive substances through the buffer material. The dispersal processes in the geosphere and the biosphere are also described together with the transfer mechanisms to the ecological systems as well as radiation doses. Finally a summary is given of the safety analysis of the proposed method for the handling and final storage of the spent fuel. (E.R.)

  4. Disposal of spent nuclear fuel in geological repositories

    International Nuclear Information System (INIS)

    Plutonium has been and still is produced in the world's many civil and military nuclear programs. Although the nuclear establishments of several countries, most noticeably, Japan, France, Greta Britain and Russia, are advocating the recycling of plutonium, there are also two nuclear waste disposition 'strategies' that involve the direct final disposal of plutonium in geological repositories: the direct final disposal of spent fuel or spent civil mixed oxide fuel when option to reprocess it has been rejected; and the final disposal of 'spent fuel standard' excess weapon plutonium when it has been 'anti-reprocessed' or burned as 'military mixed oxide fuel. It is important to understand that the magnitude of long-term safeguards concern of plutonium disposal in geological repositories depends very much on the future development of nuclear energy application. It might be decided to solve the global plutonium predicament by making global stocks of plutonium 'irretrievable', thus removing the needs for safeguards

  5. Neutron resonance transmission analysis of reactor spent fuel assemblies

    International Nuclear Information System (INIS)

    A method called Neutron Resonance Transmission Analysis (NRTA) is under study which would use a pulsed neutron beam for nondestructive isotopic assay of a complete spent fuel assembly. Neutrons removed from the collimated beam by absorption or scattering in the resonances of the various isotopes in the spent fuel appear as dips in the neutron transmission. The method is completely insensitive to matrix materials such as oxide, fuel cladding, and other structural members. Measurements on spent fuel buttons using the NBS linac as a pulsed neutron source demonstrate a high accuracy capability for the isotopes 234235236238U, 239240241242Pu, 241Am, 243Am, and several fission products. The NRTA method offers high speed and modest operational cost, and it can be implemented with commercially available medical or radiographic γ-ray generators adapted for neutron production. (Auth.)

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

    International Nuclear Information System (INIS)

    term issues involved in spent fuel storage, including long term caretaking/refurbishment of facilities or transition from one option to another in consideration of fuel technology evolution, changes in long term policies, market-based influences, as well as changes in regulatory criteria, deserve careful consideration. Although it can be said that competitive services for AFR storage are currently available from the market, it is often not evident how to choose the suitable option or technology for storage because of the complex issues involved in the decision, including a range of future uncertainties. Furthermore, focal issues in selecting an AFR storage facility can shift with time as spent fuel management policies, strategies and technologies advance and can change from one country to another due to considerations particular to those countries. In addition, as some common issues such as the trend toward privatization of former public enterprises in this sector, together with the issue of public involvement, some profound impacts on the nuclear industry in general and spent fuel management in particular may be expected. This TECDOC attempts to provide information in the approach to select AFR facilities for spent fuel storage. This TECDOC is an output of a series of meetings, which started with an Advisory Group Meeting in 2000, and was followed by three consultants meetings in 2001, 2002 and 2003

  7. Characterization of spent fuel approved testing material---ATM-105

    International Nuclear Information System (INIS)

    The characterization data obtained to data are described for Approved Testing Material 105 (ATM-105), which is spent fuel from Bundles CZ346 and CZ348 of the Cooper Nuclear Power Plant, a boiling-water reactor. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) Program. ATM-105 consists of 88 full-length irradiated fuel rods with rod-average burnups of about 2400 GJ/kgM (28 MWd/kgM) and expected fission gas release of about 1%. Characterization data include (1) descriptions of as-fabricated fuel design, irradiation history, and subsequent storage and handling; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) special fuel studies involving analytical transmission electron microscopy (AEM); (6) calculated nuclide inventories and radioactivities in the fuel and cladding; and (7) radiochemical analyses of the fuel and cladding. Additional analyses of the fuel are being conducted and will be included in planned revisions of this report

  8. Characterization of spent fuel approved testing material---ATM-105

    Energy Technology Data Exchange (ETDEWEB)

    Guenther, R.J.; Blahnik, D.E.; Campbell, T.K.; Jenquin, U.P.; Mendel, J.E.; Thomas, L.E.; Thornhill, C.K.

    1991-12-01

    The characterization data obtained to data are described for Approved Testing Material 105 (ATM-105), which is spent fuel from Bundles CZ346 and CZ348 of the Cooper Nuclear Power Plant, a boiling-water reactor. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) Program. ATM-105 consists of 88 full-length irradiated fuel rods with rod-average burnups of about 2400 GJ/kgM (28 MWd/kgM) and expected fission gas release of about 1%. Characterization data include (1) descriptions of as-fabricated fuel design, irradiation history, and subsequent storage and handling; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) special fuel studies involving analytical transmission electron microscopy (AEM); (6) calculated nuclide inventories and radioactivities in the fuel and cladding; and (7) radiochemical analyses of the fuel and cladding. Additional analyses of the fuel are being conducted and will be included in planned revisions of this report.

  9. Analysis of spent fuel under long term dry storage conditions

    International Nuclear Information System (INIS)

    The dry storage is a safe and economic intermediate solution contributing to the reduction of the inventory in the used fuel pools while awaiting the development of a back-end fuel strategy: direct disposal or reprocessing and recycling. The dry storage facility is considered as a highly resistant and passive system and experience accumulated over the last years confirmed this statement. One of the prerequisites for confirming the safety of dry spent fuel storage technologies is the ability to predict the spent fuel performance during period of storage. The paper presents the methodology for evaluation of spent fuel behavior during the long term dry storage. The methodology includes two world wide used code systems: SCALE and TRANSURANUS. The cladding outer temperature in the closed cask has to be calculated by some thermo-hydraulic code and then used by the TRANSURANUS as initial conditions. This could be an object of future studies. The residual heat rating and fast neutron flux were assessed by the SCALE 6 code system and taken as an input data for the fuel performance calculations. During the three periods of the fuel life (irradiation, cooling and dry storage), the fast neutron flux, linear heat rate and the cooling conditions are considerably different. The TRANSURANUS code allows accounting for the change of the coolant nature (RESTART mode) which is important advantage of the code. The developed approach of dry storage treatment was demonstrated on the base of a WWER-440 fuel rod, irradiated four cycles in the Kozloduy NPP, Unit 4. The results, obtained in these first analyses, give grounds to accept that the preliminary assessment of the spent fuel properties are applicable to the problem of analyzing spent fuel behavior under long term storage - both wet and dry one. (authors)

  10. Track 9: fuel cycle, spent fuel, decommissioning, and waste management. Dry storage of commercial spent nuclear fuel. Panel Discussion

    International Nuclear Information System (INIS)

    Full text of publication follows: Currently, dry storage of spent nuclear fuel is a mature technology with a firm technical basis when the burnup of fuel is 45 GWd/tonne U for up to 100-yr duration. A technical basis for the safe extension of dry storage to the higher burnups and longer times needs to be established. In particular, the expected behavior of cask and fuel materials must be established. Experts from the U.S. Nuclear Regulatory Commission, the American Society for Testing and Materials, utilities storing fuel, the Electric Power Research Institute, and national laboratories will discuss the current status of that database, what data need to be obtained, what programs are in place to obtain data to augment the basis, and the results of those programs. (authors)

  11. Spent fuel sabotage aerosol ratio program : FY 2004 test and data summary.

    Energy Technology Data Exchange (ETDEWEB)

    Brucher, Wenzel (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Koch, Wolfgang (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Pretzsch, Gunter Guido (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Loiseau, Olivier (Institut de Radioprotection et de Surete Nucleaire, France); Mo, Tin (U.S. Nuclear Regulatory Commission, Washington, DC); Billone, Michael C. (Argonne National Laboratory, Argonne, IL); Autrusson, Bruno A. (Institut de Radioprotection et de Surete Nucleaire, France); Young, F. I. (U.S. Nuclear Regulatory Commission, Washington, DC); Coats, Richard Lee; Burtseva, Tatiana (Argonne National Laboratory, Argonne, IL); Luna, Robert Earl; Dickey, Roy R.; Sorenson, Ken Bryce; Nolte, Oliver (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Thompson, Nancy Slater (U.S. Department of Energy, Washington, DC); Hibbs, Russell S. (U.S. Department of Energy, Washington, DC); Gregson, Michael Warren; Lange, Florentin (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Molecke, Martin Alan; Tsai, Han-Chung (Argonne National Laboratory, Argonne, IL)

    2005-07-01

    This multinational, multi-phase spent fuel sabotage test program is quantifying the aerosol particles produced when the products of a high energy density device (HEDD) interact with and explosively particulate test rodlets that contain pellets of either surrogate materials or actual spent fuel. This program has been underway for several years. This program provides data that are relevant to some sabotage scenarios in relation to spent fuel transport and storage casks, and associated risk assessments. The program also provides significant technical and political benefits in international cooperation. We are quantifying the Spent Fuel Ratio (SFR), the ratio of the aerosol particles released from HEDD-impacted actual spent fuel to the aerosol particles produced from surrogate materials, measured under closely matched test conditions, in a contained test chamber. In addition, we are measuring the amounts, nuclide content, size distribution of the released aerosol materials, and enhanced sorption of volatile fission product nuclides onto specific aerosol particle size fractions. These data are the input for follow-on modeling studies to quantify respirable hazards, associated radiological risk assessments, vulnerability assessments, and potential cask physical protection design modifications. This document includes an updated description of the test program and test components for all work and plans made, or revised, during FY 2004. It also serves as a program status report as of the end of FY 2004. All available test results, observations, and aerosol analyses plus interpretations--primarily for surrogate material Phase 2 tests, series 2/5A through 2/9B, using cerium oxide sintered ceramic pellets are included. Advanced plans and progress are described for upcoming tests with unirradiated, depleted uranium oxide and actual spent fuel test rodlets. This spent fuel sabotage--aerosol test program is coordinated with the international Working Group for Sabotage Concerns of

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

    International Nuclear Information System (INIS)

    Full text: The properties of light water reactor (LWR) fuel rods and assemblies are altered in service due to irradiation. Some of these alterations render the fuel unsuitable for emplacement in casks used for storage or transportation without special handling. Title 10 (Energy) of the U.S. Code of Federal Regulations Part 72 (storage) and Part 71 (transportation) establish direct requirements for the behavior expected of spent fuel. In particular, retrievability and prevention of gross breaches are required in storage and no reconfiguration of the fuel is allowed during normal transport. In addition, in the process of meeting other regulations related to criticality, shielding, and containment, the cask designers may need to place additional requirements on the behavior of the fuel. The definition of damaged fuel might be based on the ability of the fuel to perform in a manner such that the direct regulatory requirements and the onus placed on the fuel by the cask designer are met. Fuels that have alterations that do not permit it to perform its required safety function, without special handling, should be regarded as damaged. Since the requirements placed on the fuel may vary during phases of the fuel cycle, the potential exists for independent definitions to co-exist for interim dry storage, transport, and final disposal in a geologic repository. The United States Nuclear Regulatory Commission's (USNRC) Spent Fuel Program Office (SFPO) has provided guidance in defining damaged fuel in Interim Staff Guidance (ISG) -1. This guidance is similar to that being developed by the American National Standards Institute (ANSI). Neither of these documents provides the logic behind the definition of damaged fuel. This paper will discuss the requirements placed on the fuel for dry interim storage and transportation and the ways that these requirements drive the definition of damaged spent fuel. Examples will be given illustrating the methodology. (author)

  13. Studies and research concerning BNFP. Nuclear spent fuel transportation studies

    International Nuclear Information System (INIS)

    Currently, there are a number of institutional problems associated with the shipment of spent fuel assemblies from commercial nuclear power plants: new and conflicting regulations, embargoing of certain routes, imposition of transport safeguards, physical security in-transit, and a lack of definition of when and where the fuel will be moved. This report presents a summary of these types and kinds of problems. It represents the results of evaluations performed relative to fuel receipt at the Barnwell Nuclear Fuel Plant. Case studies were made which address existing reactor sites with near-term spent fuel transportation needs. Shipment by either highway, rail, water, or intermodal water-rail was considered. The report identifies the impact of new regulations and uncertainty caused by indeterminate regulatory policy and lack of action on spent fuel acceptance and storage. This stagnant situation has made it impossible for industry to determine realistic transportation scenarios for business planning and financial risk analysis. A current lack of private investment in nuclear transportation equipment is expected to further prolong the problems associated with nuclear spent fuel and waste disposition. These problems are expected to intensify in the 1980's and in certain cases will make continuing reactor plant operation difficult or impossible

  14. Long term Integrity of PWR Spent Fuel in Dry Storage

    International Nuclear Information System (INIS)

    The newly established organization KRMC (Korea radioactive waste management corporation) which is responsible for all kinds of radioactive waste generated in the Republic of Korea launched the PWR spent fuel dry storage research project in June 2009. This project has objectives to develop a storage system and evaluate the integrity of PWR fuel in dry storage. The project consists of three steps. At first step, it would develop own degradation models by referring to pre-exist good models and develop the hot test scenarios. Second step, test facilities would be constructed and used for testing the degradation behaviour in each mechanisms and in total. As a final step, total evaluation code would be developed by integrating each degradation model produced in the first step and the test data produced in the second step. All the activities would be summarized into a report and applied to licensing work. The Republic of Korea PWR spent fuels have unique characteristics of various fuel types (array type, clad material) and high capacity factor (maximum usage of fuel which is bad for integrity). These facts could impact on the research ranges of experimental data needed for degradation evaluation. In this research, spent fuel performance data concerning long term dry storage will be analysed and the major degradation mechanisms like creep and hydride behaviour will be studied and proposed for Korean PWR spent fuels

  15. Separation of actinides from spent nuclear fuel: A review.

    Science.gov (United States)

    Veliscek-Carolan, Jessica

    2016-11-15

    This review summarises the methods currently available to extract radioactive actinide elements from solutions of spent nuclear fuel. This separation of actinides reduces the hazards associated with spent nuclear fuel, such as its radiotoxicity, volume and the amount of time required for its' radioactivity to return to naturally occurring levels. Separation of actinides from environmental water systems is also briefly discussed. The actinide elements typically found in spent nuclear fuel include uranium, plutonium and the minor actinides (americium, neptunium and curium). Separation methods for uranium and plutonium are reasonably well established. On the other hand separation of the minor actinides from lanthanide fission products also present in spent nuclear fuel is an ongoing challenge and an area of active research. Several separation methods for selective removal of these actinides from spent nuclear fuel will be described. These separation methods include solvent extraction, which is the most commonly used method for radiochemical separations, as well as the less developed but promising use of adsorption and ion-exchange materials.

  16. Microbial biofilm growth on irradiated, spent nuclear fuel cladding

    International Nuclear Information System (INIS)

    A fundamental criticism regarding the potential for microbial influenced corrosion in spent nuclear fuel cladding or storage containers concerns whether the required microorganisms can, in fact, survive radiation fields inherent in these materials. This study was performed to unequivocally answer this critique by addressing the potential for biofilm formation, the precursor to microbial-influenced corrosion, in radiation fields representative of spent nuclear fuel storage environments. This study involved the formation of a microbial biofilm on irradiated spent nuclear fuel cladding within a hot cell environment. This was accomplished by introducing 22 species of bacteria, in nutrient-rich media, to test vessels containing irradiated cladding sections and that was then surrounded by radioactive source material. The overall dose rate exceeded 2 Gy/h gamma/beta radiation with the total dose received by some of the bacteria reaching 5 x 103 Gy. This study provides evidence for the formation of biofilms on spent-fuel materials, and the implication of microbial influenced corrosion in the storage and permanent deposition of spent nuclear fuel in repository environments

  17. Separation of actinides from spent nuclear fuel: A review.

    Science.gov (United States)

    Veliscek-Carolan, Jessica

    2016-11-15

    This review summarises the methods currently available to extract radioactive actinide elements from solutions of spent nuclear fuel. This separation of actinides reduces the hazards associated with spent nuclear fuel, such as its radiotoxicity, volume and the amount of time required for its' radioactivity to return to naturally occurring levels. Separation of actinides from environmental water systems is also briefly discussed. The actinide elements typically found in spent nuclear fuel include uranium, plutonium and the minor actinides (americium, neptunium and curium). Separation methods for uranium and plutonium are reasonably well established. On the other hand separation of the minor actinides from lanthanide fission products also present in spent nuclear fuel is an ongoing challenge and an area of active research. Several separation methods for selective removal of these actinides from spent nuclear fuel will be described. These separation methods include solvent extraction, which is the most commonly used method for radiochemical separations, as well as the less developed but promising use of adsorption and ion-exchange materials. PMID:27427893

  18. Spent nuclear fuel management. Moving toward a century of spent fuel management: A view from the halfway mark

    International Nuclear Information System (INIS)

    Full text: A half-century ago, President Eisenhower in his 1953 'Atoms for Peace' speech, offered nuclear technology to other nations as part of a broad nuclear arms control initiative. In the years that followed, the nuclear power generation capabilities of many nations has helped economic development and contributed to the prosperity of the modern world. The growth of nuclear power, while providing many benefits, has also contributed to an increasing global challenge over safe and secure spent fuel management. Over 40 countries have invested in nuclear energy, developing over 400 nuclear power reactors. Nuclear power supplies approximately 16% of the global electricity needs. With the finite resources and challenges of fossil fuels, nuclear power will undoubtedly become more prevalent in the future, both in the U.S. and abroad. We must address this inevitability with new paradigms for managing a global nuclear future. Over the past fifty years, the world has come to better understand the strong interplay between all elements of the nuclear fuel cycle, global economics, and global security. In the modern world, the nuclear fuel cycle can no longer be managed as a simple sequence of technological, economic and political challenges. Rather it must be seen, and managed, as a system of strongly interrelated challenges. Spent fuel management, as one element of the nuclear fuel system, cannot be relegated to the back-end of the fuel cycle as only a disposal or storage issue. There exists a wealth of success and experience with spent fuel management over the past fifty years. We must forge this experience with a global systems perspective, to reshape the governing of all aspects of the nuclear fuel cycle, including spent fuel management. This session will examine the collective experience of spent fuel management enterprises, seeking to shape the development of new management paradigms for the next fifty years. (author)

  19. Present status and prospect of spent fuel transportation

    International Nuclear Information System (INIS)

    Problems linked with spent fuel transportation in Japan, where there are 35 NPPs in operation, are considered. Every year about 500 t U are shipped to fuel reprocessing plants in Japan, as well as in France and UK. Four kinds of casks: HZ, EXCELLOX, TN and TK - are used for this purpose. By the mid-1990's it is suggested to build in Japan fuel reprocessing plant with capacity of 800 t U per year

  20. Characterization of spent fuel approved testing material--ATM-104

    Energy Technology Data Exchange (ETDEWEB)

    Guenther, R.J.; Blahnik, D.E.; Jenquin, U.P.; Mendel, J.E.; Thomas, L.E.; Thornhill, C.K.

    1991-12-01

    The characterization data obtained to date are described for Approved Testing Material 104 (ATM-104), which is spent fuel from Assembly DO47 of the Calvert Cliffs Nuclear Power Plant (Unit 1), a pressurized-water reactor. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) Program. ATM-104 consists of 128 full-length irradiated fuel rods with rod-average burnups of about 42 MWd/kgM and expected fission gas release of about 1%. A variety of analyses were performed to investigate cladding characteristics, radionuclide inventory, and redistribution of fission products. Characterization data include (1) fabricated fuel design, irradiation history, and subsequent storage and handling history; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) special fuel studies involving analytical transmission electron microscopy (AEM) and electron probe microanalyses (EPMA); (6) calculated nuclide inventories and radioactivities in the fuel and cladding; and (7) radiochemical analyses of the fuel and cladding.

  1. 77 FR 37937 - License Renewal Application for Prairie Island Nuclear Generating Plant Independent Spent Fuel...

    Science.gov (United States)

    2012-06-25

    ..., Division of Spent Fuel Storage and Transportation, Office of Nuclear Material Safety and Safeguards, U.S..., possession, storage and transfer of spent fuel, reactor-related Greater than Class C (GTCC) waste and other radioactive materials associated with spent fuel storage at the PINGP site-specific Independent Spent......

