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Sample records for storage pools fuel

  1. Fuel assembly storage pool

    International Nuclear Information System (INIS)

    Hiranuma, Hiroshi.

    1976-01-01

    Object: To remove limitation of the number of storage of fuel assemblies to increase the number of storage thereof so as to relatively reduce the water depth required for shielding radioactive rays. Structure: Fuel assembly storage rack containers for receiving a plurality of spent fuel assembly racks are stacked in multi-layer fashion within a storage pool filled with water for shielding radioactive rays and removing heat. (Furukawa, Y.)

  2. Spent fuel storage pool

    International Nuclear Information System (INIS)

    Murakami, Naoshi.

    1996-01-01

    Fences are disposed to a fuel exchange floor surrounding the upper surface of a fuel pool for preventing overflow of pool water. The fences comprise a plurality of flat boards arranged in parallel with each other in the longitudinal direction while being vertically inclined, and slits are disposed between the boards for looking down the pool. Further, the fences comprise wide boards and are constituted so as to be laid horizontally on the fuel exchange floor in a normal state and uprisen by means of the signals from an earthquake sensing device. Even if pool water is overflow from the fuel pool by the vibrations occurred upon earthquake and flown out to the floor of the fuel exchange floor, the overflow from the fuel exchange floor is prevented by the fences. An operator who monitors the fuel pool can observe the inside of the fuel pool through the slits formed to the fences during normal operation. The fences act as resistance against overflowing water upon occurrence of an earthquake thereby capable of reducing the overflowing amount of water due to the vibrations of pool water. The effect of preventing overflowing water can be enhanced. (N.H.)

  3. Compact fuel storage rack for fuel pools

    International Nuclear Information System (INIS)

    Parras, F.; Louvat, J.P.

    1986-01-01

    ETS LEMER and FRAMATOME propose a new compact storage rack. This rack permits a considerable increase of the storage capacity of cooling pools. A short description of the structure and the components is presented, to propose racks that are: . Inalterable, . Compact, . Insensitive to earthquakes. Installation in pools already in operation is simplified by their light structure and the bearing device [fr

  4. Laser surveillance systems for fuel storage pools

    International Nuclear Information System (INIS)

    Boeck, H.

    1985-06-01

    A Laser Surveillance System (LASSY) as a new safeguards device has been developed under the IAEA research contract No. 3458/RB at the Atominstitut Wien using earlier results by S. Fiarman. This system is designed to act as a sheet of light covering spent fuel assemblies in spent fuel storage pools. When movement of assemblies takes place, LASSY detects and locates the position of the movement in the pool and when interrogated, presents a list of pool positions and times of movement to the safeguards inspector. A complete prototype system was developed and built. Full scale tests showed the principal working capabilities of a LASSY underwater

  5. Structure for nuclear fuel storage pools

    International Nuclear Information System (INIS)

    Ebata, Sakae; Nichiei, Shinji.

    1979-01-01

    Purpose: To enable leak detection in nuclear fuel storage pools, as well as prevent external leakages while keeping the strength of the constructional structures. Constitution: Protection plates are provided around pool linear plates and a leak reception is provided to the bottom. Leakages are detected by leak detecting pipeways and the external leakages are prevented by collecting them in a detection area provided in the intermediate layer. Since ferro-reinforcements at the bottom wall of the pool are disconnected by the protection plate making it impossible to form the constructional body, body hunches are provided to the bottom wall of the pool for processing the ferro-reinforcements. (Yoshino, Y.)

  6. Behavior of spent nuclear fuel in water pool storage

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.

    1977-09-01

    Storage of irradiated nuclear fuel in water pools (basins) has been standard practice since nuclear reactors first began operation approximately 34 years ago. Pool storage is the starting point for all other fuel storage candidate processes and is a candidate for extended interim fuel storage until policy questions regarding reprocessing and ultimate disposal have been resolved. This report assesses the current performance of nuclear fuel in pool storage, the range of storage conditions, and the prospects for extending residence times. The assessment is based on visits to five U.S. and Canadian fuel storage sites, representing nine storage pools, and on discussions with operators of an additional 21 storage pools. Spent fuel storage experience from British pools at Winfrith and Windscale and from a German pool at Karlsruhe (WAK) also is summarized

  7. Corrosion surveillance in spent fuel storage pools

    International Nuclear Information System (INIS)

    Howell, J.P.