  2. The State of the Art of the Decladding Method for the Spent Nuclear Fuel Rods

    International Nuclear Information System (INIS)

    Our country's energy consumption is increasing day after day even though it relies on imports for more than 95 percent of its energy needs. In this circumstances, the atomic energy is a promising alternative to solve the problem of an energy security and an environmental preservation simultaneously. However, nuclear power produces spent fuel which is a highly radioactive waste. For a reliable and effective management of the spent fuel, the ACP(Advanced Spent Fuel Conditioning Process) is being developed at the KAERI. As a first state equipment of the ACP, a decladding machine is used to separate spent fuel rod into the UO2 pellets and hulls. This technical report aims to analyze existing decladding methods, and then, find a suitable decladding mechanism for the ACP. Many studies on the decladding of spent fuel can be categorized two approaches: chemical approach and mechanical one. In this report, we concentrated on the mechanical decladding approach. We developed engineering scale decladding device(20 kgHM/batch) and evaluated the performance through the verification experiments. We expect that this technical report helps in developing a scale-up equipment and technology

  3. K Basin spent nuclear fuel characterization

    International Nuclear Information System (INIS)

    The results of the characterization efforts completed for the N Reactor fuel stored in the Hanford K Basins were Collected and summarized in this single referencable document. This summary provides a ''road map'' for what was done and the results obtained for the fuel characterization program initiated in 1994 and scheduled for completion in 1999 with the fuel oxidation rate measurement under moist inert atmospheres

  4. Criticality safety analysis of WWER-1000 spent nuclear fuel storage

    International Nuclear Information System (INIS)

    Nuclear safety of spent nuclear fuel management is ensured by implementation of the basic safety functions: providing subcriticality, residual heat removal and retention of radioactive products within the physical barriers. To ensure subcriticality during both normal operation and design basis accidents the effective multiplication factor of neutrons Keff must be lower than 0.95. An evaluation of criticality of spent fuel facilities have been made by the modular code system SCALE. The basic calculations are performed with version 6.1 and are validated with version 6.0 of the code system. Spent fuel assemblies type TVSA are modeled as they are representative for WWER-1000 nuclear fuel and cover the characteristics of the earlier modifications of the fuel assemblies. The modeling of the spent fuel containers and equipment is in accordance with actual geometric dimensions and material composition. In all performed calculations, the results demonstrate that the criticality safety criteria are achieved and the effective multiplication factor Keff is lower than the regulatory requirements. (authors)

  5. Spent fuel and radioactive waste inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    Current inventories and characteristics of commercial spent fuels and both commercial and US Department of Energy (DOE) radioactive wastes were compiled through December 31, 1983, based on the most reliable information available from government sources and the open literature, technical reports, and direct contacts. Future waste and spent fuel to be generated over the next 37 years and characteristics of these materials are also presented, consistent with the latest DOE/Energy Information Administration (EIA) or projection of US commercial nuclear power growth and expected defense-related and private industrial and institutional activities. Materials considered, on a chapter-by-chapter basis, are: spent fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, airborne waste, remedial action waste, and decommissioning waste. For each category, current and projected inventories are given through the year 2020, and the radioactivity and thermal power are calculated, based on reported or calculated isotopic compositions. 48 figures, 107 tables

  6. Mission Need Statement: Idaho Spent Fuel Facility Project

    Energy Technology Data Exchange (ETDEWEB)

    Barbara Beller

    2007-09-01

    Approval is requested based on the information in this Mission Need Statement for The Department of Energy, Idaho Operations Office (DOE-ID) to develop a project in support of the mission established by the Office of Environmental Management to "complete the safe cleanup of the environmental legacy brought about from five decades of nuclear weapons development and government-sponsored nuclear energy research". DOE-ID requests approval to develop the Idaho Spent Fuel Facility Project that is required to implement the Department of Energy's decision for final disposition of spent nuclear fuel in the Geologic Repository at Yucca Mountain. The capability that is required to prepare Spent Nuclear Fuel for transportation and disposal outside the State of Idaho includes characterization, conditioning, packaging, onsite interim storage, and shipping cask loading to complete shipments by January 1,2035. These capabilities do not currently exist in Idaho.

  7. Licensing Procedures for Interim Storage of Spent Fuel in Germany

    International Nuclear Information System (INIS)

    In accordance with the waste management concept in Germany spent fuel is stored in interim storage facilities for 40 years until disposal in a geological repository. The storage concept bases on dry storage of the spent fuel in metallic transport and storage casks, standing upright in halls of reinforced concrete. Storage of spent fuel as well as significant modifications of the storage require a license according to art. 6 of the Atomic Energy Act. The Federal Office for Radiation Protection (Bundesamt für Strahlenschutz - BfS) is the competent licensing authority. The mode of the licensing procedure — whether formalized or non-formalized — depends on the necessity to carry out an environmental impact assessment. Formalized licensing procedures include a public participation procedure. In the following, the licensing prodecures are illustrated and a short overview over the current licensing procedures conducted by BfS is given. (author)

  8. A GIS based methodology for nuclear reactor spent fuel disposal

    International Nuclear Information System (INIS)

    This article aims at studying the use of Geographical Information Systems for selecting a site for radioactive waste disposal of spent fuel generated by operation of Angra 1 and 2 nuclear power stations, in order to provide additional means for solving this problem in Brazil. This spent fuel continues to generate decay heat and radiation after its use in power stations. The disposal should be done in such a way as to isolate the nuclear spent fuel from people and the environment, protecting them from the heat and radioactivity for a long period of time. After elaboration of a database containing geological, hydrological, tectonic, weather, transport, conservation unit, amongst other information, one intends to combine these information, and make comparisons using preset criteria, in order to indicate the most adequate sites for disposal. (author)

  9. Expected reaction to various options for spent-fuel storage

    International Nuclear Information System (INIS)

    This paper discusses the reactions that might be expected in response to the selection of various options for the storage of spent nuclear fuel. The need for interim dry storage of commercial spent nuclear fuel is increasing as the reactor spent fuel pools become filled. Until the repository for final disposal becomes available, there are several interim storage options available. The paper discusses the features which make each option attractive or unattractive to the public. This paper draws on the experience of the DOE with the Indian Tribes and communities that expressed interest in an MRS facility, past experience in attempting to site an MRS facility at Oak Ridge, Tennessee, the experience of utilities in developing dry at-reactor storage, and other relevant experience. For some options, the discussion of potential reactions is of necessity partly speculative

  10. A burnup credit calculation methodology for PWR spent fuel transportation

    International Nuclear Information System (INIS)

    A burnup credit calculation methodology for PWR spent fuel transportation has been developed and validated in CEA/Saclay. To perform the calculation, the spent fuel composition are first determined by the PEPIN-2 depletion analysis. Secondly the most important actinides and fission product poisons are automatically selected in PEPIN-2 according to the reactivity worth and the burnup for critically consideration. Then the 3D Monte Carlo critically code TRIMARAN-2 is used to examine the subcriticality. All the resonance self-shielded cross sections used in this calculation system are prepared with the APOLLO-2 lattice cell code. The burnup credit calculation methodology and related PWR spent fuel transportation benchmark results are reported and discussed. (authors)

  11. Information handbook on independent spent fuel storage installations

    Energy Technology Data Exchange (ETDEWEB)

    Raddatz, M.G.; Waters, M.D.

    1996-12-01

    In this information handbook, the staff of the U.S. Nuclear Regulatory Commission describes (1) background information regarding the licensing and history of independent spent fuel storage installations (ISFSIs), (2) a discussion of the licensing process, (3) a description of all currently approved or certified models of dry cask storage systems (DCSSs), and (4) a description of sites currently storing spent fuel in an ISFSI. Storage of spent fuel at ISFSIs must be in accordance with the provisions of 10 CFR Part 72. The staff has provided this handbook for information purposes only. The accuracy of any information herein is not guaranteed. For verification or for more details, the reader should refer to the respective docket files for each DCSS and ISFSI site. The information in this handbook is current as of September 1, 1996.

  12. Alteration of spent fuel matrix under unsaturated water conditions

    International Nuclear Information System (INIS)

    Drip tests which simulate the unsaturated conditions expected in the potential repository at Yucca Mountain are in progress to evaluate the long-term performance of spent fuel. This paper examines the corrosion behavior of the spent fuel matrix under conditions in which water is introduced at a rate of 1.5 mL every 7 days. Our recent results suggest a rapid reaction rate of the spent fuel matrix, the formation of alteration products that are similar to the sequence found in ore deposits in uranium mines, and the presence of colloidal species in the leachate. These results are compared to results from two models developed for a potential repository in an unsaturated zone

  13. Spent fuel and high-level radioactive waste storage

    International Nuclear Information System (INIS)

    The subject of spent fuel and high-level radioactive waste storage, is bibliographically reviewed. The review shows that in the majority of the countries, spent fuels and high-level radioactive wastes are planned to be stored for tens of years. Sites for final disposal of high-level radioactive wastes have not yet been found. A first final disposal facility is expected to come into operation in the United States of America by the year 2010. Other final disposal facilities are expected to come into operation in Germany, Sweden, Switzerland and Japan by the year 2020. Meanwhile , stress is placed upon the 'dry storage' method which is carried out successfully in a number of countries (Britain and France). In the United States of America spent fuels are stored in water pools while the 'dry storage' method is still being investigated. (Author)

  14. Application of spent fuel treatment technology to plutonium immobilization

    International Nuclear Information System (INIS)

    The purpose of the electrometallurgical treatment technology being developed at Argonne National Laboratory (ANL) is to convert certain spent nuclear fuels into waste forms that are suitable for disposal in a geological repository for nuclear waste. The spent fuels of interest are those that cannot be safely stored for a long time in their current condition, and those that cannot be qualified for repository disposal. This paper explores the possibility of applying this electrometallurgical treatment technology to immobilization of surplus fissile materials, primarily plutonium. Immobilization of surplus fissile materials by electrometallurgical treatment could be done in the same facilities, at the same time. and in the same equipment as the proposed treatment of the present inventory of spent nuclear fuel. The cost and schedule savings of this simultaneous treatment scheme would be significant

  15. Status of Proposed Repository for Latin-American Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ferrada, J.J.

    2004-10-04

    This report compiles preliminary information that supports the premise that a repository is needed in Latin America and analyzes the nuclear situation (mainly in Argentina and Brazil) in terms of nuclear capabilities, inventories, and regional spent-fuel repositories. The report is based on several sources and summarizes (1) the nuclear capabilities in Latin America and establishes the framework for the need of a permanent repository, (2) the International Atomic Energy Agency (IAEA) approach for a regional spent-fuel repository and describes the support that international institutions are lending to this issue, (3) the current situation in Argentina in order to analyze the Argentinean willingness to find a location for a deep geological repository, and (4) the issues involved in selecting a location for the repository and identifies a potential location. This report then draws conclusions based on an analysis of this information. The focus of this report is mainly on spent fuel and does not elaborate on other radiological waste sources.

  16. Recent developments in spent fuel management in Norway - 59260

    International Nuclear Information System (INIS)

    Spent Nuclear Fuel (SNF) in Norway has arisen from irradiation of fuel in the NORA, Jeep I and Jeep II reactors at Kjeller, and in the Heavy Boiling Water Reactor (HBWR) in Halden. In total there is some 16 tonnes of SNF, with 12 tons of aluminium-clad fuel, of which 10 tonnes is metallic uranium fuel and the remainder oxide (UO2). The portion of this fuel that is similar to commercial fuel (UO2 clad in Zircaloy) may be suitable for direct disposal on the Swedish model or in other repository designs. However, metallic uranium and/or fuels clad in aluminium are chemically reactive and there would be risks associated with direct disposal. Two committees were established by the Government of Norway in January 2009 to make recommendations for the interim storage and final disposal of spent fuel in Norway. The Technical Committee on Storage and Disposal of Metallic Uranium Fuel and Al-clad Fuels was formed with the mandate to recommend treatment (i.e. conditioning) options for metallic uranium fuel and aluminium-clad fuel to render them stable for long term storage and disposal. This committee, whose members were drawn from the nuclear industry, reported in January 2010, and recommended commercial reprocessing as the best option for these fuels. The Phase-2 committee, which in part based its work on the work of previous committees and on the report of the Technical Committee, had the mandate to find the most suitable technical solution and localisation for intermediate storage for spent nuclear fuel and long-lived waste. The membership of this committee was chosen to represent a broad cross section of stakeholders. The committee evaluated different solutions and their associated costs, and recommended one of the options. The committee's report published in early 2011. This paper summarises the conclusions of the two committees, and thereby illustrates the steps taken by one country to formulate a strategy for the long-term management of its SNF. (authors)

  17. Spent fuel dry storage technology development: thermal evaluation of isolated drywell containing spent fuel (1.25 kW PWR spent fuel assembly)

    International Nuclear Information System (INIS)

    A spent fuel Isolated Drywell Test was conducted at the Engine-Maintenance, Assembly and Disassembly (E-MAD) facility on the Nevada Test Site utilizing a pressurized water reactor spent fuel assembly having a decay heat level of approximately 1.25 kW. The fuel assembly was encapsulated inside the E-MAD Hot Bay and placed in an instrumented near-surface drywell is located 50 feet from an adjacent drywell. Instrumentation provided to measure canister, liner and soil temperatures consisted of thermocouples which were inserted into tubes on the outside of the canister and drywell liner and thermocouples which were attached to plastic pipe and grouted into holes in the soil. Temperatures from the isolated drywell and the adjacent soil were recorded throughout the nine month Isolated Drywell Test. Canister and drywell liner temperatures reached their peak values (3230F and 2620F, respectively) during October, 1980. Thereafter, all temperatures began to decrease in response to the decay heat and seasonal atmospheric temperature changes. This thermal response is comparable to that of the approximately 1.0 kW spent fuel assemblies previously tested at E-MAD where peak canister and liner temperatures of 2540F and 2030F were recorded. A previously developed computer model was utilized to predict the thermal response of the surrounding soil are presented and are compared with the test data

  18. Indian Strategy for Management of Spent Fuel from Power Reactors

    International Nuclear Information System (INIS)

    The boom of India’s economy and population, and the determination to increase the nuclear share in the energy mix, will lead to a growing nuclear fuel demand. Regarding the limited known uranium resource within its territory, India will prefer to operate with a closed fuel cycle in order to improve the use of the already acquired uranium. This cycle will be based on the strategy to make efficient use of spent fuel in 3 consecutive steps: in PHWRs, in sodium cooled FBRs and with the thorium-uranium 233 cycles. However, in order to be efficient, India requires constructing infrastructure to reprocess fuel for those 3 stages, but also strategies to convey spent fuel to the reprocessing facilities. (author)

  19. GT-MHR spent fuel storage disposal without processing

    International Nuclear Information System (INIS)

    Possibility of GT-MHR spent fuel storage during long time without additional processing is discussed in this paper. Spent fuel elements discharged from this reactor type are ideal waste forms for permanent disposal in a geologic repository. The graphite fuel elements and the ceramic coatings on the fuel particles are as-manufactured engineered barriers that provide excellent near field containment of radionuclides and minimize reliance on the waste package and surrounding geologic media for long-term containment. Because of the high level of plutonium destruction and degradation achieved by GT-MHR, the isotopic composition of residual plutonium in spent fuel elements would not be practical for use in nuclear weapons and for energy production. Dilution of plutonium within the relatively large volume of GT-MHR fuel elements provides excellent resistance to diversion throughout the fuel cycle. This is accomplished without adversely impacting repository land requirements, since repository loading is determined by decay heat load and not by physical volume. These conditions of safe fuel storage: criticality conditions, conditions of decay heat removing and radiation doses are discussed as well. (author)

  20. Shielding Performance Measurements of Spent Fuel Transportation Container

    Directory of Open Access Journals (Sweden)

    SUN Hong-chao

    2015-11-01

    Full Text Available The safety supervision of radioactive material transportation package has been further stressed and implemented. The shielding performance measurements of spent fuel transport container is the important content of supervision. However, some of the problems and difficulties reflected in practice need to be solved, such as the neutron dose rate on the surface of package is too difficult to measure exactly, the monitoring results are not always reliable, etc. The monitoring results using different spectrometers were compared and the simulation results of MCNP runs were considered. An improvement was provided to the shielding performance measurements technique and management of spent fuel transport.

  1. International management and storage of plutonium and spent fuel

    International Nuclear Information System (INIS)

    The first part of this study discusses certain questions that may arise from the disseminated production and storage of plutonium and, in the light of the relevant provisions of the Agency's Statute, examines possible arrangements for the storage of separated plutonium under international auspices and its release to meet energy or research requirements. The second part of the study deals similarly with certain problems presented by growing accumulations of spent fuel from light-water reactors in various countries and examines possible solutions, including the establishment of regional or multinational spent fuel storage facilities

  2. Criticality safety aspects of spent fuel arrays from emerging nuclear fuel cycles

    Energy Technology Data Exchange (ETDEWEB)

    Nicolaou, G. [University of Thrace, Department of Electrical and Computer Engineering, Laboratory of Nuclear Technology, Kimmerria Campus, 67100 Xanthi (Greece)

    2010-07-01

    Emerging nuclear fuel cycles: fuels with Pu or minor actinides (MA) for their self-generated recycling or transmutation in PWR or FR {yields} reduction of radiotoxicity of HLW. The aim of work is to assess criticality (k{sub {infinity}}) of arrays of spent nuclear fuels from these emerging fuel cycles. Procedures: Calculations of - k{sub {infinity}}, using MCNP5 based on fresh and spent fuel compositions (infinite arrays), - spent fuel compositions using ORIGEN. Fuels considered: - commercial PWR-UO{sub 2} (R1) and -MOX (R2), [45 GWd/t] and fast reactor [100 GWd/t] (R3), - PWR self-generated Pu recycling (S1) and MA recycling (S2), FR self-generated MA recycling (S3), FR with 2% {sup 237}Np for transmutation purposes (T). Results: k{sub {infinity}} based on fresh and spent fuel compositions is shown. Fuels are clustered in two distinct families: - fast reactor fuels, - thermal reactor fuels; k{sub {infinity}} decreases when calculated on the basis of actinide and fission product inventory. In conclusions: - Emerging fuels considered resemble their corresponding commercial fuels; - k{sub {infinity}} decreases in all cases when calculated on the basis of spent fuel compositions (reactivity worth {approx}-20%{Delta}k/k), hence improving the effectiveness of packaging. (author)

  3. TRIGA LEU and HEU fuel shielding analysis during spent fuel transport

    International Nuclear Information System (INIS)

    The paper goal is a comparative study on the effects of TRIGA LEU and HEU fuel for the shielding analysis during spent fuel transport. All geometrical and material data for the shipping cask were considered according to NAC-LWT Cask approved model. The shielding analysis estimates the radiation doses to the shipping cask wall, and in air at 1 m and 2 m, respectively, from the cask, by means of 3D Monte Carlo MORSE-SGC code. Before loading into the shipping cask, TRIGA spent fuel source terms and spent fuel parameters have been obtained by means of ORIGEN-S code. Both codes are included in ORNL's SCALE 5 programs package. 60Co radioactivity is important for HEU spent fuel; actinides contribution to total fuel 60 radioactivity is low. For LEU spent fuel 60Co radioactivity is insignificant; actinides contribution to total fuel radioactivity is high. Comparison of TRIGA fuels gamma source terms shows that the LEU source terms are higher then the HEU ones. LEU spent fuel photon dose rates are greater than the HEU ones. Dose rates for both HEU and LEU fuel contents are below regulatory limits. (author)

  4. Thermal analysis of spent fuel shipping cask for application of metalized fuel

    International Nuclear Information System (INIS)

    Thermal analysis of spent fuel shipping cask loaded with 4 spent PWR fuel assemblies has been carried out using the fluent code. And the temperature distribution of cask for application of 4 metalized fuels equivalent to 16 PWR fuels has been also calculated. Total decay heat from 4 spent PWR fuels and 4 metalized spent fuels are 2.2 kW and 4.4 kW, respectively. The calculated temperatures for 4 spent PWR fuels were compared with the proven data presented from the safety analysis report of shipping cask. It has good agreement between two results. The maximum fuel rod temperatures inside the canisters of square and hexagonal types are estimated to be 269 .deg. C and 212 .deg. C, respectively. Therefore, it is found that the hexagonal canister loaded with metalized fuel rods is more advantageous in aspect of thermal characteristics and storage efficiency. Fuel temperature in the cavity of helium gas for hexagonal canister is lower than the temperature for spent PWR fuel

  5. Reactor-specific spent fuel discharge projections, 1984 to 2020

    Energy Technology Data Exchange (ETDEWEB)

    Heeb, C.M.; Libby, R.A.; Holter, G.M.

    1985-04-01

    The original spent fuel utility data base (SFDB) has been adjusted to produce agreement with the EIA nuclear energy generation forecast. The procedure developed allows the detail of the utility data base to remain intact, while the overall nuclear generation is changed to match any uniform nuclear generation forecast. This procedure adjusts the weight of the reactor discharges as reported on the SFDB and makes a minimal (less than 10%) change in the original discharge exposures in order to preserve discharges of an integral number of fuel assemblies. The procedure used in developing the reactor-specific spent fuel discharge projections, as well as the resulting data bases themselves, are described in detail in this report. Discussions of the procedure cover the following topics: a description of the data base; data base adjustment procedures; addition of generic power reactors; and accuracy of the data base adjustments. Reactor-specific discharge and storage requirements are presented. Annual and cumulative discharge projections are provided. Annual and cumulative requirements for additional storage are shown for the maximum at-reactor (AR) storage assumption, and for the maximum AR with transshipment assumption. These compare directly to the storage requirements from the utility-supplied data, as reported in the Spent Fuel Storage Requirements Report. The results presented in this report include: the disaggregated spent fuel discharge projections; and disaggregated projections of requirements for additional spent fuel storage capacity prior to 1998. Descriptions of the methodology and the results are included in this report. Details supporting the discussions in the main body of the report, including descriptions of the capacity and fuel discharge projections, are included. 3 refs., 6 figs., 12 tabs.

  6. Alternative Measuring Approaches in Gamma Scanning on Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sihm Kvenangen, Karen

    2007-06-15

    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 gamma ray spectroscopy method called gamma scanning the gamma radiation from the fission products Cs-137, Cs-134 and Eu-154 are measured in a spent fuel assembly. From the results, conclusions can be drawn about the fuels characteristics. This degree project examines the possibilities of using alternative measuring approaches when using the gamma scanning method. The focus is on examining how to increase the quality of the measured data. How to decrease the measuring time as compared with the present measuring strategy, has also been investigated. The main part of the study comprises computer simulations of gamma scanning measurements. The simulations have been validated with actual measurements on spent nuclear fuel at the central interim storage, Clab. The results show that concerning the quality of the measuring data the conventional strategy is preferable, but with other starting positions and with a more optimized equipment. When focusing on the time aspect, the helical measuring strategy can be an option, but this needs further investigation.

  7. 10 CFR 72.236 - Specific requirements for spent fuel storage cask approval and fabrication.

    Science.gov (United States)

    2010-01-01

    ... REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR... spent fuel (i.e., intact assembly or consolidated fuel rods), the inerting atmosphere requirements. (b... removal of the stored spent fuel from a reactor site, transportation, and ultimate disposition by...

  8. Spent fuel management in the Power Engineering Ministry of USA

    International Nuclear Information System (INIS)

    There are presented the systematized data on the management with the spent fuel (SF), belonging to the USA Power Engineering Ministry, namely: 2300 t of SF of the two-purpose N reactor in Hanford; SF of the national and foreign research reactors in Savannah River (3400 fuel assemblies with uranium-aluminium fuel with aluminium can and 2000 fuel assemblies with zirconium-stainless steel can) and 265 t of SF of the Peach Bottom and Shipping-port NPPs and the TRIGA reactor in the INEEL laboratory

  9. Survey of experience with dry storage of spent nuclear fuel and update of wet storage experience

    International Nuclear Information System (INIS)

    Spent fuel storage is an important part of spent fuel management. At present about 45,000 t of spent water reactor fuel have been discharged worldwide. Only a small fraction of this fuel (approximately 7%) has been reprocessed. The amount of spent fuel arisings will increase significantly in the next 15 years. Estimates indicate that up to the year 2000 about 200,000 t HM of spent fuel could be accumulated. In view of the large quantities of spent fuel discharged from nuclear power plants and future expected discharges, many countries are involved in the construction of facilities for the storage of spent fuel and in the development of effective methods for spent fuel surveillance and monitoring to ensure that reliable and safe operation of storage facilities is achievable until the time when the final disposal of spent fuel or high level wastes is feasible. The first demonstrations of final disposal are not expected before the years 2000-2020. This is why the long term storage of spent fuel and HLW is a vital problem for all countries with nuclear power programmes. The present survey contains data on dry storage and recent information on wet storage, transportation, rod consolidation, etc. The main aim is to provide spent fuel management policy making organizations, designers, scientists and spent fuel storage facility operators with the latest information on spent fuel storage technology under dry and wet conditions and on innovations in this field. Refs, figs and tabs

  10. Loviisa NPP. Extension of spent fuel storage capacity

    International Nuclear Information System (INIS)

    In accordance with the amended Nuclear Energy Act, Imatran Voima Oy will make provisions for the storage and direct disposal of the spent nuclear fuel from units l and 2 of the Loviisa power plant in Finland. The fuel storage system was originally designed for the fuel volume generated during some five years of operation. Today, the changing situation requires that the storage capacity for spent fuel must be increased to cover the whole of the remaining operating life. The study covers all the options of increasing the storage capacity that are known and used throughout the world. In total nine options of various types have been designed and studied. (4 refs., 41 figs., 6 tabs.)

  11. Test plan for thermogravimetric analyses of BWR spent fuel oxidation

    International Nuclear Information System (INIS)

    Preliminary studies indicated the need for additional low-temperature spent fuel oxidation data to determine the behavior of spent fuel as a waste form for a tuffy repository. Short-term thermogravimetric analysis tests were recommended in a comprehensive technical approach as the method for providing scoping data that could be used to (1) evaluate the effects of variables such as moisture and burnup on the oxidation rate, (2) determine operative mechanisms, and (3) guide long-term, low-temperature oxidation testing. The initial test series studied the temperature and moisture effects on pressurized water reactor fuel as a function of particle and grain size. This document presents the test matrix for studying the oxidation behavior of boiling water reactor fuel in the temperature range of 140 to 225/degree/C. 17 refs., 7 figs., 3 tabs

  12. Test plan for spent fuel cladding containment credit tests

    International Nuclear Information System (INIS)

    Lawrence Livermore National Laboratory has chosen Westinghouse Hanford Company as a subcontractor to assist them in determining the requirements for successful disposal of spent fuel rods in the proposed Nevada Test Site repository. An initial scoping test, with the objective of determining whether or not the cladding of a breached fuel rod can be given any credit as an effective barrier to radionuclide release, is described in this test plan. 8 references, 2 figures, 4 tables

  13. Synthesis on the spent fuel long term evolution

    Energy Technology Data Exchange (ETDEWEB)

    Ferry, C.; Poinssot, Ch.; Lovera, P.; Poulesquen, A. [CEA Saclay, Dept. de Physico-Chimie (DEN/DPC), 91 - Gif sur Yvette (France); Broudic, V. [CEA Cadarache, Direction des Reacteurs Nucleaires (DRN), 13 - Saint Paul lez Durance (France); Cappelaere, Ch. [CEA Saclay, Dept. des Materiaux pour le Nucleaire(DMN), 91 - Gif-sur-Yvette (France); Desgranges, L. [CEA Cadarache, Direction des Reacteurs Nucleaires (DRN), 13 - Saint-Paul-lez-Durance (France); Garcia, Ph. [CEA Cadarache, Dept. d' Etudes des Combustibles (DEC), 13 - Saint Paul lez Durance (France); Jegou, Ch.; Roudil, D. [CEA Valrho, Dir. de l' Energie Nucleaire (DEN), 30 - Marcoule (France); Lovera, P.; Poulesquen, A. [CEA Saclay, Dept. de Physico-Chimie (DPC), 91 - Gif sur Yvette (France); Marimbeau, P. [CEA Cadarache, Dir. de l' Energie Nucleaire (DEN), 13 - Saint-Paul-lez-Durance (France); Gras, J.M.; Bouffioux, P. [Electricite de France (EDF), 75 - Paris (France)

    2005-07-01

    The French research on spent fuel long term evolution has been performed by CEA (Commissariat a l'Energie Atomique) since 1999 in the PRECCI project with the support of EDF (Electricite de France). These studies focused on the spent fuel behaviour under various conditions encountered in dry storage or in deep geological disposal. Three main types of conditions were discerned: - The evolution in a closed system which corresponds to the normal scenario in storage and to the first confinement phase in disposal; - The evolution in air which corresponds to an incidental loss of confinement during storage or to a rupture of the canister before the site re-saturation in geological disposal; - The evolution in water which corresponds to the normal scenario after the breaching of the canister in repository conditions. This document produced in the frame of the PRECCI project is an overview of the state of knowledge in 2004 concerning the long-term behavior of spent fuel under these various conditions. The state of the art was derived from the results obtained under the PRECCI project as well as from a review of the literature and of data acquired under the European project on Spent Fuel Stability under Repository Conditions. The main results issued from the French research are underlined. (authors)

  14. Spent fuel test. Climax data acquisition system integration report

    Energy Technology Data Exchange (ETDEWEB)

    Nyholm, R.A.; Brough, W.G.; Rector, N.L.

    1982-06-01

    The Spent Fuel Test - Climax (SFT-C) is a test of the retrievable, deep geologic storage of commercially generated, spent nuclear reactor fuel in granitic rock. Eleven spent fuel assemblies, together with 6 electrical simulators and 20 guard heaters, are emplaced 420 m below the surface in the Climax granite at the Nevada Test Site. On June 2, 1978, Lawrence Livermore National Laboratory (LLNL) secured funding for the SFT-C, and completed spent fuel emplacement May 28, 1980. This multi-year duration test is located in a remote area and is unattended much of the time. An extensive array of radiological safety and geotechnical instrumentation is deployed to monitor the test performance. A dual minicomputer-based data acquisition system collects and processes data from more than 900 analog instruments. This report documents the design and functions of the hardware and software elements of the Data Acquisition System and describes the supporting facilities which include environmental enclosures, heating/air-conditioning/humidity systems, power distribution systems, fire suppression systems, remote terminal stations, telephone/modem communications, and workshop areas. 9 figures.

  15. Corrosion resistance of a copper canister for spent nuclear fuel

    International Nuclear Information System (INIS)

    The report presents an evaluation of copper as canister material for spent nuclear fuel. The evaluation is made from the viewpoint of corrosion and applies to a concept of 1977. Supplementary corrosion studies have been performed. The report includes 9 appendices which deal with experimental data. (G.B.)

  16. Synthesis on the spent fuel long term evolution

    International Nuclear Information System (INIS)

    The French research on spent fuel long term evolution has been performed by CEA (Commissariat a l'Energie Atomique) since 1999 in the PRECCI project with the support of EDF (Electricite de France). These studies focused on the spent fuel behaviour under various conditions encountered in dry storage or in deep geological disposal. Three main types of conditions were discerned: - The evolution in a closed system which corresponds to the normal scenario in storage and to the first confinement phase in disposal; - The evolution in air which corresponds to an incidental loss of confinement during storage or to a rupture of the canister before the site re-saturation in geological disposal; - The evolution in water which corresponds to the normal scenario after the breaching of the canister in repository conditions. This document produced in the frame of the PRECCI project is an overview of the state of knowledge in 2004 concerning the long-term behavior of spent fuel under these various conditions. The state of the art was derived from the results obtained under the PRECCI project as well as from a review of the literature and of data acquired under the European project on Spent Fuel Stability under Repository Conditions. The main results issued from the French research are underlined. (authors)

  17. The CASCAD system: An SGN spent fuel dry storage facility

    International Nuclear Information System (INIS)

    This paper will present SGN's dry vault spent storage system. This concept is based on the CASCAD facility, designed and built by SGN for the French Atomic Energy Commission (CEA) at Cadarache, France. Cascade has been in operation since 1990 since which time SGN has customized its storage system. Because of its extensive experience in both spent fuel assembly and dry storage of high level waste, SGN is able to design solutions fully customized to fit customers' storage requirements using proven technology. Its modular approach allows for staggered investment over a period of several years for maximum flexibility. The Cascad system meets site-specific constraints and safety requirements and is able to receive a wide range of fuels and shipping casks. Since spent fuel assemblies are stored in passive cooled pits, the system is entirely passive and therefore inherently safe. Moreover, the Cascad system allows total retrievability of spent fuel after a 50-year storage period even if the reactor building no longer exists

  18. Development of a dry storage cask for PWR spent fuel

    International Nuclear Information System (INIS)

    Korea Hydro and Nuclear Power Co., Ltd.(KHNP), which operates all the nuclear power plants in Korea, is developing a new dry storage cask to store twenty four spent fuel assemblies generated from pressurized water reactors for at-reactor or away-from-reactor interim storage facility in Korea. The dry storage cask is designed and evaluated according to the requirements of the IAEA, the US NRC and the Korean regulations for the dry spent fuel storage system. It provides confinement, radiation shielding, structural integrity, subcritical control and passive heat removal for normal and accident conditions. The dry storage cask consists of a dual purpose canister providing a confinement boundary for the PWR spent fuel, and a storage overpack providing a structural and radiological boundary for long-term storage of the canister placed inside it. The overpack is constructed by a combination of steel and concrete, and is equipped with penetrating ducts near its lower and upper extremities to permit natural circulation of air to provide for the passive cooling of the canister and the contained spent fuel assemblies. This paper describes development status, description, design criteria, evaluation and demonstration tests of the dry storage cask. (authors)

  19. Method of pre-treatment of spent fuel before storage

    International Nuclear Information System (INIS)

    CO2 is compulsorily supplied to spent fuels in LMFBR type reactors to convert residual sodium deposited to spent fuels into chemically inactive form of Na2CO3. That is, spent fuels are applied with pre-treatment before storage in an underground carrier in a cell in a CO2 atmosphere in which they are transferred and accepted between an inactive gas cell and an air cell. A containing vessel is disposed on the carrier for containing spent fuels and a blower and a filter are loaded for circulating atmospheric CO2 into the vessel. Heating and cooling device are disposed as necessary. Thus, since reactions are taken place only moderately or not taken place at all in a water or air atmosphere after processing, exposed in-water storage is enabled. Accordingly, it is possible to remarkably rationalize the cleaning facilities, save exclusive vessel made of steel, reduce radioactive wastes and save relevant facilities upon dry-storage. (T.M.)

  20. Spent fuel test. Climax data acquisition system integration report

    International Nuclear Information System (INIS)

    The Spent Fuel Test - Climax (SFT-C) is a test of the retrievable, deep geologic storage of commercially generated, spent nuclear reactor fuel in granitic rock. Eleven spent fuel assemblies, together with 6 electrical simulators and 20 guard heaters, are emplaced 420 m below the surface in the Climax granite at the Nevada Test Site. On June 2, 1978, Lawrence Livermore National Laboratory (LLNL) secured funding for the SFT-C, and completed spent fuel emplacement May 28, 1980. This multi-year duration test is located in a remote area and is unattended much of the time. An extensive array of radiological safety and geotechnical instrumentation is deployed to monitor the test performance. A dual minicomputer-based data acquisition system collects and processes data from more than 900 analog instruments. This report documents the design and functions of the hardware and software elements of the Data Acquisition System and describes the supporting facilities which include environmental enclosures, heating/air-conditioning/humidity systems, power distribution systems, fire suppression systems, remote terminal stations, telephone/modem communications, and workshop areas. 9 figures

  1. Characterization of activated metals in spent fuel hardware

    International Nuclear Information System (INIS)

    Pacific Northwest Laboratory (PNL), for the U.S. Department of Energy (DOE), has been investigating the activation of spent fuel hardware in order to properly account for it in the federal waste management system. The paper presents a description and status of the program and tentative conclusions

  2. Spent Fuel Test - Climax data acquisition system operations manual

    International Nuclear Information System (INIS)

    The Spent Fuel Test-Climax (SFT-C) is a test of the retrievable, deep geologic storage of commercially generated, spent nuclear reactor fuel in granite rock. Eleven spent fuel assemblies, together with 6 electrical simulators and 20 guard heaters, are emplaced 420 m below the surface in the Climax granite at the US Department of Energy Nevada Test Site. On June 2, 1978, Lawrence Livermore National Laboratory (LLNL) secured funding for the SFT-C, and completed spent fuel emplacement May 28, 1980. The multi-year duration test is located in a remote area and is unattended much of the time. An extensive array of radiological safety and geotechnical instrumentation is deployed to monitor the test performance. A dual minicomputer-based data acquisition system (DAS) collects and processes data from more than 900 analog instruments. This report documents the software element of the LLNL developed SFT-C Data Acquisition System. It defines the operating system and hardware interface configurations, the special applications software and data structures, and support software

  3. Modelling of radiation field around spent fuel container

    NARCIS (Netherlands)

    Kryuchkov, EF; Opalovsky, VA; Tikhomirov, GV

    2005-01-01

    Operation of nuclear reactors leads to the production of spent nuclear fuel (SNF). There are two basic strategies of SNF management: ultimate disposal of SNF in geological formations and recycle or repeated utilisation of reprocessed SNF. In both options, there is an urgent necessity to study radiat

  4. Chemical reprocessing of spent nuclear fuels

    International Nuclear Information System (INIS)

    The reprocessing of nuclear fuels from atomic power stations has a twofold goal. On the one hand it is serving for fuel supply by recovering the fissile materials which have not been consumed or which have been freshly generated in the reactor. On the other hand the radioactive waste products from nuclear power generation are pretreated for long-term safe disposal. The core element of the chemical processing is the PUREX Process, a counter-current solvent extraction procedure using tributyl phosphate (TBP) as the solvent for uranium and plutonium. The chemical basis and the technological performance of the process are discussed. (orig.)

  5. Atucha I: NPP Spent Fuel Dry Storage Conceptual Design

    International Nuclear Information System (INIS)

    The present report shows a Spent Fuel Dry Storage Conceptual Design to emptying the former oldest fuel elements pool and storage them in a Dry Storage System, in order to reach the 32 (Full Power Years) FPY of Atucha I Nuclear Power Plant (CNA I) End Of Life [1]. The project consist mainly in the enlargement of one of the Pool Buildings of the Station (there are two of them) where the (Spent Fuel Elements) SFE will be stored in vertical underground silos. Each silo is composed of two storage units that contains 9 fuel elements each (18 fuel elements in each bin). This design allows a vertical storage of 2016 spent fuel elements (7 rows by 16 columns). The SFE must be transferred from the Pool Building to the Dry Storage Building through a dedicated shield for lifting and transporting the SFE. To move the shield, the actual 60 Ton capacity crane will be used. The operation time to emptying a complete pool will be approximately one year (1998 SFE). Therefore the storage system should be finished by 2013, in order not to penalize the continue operation of the Station. This conceptual design meets the basic principles of Nuclear Safety, protecting workers, public and the overall environment of ionizing radiation and radioactive contamination. This is achieved by transport and storage shielding, operation procedures and comply key conditions like subcriticality of the system, SFE monitoring and SFE heat removal. (author)

  6. A Monte Carlo based spent fuel analysis safeguards strategy assessment

    Energy Technology Data Exchange (ETDEWEB)

    Fensin, Michael L [Los Alamos National Laboratory; Tobin, Stephen J [Los Alamos National Laboratory; Swinhoe, Martyn T [Los Alamos National Laboratory; Menlove, Howard O [Los Alamos National Laboratory; Sandoval, Nathan P [Los Alamos National Laboratory

    2009-01-01

    Safeguarding nuclear material involves the detection of diversions of significant quantities of nuclear materials, and the deterrence of such diversions by the risk of early detection. There are a variety of motivations for quantifying plutonium in spent fuel assemblies by means of nondestructive assay (NDA) including the following: strengthening the capabilities of the International Atomic Energy Agencies ability to safeguards nuclear facilities, shipper/receiver difference, input accountability at reprocessing facilities and burnup credit at repositories. Many NDA techniques exist for measuring signatures from spent fuel; however, no single NDA technique can, in isolation, quantify elemental plutonium and other actinides of interest in spent fuel. A study has been undertaken to determine the best integrated combination of cost effective techniques for quantifying plutonium mass in spent fuel for nuclear safeguards. A standardized assessment process was developed to compare the effective merits and faults of 12 different detection techniques in order to integrate a few techniques and to down-select among the techniques in preparation for experiments. The process involves generating a basis burnup/enrichment/cooling time dependent spent fuel assembly library, creating diversion scenarios, developing detector models and quantifying the capability of each NDA technique. Because hundreds of input and output files must be managed in the couplings of data transitions for the different facets of the assessment process, a graphical user interface (GUI) was development that automates the process. This GUI allows users to visually create diversion scenarios with varied replacement materials, and generate a MCNPX fixed source detector assessment input file. The end result of the assembly library assessment is to select a set of common source terms and diversion scenarios for quantifying the capability of each of the 12 NDA techniques. We present here the generalized

  7. Advanced coated particle fuels

    International Nuclear Information System (INIS)

    The coated particle fuel (cpf) has been developed for use in high-temperature gas-cooled reactors, but it may find applications in other types of reactors. In JAERI, besides the development of cpf for High Temperature Engineering Test Reactor, conceptual studies of the cpf applications in actinide burner reactors and space reactors have been made. The conceptual design studies as well as the research and development of advanced coatings, ZrC and TiN, are reviewed. (author)

  8. Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation

    International Nuclear Information System (INIS)

    Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spent fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to the age of spent fuels. This paper intends to evaluate the typical spent fuel compositions of light water (LWR) and fast breeder reactors (FBR) from the view point of some foot prints of nuclear forensics. An established depletion code of ORIGEN is adopted to analyze LWR spent fuel (SF) for several burnup constants and decay times. For analyzing some spent fuel compositions of FBR, some coupling codes such as SLAROM code, JOINT and CITATION codes including JFS-3-J-3.2R as nuclear data library have been adopted. Enriched U-235 fuel composition of oxide type is used for fresh fuel of LWR and a mixed oxide fuel (MOX) for FBR fresh fuel. Those MOX fuels of FBR come from the spent fuels of LWR. Some typical spent fuels from both LWR and FBR will be compared to distinguish some typical foot-prints of SF based on nuclear forensic analysis.