    1996-01-01

    In mid-1991, corrosion of aluminum-clad spent nuclear fuel was observed in the light-water filled basins at the Savannah River site. A corrosion surveillance program was initiated in the P, K, L-Reactor basins and in the Receiving Basin for Offsite Fuels (RBOF). This program verified the aggressive nature of the pitting corrosion and provided recommendations for changes in basin operations to permit extended longer term interim storage. The changes were implemented during 1994--1996 and have resulted in significantly improved basin water quality with conductivity in the 1--3 microS/cm range. Under these improved conditions, no new pitting has been observed over the last three years. This paper describes the corrosion surveillance program at SRS and what has been learned about the corrosion of aluminum-clad in spent fuel storage pools

  8. Spent nuclear fuel storage pool thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    Gay, R.R.

    1984-01-01

    Storage methods and requirements for spent nuclear fuel at U.S. commercial light water reactors are reviewed in Section 1. Methods of increasing current at-reactor storage capabilities are also outlined. In Section 2 the development of analytical methods for the thermal-hydraulic analysis of spent fuel pools is chronicled, leading up to a discussion of the GFLOW code which is described in Section 3. In Section 4 the verification of GFLOW by comparisons of the code's predictions to experimental data taken inside the fuel storage pool at the Maine Yankee nuclear power plant is presented. The predictions of GFLOW using 72, 224, and 1584 node models of the storage pool are compared to each other and to the experimental data. An example of thermal licensing analysis for Maine Yankee using the GFLOW code is given in Section 5. The GFLOW licensing analysis is compared to previous licensing analysis performed by Yankee Atomic using the RELAP-4 computer code

  9. Immobilization of radioactive waste sludge from spent fuel storage pool

    International Nuclear Information System (INIS)

    Pavlovic, R.; Plecas, I.

    1998-01-01

    In the last forty years, in FR Yugoslavia, as result of the research reactors' operation and radionuclides application in medicine, industry and agriculture, radioactive waste materials of the different categories and various levels of specific activities were generated. As a temporary solution, these radioactive waste materials are stored in the two hanger type interim storages for solid waste and some type of liquid waste packed in plastic barrels, and one of three stainless steal underground containers for other types of liquid waste. Spent fuel elements from nuclear reactors in the Vinca Institute have been temporary stored in water filled storage pool. Due to the fact that the water in the spent fuel elements storage pool have not been purified for a long time, all metallic components submerged in the water have been hardly corroded and significant amount of the sludge has been settled on the bottom of the pool. As a first step in improving spent fuel elements storage conditions and slowing down corrosion in the storage spent fuel elements pool we have decided to remove the sludge from the bottom of the pool. Although not high, but slightly radioactive, this sludge had to be treated as radioactive waste material. Some aspects of immobilisation, conditioning and storage of this sludge are presented in this paper. (author

  10. Bases for extrapolating materials durability in fuel storage pools

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.

    1994-12-01

    A major body of evidence indicates that zirconium alloys have the most consistent and reliable durability in wet storage, justifying projections of safe wet storage greater than 50 y. Aluminum alloys have the widest range of durabilities in wet storage; systematic control and monitoring of water chemistry have resulted in low corrosion rates for more than two decades on some fuels and components. However, cladding failures have occurred in a few months when important parameters were not controlled. Stainless steel is extremely durable when stress, metallurgical and water chemistry factors are controlled. LWR SS cladding has survived for 25 y in wet storage. However, sensitized, stressed SS fuels and components have seriously degraded in fuel storage pools (FSPs) at ∼ 30 C. Satisfactory durability of fuel assembly and FSP component materials in extended wet storage requires investments in water quality management and surveillance, including chemical and biological factors. The key aspect of the study is to provide storage facility operators and other decision makers a basis to judge the durability of a given fuel type in wet storage as a prelude to basing other fuel management plans (e.g. dry storage) if wet storage will not be satisfactory through the expected period of interim storage

  11. On the neutronics of spent fuel storage pools

    International Nuclear Information System (INIS)

    Caro, R.; Martinez-Val, J.M.; Donoso, E.