  9. Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation

    Science.gov (United States)

    Permana, Sidik; Suzuki, Mitsutoshi; Su'ud, Zaki

    2012-06-01

    Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spent fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to the age of spent fuels. This paper intends to evaluate the typical spent fuel compositions of light water (LWR) and fast breeder reactors (FBR) from the view point of some foot prints of nuclear forensics. An established depletion code of ORIGEN is adopted to analyze LWR spent fuel (SF) for several burnup constants and decay times. For analyzing some spent fuel compositions of FBR, some coupling codes such as SLAROM code, JOINT and CITATION codes including JFS-3-J-3.2R as nuclear data library have been adopted. Enriched U-235 fuel composition of oxide type is used for fresh fuel of LWR and a mixed oxide fuel (MOX) for FBR fresh fuel. Those MOX fuels of FBR come from the spent fuels of LWR. Some typical spent fuels from both LWR and FBR will be compared to distinguish some typical foot-prints of SF based on nuclear forensic analysis.

  10. German Spent Nuclear Fuel Legacy: Characteristics and High-Level Waste Management Issues

    Directory of Open Access Journals (Sweden)

    A. Schwenk-Ferrero

    2013-01-01

    Full Text Available Germany is phasing-out the utilization of nuclear energy until 2022. Currently, nine light water reactors of originally nineteen are still connected to the grid. All power plants generate high-level nuclear waste like spent uranium or mixed uranium-plutonium dioxide fuel which has to be properly managed. Moreover, vitrified high-level waste containing minor actinides, fission products, and traces of plutonium reprocessing loses produced by reprocessing facilities has to be disposed of. In the paper, the assessments of German spent fuel legacy (heavy metal content and the nuclide composition of this inventory have been done. The methodology used applies advanced nuclear fuel cycle simulation techniques in order to reproduce the operation of the German nuclear power plants from 1969 till 2022. NFCSim code developed by LANL was adopted for this purpose. It was estimated that ~10,300 tonnes of unreprocessed nuclear spent fuel will be generated until the shut-down of the ultimate German reactor. This inventory will contain ~131 tonnes of plutonium, ~21 tonnes of minor actinides, and 440 tonnes of fission products. Apart from this, ca.215 tonnes of vitrified HLW will be present. As fission products and transuranium elements remain radioactive from 104 to 106 years, the characteristics of spent fuel legacy over this period are estimated, and their impacts on decay storage and final repository are discussed.

  11. Nuclear spent fuel management scenarios. Status and assessment report

    Energy Technology Data Exchange (ETDEWEB)

    Dufek, J.; Arzhanov, V.; Gudowski, W. [Royal Inst. of Technology, Stockholm (Sweden). Dept. of Nuclear and Reactor Physics

    2006-06-15

    The strategy for management of spent nuclear fuel from the Swedish nuclear power programme is interim storage for cooling and decay for about 30 years followed by direct disposal of the fuel in a geologic repository. In various contexts it is of interest to compare this strategy with other strategies that might be available in the future as a result of ongoing research and development. In particular partitioning and transmutation is one such strategy that is subject to considerable R and D-efforts within the European Union and in other countries with large nuclear programmes. To facilitate such comparisons for the Swedish situation, with a planned phase out of the nuclear power programme, SKB has asked the team at Royal Inst. of Technology to describe and explore some scenarios that might be applied to the Swedish programme. The results of this study are presented in this report. The following scenarios were studied by the help of a specially developed computer programme: Phase out by 2025 with direct disposal. Burning plutonium and minor actinides as MOX in BWR. Burning plutonium and minor actinides as MOX in PWR. Burning plutonium and minor actinides in ADS. Combined LWR-MOX plus ADS. For the different scenarios nuclide inventories, waste amounts, costs, additional electricity production etc have been assessed. As a general conclusion it was found that BWR is more efficient for burning plutonium in MOX fuel than PWR. The difference is approximately 10%. Furthermore the BWR produces about 10% less americium inventory. An ADS reactor park can theoretically in an ideal case burn (transmute) 99% of the transuranium isotopes. The duration of such a scenario heavily depends on the interim time needed for cooling the spent fuel before reprocessing. Assuming 10 years for cooling of nuclear fuel from ADS, the duration will be at least 200 years under optimistic technical assumptions. The development and use of advanced pyro-processing with an interim cooling time of only

  12. Spent fuel management options for research reactors in Latin America

    International Nuclear Information System (INIS)

    Research reactors (RRs) have been operated in Latin America since the late 1950s, and a total of 23 RRs have been built in the region. At the time of writing (November 2005), 18 RRs are in operation, 4 have been shut down and 1 has been decommissioned. The number of operating RRs in Latin America represents around 6% of the existing operational RRs worldwide and around 21% of the RRs operating in developing countries. Common to all RRs in the region is a consistent record of safe and successful operation. With the purpose of carrying out a collaborative study of different aspects of the management of spent fuel from RRs, some countries from the region proposed to the IAEA in 2000 the organization of a Regional Project. The project (IAEA TC Regional Project RLA/4/018) that was approved for the biennium 2001-2002 and extended for 2003-2004 included the participation of Argentina, Brazil, Chile, Mexico and Peru. The main objectives of this project were: (a) to define the basic conditions for a regional strategy for managing spent fuel that will provide solutions compatible with the economic and technological realities of the countries involved; and (b) to determine what is needed for the temporary wet and dry storage of spent fuel from the research reactors in the countries of the Latin American region that participated in the project. This TECDOC is based on the results of TC Regional Project RLA/4/018. This project was successful in identifying and assessing a number of viable alternatives for RRSF management in the Latin American region. Options for operational and interim storage, spent fuel conditioning and final disposal have been carefully considered. This report presents the views of Latin American experts on RR spent fuel management and will be useful as reference material for the Latin American RR community, decision making authorities in the region and the public in general

  13. Thai Strategic Plan for Spent Fuel Management

    International Nuclear Information System (INIS)

    In the past few years, nuclear power has become a key part of the global energy solution. Many countries have been planning on expansion and embarking of the nuclear power in their countries. It is not surprising that Thailand is one of those countries recognizing that nuclear energy is one of the promising options responses to Thai energy policy on energy security, fuel diversification, and greenhouse-gas emission reduction

  14. Size Design of CdZnTe Detector Shield for Measuring Burnup of Spent Fuel

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    <正>It is important to measure the burnup of spent fuel for nuclear safeguards, burnup credit and critical safety in spent-fuel reprocessing process. The purpose of this work is designing a portable device to

  15. Seismic analysis of spent nuclear fuel storage racks

    Energy Technology Data Exchange (ETDEWEB)

    Shah, S.J.; Biddle, J.R.; Bennett, S.M.; Schechter, C.B. [Framatome Cogema Fuels, Lynchburg, VA (United States); Harstead, G.A. [Harstead Engineering Associates, Inc., Old Tappan, NJ (United States); Marquet, F. [ATEA/FRAMATOME, Carquefou (France)

    1996-06-01

    In many nuclear power plants, existing storage racks are being replaced with high-density racks to accommodate the increasing inventory of spent fuel. In the hypothetical design considered here, the high-density arrangement of fuel assemblies, or consolidated fuel canisters, is accomplished through the use of borated stainless steel (BSS) plates acting as neutron absorbers. No structural benefit from the BSS is assumed. This paper describes the methods used to perform seismic analysis of high density spent fuel storage racks. The sensitivity of important parameters such as the effect of variation of coefficients of friction between the rack legs and the pool floor and fuel loading conditions (consolidated and unconsolidated) are also discussed in the paper. Results of this study are presented. The high-density fuel racks are simply supported by the pool floor with no structural connections to adjacent racks or to the pool walls or floor. Therefore, the racks are free standing and may slide and tip. Several time history, nonlinear, seismic analyses are required to account for variations in the coefficient of friction, rack loading configuration, and the type of the seismic event. This paper presents several of the mathematical models usually used. Friction cannot be precisely predicted, so a range of friction coefficients is assumed. The range assumed for the analysis is 0.2 to 0.8. A detailed model representing a single rack is used to evaluate the 3-D loading effects. This model is a controlling case for the stress analysis. A 2-D multi-rack model representing a row of racks between the spent fuel pool walls is used to evaluate the change in gaps between racks. The racks are normally analyzed for the fuel loading conditions of consolidated, full, empty, and half-loaded with fuel assemblies.

  16. Storage of LWR spent fuel in air. Volume 3, Results from exposure of spent fuel to fluorine-contaminated air

    Energy Technology Data Exchange (ETDEWEB)

    Cunningham, M.E.; Thomas, L.E.

    1995-06-01

    The Behavior of Spent Fuel in Storage (BSFS) Project has conducted research to develop data on spent nuclear fuel (irradiated U0{sub 2}) that could be used to support design, licensing, and operation of dry storage installations. Test Series B conducted by the BSFS Project was designed as a long-term study of the oxidation of spent fuel exposed to air. It was discovered after the exposures were completed in September 1990 that the test specimens had been exposed to an atmosphere of bottled air contaminated with an unknown quantity of fluorine. This exposure resulted in the test specimens reacting with both the oxygen and the fluorine in the oven atmospheres. The apparent source of the fluorine was gamma radiation-induced chemical decomposition of the fluoro-elastomer gaskets used to seal the oven doors. This chemical decomposition apparently released hydrofluoric acid (HF) vapor into the oven atmospheres. Because the Test Series B specimens were exposed to a fluorine-contaminated oven atmosphere and reacted with the fluorine, it is recommended that the Test Series B data not be used to develop time-temperature limits for exposure of spent nuclear fuel to air. This report has been prepared to document Test Series B and present the collected data and observations.

  17. Standard guide for drying behavior of spent nuclear fuel

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2008-01-01

    1.1 This guide is organized to discuss the three major components of significance in the drying behavior of spent nuclear fuel: evaluating the need for drying, drying spent nuclear fuel, and confirmation of adequate dryness. 1.1.1 The guide addresses drying methods and their limitations in drying spent nuclear fuels that have been in storage at water pools. The guide discusses sources and forms of water that remain in SNF, its container, or both, after the drying process and discusses the importance and potential effects they may have on fuel integrity, and container materials. The effects of residual water are discussed mechanistically as a function of the container thermal and radiological environment to provide guidance on situations that may require extraordinary drying methods, specialized handling, or other treatments. 1.1.2 The basic issue in drying is to determine how dry the SNF must be in order to prevent issues with fuel retrievability, container pressurization, or container corrosion. Adequate d...

  18. Thermal test and analysis of a spent fuel storage cask

    International Nuclear Information System (INIS)

    A thermal test simulated with full-scale cask model for the normal storage was performed to verify the storage skill of the spent fuels of the cask. The maximum temperature at each point in the test was lower than the allowable temperature. The integrity of the cask was maintained. It was observed that the safety of containment system was also kept according to the check of the seal before and after the thermal test. Therefore it was shown that using the present skill, it is possible to store spent fuels in the dry-type cask safely. Moreover, because of the good agreement between analysis and experimental results, it was shown that the analysis model was successfully established to estimate the temperature distribution of the fuel cladding and the seal portion. (J.P.N.)

  19. Spent nuclear fuel for disposal in the KBS-3 repository

    Energy Technology Data Exchange (ETDEWEB)

    Grahn, Per; Moren, Lena; Wiborgh, Maria

    2010-12-15

    The report is included in a set of Production reports, presenting how the KBS-3 repository is designed, produced and inspected. The set of reports is included in the safety report for the KBS-3 repository and repository facility. The report provides input to the assessment of the long-term safety, SR-Site as well as to the operational safety report, SR-Operation. The report presents the spent fuel to be deposited, and the requirements on the handling and selection of fuel assemblies for encapsulation that follows from that it shall be deposited in the KBS-3 repository. An overview of the handling and a simulation of the encapsulation and the resulting canisters to be deposited are presented. Finally, the initial state of the encapsulated spent nuclear fuel is given. The initial state comprises the radionuclide inventory and other data required for the assessment of the long-term safety

  20. Equipment for the management of spent fuels and radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Bower, C. C. F.; Carter, C. C.; Doubt, H. A. [GEC Alsthom Engineering System Ltd., Leicester (United Kingdom)

    1996-04-15

    UK experience over the last thirty years with the design and implementation of equipment for the management of spent fuels and radioactive wastes has ranged from remote handling, through encapsulation and containerisation, to the medium-term storage of heat-producing fuels and wastes in the dry state. The design principles involved in handling, transporting and storing hazardous materials safely and reliably, while ensuring biological shielding, containment and cooling of radioactive materials, are common to the various kinds of equipment presented in this paper, even though the individual requirements may be very different. The UK nuclear programme over the last thirty years has encouraged the development of extensive expertise in the engineering of equipment for the management of spent fuel and radioactive waste. This expertise can be applied with benefit to the Korean nuclear programme.

  1. Spent fuel management strategy in the United Kingdom

    International Nuclear Information System (INIS)

    The spent fuel management strategy for commercial reactors, Magnox and AGR, in the United Kingdom is summarized. The reasons for following a reprocessing strategy for spent fuel from each reactor type are discussed, and also the rationale for the UK Generating Boards' proposal to build an AGR dry buffer store. Current developments in reprocessing facilities are described, notably waste treatment plants at the British Nuclear Fuels' Sellafield site. By the 1990s this investment programme will be complete, discharges to the sea will be near zero and solid waste will be converted to a more stable form, easier for long term storage. Also in the 1990s an integrated uranium recycling service will be established in the UK. (author). 1 fig

  2. Spent Fuel and Waste Management Technology Development Program

    International Nuclear Information System (INIS)

    This report provides information on the progress of activities during fiscal year 1993 in the Spent Fuel and Waste Management Technology Development Program (SF ampersand WMTDP) at the Idaho Chemical Processing Plant (ICPP). As a new program, efforts are just getting underway toward addressing major issues related to the fuel and waste stored at the ICPP. The SF ampersand WMTDP has the following principal objectives: Investigate direct dispositioning of spent fuel, striving for one acceptable waste form; determine the best treatment process(es) for liquid and calcine wastes to minimize the volume of high level radioactive waste (HLW) and low level waste (LLW); demonstrate the integrated operability and maintainability of selected treatment and immobilization processes; and assure that implementation of the selected waste treatment process is environmentally acceptable, ensures public and worker safety, and is economically feasible

  3. Multi-pack Disposal Concepts for Spent Fuel (Rev. 0)

    Energy Technology Data Exchange (ETDEWEB)

    Hadgu, Teklu [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hardin, Ernest [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Matteo, Edward N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-12-01

    At the initiation of the Used Fuel Disposition (UFD) R&D campaign, international geologic disposal programs and past work in the U.S. were surveyed to identify viable disposal concepts for crystalline, clay/shale, and salt host media (Hardin et al., 2012). Concepts for disposal of commercial spent nuclear fuel (SNF) and high-level waste (HLW) from reprocessing are relatively advanced in countries such as Finland, France, and Sweden. The UFD work quickly showed that these international concepts are all “enclosed,” whereby waste packages are emplaced in direct or close contact with natural or engineered materials . Alternative “open” modes (emplacement tunnels are kept open after emplacement for extended ventilation) have been limited to the Yucca Mountain License Application Design (CRWMS M&O, 1999). Thermal analysis showed that, if “enclosed” concepts are constrained by peak package/buffer temperature, waste package capacity is limited to 4 PWR assemblies (or 9-BWR) in all media except salt. This information motivated separate studies: 1) extend the peak temperature tolerance of backfill materials, which is ongoing; and 2) develop small canisters (up to 4-PWR size) that can be grouped in larger multi-pack units for convenience of storage, transportation, and possibly disposal (should the disposal concept permit larger packages). A recent result from the second line of investigation is the Task Order 18 report: Generic Design for Small Standardized Transportation, Aging and Disposal Canister Systems (EnergySolution, 2015). This report identifies disposal concepts for the small canisters (4-PWR size) drawing heavily on previous work, and for the multi-pack (16-PWR or 36-BWR).

  4. Multi-Pack Disposal Concepts for Spent Fuel (Revision 1)

    Energy Technology Data Exchange (ETDEWEB)

    Hardin, Ernest [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Matteo, Edward N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hadgu, Teklu [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-01-01

    At the initiation of the Used Fuel Disposition (UFD) R&D campaign, international geologic disposal programs and past work in the U.S. were surveyed to identify viable disposal concepts for crystalline, clay/shale, and salt host media. Concepts for disposal of commercial spent nuclear fuel (SNF) and high-level waste (HLW) from reprocessing are relatively advanced in countries such as Finland, France, and Sweden. The UFD work quickly showed that these international concepts are all “enclosed,” whereby waste packages are emplaced in direct or close contact with natural or engineered materials . Alternative “open” modes (emplacement tunnels are kept open after emplacement for extended ventilation) have been limited to the Yucca Mountain License Application Design. Thermal analysis showed that if “enclosed” concepts are constrained by peak package/buffer temperature, that waste package capacity is limited to 4 PWR assemblies (or 9 BWR) in all media except salt. This information motivated separate studies: 1) extend the peak temperature tolerance of backfill materials, which is ongoing; and 2) develop small canisters (up to 4-PWR size) that can be grouped in larger multi-pack units for convenience of storage, transportation, and possibly disposal (should the disposal concept permit larger packages). A recent result from the second line of investigation is the Task Order 18 report: Generic Design for Small Standardized Transportation, Aging and Disposal Canister Systems. This report identifies disposal concepts for the small canisters (4-PWR size) drawing heavily on previous work, and for the multi-pack (16-PWR or 36-BWR).

  5. PWR and BWR spent fuel assembly gamma spectra measurements

    Science.gov (United States)

    Vaccaro, S.; Tobin, S. J.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Hu, J.; Schwalbach, P.; Sjöland, A.; Trellue, H.; Vo, D.