    1980-01-01

    The neutron physics of light-water-reactor fuel elements storage is analyzed for reviewing the calculation methodologies and pointing out its characteristics, specially those related to the safety analysis report. Some numerical results are presented, involving both clean and poisoned storage pools. Besides the conventional criticality calculations in nominal and accidental circumstances, the so-called optimum moderation phenomenon is dealt with special emphasis. (author)

  12. Application of burnup credit for PWR spent fuel storage pool

    International Nuclear Information System (INIS)

    Shin, Hee Sung; Ro, Seung-Gy; Bae, Kang Mok; Kim, Ik Soo; Shin, Young Joon

    1999-01-01

    A study on the application of burnup credit for a PWR spent fuel storage pool has been investigated using a computer code system such as CSAS6 module of SCALE 4.3 in association with 44-group SCALE cross-section library. The calculation bias of the code system at a 95% probability with a 95% confidence level seems to be 0.00951 by benchmarking the system for forty six experimental data. With the aid of this computer code system, criticality analysis has been performed for the PWR spent fuel storage pool. Uncertainties due to postulated abnormal and accidental conditions, and manufacturing tolerance such as stainless steel thickness of storage rack, fuel enrichment, fuel density and box size have statistically been combined and resulted in 0.00674. Also, isotopic correction factor which was based on the calculated and measured concentration of 43 isotopes for both selected actinides and fission products important in burnup credit application has been taken into account in the criticality analysis. It is revealed that the minimum burnup with the corrected isotopic concentrations as required for the safe storage is 5,730 MWd/tU in enriched fuel of 5.0 wt%. (author)

  13. Stabilization of reactor fuel storage pool-TTP

    International Nuclear Information System (INIS)

    Sevigny, G.

    1994-10-01

    The proposed work includes evaluating standard and improved technologies an designing an integrated demonstration system to clean the water and sludge the fuel storage pools. The water released would meet drinking water standards and tritium standards. The volume of radioactive sludge would be reduced by partial separation of the sludge and radionuclides and eventual solidification of the hazardous and radioactive waste. The scope of the wo includes a survey of needs and applicable technologies, system engineering evaluation, conceptual design, detailed design, fabrication of the integrat demonstration system, and testing of the system. The survey task will locate potential specific customers within the DOE complex, and outside of the DOE complex throughout the United States, that be able to utilize the narrowly focused technology to stabilize/shutdown reactor fuel storage pools, responsible parties will be located and asked respond to a survey about their specific process requirements. Literature searches will be run through technical and scientific databases to locate technologies that may be an improvement over the standard baselined technol for cleanup of radioactively-contaminated pools. Systems engineering will provide decision analysis support for the development, evaluation, design, test functions of the treatment of pool water and sludge

  14. Stabilization of reactor fuel storage pool-TTP

    Energy Technology Data Exchange (ETDEWEB)

    Sevigny, G.

    1994-10-01

    The proposed work includes evaluating standard and improved technologies an designing an integrated demonstration system to clean the water and sludge the fuel storage pools. The water released would meet drinking water standards and tritium standards. The volume of radioactive sludge would be reduced by partial separation of the sludge and radionuclides and eventual solidification of the hazardous and radioactive waste. The scope of the wo includes a survey of needs and applicable technologies, system engineering evaluation, conceptual design, detailed design, fabrication of the integrat demonstration system, and testing of the system. The survey task will locate potential specific customers within the DOE complex, and outside of the DOE complex throughout the United States, that be able to utilize the narrowly focused technology to stabilize/shutdown reactor fuel storage pools, responsible parties will be located and asked respond to a survey about their specific process requirements. Literature searches will be run through technical and scientific databases to locate technologies that may be an improvement over the standard baselined technol for cleanup of radioactively-contaminated pools. Systems engineering will provide decision analysis support for the development, evaluation, design, test functions of the treatment of pool water and sludge.

  15. Current perceptions of spent nuclear fuel behavior in water pool storage

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.

    1977-06-01

    A survey was conducted of a cross section of U.S. and Canadian fuel storage pool operators to define the spent fuel behavior and to establish the range of pool storage environments. There is no evidence for significant corrosion degradation. Fuel handling causes only minimal damage. Most fuel bundles with defects generally are stored without special procedures. Successful fuel storage up to 18 years with benign water chemistry has been demonstrated. 2 tables

  16. Nonlinear analysis and evaluation of a reinforced concrete spent fuel storage pool for accidental thermal loads

    International Nuclear Information System (INIS)

    Kabir, A.F.; Bolourchi, S.