    2016-10-01

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative-Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

  6. Leaching of spent fuel in the presence of environmental material

    International Nuclear Information System (INIS)

    The aim of this work is the study of the alteration kinetics of spent fuels and the making of a status of the radioactivity released by spent fuels in conditions of direct disposal in deep underground. A system has been fitted inside a shielded cell to study the leaching by synthetic groundwater of fuel powder irradiated at 60 GWJ.tU-1 in the presence of environmental material (clay or granite) at 40 bars and 90 deg. C. This system allows to reach and keep reductive conditions characteristic of the redox conditions of a deep geological repository. The preparation of calibrated spent fuel powders and the recovery of the activity fixed by the environmental materials has required the implementation of specific procedures. Similar experiments have been performed in parallel with Simfuel in a controlled area. A first series of experiments has been carried out in 4 environments for each fuel. Important sorption phenomena take place in the environmental materials and the actinide concentrations stabilize rapidly at low values: 10-8 mol/l for U, 10-12 mol/l for Pu and 10-13-10-14 mol/l for Cm. The activity released by 90Sr at the end of each experiment is about two times higher in the presence of clay than in the presence of granite. The average alteration rates are of about 0.2 mg.m-2/day in the presence of granite and 0.4 to 0.6 mg.m-2/day in the presence of clay. They are comparable to those reported in the literature for reducing conditions. Such tests are necessary to determine the leaching rate of spent fuels in reducing conditions and in the absence of environmental materials in order to show the possible effects of these materials. (J.S.)

  7. Report on interim storage of spent nuclear fuel. Midwestern high-level radioactive waste transportation project

    Energy Technology Data Exchange (ETDEWEB)

    1993-04-01

    The report on interim storage of spent nuclear fuel discusses the technical, regulatory, and economic aspects of spent-fuel storage at nuclear reactors. The report is intended to provide legislators state officials and citizens in the Midwest with information on spent-fuel inventories, current and projected additional storage requirements, licensing, storage technologies, and actions taken by various utilities in the Midwest to augment their capacity to store spent nuclear fuel on site.

  8. Final Generic Environmental Impact Statement. Handling and storage of spent light water power reactor fuel. Volume 2. Appendices

    International Nuclear Information System (INIS)

    This volume contains the following appendices: LWR fuel cycle, handling and storage of spent fuel, termination case considerations (use of coal-fired power plants to replace nuclear plants), increasing fuel storage capacity, spent fuel transshipment, spent fuel generation and storage data, characteristics of nuclear fuel, away-from-reactor storage concept, spent fuel storage requirements for higher projected nuclear generating capacity, and physical protection requirements and hypothetical sabotage events in a spent fuel storage facility

  9. Storage of Spent Nuclear Fuel in Norway: Status and Prospects

    International Nuclear Information System (INIS)

    Spent Nuclear Fuel (SNF) in Norway has arisen from irradiation of fuel in the JEEP I and JEEP II reactors at Kjeller, and in the Halden Boiling Water Reactor (HBWR) in Halden. In total there are some 16 tonnes of SNF, all of which is currently stored on-site, in either wet or dry storage facilities. The greater part of the SNF, 12 tonnes, consists of aluminium-clad fuel, of which 10 tonnes is metallic uranium fuel and the remainder oxide (UO2). Such fuel presents significant challenges with respect to long-term storage and disposal. Current policy is that existing spent fuel will, as far as possible considering its suitability for later direct disposal, be stored until final disposal is possible. Several committees have advised the Government of Norway on, among others, policy issues, storage methods and localisation of a storage facility. Both experts and stakeholders have participated in these committees. This paper presents an overview of the spent fuel in Norway and a description of current storage arrangements. The prospects for long-term storage are then described, including a summary of recommendations made to government, the reactions of various stakeholders to these recommendations, the current status, and the proposed next steps. A recommended policy is to construct a new storage facility for the fuel to be stored for a period of at least 50 years. In the meantime a national final disposal facility should be constructed and taken into operation. It has been recommended that the aluminium-clad fuel be reprocessed in an overseas commercial facility to produce a stable waste form for storage and disposal. This recommendation is controversial, and a decision has not yet been taken on whether to pursue this option. An analysis of available storage concepts for the more modern fuel types resulted in the recommendation to use dual-purpose casks. In addition, it was recommended to construct a future storage facility in a rock hall instead of a free

  10. 77 FR 9591 - List of Approved Spent Fuel Storage Casks: HI-STORM 100, Revision 8

    Science.gov (United States)

    2012-02-17

    ... COMMISSION 10 CFR Part 72 RIN 3150-AJ05 List of Approved Spent Fuel Storage Casks: HI-STORM 100, Revision 8... Commission (NRC or the Commission) is proposing to amend its spent fuel storage cask regulations by revising... Spent Fuel Storage Casks'' to include Amendment No. 8 to Certificate of Compliance (CoC) No....

  11. 77 FR 7211 - Pacific Gas and Electric Company, Diablo Canyon Independent Spent Fuel Storage Installation...

    Science.gov (United States)

    2012-02-10

    ... COMMISSION Pacific Gas and Electric Company, Diablo Canyon Independent Spent Fuel Storage Installation... Manager, Division of Spent Fuel Storage and Transportation, Office of Nuclear Material Safety and... Pacific Gas and Electric Company (PG&E) for the Diablo Canyon (DC) Independent Spent Fuel...

  12. 75 FR 12315 - Pacific Gas and Electric Company; Diablo Canyon Independent Spent Fuel Storage Installation...

    Science.gov (United States)

    2010-03-15

    ... COMMISSION Pacific Gas and Electric Company; Diablo Canyon Independent Spent Fuel Storage Installation..., Division of Spent Fuel Storage and Transportation, Office of Nuclear Material Safety and Safeguards, Mail... Gas and Electric Company (PG&E) for the Diablo Canyon Independent Spent Fuel Storage...

  13. 78 FR 40199 - Draft Spent Fuel Storage and Transportation Interim Staff Guidance

    Science.gov (United States)

    2013-07-03

    ... COMMISSION Draft Spent Fuel Storage and Transportation Interim Staff Guidance AGENCY: Nuclear Regulatory... Regulatory Commission (NRC) requests public comment on Draft Spent Fuel Storage and Transportation Interim... given in NUREG-1927 ``Standard Review Plan for Renewal of Spent Fuel Dry Cask Storage System...

  14. 78 FR 16619 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System

    Science.gov (United States)

    2013-03-18

    ...; ] NUCLEAR REGULATORY COMMISSION 10 CFR Part 72 RIN 3150-AJ22 List of Approved Spent Fuel Storage Casks... Regulatory Commission (NRC) is proposing to amend its spent fuel storage regulations by revising the NAC... within the ``List of Approved Spent Fuel Storage Casks'' to include Amendment No. 3 to Certificate...

  15. 76 FR 29280 - Diablo Canyon Independent Spent Fuel Storage Installation; Notice of Docketing for Amendment...

    Science.gov (United States)

    2011-05-20

    ... COMMISSION Diablo Canyon Independent Spent Fuel Storage Installation; Notice of Docketing for Amendment...: John M. Goshen, Project Manager, Licensing Branch, Division of Spent Fuel Storage and Transportation..., possession, storage and transfer of spent fuel, reactor-related Greater than Class C waste and...

  16. 75 FR 60147 - Calvert Cliffs Nuclear Power Plant, LLC; Independent Spent Fuel Storage Installation; Notice of...

    Science.gov (United States)

    2010-09-29

    ... COMMISSION Calvert Cliffs Nuclear Power Plant, LLC; Independent Spent Fuel Storage Installation; Notice of... Branch, Division of Spent Fuel Storage and Transportation, Office of Nuclear Material Safety and... Independent Spent Fuel Storage Installation (ISFSI) Technical Specifications (TS) be revised as follows: 1....

  17. 78 FR 3454 - Prairie Island, Independent Spent Fuel Storage Installation; Notice of Docketing of Amendment...

    Science.gov (United States)

    2013-01-16

    ... COMMISSION Prairie Island, Independent Spent Fuel Storage Installation; Notice of Docketing of Amendment... Technical Specifications of the TN-40HT cask utilized at its Prairie Island independent spent fuel storage... transfer spent fuel from Prairie Island Nuclear Station Units 1 and 2. Specifically, the amendment seeks...

  18. 76 FR 30980 - Pacific Gas and Electric Company; Humboldt Bay Independent Spent Fuel Storage Installation...

    Science.gov (United States)

    2011-05-27

    ... COMMISSION Pacific Gas and Electric Company; Humboldt Bay Independent Spent Fuel Storage Installation... INFORMATION CONTACT: William Allen, Project Manager, Licensing Branch, Division of Spent Fuel Storage and... a modification to License No. SNM-2514 at its Humboldt Bay Independent Spent Fuel...

  19. 78 FR 61401 - Entergy Nuclear Operations, Inc.; Big Rock Point; Independent Spent Fuel Storage Installation

    Science.gov (United States)

    2013-10-03

    ... COMMISSION Entergy Nuclear Operations, Inc.; Big Rock Point; Independent Spent Fuel Storage Installation..., Inc. (ENO) on June 20, 2012, for the Big Rock Point (BRP) Independent Spent Fuel Storage Installation... Regulatory Evaluation In the Final Rule for Storage of Spent Fuel in NRC-Approved Storage Casks at...

  20. 77 FR 9515 - List of Approved Spent Fuel Storage Casks: HI-STORM 100, Revision 8

    Science.gov (United States)

    2012-02-17

    ... Storage of Spent Fuel at Power Reactor Sites'' (55 FR 29181; July 18, 1990). This rule also established a... RIN 3150-AJ05 List of Approved Spent Fuel Storage Casks: HI-STORM 100, Revision 8 AGENCY: Nuclear... Commission) is amending its spent fuel storage regulations by revising the Holtec International HI-STORM...

  1. 78 FR 78693 - List of Approved Spent Fuel Storage Casks: Transnuclear, Inc. Standardized NUHOMS® Cask System

    Science.gov (United States)

    2013-12-27

    ... RIN 3150-AJ10 List of Approved Spent Fuel Storage Casks: Transnuclear, Inc. Standardized NUHOMS Cask... final rule amended the NRC's spent fuel storage regulations by revising the Transnuclear, Inc. Standardized NUHOMS Cask System listing within the ``List of Approved Spent Fuel Storage Casks'' to...

  2. 76 FR 70331 - List of Approved Spent Fuel Storage Casks: MAGNASTOR ® System, Revision 2

    Science.gov (United States)

    2011-11-14

    ... part 72, entitled ``General License for Storage of Spent Fuel at Power Reactor Sites'' (55 FR 29181... spent fuel storage cask designs. The NRC subsequently issued a final rule on November 21, 2008 (73 FR... COMMISSION 10 CFR Part 72 RIN 3150-AI91 List of Approved Spent Fuel Storage Casks: MAGNASTOR System,......

  3. 75 FR 42339 - List of Approved Spent Fuel Storage Casks: NAC-MPC System, Revision 6

    Science.gov (United States)

    2010-07-21

    ...; ] NUCLEAR REGULATORY COMMISSION 10 CFR Part 72 RIN 3150--AI88 List of Approved Spent Fuel Storage Casks: NAC.... Nuclear Regulatory Commission (NRC) is proposing to amend its spent fuel storage cask regulations by... 72. PART 72--LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL,...

  4. 78 FR 123 - Diablo Canyon, Independent Spent Fuel Storage Installation; License Amendment Request...

    Science.gov (United States)

    2013-01-02

    ... COMMISSION Diablo Canyon, Independent Spent Fuel Storage Installation; License Amendment Request, Opportunity... Spent Fuel Storage and Transportation, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear... Canyon Independent Spent Fuel Storage Installation (ISFSI) site located in San Luis Obispo...

  5. 78 FR 69460 - Proposed License Renewal of the Prairie Island Independent Spent Fuel Storage Installation

    Science.gov (United States)

    2013-11-19

    ... COMMISSION Proposed License Renewal of the Prairie Island Independent Spent Fuel Storage Installation AGENCY... Prairie Island Nuclear Generating Plant (PINGP) site-specific Independent Spent Fuel Storage Installation... Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater...

  6. 75 FR 33678 - List of Approved Spent Fuel Storage Casks: MAGNASTOR System, Revision 1

    Science.gov (United States)

    2010-06-15

    ... COMMISSION 10 CFR Part 72 RIN 3150-AI86 List of Approved Spent Fuel Storage Casks: MAGNASTOR System, Revision... Regulatory Commission (NRC) is amending its spent fuel storage regulations by revising the NAC International Inc. (NAC) MAGNASTOR System listing within the ``List of Approved Spent Fuel Storage Casks''...

  7. 78 FR 56947 - Prairie Island; Independent Spent Fuel Storage Installation; Notice of Docketing of Amendment...

    Science.gov (United States)

    2013-09-16

    ... COMMISSION Prairie Island; Independent Spent Fuel Storage Installation; Notice of Docketing of Amendment... spent fuel storage installation located in Welch, Minnesota. DATES: Requests for a hearing or petition... Prairie Island independent spent fuel storage installation located in Welch, Minnesota (ADAMS Accession...

  8. 75 FR 42292 - List of Approved Spent Fuel Storage Casks: NAC-MPC System, Revision 6

    Science.gov (United States)

    2010-07-21

    ... for Storage of Spent Fuel at Power Reactor Sites'' (55 FR 29181; July 18, 1990). This rule also... COMMISSION 10 CFR Part 72 RIN 3150-AI88 List of Approved Spent Fuel Storage Casks: NAC-MPC System, Revision 6... Commission (NRC) is amending its spent fuel storage regulations by revising the NAC International Inc....

  9. 76 FR 9381 - Notice of Availability of Interim Staff Guidance Documents for Spent Fuel Storage Casks

    Science.gov (United States)

    2011-02-17

    ... COMMISSION Notice of Availability of Interim Staff Guidance Documents for Spent Fuel Storage Casks AGENCY... Gordon, Structural Mechanics and Materials Branch, Division of Spent Fuel Storage and Transportation... performing technical reviews of spent fuel storage and transportation packaging licensing actions.'' This...

  10. 78 FR 59073 - Pacific Gas and Electric Company; Humboldt Bay Independent Spent Fuel Storage Installation...

    Science.gov (United States)

    2013-09-25

    ... COMMISSION Pacific Gas and Electric Company; Humboldt Bay Independent Spent Fuel Storage Installation...) independent spent fuel storage installation (ISFSI). ADDRESSES: Please refer to Docket ID NRC-2011-0115 when... storage of spent fuel and associated radioactive materials at the HB ISFSI. Pursuant to 10 CFR...

  11. 10 CFR 72.214 - List of approved spent fuel storage casks.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false List of approved spent fuel storage casks. 72.214 Section... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General License for Storage of Spent Fuel at Power Reactor Sites § 72.214 List of approved...

  12. 75 FR 33736 - List of Approved Spent Fuel Storage Casks: MAGNASTOR System, Revision 1

    Science.gov (United States)

    2010-06-15

    ... COMMISSION 10 CFR Part 72 RIN 3150--AI86 List of Approved Spent Fuel Storage Casks: MAGNASTOR System... Regulatory Commission (NRC) is proposing to amend its spent fuel storage cask regulations by revising the NAC International, Inc. (NAC), MAGNASTOR System listing within the ``List of Approved Spent Fuel Storage Casks''...

  13. 78 FR 63408 - List of Approved Spent Fuel Storage Casks: Transnuclear, Inc. Standardized NUHOMS® Cask System

    Science.gov (United States)

    2013-10-24

    ... COMMISSION 10 CFR Part 72 RIN 3150-AJ10 List of Approved Spent Fuel Storage Casks: Transnuclear, Inc.... Nuclear Regulatory Commission (NRC) is amending its spent fuel storage regulations by revising the Transnuclear, Inc. Standardized NUHOMS Cask System listing within the ``List of Approved Spent Fuel...

  14. 77 FR 4203 - List of Approved Spent Fuel Storage Casks: MAGNASTOR® System, Revision 2

    Science.gov (United States)

    2012-01-27

    ... 3150-AI91 List of Approved Spent Fuel Storage Casks: MAGNASTOR System, Revision 2 AGENCY: Nuclear... amended the NRC's spent fuel storage regulations by revising the NAC International, Inc. (NAC) MAGNASTOR System listing within the ``List of Approved Spent Fuel Storage Casks'' to include Amendment No. 2...

  15. 76 FR 22935 - Calvert Cliffs Nuclear Power Plant, LLC Independent Spent Fuel Storage Installation; Notice of...

    Science.gov (United States)

    2011-04-25

    ... Class C (GTCC) waste and other radioactive materials associated with spent fuel storage at the CCNPP... granted, the renewed license will authorize the applicant to continue to store spent fuel in a dry cask... COMMISSION Calvert Cliffs Nuclear Power Plant, LLC Independent Spent Fuel Storage Installation; Notice...

  16. 75 FR 77017 - Nextera Energy Seabrook, LLC Seabrook Station Independent Spent Fuel Storage Installation; Exemption

    Science.gov (United States)

    2010-12-10

    ... COMMISSION Nextera Energy Seabrook, LLC Seabrook Station Independent Spent Fuel Storage Installation; Exemption 1.0 Background NextEra Energy Seabrook, LLC (NextEra, the licensee) is the holder of Facility..., subpart K, a general license is issued for the storage of spent fuel in an independent spent fuel...

  17. Radiochemical analyses of several spent fuel Approved Testing Materials

    Energy Technology Data Exchange (ETDEWEB)

    Guenther, R.J.; Blahnik, D.E.; Wildung, N.J.

    1994-09-01

    Radiochemical characterization data are described for UO{sub 2} and UO{sub 2} plus 3 wt% Gd{sub 2}O{sub 3} commercial spent nuclear fuel taken from a series of Approved Testing Materials (ATMs). These full-length nuclear fuel rods include MLA091 of ATM-103, MKP070 of ATM-104, NBD095 and NBD131 of ATM-106, and ADN0206 of ATM-108. ATMs 103, 104, and 106 were all irradiated in the Calvert Cliffs Nuclear Power Plant (Reactor No.1), a pressurized-water reactor that used fuel fabricated by Combustion Engineering. ATM-108 was part of the same fuel bundle designed as ATM-105 and came from boiling-water reactor fuel fabricated by General Electric and irradiated in the Cooper Nuclear Power Plant. Rod average burnups and expected fission gas releases ranged from 2,400 to 3,700 GJ/kgM. (25 to 40 Mwd/kgM) and from less than 1% to greater than 10%, respectively, depending on the specific ATM. The radiochemical analyses included uranium and plutonium isotopes in the fuel, selected fission products in the fuel, fuel burnup, cesium and iodine on the inner surfaces of the cladding, {sup 14}C in the fuel and cladding, and analyses of the gases released to the rod plenum. Supporting examinations such as fuel rod design and material descriptions, power histories, and gamma scans used for sectioning diagrams are also included. These ATMs were examined as part of the Materials Characterization Center Program conducted at Pacific Northwest Laboratory provide a source of well-characterized spent fuel for testing in support of the US Department of Energy Office of Civilian Radioactive Waste Management Program.

  18. Disposal of spent fuel from German nuclear power plants - 16028

    International Nuclear Information System (INIS)

    The 'direct disposal of spent fuel' as a part of the current German reference concept was developed as an alternative to spent fuel reprocessing and vitrified HLW disposal. The technical facilities necessary for the implementation of this part of the reference concept, the so called POLLUXR concept, i.e. interim storage buildings for casks containing spent fuel, a pilot conditioning facility, and a special cask 'POLLUX' for final disposal have been built. With view to a geological salt formation all handling procedures for the direct disposal of spent fuel were tested aboveground in full-scale test facilities. To optimise the reference concept, all operational steps have been reviewed for possible improvements. The two additional concepts for the direct disposal of SF are the BSK 3 concept and the DIREGT concept. Both concepts rely on borehole emplacement technology, vertical boreholes for the BSK 3 concept und horizontal boreholes for the DIREGT concept. Supported by the EU and the German Federal Ministry of Economics and Technology (BMWi), DBE TECHNOLOGY built an aboveground full-scale test facility to simulate all relevant handling procedures for the BSK 3 disposal concept. GNS (Company for Nuclear Service), representing the German utilities, provided the main components and its know-how concerning cask design and manufacturing. The test program was concluded recently after more than 1.000 emplacement operations had been performed successfully. The BSK 3 emplacement system in total comprises an emplacement device, a borehole lock, a transport cart, a transfer cask which will shuttle between the aboveground conditioning facility and the underground repository, and the BSK 3 canister itself, designed to contain the fuel rods of three PWR-fuel assemblies with a total of about 1.6 tHM. The BSK 3 concept simplifies the operation of the repository because the handling procedures and techniques can also be applied for the disposal of reprocessing residues. In addition

  19. Modelling spent fuel and HLW behaviour in repository conditions

    Energy Technology Data Exchange (ETDEWEB)

    Esparza, A. M.; Esteban, J. A.

    2003-07-01

    The aim of this report is to give the reader an overall insight of the different models, which are used to predict the long-term behaviour of the spent fuels and HLW disposed in a repository. The models must be established on basic data and robust kinetics describing the mechanisms controlling spent fuel alteration/dissolution in a repository. The UO2 matrix, or source term, contains embedded in it the , majority of radionuclides of the spent fuel (some are in the gap cladding). For this reason the SF radionuclides release models play a significant role in the performance assessment of radioactive waste disposal. The differences existing between models published in the literature are due to the conceptual understanding of the processes and the degree of the conservatism used with the parameter values, and the boundary conditions. They mainly differ in their level of simplification and their final objective. Sometimes are focused the show compliance with regulatory requirements, other to support decision making, to increase the level of confidence of public and scientific community, could be empirical, semi-empirical or analytical. The models take into account the experimental results from radionuclides releases and their extrapolation to the very long term. Its necessary a great statistics for have a representative dissolution rate, due at the number of experimental results is not very high and many of them show a great scatter, independently of theirs different compositions by axial and radial variations, due to linear power or local burnup. On the other hand, it is difficult to predict the spent fuel behaviour over the long term, based in short term experiments. In this report is given a little description of the radionuclides distribution in the spent fuel and also in the cladding/pellet gap, grain boundary, cracks and rim zones (the matrix rim zone can be considered with an especial characteristics very different to the rest of the spent fuel), and structural

  20. The Study on Regulations of Interim Spent Fuel Storage Facility in Japan

    International Nuclear Information System (INIS)

    The expansion of nuclear power generation inevitably result in the increase of spent nuclear fuel generation. Every year ∼700t of spent nuclear fuels are accumulated and total of 10,761t of spent nuclear fuels generated from 20 reactors are stored within the reactor sites by the end of 2009. If this trend is maintained, it is expected that Korea will have ∼30,000t of spent nuclear fuel by 2030. The capacity of pools in reactor sites is expected to be reached from 2016 and Korea need to prepare measures including the introduction of interim storage facility urgently. Considering the domestic situation described so far, it is be very necessary to survey and analyze the regulations of interim storage facilities of advanced countries. This report excerpts and analyzes the report of Japanese Nuclear Energy Safety Organization (JNES) on the regulations of interim storage facilities which was issues before the construction of the Mutz Interim Storage Facility. This report will be effectively used for the preparation of interim storage facility of Korea

  1. Impact of dissolution on the uncertainty of spent fuel analysis

    International Nuclear Information System (INIS)

    One of the objectives of the French Alternative Energies and Atomic Energy Commission in the Marcoule Centre is to accurately quantify the composition of nuclear spent fuel, i.e. to determine the concentration of each isotope with suitable measurement uncertainty. These analysis results are essential for the validation of calculation codes used for the simulation of fuel behaviour in nuclear reactors and for nuclear matter accountancy. The different experimental steps are first the reception of a piece of spent fuel rod at the laboratory of dissolution studies, and then dissolution in a hot cell of a sample of the spent fuel rod received. Several steps are necessary to obtain a complete dissolution. Taking into account these process steps, and not only those of analysis for the evaluation of measurement uncertainties, is new, and is described in this paper. The uncertainty estimation incorporating the process has been developed following the approach proposed by the Guide to the Expression of Uncertainty in Measurement (GUM). The mathematical model of measurement was established by examining the dissolution process step by step. The law of propagation of uncertainty was applied to this model. A point by point examination of each step of the process permitted the identification of all sources of uncertainties considered in this propagation for each input variable. The measurement process presented involves the process and the analysis. The contents of this document show the importance of taking the process into account in order to give a more reliable uncertainty assessment to the result of a concentration or isotope ratio of two isotopes in spent fuel. (author)

  2. Options for the management of spent fuel from RA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Aden, V.G.; Bulkin, S.Yu.; Sokolov, A.V. [Research and Development Institute of Power Engineering, Moscow (Russian Federation); Matausek, M.V.; Vukadin, Z. [VINCA Institute of Nuclear Science, Belgrade (Yugoslavia)

    2000-07-01

    Leaking fuel elements can be placed in a transportation cask or stored in a long-term dry storage facility if they are placed within an additional metal can. Two different ways of RA reactor spent fuel elements preparation for transportation or long-term storage are considered: 'all fuel elements canning without leak-tightness testing' and 'all fuel elements leak-tightness testing'. Comparison of these two options is performed according to the following criteria: the radiological influence on workers and environment; the need for non-standard equipment fabrication; the time required for work performance; the possibility for fuel elements deterioration during transportation or storage. It is believed that the first option offers several distinct advantages, which can be summarized as: greater reliability in the course of transportation or dry storage; higher safety for workers; lower expenditures for non-standard equipment manufacturing; shorter duration of work. (author)

  3. Exploratory Design of a Reactor/Fuel Cycle Using Spent Nuclear Fuel Without Conventional Reprocessing - 13579

    Energy Technology Data Exchange (ETDEWEB)

    Bertch, Timothy C.; Schleicher, Robert W.; Rawls, John D. [General Atomics 3550 General Atomics Court San Diego, CA 92130 (United States)

    2013-07-01

    General Atomics has started design of a waste to energy nuclear reactor (EM2) that can use light water reactor (LWR) spent nuclear fuel (SNF). This effort addresses two problems: using an advanced small reactor with long core life to reduce nuclear energy overnight cost and providing a disposal path for LWR SNF. LWR SNF is re-fabricated into new EM2 fuel using a dry voloxidation process modeled on AIROX/ OREOX processes which remove some of the fission products but no heavy metals. By not removing all of the fission products the fuel remains self-protecting. By not separating heavy metals, the process remains proliferation resistant. Implementation of Energy Multiplier Module (EM2) fuel cycle will provide low cost nuclear energy while providing a long term LWR SNF disposition path which is important for LWR waste confidence. With LWR waste confidence recent impacts on reactor licensing, an alternate disposition path is highly relevant. Centered on a reactor operating at 250 MWe, the compact electricity generating system design maximizes site flexibility with truck transport of all system components and available dry cooling features that removes the need to be located near a body of water. A high temperature system using helium coolant, electricity is efficiently produced using an asynchronous high-speed gas turbine while the LWR SNF is converted to fission products. Reactor design features such as vented fuel and silicon carbide cladding support reactor operation for decades between refueling, with improved fuel utilization. Beyond the reactor, the fuel cycle is designed so that subsequent generations of EM2 reactor fuel will use the previous EM2 discharge, providing its own waste confidence plus eliminating the need for enrichment after the first generation. Additional LWR SNF is added at each re-fabrication to replace the removed fission products. The fuel cycle uses a dry voloxidation process for both the initial LWR SNF re-fabrication and later for EM2

  4. Interim Dry Storage of Spent Fuel in Casks

    International Nuclear Information System (INIS)

    French option for the back end of the fuel cycle is reprocessing of used fuel and recycling the fissile material, except some very specific fuel stored in vaults (dry conditions). Used fuel management solutions studied by AREVA for various countries allow for either direct transport to the reprocessing plant, or interim storage and transport after storage of used fuel. Interim storage solutions are wet storage or dry storage (DSC, metal casks or vault systems). When the decision on used fuel management has been postponed, some extension of interim storage duration is considered, therefore it becomes necessary to study used fuel and cask material behaviour and deterioration mechanisms. One objective of this R&D was to review research efforts on spent fuel behaviour and Dry storage experience in casks. Particularly we were interested in the assessment of retrievability of fuel after storage for further use. A review therefore, was made of the effect of storage time/ temperatures and of loading/ drying operation on used fuel integrity. R&D programmes were also carried out on the evaluation of cask materials in long term, especially materials susceptible to degradation

  5. TRIGA reactor spent fuel pool under severe earthquake conditions

    International Nuclear Information System (INIS)

    Supplemental criticality safety analysis of a pool type storage for TRIGA spent fuel at 'Jozef Stefan' Institute in Ljubljana, Slovenia, is presented. Previous results (Ravnik, M, Glumac, B., 1996) have shown that subcriticality is not guaranteed for some postulated accidents. To mitigate this deficiency, a study was made about replacing a certain number of fuel elements in the rack with absorber rods (Glumac, B., Ravnik, M., Logar, M., 1997) to lower the supercriticality probability, when the pitch is decreased to contact (as a consequence of a severe earthquake) in a square arrangement. The criticality analysis for the hexagonal contact pitch is presented in this paper, following the same scenario as outlined above. The Monte Carlo computer code MCNP4B with ENDF-B/VI library and detailed three dimensional geometry was used. First, the analysis about the influence of the number of triangular fuel piles on the bottom that could appear, if the fuel rack, made of three segments, disintegrates, is presented. Next, the number of uniformly mixed absorber rods in the lattice needed to sustain the subcriticality of the storage for hexagonal contact pitch is studied. Because of supercriticality possibility due to random mixing of the absorber rods in the case of lattice compaction, a probabilistic study was made in order to sample the probability density functions for random lattice loadings of the absorber rods. The results show that reasonably low probabilities for supercriticality can be achieved even when fresh 12 wt.% standard TRIGA fuel is stored in the spent fuel pool. (orig.)

  6. Integrated Strategy for Spent Fuel Management

    International Nuclear Information System (INIS)

    Today, the United States of America rely on the Yucca Mountain geological repository for the storage of SF. In order to replace it, an integrated strategy is being elaborated by NRC, DOE and the new blue ribbon commission, initiated by the Obama administration. The strategy aims to update licensing, regulation and enhance research efforts in 3 inter dependant cores: transportation, reprocessing and waste disposal. The main goal is to bring a solution to environmental, social, security and safeguards issues. However, this integrated strategy is still under development, as many questions are yet to be answered. I recently began serving in my present capacity as the Deputy Executive Director for Materials, Waste, Research, State, Tribal, and Compliance Programs. Before this, I served for three years as the Director of the Nuclear Regulatory Commission’s Office of Nuclear Material Safety and Safeguards, where I lead regulatory programs for SF storage, transportation, fuel cycle facilities, high-level waste disposal, and domestic and international safeguards. Representatives from the NRC will be making several presentations throughout this conference. We welcome your candid feedback and suggestions on our regulatory programs, as well as opportunities to collaborate with our international partners on a bilateral and multilateral basis. (author)

  7. Spent fuel assembly hardware: Characterization and 10 CFR 61 classification for waste disposal: Volume 1, Activation measurements and comparison with calculations for spent fuel assembly hardware

    Energy Technology Data Exchange (ETDEWEB)

    Luksic, A.

    1989-06-01

    Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel. During consolidation, the fuel pins are removed from an intact fuel assembly and repackaged into a more compact configuration. After repackaging, approximately 30 kg of residual spent fuel assembly hardware per assembly remains that is also radioactive and requires disposal. Understanding the nature of this secondary waste stream is critical to designing a system that will properly handle, package, store, and dispose of the waste. This report presents a methodology for estimating the radionuclide inventory in irradiated spent fuel hardware. Ratios are developed that allow the use of ORIGEN2 computer code calculations to be applied to regions that are outside the fueled region. The ratios are based on the analysis of samples of irradiated hardware from spent fuel assemblies. The results of this research are presented in three volumes. In Volume 1, the development of scaling factors that can be used with ORIGEN2 calculations to estimate activation of spent fuel assembly hardware is documented. The results from laboratory analysis of irradiated spent-fuel hardware samples are also presented in Volume 1. In Volumes 2 and 3, the calculated flux profiles of spent nuclear fuel assemblies are presented for pressurized water reactors and boiling water reactors, respectively. The results presented in Volumes 2 and 3 were used to develop the scaling factors documented in Volume 1. 5 refs., 4 figs., 21 tabs.

  8. Spent fuel assembly hardware: Characterization and 10 CFR 61 classification for waste disposal: Volume 2, Calculated activity profiles of spent nuclear fuel assembly hardware for pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Short, S.M.; Luksic, A.T.; Lotz, T.L.; Schutz, M.E.

    1989-06-01

    Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel as required by the Nuclear Waste Policy Act of 1982. During consolidation, the fuel pins are removed from an intact fuel assembly and repackaged into a more compact configuration. After repackaging, approximately 30 kg of residual spent fuel assembly hardware per assembly remains that is also radioactive and requires disposal. Understanding the nature of this secondary waste stream is critical to designing a system that will properly handle, package, store, and dispose of the waste. This report present a methodology for estimating the radionuclide inventory in irradiated spent fuel hardware. Ratios are developed that allow the use of ORIGEN2 computer code calculations to be applied to regions that are outside the fueled region. The ratios are based on the analysis of samples of irradiated hardware from spent fuel assemblies. The results of this research are presented in three volumes. In Volume 1, the development of scaling factors that can be used with ORIGEN2 calculations to estimate activation of spent fuel assembly hardware is documented. The results from Laboratory analysis of irradiated spent-fuel hardware samples are also presented in Volume 1. In Volumes 2 and 3, the calculated flux profiles of spent nuclear fuel assemblies are presented for pressurized water reactors and boiling water reactors, respectively. The results presented in Volumes 2 and 3 were used to develop the scaling factors documented in Volume 1.

  9. Spent fuel assembly hardware: Characterization and 10 CFR 61 classification for waste disposal: Volume 3, Calculated activity profiles of spent nuclear fuel assembly hardware for boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Short, S.M.; Luksic, A.T.; Schutz, M.E.

    1989-06-01

    Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel as required by the Nuclear Waste Policy Act of 1982. During consolidation, the fuel pins are removed from an intact fuel assembly and repackaged into a more compact configuration. After repackaging, approximately 30 kg of residual spent fuel assembly hardware per assembly that is also radioactive and required disposal. Understanding the nature of this secondary waste stream is critical to designing a system that will properly handle, package, store, and dispose of the waste. This report presents a methodology for estimating the radionuclide inventory in irradiated spent fuel hardware. Ratios are developed that allow the use of ORIGEN2 computer code calculations to be applied to regions that are outside the fueled region. The ratios are based on the analysis of samples of irradiated hardware from spent fuel assemblies. The results of this research are presented in three volumes. In Volume 1, the development of scaling factors that can be used with ORIGEN2 calculations to estimate activation of spent fuel assembly hardware is documented. The results from laboratory analysis of irradiated spent-fuel hardware samples are also presented in Volume 1. In Volume 2 and 3, the calculated flux profiles of spent nuclear fuel assemblies are presented for pressurized water reactors and boiling water reactors, respectively. The results presented in Volumes 2 and 3 were used to develop the scaling factors documented in Volume 1.

  10. Spent Nuclear Fuel Management: Levelized Cost of Electricity Generation and Analysis of Various Production Scenarios

    Directory of Open Access Journals (Sweden)

    Laura Rodriguez-Penalonga

    2016-03-01

    Full Text Available This article aims to analyze the results of an economic study carried out to compare the influence of nuclear production capacity in different countries. The analysis is based on LCOEs (levelized cost of electricity for three back-end strategies: open cycle, closed cycle and advanced closed cycle. The results show that costs are not a relevant criteria in order to select an energy policy for the spent nuclear fuel management.

  11. International symposium on storage of spent fuel from power reactors. Book of extended synopses

    International Nuclear Information System (INIS)

    This book of extended synopses includes papers presented at the International Symposium on Storage of Spent Fuel from Power Reactors organized by IAEA and held in Vienna from 9 to 13 November 1998. It deals with the problems of spent fuel management being an outstanding stage in the nuclear fuel cycle, strategy of interim spent fuel storage, transportation and encapsulation of spent fuel elements from power reactors. Spent fuel storage facilities at reactor sites are always wet while spent fuel storage facilities away from reactor are either wet or dry including casks and vaults. Different design solutions and constructions of storage or transportation casks as well as storing facilities are presented, as well as status of spent fuel storage together with experiences achieved in a number of member states, in the frame of safety, licensing and regulating procedures

  12. Experience with failed or damaged spent fuel and its impacts on handling

    International Nuclear Information System (INIS)

    Spent fuel management planning needs to include consideration of failed or damaged spent light-water reactor (LWR) fuel. Described in this paper, which was prepared under the Commercial Spent Fuel Management (CSFM) Program that is sponsored by the US Department of Energy (DOE), are the following: the importance of fuel integrity and the behavior of failed fuel, the quantity and burnup of failed or damaged fuel in storage, types of defects, difficulties in evaluating data on failed or damaged fuel, experience with wet storage, experience with dry storage, handling of failed or damaged fuel, transporting of fuel, experience with higher burnup fuel, and conclusions. 15 refs

  13. Advanced Fuels Campaign 2012 Accomplishments

    Energy Technology Data Exchange (ETDEWEB)

    Not Listed

    2012-11-01

    The Advanced Fuels Campaign (AFC) under the Fuel Cycle Research and Development (FCRD) program is responsible for developing fuels technologies to support the various fuel cycle options defined in the DOE Nuclear Energy Research and Development Roadmap, Report to Congress, April 2010. The fiscal year 2012 (FY 2012) accomplishments are highlighted below. Kemal Pasamehmetoglu is the National Technical Director for AFC.

  14. Hot isostatic pressing of ceramic waste from spent nuclear fuel

    International Nuclear Information System (INIS)

    Argonne National Laboratory has developed a process to immobilize waste salt containing fission products, uranium, and transuranic elements as chlorides in a glass-bonded ceramic waste form. This salt was generated in the electrorefining operation used in electrometallurgical treatment of spent Experimental Breeder Reactor-II fuel. The ceramic waste process culminated with a hot isostatic pressing operation. This paper reviews the installation and operation of a hot isostatic press in a radioactive environment. Processing conditions for the hot isostatic press are presented for non-irradiated material and irradiated material. Sufficient testing was performed to demonstrate that a hot isostatic press could be used as the final step of the processing of ceramic waste for the electrometallurgical spent fuel treatment process

  15. Risk of transporting spent nuclear fuel by truck

    International Nuclear Information System (INIS)

    The risk methodology used to evaluate the risk in shipping spent fuel includes: 1) a description of the spent fuel transport system, 2) identification of potential release sequences, 3) evaluation of the probabilities and consequences of the releases, and 4) calculation and assessment of the risk. The system description includes projected industry characteristics, amounts to be shipped, shipping package descriptions, material characteristics, transport mode, transport routes used and weather and population distribution information. Release sequences are identified by fault tree analysis tehniques. Releases are evaluated using package failure data, normal transport and transport accident environment data and mathematical models for material dispersion and resultant health effects. This information is combined to calculate the shipping system risk which is compared to other known risks. The data may be further analyzed to determine the primary contributors to the risk and identify possible methods for reducing the risk, if the current risk level is judged by society to be unacceptable

  16. Thermomechanical modeling of the Spent Fuel Test-Climax

    International Nuclear Information System (INIS)

    The Spent Fuel Test-Climax (SFT-C) was conducted to evaluate the feasibility of retrievable deep geologic storage of commercially generated spent nuclear-reactor fuel assemblies. One of the primary aspects of the test was to measure the thermomechanical response of the rock mass to the extensive heating of a large volume of rock. Instrumentation was emplaced to measure stress changes, relative motion of the rock mass, and tunnel closures during three years of heating from thermally decaying heat sources, followed by a six-month cooldown period. The calculations reported here were performed using the best available input parameters, thermal and mechanical properties, and power levels which were directly measured or inferred from measurements made during the test. This report documents the results of these calculations and compares the results with selected measurements made during heating and cooling of the SFT-C

  17. Nuclear Waste Imaging and Spent Fuel Verification by Muon Tomography

    CERN Document Server

    Jonkmans, G; Jewett, C; Thompson, M

    2012-01-01

    This paper explores the use of cosmic ray muons to image the contents of shielded containers and detect high-Z special nuclear materials inside them. Cosmic ray muons are a naturally occurring form of radiation, are highly penetrating and exhibit large scattering angles on high Z materials. Specifically, we investigated how radiographic and tomographic techniques can be effective for non-invasive nuclear waste characterization and for nuclear material accountancy of spent fuel inside dry storage containers. We show that the tracking of individual muons, as they enter and exit a structure, can potentially improve the accuracy and availability of data on nuclear waste and the contents of Dry Storage Containers (DSC) used for spent fuel storage at CANDU plants. This could be achieved in near real time, with the potential for unattended and remotely monitored operations. We show that the expected sensitivity, in the case of the DSC, exceeds the IAEA detection target for nuclear material accountancy.