    1991-01-01

    A feasibility study was conducted for addition of consolidated fuel racks to an existing reinforced concrete spent fuel storage pool of a Mark I BWR plant. Nonlinear analysis of a detailed three-dimensional model of the fuel pool, considering cracking in concrete under gravity and thermal load conditions, showed that the pool has reserve capacities to carry the additional loads. (author)

  17. Storage of water reactor spent fuel in water pools. Survey of world experience

    International Nuclear Information System (INIS)

    1982-01-01

    Following discharge from a nuclear reactor, spent fuel has to be stored in water pools at the reactor site to allow for radioactive decay and cooling. After this initial storage period, the future treatment of spent fuel depends on the fuel cycle concept chosen. Spent fuel can either be treated by chemical processing or conditioning for final disposal at the relevant fuel cycle facilities, or be held in interim storage - at the reactor site or at a central storage facility. Recent forecasts predict that, by the year 2000, more than 150,000 tonnes of heavy metal from spent LWR fuel will have been accumulated. Because of postponed commitments regarding spent fuel treatment, a significant amount of spent fuel will still be held in storage at that time. Although very positive experience with wet storage has been gained over the past 40 years, making wet storage a proven technology, it appears desirable to summarize all available data for the benefit of designers, storage pool operators, licensing agenices and the general public. Such data will be essential for assessing the viability of extended water pool storage of spent nuclear fuel. In 1979, the International Atomic Energy Agency and the Nuclear Energy Agency of the OECD jointly issued a questionnaire dealing with all aspects of water pool storage. This report summarizes the information received from storage pool operators

  18. Irradiation of Microbes from Spent Nuclear Fuel Storage Pool Environments

    International Nuclear Information System (INIS)

    Breckenridge, C.R.; Watkins, C.S.; Bruhn, D.F.; Roberto, F.F.; Tsang, M.N.; Pinhero, P.J.; Brey, R.F.; Wright, R.N.; Windes, W.F.

    1999-01-01

    Microbes have been isolated and identified from spent nuclear fuel storage pools at the Idaho National Engineering and Environmental Laboratory (INEEL). Included among these are Corynebacterium aquaticum, Pseudomonas putida, Comamonas acidovorans, Gluconobacter cerinus, Micrococcus diversus, Rhodococcus rhodochrous, and two strains of sulfate-reducing bacteria (SRB). We examined the sensitivity of these microbes to a variety of total exposures of radiation generated by a 6-MeV linear accelerator (LINAC). The advantage of using a LINAC is that it provides a relatively quick screen of radiation tolerance. In the first set of experiments, we exposed each of the aforementioned microbes along with four additional microbes, pseudomonas aeruginosa, Micrococcus luteus, Escherchia coli, and Deinococcus radiodurans to exposures of 5 x 10 3 and 6 x 10 4 rad. All microbial specimens withstood the lower exposure with little or no reduction in cell population. Upon exposing the microbes to the larger dose of 6 x 10 4 rad, we observed two distinct groupings: microbes that demonstrate resistance to radiation, and microbes that display intolerance through a dramatic reduction from their initial population. Microbes in the radiation tolerant grouping were exposed to 1.1 x 10 5 rad to examine the extent of their resistance. We observe a correlation between radiation resistance and gram stain. The gram-positive species we examined seem to demonstrate a greater radiation resistance

  19. Irradiation of Microbes from Spent Nuclear Fuel Storage Pool Environments

    Energy Technology Data Exchange (ETDEWEB)

    Breckenridge, C.R.; Watkins, C.S.; Bruhn, D.F.; Roberto, F.F.; Tsang, M.N.; Pinhero, P.J. [INEEL (US); Brey, R.F. [ISU (US); Wright, R.N.; Windes, W.F.