  18. Safety analysis of disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    The spent fuel from the Olkiluoto NPP (TVO I and II) is planned to be disposed of in a repository to be constructed at a depth of about 500 meters in the crystalline bedrock. The thesis is dealing with the safety analysis of the disposal. The main topics presented in the thesis are: (1) The amount of radioactive properties of the spent fuel, (2) The canister design and the planned disposal concept, (3) The results of the preliminary site investigations, (4) Discussion of the multi-barrier principle, (5) The general principles and methodology of the TVO-92 safety analysis, (6) Groundwater flow analysis, (7) Durability and behaviour of the canister, (8) Biosphere analysis and reference scenario, and (9) The sensitivity and uncertainty analyses. (246 refs., 75 figs., 44 tabs.)

  19. Recommendations on Fuel Parameters for Standard Technical Specifications for Spent Fuel Storage Casks

    Energy Technology Data Exchange (ETDEWEB)

    Bowman, S.M.

    2001-03-08

    The U.S. Nuclear Regulatory Commission (NRC) is currently reviewing the technical specifications for spent fuel storage casks in an effort to develop standard technical specifications (STS) that define the allowable spent nuclear fuel (SNF) contents. One of the objectives of the review is to minimize the level of detail in the STS that define the acceptable fuel types. To support this initiative, this study has been performed to identify potential fuel specification parameters needed for criticality safety and radiation shielding analysis and rank their importance relative to a potential compromise of the margin of safety.

  20. Review of Drying Methods for Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Large, W.S.

    1999-10-21

    SRTC is developing technology for direct disposal of aluminum spent nuclear fuel (SNF). The development program includes analyses and tests to support design and safe operation of a facility for ''road ready'' dry storage of SNF-filled canisters. The current technology development plan includes review of available SNF drying methods and recommendation of a drying method for aluminum SNF.

  1. Review of partitioning proposals for spent nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Bowersox, D.F.

    1976-07-01

    The initial phase of a study about recovery of valuable fission products from spent nuclear fuels has been to review various partitioning proposals. This report briefly describes the aqueous Purex process, the salt transport process, melt refining, fluoride volatility process, and gravimetric separations. All these processes appear to be possible technically, but further research will be necessary to determine which are most feasible. This review includes general recommendations for experimental research and development of several partitioning options.

  2. Handling encapsulated spent fuel in a geologic repository environment

    International Nuclear Information System (INIS)

    In support of the Spent Fuel Test-Climate at the U.S. Department of Energy's Nevada Test Site, a spent-fuel canister handling system has been designed, deployed, and operated successfully during the past five years. This system transports encapsulated commercial spent-fuel assemblies between the packaging facility and the test site (approx. 100 km), transfers the canisters 420 m vertically to and from a geologic storage drift, and emplaces or retrieves the canisters from the storage holes in the floor of the drift. The spent-fuel canisters are maintained in a fully shielded configuration at all times during the handling cycle, permitting manned access at any time for response to any abnormal conditions. All normal operations are conducted by remote control, thus assuring as low as reasonably achievable exposures to operators; specifically, we have had no measurable exposure during 30 canister transfer operations. While not intended to be prototypical of repository handling operations, the system embodies a number of concepts, now demonstrated to be safe, reliable, and economical, which may be very useful in evaluating full-scale repository handling alternatives in the future. Among the potentially significant concepts are: Use of an integral shielding plug to minimize radiation streaming at all transfer interfaces. Hydraulically actuated transfer cask jacking and rotation features to reduce excavation headroom requirements. Use of a dedicated small diameter (0.5 m) drilled shaft for transfer between the surface and repository workings. A wire-line hoisting system with positive emergency braking device which travels with the load. Remotely activated grapples - three used in the system - which are insensitive to load orientation. Rail-mounted underground transfer vehicle operated with no personnel underground

  3. Spent nuclear fuel canister storage building conceptual design report

    Energy Technology Data Exchange (ETDEWEB)

    Swenson, C.E. [Westinghouse Hanford Co., Richland, WA (United States)

    1996-01-01

    This Conceptual Design Report provides the technical basis for the Spent Nuclear Fuels Project, Canister Storage Building, and as amended by letter (correspondence number 9555700, M.E. Witherspoon to E.B. Sellers, ``Technical Baseline and Updated Cost Estimate for the Canister Storage Building``, dated October 24, 1995), includes the project cost baseline and Criteria to be used as the basis for starting detailed design in fiscal year 1995.

  4. Nuclear Waste Imaging and Spent Fuel Verification by Muon Tomography

    OpenAIRE

    Jonkmans, G.; Anghel, V. N. P.; Jewett, C.; Thompson, M.

    2012-01-01

    This paper explores the use of cosmic ray muons to image the contents of shielded containers and detect high-Z special nuclear materials inside them. Cosmic ray muons are a naturally occurring form of radiation, are highly penetrating and exhibit large scattering angles on high Z materials. Specifically, we investigated how radiographic and tomographic techniques can be effective for non-invasive nuclear waste characterization and for nuclear material accountancy of spent fuel inside dry stor...

  5. Dry store for spent fuel elements from nuclear reactors

    International Nuclear Information System (INIS)

    In the dry store for spent fuel elements from nuclear reactors which are enclosed in storage tubes and cooled with air, the storage tubes being arranged in shafts of a storage building, a loading device is provided underneath the shafts and in a cooling air shaft designed for transporting. The loading device therefore requires only a small lifting height and the chances of storage tubes falling from great heights are excluded. This invention is applicable in particular for intermediate stores. (orig./RW)

  6. Review of decommissioning, spent fuel and radwaste management in Slovakia

    International Nuclear Information System (INIS)

    Two nuclear power plants with two WWER reactors are currently under operation in Jaslovske Bohunice and NPP A-1 is under decommissioning on the same site. At the second nuclear site in the Slovak Republic in Mochovce third nuclear power plant with two units is in operation. In accordance with the basic Slovak legislation (Act on Peaceful Utilisation of Nuclear Energy) defining the responsibilities, roles and authorities for all organisations involved in the decommissioning of nuclear installations Nuclear Regulatory Authority requires submission of conceptual decommissioning plans by the licensee. The term 'decommissioning' is used to describe the set of actions to be taken at the end of the useful life of a facility, in order to retire the facility from service while, simultaneously, ensuring proper protection of the workers, the general public and the environment. This set of activities is in principle comprised of planning and organisation of decommissioning inclusive strategy development, post-operational activities, implementation of decommissioning (physical and radiological characterisation, decontamination, dismantling and demolition, waste and spent fuel management), radiological, aspects, completion of decommissioning as well as ensuring of funding for these activities. Responsibility for nuclear installations decommissioning, radwaste and spent fuel, management in Slovakia is with a subsidiary of Slovak Electric called Nuclear Installations Decommissioning Radwaste and Spent Fuel Management (acronym SE VYZ), established on January 1, 1996. This paper provides description of an approach to planning of the NPP A-1 and NPPs with WWER reactors decommissioning, realisation of treatment, conditioning and disposal of radwaste, as well as spent fuel management in Slovakia. It takes into account that detail papers on all these issues will follow later during this meeting. (author)

  7. Certification of a spent fuel cask for storage and transportation

    International Nuclear Information System (INIS)

    This paper addresses the US Nuclear Regulatory Commission's requirements for the dry storage and transportation of spent fuel, focusing on how the performance standards differ between storage and transportation. The paper also discusses the NRC cask review process, and some current issues in each area of certification. In addition, some of the issues associated with the US Department of Energy's proposed multi-purpose canister are discussed

  8. Structural evaluation of spent fuel dry storage cask

    International Nuclear Information System (INIS)

    In a various regulations and standards related to the spent fuel storage, the storage casks should be designed to sustain the structural integrity under the accident conditions of predicted operation and design criteria. These conditions for the structural evaluation requires the drop, tip-over, wind like tornado and typhoon, flood and earthquake. This paper describes the load cases and conceptual evaluation method for the structural evaluation. Preliminary safety analysis of the concrete storage system were performed

  9. Characterization of alloy particles extracted from spent nuclear fuel

    Science.gov (United States)

    Cui, D.; Rondinella, V. V.; Fortner, J. A.; Kropf, A. J.; Eriksson, L.; Wronkiewicz, D. J.; Spahiu, K.

    2012-01-01

    We characterized, for the first time, submicro- and nanosized fission product-alloy particles that were extracted nondestructively from spent nuclear fuel, in terms of noble metal (Mo-Ru-Tc-Rh-Pd-Te) composition, atomic level homogeneity and lattice parameters. The evidences obtained in this work contribute to an improved understanding of the redox chemistry of radionuclides in nuclear waste repository environments and, in particular, of the catalytic properties of these unique metal alloy particles.

  10. Spent nuclear fuel canister storage building conceptual design report

    International Nuclear Information System (INIS)

    This Conceptual Design Report provides the technical basis for the Spent Nuclear Fuels Project, Canister Storage Building, and as amended by letter (correspondence number 9555700, M.E. Witherspoon to E.B. Sellers, ''Technical Baseline and Updated Cost Estimate for the Canister Storage Building'', dated October 24, 1995), includes the project cost baseline and Criteria to be used as the basis for starting detailed design in fiscal year 1995

  11. Naval Spent Nuclear Fuel disposal Container System Description Document

    Energy Technology Data Exchange (ETDEWEB)

    N. E. Pettit

    2001-07-13

    The Naval Spent Nuclear Fuel Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers/waste packages are loaded and sealed in the surface waste handling facilities, transferred underground through the access drifts using a rail mounted transporter, and emplaced in emplacement drifts. The Naval Spent Nuclear Fuel Disposal Container System provides long term confinement of the naval spent nuclear fuel (SNF) placed within the disposal containers, and withstands the loading, transfer, emplacement, and retrieval operations. The Naval Spent Nuclear Fuel Disposal Container System provides containment of waste for a designated period of time and limits radionuclide release thereafter. The waste package maintains the waste in a designated configuration, withstands maximum credible handling and rockfall loads, limits the waste form temperature after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Each naval SNF disposal container will hold a single naval SNF canister. There will be approximately 300 naval SNF canisters, composed of long and short canisters. The disposal container will include outer and inner cylinder walls and lids. An exterior label will provide a means by which to identify a disposal container and its contents. Different materials will be selected for the waste package inner and outer cylinders. The two metal cylinders, in combination with the Emplacement Drift System, drip shield, and the natural barrier will support the design philosophy of defense-in-depth. The use of materials with different properties prevents a single mode failure from breaching the waste package. The inner cylinder and inner cylinder lids will be constructed of stainless steel while the outer cylinder and outer cylinder lids will be made of high-nickel alloy.

  12. Safety research activities for Japanese regulations of spent fuel interim storage facilities

    International Nuclear Information System (INIS)

    Japan Nuclear Energy Safety Organization (JNES) carries out (a) preparation of technical documents, (b) technical evaluations of standards (prepared by academic societies), etc. and (c) other R and D activities, to support Nuclear Regulation Authority (NRA: which controls the regulations for Spent Fuel Interim Storage Facilities). In 2012 fiscal year, JNES carried out dynamic test of spent fuel to examine the integrity of spent fuel under cask drop accidents, and preparation for PWR spent fuel storage test to prove long term integrity of spent fuel and cask itself. Some of these tests will be also carried out in 2013 fiscal year and after. (author)

  13. Alternatives for water basin spent fuel storage using pin storage

    International Nuclear Information System (INIS)

    The densest tolerable form for storing spent nuclear fuel is storage of only the fuel rods. This eliminates the space between the fuel rods and frees the hardware to be treated as non-fuel waste. The storage density can be as much as 1.07 MTU/ft2 when racks are used that just satisfy the criticality and thermal limitations. One of the major advantages of pin storage is that it is compatible with existing racks; however, this reduces the storage density to 0.69 MTU/ft2. Even this is a substantial increase over the 0.39 MTU/ft2 that is achievable with current high capacity stainless steel racks which have been selected as the bases for comparison. Disassembly requires extensive operation on the fuel assembly to remove the upper end fitting and to extract the fuel rods from the assembly skeleton. These operations will be performed with the aid of an elevator to raise the assembly where each fuel rod is grappled. Lowering the elevator will free the fuel rod for transfer to the storage canister. A storage savings of $1510 per MTU can be realized if the pin storage concept is incorporated at a new away-from-reactor facility. The storage cost ranges from $3340 to $7820 per MTU of fuel stored with the lower cost applying to storage at an existing away-from-reactor storage facility and the higher cost applying to at-reactor storage

  14. Safety Analysis Report for the PWR Spent Fuel Canister

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Heui Joo; Choi, Jong Won; Cho, Dong Keun; Chun, Kwan Sik; Lee, Jong Youl; Kim, Seong Ki; Kim, Seong Soo; Lee, Yang

    2005-11-15

    This report outlined the results of the safety assessment of the canisters for the PWR spent fuels which will be used in the KRS. All safety analyses including criticality and radiation shielding analyses, mechanical analyses, thermal analyses, and containment analyses were performed. The reference PWR spent fuels were in the 17x17 and determined to have 45,000 MWD/MTU burnup. The canister consists of copper outer shell and nodular cast iron inner structure with diameter of 102 cm and height of 483 cm. Criticality safety was checked for normal and abnormal conditions. It was assumed that the integrity of engineered barriers is preserved and saturated with water of 1.0g/cc for normal condition. For the abnormal condition container and bentonite was assumed to disappear, which allows the spent fuel to be surrounded by water with the most reactive condition. In radiation shielding analysis it was investigated that the absorbed dose at the surface of the canister met the safety limit. The structural analysis was conducted considering three load conditions, normal, extreme, and rock movement condition. Thermal analysis was carried out for the case that the canister with four PWR assemblies was deposited in the repository 500 meter below the surface with 40 m tunnel spacing and 6 m deposition hole spacing. The results of the safety assessment showed that the proposed KDC-1 canister met all the safety limits.

  15. TVO-92 safety analysis of spent fuel disposal

    Energy Technology Data Exchange (ETDEWEB)

    Vieno, T.; Hautojaervi, A.; Koskinen, L.; Nordman, H. [Technical Research Centre of Finland, Espoo (Finland). Nuclear Engineering Lab.

    1993-08-01

    The spent fuel from the TVO I and TVO II reactors at the Olkiluoto nuclear power plant is planned to be disposed in a repository constructed at a depth of about 500 meters in crystalline bedrock. Teollisuuden Voima Oy (TVO) has carried out preliminary site investigations for spent fuel disposal between 1987 and 1992 at five areas in Finland (Olkiluoto, Kivetty, Romuvaara, Syyry and Veitsivaara). The Safety analysis of the disposal system is presented in the report. Spent fuel will be encapsulated in composite copper-steel canisters. The canister design (ACP canister) consists of an inner container of steel as a load-bearing element and an outer container of oxygen-free copper to provide a shield against corrosion. In the repository the canisters will be emplaced in vertical holes drilled in the floors of horizontal deposition tunnels. The annulus between the canister and the rock is filled with compacted bentonite. The results of the safety analysis attest that the planned disposal system fulfils the safety requirements. Suitable places for the repository can be found at each of the five investigation sites.

  16. The Performance of Spent Fuel Casks in Severe Tunnel Fires

    International Nuclear Information System (INIS)

    The Nuclear Regulatory Commission (NRC), working with the National Institute of Standards and Technology (NIST), Pacific Northwest National Laboratory (PNNL), and the National Transportation Safety Board (NTSB), performed analyses to predict the response of various spent fuel transportation cask designs when exposed to a fire similar to that which occurred in the Howard Street railroad tunnel in downtown Baltimore, Maryland on July 18, 2001. The thermal performance of three different spent fuel cask designs (HOLTEC HI-STAR 100, TransNuclear TN-68, and NAC-LWT) was evaluated with the ANSYSR and COBRA-SFS analysis codes, utilizing boundary conditions for the tunnel fire obtained using NIST's Fire Dynamics Simulator (FDS) code. NRC Staff evaluated the potential for a release of radioactive material from each of the three transportation casks analyzed for the Baltimore tunnel fire scenario. The results of these analyses are described in detail in Spent Fuel Transportation Package Response to the Baltimore Tunnel Fire Scenario, NUREG/CR-6886, published in draft for comment in November 2005. Comments received by the NRC on NUREG/CR-6886 will be addressed in the final version of the report. (authors)

  17. Digital mock-up for the spent fuel disassembly processes

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. Y.; Kim, S. H.; Song, T. G.; Kim, Y. H.; Hong, D. H.; Yoon, J. S

    2000-12-01

    In this study, the graphical design system is developed and the digital mock-up is implemented for designing the spent fuel handling and disassembly processes. The system consists of a 3D graphical modeling system, a devices assembling system, and a motion simulation system. This system is used throughout the design stages from the conceptual design to the motion analysis. By using this system, all the process involved in the spent fuel handling and disassembly processes are analyzed and optimized. Also, this system is used in developing the on-line graphic simulator which synchronously simulates the motion of the equipment in a real time basis by connecting the device controllers with the graphic server through the TCP/IP network. This simulator can be effectively used for detecting the malfunctions of the process equipment which is remotely operated. Thus, the simulator enhances the reliability and safety of the spent fuel handling process by providing the remote monitoring function of the process. The graphical design system and the digital mock-up system can be effectively used for designing the process equipment, as well as the optimized process and maintenance process. And the on-line graphic simulator can be an alternative of the conventional process monitoring system which is a hardware based system.

  18. TVO-92 safety analysis of spent fuel disposal

    International Nuclear Information System (INIS)

    The spent fuel from the TVO I and TVO II reactors at the Olkiluoto nuclear power plant is planned to be disposed in a repository constructed at a depth of about 500 meters in crystalline bedrock. Teollisuuden Voima Oy (TVO) has carried out preliminary site investigations for spent fuel disposal between 1987 and 1992 at five areas in Finland (Olkiluoto, Kivetty, Romuvaara, Syyry and Veitsivaara). The Safety analysis of the disposal system is presented in the report. Spent fuel will be encapsulated in composite copper-steel canisters. The canister design (ACP canister) consists of an inner container of steel as a load-bearing element and an outer container of oxygen-free copper to provide a shield against corrosion. In the repository the canisters will be emplaced in vertical holes drilled in the floors of horizontal deposition tunnels. The annulus between the canister and the rock is filled with compacted bentonite. The results of the safety analysis attest that the planned disposal system fulfils the safety requirements. Suitable places for the repository can be found at each of the five investigation sites

  19. Thermal analysis of cold vacuum drying of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Piepho, M.G.

    1998-07-20

    The thermal analysis examined transient thermal and chemical behavior of the Multi canister Overpack (MCO) container for a broad range of cases that represent the Cold Vacuum Drying (CVD) processes. The cases were defined to consider both normal and off-normal operations at the CVD Facility for an MCO with Mark IV N, Reactor spent fuel in four fuel baskets and one scrap basket. This analysis provides the basis for the MCO thermal behavior at the CVD Facility for its Phase 2 Safety Analysis Report (revision 4).

  20. Dosimetry at an interim storage for spent nuclear fuel.

    Science.gov (United States)

    Králík, M; Kulich, V; Studeny, J; Pokorny, P

    2007-01-01

    The Czech nuclear power plant Dukovany started its operation in 1985. All fuel spent from 1985 up to the end of 2005 is stored at a dry interim storage, which was designed for 60 CASTOR-440/84 casks. Each of these casks can accommodate 84 fuel assemblies from VVER 440 reactors. Neutron-photon mixed fields around the casks were characterized in terms of ambient dose equivalent measured by standard area dosemeters. Except this, neutron spectra were measured by means of a Bonner sphere spectrometer, and the measured spectra were used to derive the corresponding ambient dose equivalent due to neutrons. PMID:17526479

  1. A spent fuel assemblies monitoring device by nondestructive analysis 'PYTHON'

    International Nuclear Information System (INIS)

    The monitoring of spent fuel assemblies (16 x 16 UOX) in KWG-reactor pool with the use of non-destructive methods (total Gamma and neutron counting) allow the control of average burn-up and the extremity burn-up. The measurements allow a safety-criticality control before loading the fuel assemblies into the transport casks. A device called PYTHON has been tested and qualified in France. This paper presents a description of the industrial PYTHON device and the results of the measurements. (orig.)

  2. Dosimetry at an interim storage for spent nuclear fuel.

    Science.gov (United States)

    Králík, M; Kulich, V; Studeny, J; Pokorny, P

    2007-01-01

    The Czech nuclear power plant Dukovany started its operation in 1985. All fuel spent from 1985 up to the end of 2005 is stored at a dry interim storage, which was designed for 60 CASTOR-440/84 casks. Each of these casks can accommodate 84 fuel assemblies from VVER 440 reactors. Neutron-photon mixed fields around the casks were characterized in terms of ambient dose equivalent measured by standard area dosemeters. Except this, neutron spectra were measured by means of a Bonner sphere spectrometer, and the measured spectra were used to derive the corresponding ambient dose equivalent due to neutrons.

  3. International tracking and monitoring of nuclear spent fuel transport

    International Nuclear Information System (INIS)

    A cooperative monitoring project was recently initiated to track and monitor shipments of spent fuel from the Lucas Heights HiFAR Research Reactor, located near Sydney, Australia, to either the United States or France. Partners in the project are the US Department of Energy and the Australian Safeguards and Non-Proliferation Office. This project could satisfy a need to have near-real-time continuity of knowledge of a shipment of nuclear material. The benefits of demonstrating a fully operational and sustainable system are significant, because the envisioned system could significantly enhance global safety and security of nuclear material in the transport phases of the nuclear fuel cycle. (author)

  4. Lessons learned from a review of international approaches to spent fuel management

    Directory of Open Access Journals (Sweden)

    Hambley David

    2016-01-01

    Full Text Available Worldwide, a variety of approaches to the management of spent fuel have been adopted. A review of approaches adopted internationally was undertaken to inform decision making on spent fuel management in UK. The review surveyed spent fuel storage and disposal practices, standards, trends and recent developments in 16 countries and carried out more detailed studies into the evolution of spent fuel storage and disposal strategies in four countries. The review highlighted that: (1 spent fuel management should be aligned to the national policy for final dispositioning of the fuel; (2 national spent fuel storage arrangements should deliver efficiency across all spent fuel management activities; (3 commercial and financial arrangements should ensure that spent fuel management decisions do not unnecessarily limit future fuel handling, packaging and disposal activities; (4 extended storage of spent fuel prior to packaging provides increased flexibility in the design of future packaging and disposal concepts. Storage of spent fuel over 100 years or more using existing technologies is technically feasible and operationally credible. Local factors such as existing infrastructure, approach to fuel cycle management, existing experience/capability and short-term cash flow considerations all influence technology selection. Both wet and dry storage systems continue to receive regulatory approval and are acceptable.

  5. Spent fuel disassembly and canning programs at the Barnwell Nuclear Fuel Plant (BNFP)

    International Nuclear Information System (INIS)

    Methods of disassembling and canning spent fuel to allow more efficient storage are being investigated at the BNFP. Studies and development programs are aimed at dry disassembly of fuel to allow storage and shipment of fuel pins rather than full fuel assemblies. Results indicate that doubling existing storage capacity or tripling the carrying capacity of existing transportation equipment is achievable. Disassembly could be performed in the BNFP hot cells at rates of about 12 to 15 assemblies per day

  6. Spent fuel disassembly and canning programs at the Barnwell Nuclear Fuel Plant (BNFP)

    International Nuclear Information System (INIS)

    Methods of disassembling and canning spent fuel to allow more efficient storage are being investigated at the BNFP. Studies and development programs are aimed at dry disassembly of fuel to allow storage and shipment of fuel pins rather than complete fuel assemblies. Results indicate that doubling existing storage capacity or tripling the carrying capacity of existing transportation equipment is achievable. Disassembly could be performed in the BNFP hot cells at rates of about 12 to 15 assemblies per day

  7. Proposed high throughput electrorefining treatment for spent N- Reactor fuel

    International Nuclear Information System (INIS)

    A high-throughput electrorefining process is being adapted to treat spent N-Reactor fuel for ultimate disposal in a geologic repository. Anodic dissolution tests were made with unirradiated N-Reactor fuel to determine the type of fragmentation necessary to provide fuel segments suitable for this process. Based on these tests, a conceptual design was produced of a plant-scale electrorefiner. In this design, the diameter of an electrode assembly is about 1.07 m (42 in.). Three of these assemblies in an electrorefiner would accommodate a 3-metric-ton batch of N-Reactor fuel that would be processed at a rate of 42 kg of uranium per hour

  8. Storing the Spent Nuclear Fuel in Dry Casks Licensed for a Century as an Alternative to Recycling Solution

    Science.gov (United States)

    Milincic, Radovan

    2010-02-01

    Management of spent nuclear power reactor fuels is one of the most urgent problems in nuclear technology. Yearly production of new spent fuel is in the range of thousands of tons, topping a couple of hundred thousand tons of spent fuel already. This material is extremely radioactive and currently there is no adequate international policy, control or management regarding it. I propose here an intermediate term solution to this problem, which will be technologically and economically sustainable: interim spent-fuel storage as an alternative to reprocessing. The reprocessing inherently increases the net amount of the plutonium, which can be used for production of nuclear arms. Moreover, it is an expensive process with the net effect of producing different type of radioactive waste. In particular, the development of a dry cask for nuclear waste storage on site and transport, licensed for a period of hundred years would provide a significantly less expensive solution in the recent future, giving a needed relief to crowded spent-fuel storage pools. Currently in the U.S, NRC licenses existing storage casks for 20 years; and licenses for some of the dry cask storage facilities in the U.S. are about to expire. The extended life dry casks will provide sufficient intermediate period toward a more efficient and/or technologically advanced solution for spent fuel. )

  9. Corrosion of Spent Nuclear Fuel: The Long-Term Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Ewing, Rodney C.

    2003-09-14

    The successful disposal of spent nuclear fuel (SNF) is one of the most serious challenges to the successful completion of the nuclear fuel cycle and the future of nuclear power generation. In the United States, 21 percent of the electricity is generated by 107 commercial nuclear power plants (NPP), each of which generates 20 metric tons of spent nuclear fuel annually. In 1996, the total accumulation of spent nuclear fuel was 33,700 metric tons of heavy metal (MTHM) stored at 70 sites around the country. The end-of-life projection for current nuclear power plants (NPP) is approximately 86,000 MTHM. In the proposed nuclear waste repository at Yucca Mountain over 95% of the radioactivity originates from spent nuclear fuel. World-wide in 1998, approximately 130,000 MTHM of SNF have accumulated, most of it located at 236 NPP in 36 countries. Annual production of SNF is approximately 10,000 MTHM, containing about 100 tons of ''reactor grade'' plutonium. Any reasonable increase in the proportion of energy production by NPP, i.e., as a substitute for hydrocarbon-based sources of energy, will significantly increase spent nuclear fuel production. Spent nuclear fuel is essentially UO{sub 2} with approximately 4-5 atomic percent actinides and fission product elements. A number of these elements have long half-lives hence, the long-term behavior of the UO{sub 2} is an essential concern in the evaluation of the safety and risk of a repository for spent nuclear fuel. One of the unique and scientifically most difficult aspects of the successful disposal of spent nuclear fuel is the extrapolation of short-term laboratory data (hours to years) to the long time periods (10{sup 3} to 10{sup 5} years) as required by the performance objectives set in regulations, i.e. 10 CFR 60. The direct verification of these extrapolations or interpolations is not possible, but methods must be developed to demonstrate compliance with government regulations and to satisfy the

  10. Behavior of iodine in the dissolution of spent nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Sakurai, Tsutomu; Komatsu, Kazunori; Takahashi, A. [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)

    1997-08-01

    The results of laboratory-scale experiments concerning the behavior of iodine in the dissolution of spent nuclear fuels, which were carried out at the Japan Atomic Energy Research Institute, are summarized. Based on previous and new experimental results, the difference in quantity of residual iodine in the fuel solution between laboratory-scale experiments and reprocessing plants is discussed, Iodine in spent fuels is converted to the following four states: (1) oxidation into I{sub 2} by nitric acid, (2) oxidation into I{sub 2} by nitrous acid generated in the dissolution, (3) formation of a colloid of insoluble iodides such as AgI and PdI{sub 2}, and (4) deposition on insoluble residue. Nitrous acid controls the amount of colloid formed. As a result, up to 10% of iodine in spent fuels is retained in the fuel solution, up to 3% is deposited on insoluble residue, and the balance volatilizes to the off-gas, Contrary to earlier belief, when the dissolution is carried out in 3 to 4 M HNO{sub 3} at 100{degrees}C, the main iodine species in a fuel solution is a colloid, not iodate, Immediately after its formation, the colloid is unstable and decomposes partially in the hot nitric acid solution through the following reaction: AgI(s) + 2HNO{sub 3}(aq) = {1/2}I{sub 2}(aq) + AgNO{sub 3}(aq) + NO{sub 2}(g) + H{sub 2}O(1). For high concentrations of gaseous iodine, I{sub 2}(g), and NO{sub 2}, this reaction is reversed towards formation of the colloid (AgI). Since these concentrations are high near the liquid surface of a plant-scale dissolver, there is a possibility that the colloid is formed there through this reversal, Simulations performed in laboratory-scale experiments demonstrated this reversal, This phenomenon can be one reason the quantity of residual iodine in spent fuels is higher in reprocessing plants than in laboratory-scale experiments. 17 refs., 5 figs., 3 tabs.

  11. An approach to determine a defensible spent fuel ratio.

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, Samuel G.; Lindgren, Eric Richard

    2014-03-01

    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 (DUO2), have been conducted in the interim to more definitively determine the source term from these postulated events. In all the previous studies, the postulated attack of greatest interest was by a conical shape charge (CSC) that focuses the explosive energy much more efficiently than bulk explosives. 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 DUO2 surrogate. Previous attempts to define the SFR have resulted in estimates ranging from 0.42 to 12 and include suboptimal experimental techniques and data comparisons. Different researchers have suggested using SFR values of 3 to 5.6. Sound technical 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 dry storage of spent nuclear fuel. Currently, Oak Ridge National Laboratory (ORNL) is in possession of several samples of spent nuclear fuel (SNF) that were used in the original SFR studies in the 1980s and were intended for use in a modern effort at Sandia National Laboratories (SNL) in the 2000s. A portion of these samples are being used for a variety of research efforts. However, the entirety of SNF samples at ORNL is scheduled for disposition at the Waste Isolation Pilot Plant (WIPP) by approximately the end of 2015. If a defensible SFR is to be determined for use in storage and transportation security analyses, the need to begin this effort

  12. Effects of high burnup on spent-fuel casks

    International Nuclear Information System (INIS)

    Utility fuel managers have become very interested in higher burnup fuels as a means to reduce the impact of refueling outages. High-burnup fuels have significant effects on spent-fuel storage or transportation casks because additional heat rejection and shielding capabilities are required. Some existing transportation casks have useful margins that allow shipment of high-burnup fuel, especially the NLI-1/2 truck cask, which has been relicensed to carry pressurized water reactor (PWR) fuel with 56,000 MWd/ton U burnup at 450 days of cooling time. New cask designs should consider the effects of high burnup for future use, even though it is not commercially desirable to include currently unneeded capability. In conclusion, the increased heat and gamma radiation of high-burnup fuels can be accommodated by additional cooling time, but the increased neutron radiation source cannot be accommodated unless the balance of neutron and gamma contributions to the overall dose rate is properly chosen in the initial cask design. Criticality control of high-burnup fuels is possible with heavily poisoned baskets, but burnup credit in licensing is a much more direct means of demonstrating criticality safety

  13. Analysis of the Reuse of Uranium Recovered from the Reprocessing of Commercial LWR Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    DelCul, Guillermo Daniel [ORNL; Trowbridge, Lee D [ORNL; Renier, John-Paul [ORNL; Ellis, Ronald James [ORNL; Williams, Kent Alan [ORNL; Spencer, Barry B [ORNL; Collins, Emory D [ORNL

    2009-02-01

    This report provides an analysis of the factors involved in the reuse of uranium recovered from commercial light-water-reactor (LWR) spent fuels (1) by reenrichment and recycling as fuel to LWRs and/or (2) by recycling directly as fuel to heavy-water-reactors (HWRs), such as the CANDU (registered trade name for the Canadian Deuterium Uranium Reactor). Reuse is an attractive alternative to the current Advanced Fuel Cycle Initiative (AFCI) Global Nuclear Energy Partnership (GNEP) baseline plan, which stores the reprocessed uranium (RU) for an uncertain future or attempts to dispose of it as 'greater-than-Class C' waste. Considering that the open fuel cycle currently deployed in the United States already creates a huge excess quantity of depleted uranium, the closed fuel cycle should enable the recycle of the major components of spent fuel, such as the uranium and the hazardous, long-lived transuranic (TRU) actinides, as well as the managed disposal of fission product wastes. Compared with the GNEP baseline scenario, the reuse of RU in the uranium fuel cycle has a number of potential advantages: (1) avoidance of purchase costs of 11-20% of the natural uranium feed; (2) avoidance of disposal costs for a large majority of the volume of spent fuel that is reprocessed; (3) avoidance of disposal costs for a portion of the depleted uranium from the enrichment step; (4) depending on the {sup 235}U assay of the RU, possible avoidance of separative work costs; and (5) a significant increase in the production of {sup 238}Pu due to the presence of {sup 236}U, which benefits somewhat the transmutation value of the plutonium and also provides some proliferation resistance.

  14. Integrity Assessment of CANDU Spent Fuel During Interim Dry Storage in MACSTOR

    International Nuclear Information System (INIS)

    This paper presents an assessment of the integrity of CANDU spent fuel during dry storage in MACSTOR. Based on review of the safety requirements for sheath integrity during dry storage, a fuel temperature limit for spent CANDU fuel stored in MACSTOR is specified. The spent fuel conditions prior to, and during dry storage are assessed. The safety margin for spent CANDU fuel stored in MACSTOR is assessed against various failure mechanisms using the probabilistic estimation approach derived from US LWR fuel data set. (author)

  15. Commentary on spent fuel storage at Morris operation

    International Nuclear Information System (INIS)

    The General Electric Company is providing technical support to Battelle Pacific Northwest Laboratories in the analysis of the design, operation, and maintenance experience in the handling of nuclear fuel at the Independent Spent Fuel Storage Facility. The purpose of this report is to provide a description of spent fuel handling activities and systems, and an analysis of the storage performance as developed over the seven year operational history of the Morris Operation. Design considerations and performance are analyzed for both the basin and key supporting systems. The bases for this analysis are the provisions for containing radioactive by-product materials, for shielding from the radiation they emit, and for preventing the formation of a critical array. These provisions have been met effectively over the history of storage at Morris. The release of radioactive materials is minimized by the protection of the cladding integrity, the containment of the basin water, the removal of radioactive and other contaminants from the water, and by filtering and then dispersing the basin air. Four auxiliary systems are provided to accomplish this, the basin leak detection system, the filter, the coolers, and the building ventilation system. This successful history notwithstanding, action to reduce personnel exposure, to improve fuel handling reliability and to lessen the potential for accidents continues to be taken

  16. Commentary on spent fuel storage at Morris operation

    Energy Technology Data Exchange (ETDEWEB)

    Eger, K.J.; Zima, G.E.

    1979-10-01

    The General Electric Company is providing technical support to Battelle Pacific Northwest Laboratories in the analysis of the design, operation, and maintenance experience in the handling of nuclear fuel at the Independent Spent Fuel Storage Facility. The purpose of this report is to provide a description of spent fuel handling activities and systems, and an analysis of the storage performance as developed over the seven year operational history of the Morris Operation. Design considerations and performance are analyzed for both the basin and key supporting systems. The bases for this analysis are the provisions for containing radioactive by-product materials, for shielding from the radiation they emit, and for preventing the formation of a critical array. These provisions have been met effectively over the history of storage at Morris. The release of radioactive materials is minimized by the protection of the cladding integrity, the containment of the basin water, the removal of radioactive and other contaminants from the water, and by filtering and then dispersing the basin air. Four auxiliary systems are provided to accomplish this, the basin leak detection system, the filter, the coolers, and the building ventilation system. This successful history notwithstanding, action to reduce personnel exposure, to improve fuel handling reliability and to lessen the potential for accidents continues to be taken.

  17. K-Basin spent nuclear fuel characterization data report 2

    International Nuclear Information System (INIS)

    An Integrated Process Strategy has been developed to package, condition, transport, and store in an interim storage facility the spent nuclear fuel (SNF) currently residing in the K-Basins at Hanford. Information required to support the development of the condition process and to support the safety analyses must be obtained from characterization testing activities conducted on fuel samples from the Basins. Some of the information obtained in the testing was reported in PNL-10778, K-Basin Spent Nuclear Fuel Characterization Data Report (Abrefah et al. 1995). That report focused on the physical, dimensional, metallographic examinations of the first K-West (KW) Basin SNF element to be examined in the Postirradiation Testing Laboratory (PTL) hot cells; it also described some of the initial SNF conditioning tests. This second of the series of data reports covers the subsequent series of SNF tests on the first fuel element. These tests included optical microscopy analyses, conditioning (drying and oxidation) tests, ignition tests, and hydrogen content tests

  18. Dose and temperature distribution in spent fuel containing material

    Directory of Open Access Journals (Sweden)

    Viererbl Ladislav

    2016-01-01

    Full Text Available Spent fuel containing material (SFCM can arise during severe nuclear reactor accident by melting of a reactor core and surrounding material (corium or during accident in spent fuel storage. It consists of nuclear fuel, fission products, activation products and materials from fuel cladding, concrete, etc. The paper deals with dose and temperature characteristics inside the SFCM after transition of the molten mixture to solid state. Calculations were made on simplified spherical models, without connection to some specific nuclear accident. The dose rate was estimated for alpha, beta and gamma radiation in times over the course of 30 years from the end of the fission chain reaction. Concentration of helium generated in the material by alpha decay was calculated. For the dose rate values estimation, computation code ORIGEN 2.2 with dosimetric library ENDF/B-IV were used. Temperature distribution inside the solid SFCM was calculated by FLUENT code. As source of heating, energy of radioactive decays was taken. Estimated dose and temperature characteristics can be used, e.g. for evaluation of radiation damage and temperature behaviour of SFCM or for radiation test design of corium simulating materials.

  19. The synchronous active neutron detection system for spent fuel assay

    Energy Technology Data Exchange (ETDEWEB)

    Pickrell, M.M.; Kendall, P.K.

    1994-10-01

    The authors have begun to develop a novel technique for active neutron assay of fissile material in spent nuclear fuel. This approach will exploit the unique operating features of a 14-MeV neutron generator developed by Schlumberger. This generator and a novel detection system will be applied to the direct measurement of the fissile material content in spent fuel in place of the indirect measures used at present. The technique they are investigating is termed synchronous active neutron detection (SAND). It closely follows a method that has been used routinely in other branches of physics to detect very small signals in the presence of large backgrounds. Synchronous detection instruments are widely available commercially and are termed {open_quotes}lock-in{close_quotes} amplifiers. The authors have implemented a digital lock-in amplifier in conjunction with the Schlumberger neutron generator to explore the possibility of synchronous detection with active neutrons. This approach is possible because the Schlumberger system can operate at up to a 50% duty factor, in effect, a square wave of neutron yield. The results to date are preliminary but quite promising. The system is capable of resolving the fissile material contained in a small fraction of the fuel rods in a cold fuel assembly. It also appears to be quite resilient to background neutron interference. The interrogating neutrons appear to be nonthermal and penetrating. Although a significant amount of work remains to fully explore the relevant physics and optimize the instrument design, the underlying concept appears sound.

  20. HTGR spent fuel element decay heat and source term analysis

    International Nuclear Information System (INIS)

    Decay heat, gamma dose rates, and neutron source strengths were determined for spent fuel elements from a High-Temperature Gas-Cooled Reactor (HTGR). The calculations were based on curie values reported in General Atomic Report GA-A13886 for the earlier commercial version of a 3000-MW(t) HTGR utilizing the thorium-uranium four-year fuel cycle. The reactor core was designed for an average thermal power density of 8.5 watts per cm3 and a carbon-to-thorium atom ratio which varies between 210:1 and 240:1. Calculations of decay heat, gamma dose rates, and neutron source strengths were made for spent fuel elements from the initial core and from representative nonrecycle and recycle reloads. The study was performed for decay times from 180 days to 10 years. Tables of the isotopic results are given for both the fertile and fissile particles in the fuel elements. In addition, ordered tables of the important isotopic contributors are presented. Graphical presentations of the results are shown and discussed; in addition, comparisons are made with previous determinations