    1999-09-03

    Microbes have been isolated and identified from spent nuclear fuel storage pools at the Idaho National Engineering and Environmental Laboratory (INEEL). Included among these are Corynebacterium aquaticum, Pseudomonas putida, Comamonas acidovorans, Gluconobacter cerinus, Micrococcus diversus, Rhodococcus rhodochrous, and two strains of sulfate-reducing bacteria (SRB). We examined the sensitivity of these microbes to a variety of total exposures of radiation generated by a 6-MeV linear accelerator (LINAC). The advantage of using a LINAC is that it provides a relatively quick screen of radiation tolerance. In the first set of experiments, we exposed each of the aforementioned microbes along with four additional microbes, pseudomonas aeruginosa, Micrococcus luteus, Escherchia coli, and Deinococcus radiodurans to exposures of 5 x 10{sup 3} and 6 x 10{sup 4} rad. All microbial specimens withstood the lower exposure with little or no reduction in cell population. Upon exposing the microbes to the larger dose of 6 x 10{sup 4} rad, we observed two distinct groupings: microbes that demonstrate resistance to radiation, and microbes that display intolerance through a dramatic reduction from their initial population. Microbes in the radiation tolerant grouping were exposed to 1.1 x 10{sup 5} rad to examine the extent of their resistance. We observe a correlation between radiation resistance and gram stain. The gram-positive species we examined seem to demonstrate a greater radiation resistance.

  20. Criteria for recladding of spent light water reactor fuel before long term pool storage

    International Nuclear Information System (INIS)

    Pettersson, K.; Jansson, L.

    1979-01-01

    The question of the need for any special treatment of failed fuel elements prior to long term pool storage has been studied. It is concluded that the main problem appears to be hydride embrittlement of failed fuel rods, which may lead to increased damage during handling and transport of the failed fuel. Some mechanisms for the degradation of failed fuel rods have been identified. They can all be considered as relatively improbable, but further experimental evidence is needed before it can be concluded that these degradation mechanisms are insignificant during pool storage. The report also contains a review of methods for identification of leaking fuel bundles and fuel rods. (Auth.)

  1. Criteria for recladding of spent light water reactor fuel before long term pool storage

    International Nuclear Information System (INIS)

    Pettersson, K.; Jansson, L.

    1979-06-01

    The question of the need for any special treatment of failed fuel elements prior to long term pool storage has been studied. It is concluded that the main problem appears to be hydride embrittlement of failed fuel rods, which may lead to increased damage during handling and transport of the failed fuel. Some mechanisms for the degradation of failed fuel rods have been identified. They can all be considered as relatively improbable, but further experimental evidence is needed before it can be concluded that thede degradation mechanisms are insignificant during pool storage. The report also contains a review of methods for identification of leaking fuel bundles and fuel rods.(author)

  2. Detection of fission products release in the research reactor 'RA' spent fuel storage pool

    International Nuclear Information System (INIS)

    Matausek, M.V.; Vukadin, Z.; Pavlovic, S.; Maksin, T.; Idakovic, Z.; Marinkovic, N.

    1997-05-01

    Spent fuel resulting from 25 years of operating the 6.5/10 MW thermal heavy water moderated and cooled research reactor RA at the VINCA Institute is presently all stored in the temporary spent fuel storage pool in the basement of the reactor building. In 1984, the reactor was shut down for refurbishment, which for a number of reasons has not yet been completed. Recent investigations show that independent of the future status of the research reactor, safe disposal of the so far irradiated fuel must be the subject of primary concern. The present status of the research reactor RA spent fuel storage pool at the VINCA Institute presents a serious safety problem. Action is therefore initiated in two directions. First, safety of the existing spent fuel storage should be improved. Second, transferring spent fuel into another, presumably dry storage space should be considered. By storing the previously irradiated fuel of the research reactor RA in a newly built storage space, sufficient free space will be provided in the existing spent fuel storage pool for the newly irradiated fuel when the reactor starts operation again. In the case that it would be decided to decommission the research reactor RA, the newly built storage space would provide safe disposal for the fuel irradiated so far

  3. Information on the feasibility study for the reracking in the fuel storage pools of the Juragua Nuclear Power Plant

    International Nuclear Information System (INIS)

    Rodriguez, J.M.; Rodriguez, I.; Lopez, D.; Guerra, R.; Rodriguez, M.; Garcia, F.

    1995-01-01

    During 1993, in the Juragua Nuclear Power Plants as engineering evaluation programme was initiated in the storage area of irradiated nuclear fuel, where work in order to determine the feasibility of capacity increase for storage of irradiated nuclear fuel at the fuel storage pools using poisoned compact close racks instead of the originally designed racks. The feasibility study is a fundamental activity of this programme for the 1994-1995 period. According to this study the prospects of assimilation of compact storage conditions in the fuel storage pools in unit number one and prolonged fuel storage pool are investigated

  4. Engineering program in order to increase the irradiated fuel storage capacity in pool facilities of Juragua

    International Nuclear Information System (INIS)

    Rodriguez R, J.

    1996-01-01

    In 1993, a technical program in the spent fuel storage area of Nuclear Plant Juragua was launched. Such a program tries to carry out an engineering assessment of the possibility of increasing the spent fuel storage capacity in pool storage facilities by using high density racks (re-racking) instead of the original (non-compact) ones. The purpose of the above-mentioned program is to evaluate possible solutions that can be applied to the construction works prior to plant operation. The first stage of the program for the 1994-95 period is an ongoing Engineering-Economic Feasibility Study (EEFS), which endeavors to examine the capabilities of the reloading pool in Unit-1 Reactor building and long-term storage pool in auxiliary building in high density storage conditions. Technical details of the EEFS and reached results and difficulties are described. (author). 5 refs., 2 figs

  5. A Study on Rack Thickness Effect for Spent Fuel Pool Storage

    International Nuclear Information System (INIS)

    Kim, Mi Jin; Lee, Hee-Jae; Sohn, Dong-Seong

    2015-01-01

    For the effective storage of used fuel, the development of high performance neutron absorbing materials is needed. One of the major concern for the used fuel storage is the assurance to keep subcriticality of the storage rack and the high performance neutron absorbing material is the vital part to assure this requirement. According to NRC guide line, the k-effective of the spent fuel storage racks must not exceed 0.95. To ensure its safety, subcriticality analysis is required. Subcriticality analysis of the used storage in spent fuel pool have been performed by different authors. Criticality calculations for light water reactor spent fuel storage rack were carried out by Jae et al. They used AMPX-KENO IV code and considered the effect of rack pitch and rack thickness for consolidated fuel. The criticality analysis has performed at Gd 0.2 and 1 wt% according to thickness change. As thickness increases, the volume of the spent fuel pool rack increases. Therefore, absorbing material also increases according to thickness

  6. A Study on Rack Thickness Effect for Spent Fuel Pool Storage

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Mi Jin; Lee, Hee-Jae; Sohn, Dong-Seong [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2015-05-15

    For the effective storage of used fuel, the development of high performance neutron absorbing materials is needed. One of the major concern for the used fuel storage is the assurance to keep subcriticality of the storage rack and the high performance neutron absorbing material is the vital part to assure this requirement. According to NRC guide line, the k-effective of the spent fuel storage racks must not exceed 0.95. To ensure its safety, subcriticality analysis is required. Subcriticality analysis of the used storage in spent fuel pool have been performed by different authors. Criticality calculations for light water reactor spent fuel storage rack were carried out by Jae et al. They used AMPX-KENO IV code and considered the effect of rack pitch and rack thickness for consolidated fuel. The criticality analysis has performed at Gd 0.2 and 1 wt% according to thickness change. As thickness increases, the volume of the spent fuel pool rack increases. Therefore, absorbing material also increases according to thickness.

  7. Design of make-up water system for Tehran research reactor spent nuclear fuels storage pool

    Energy Technology Data Exchange (ETDEWEB)

    Aghoyeh, Reza Gholizadeh [Reactor Research Group, Nuclear Science and Technology Research Institute (NSTRI), Atomic Energy Organization of Iran (AEOI), North Amirabad, P.O. Box 14155-1339, Tehran (Iran, Islamic Republic of); Khalafi, Hosein, E-mail: hkhalafi@aeoi.org.i [Reactor Research Group, Nuclear Science and Technology Research Institute (NSTRI), Atomic Energy Organization of Iran (AEOI), North Amirabad, P.O. Box 14155-1339, Tehran (Iran, Islamic Republic of)

    2010-10-15

    Spent nuclear fuels storage (SNFS) is an essential auxiliary system in nuclear facility. Following discharge from a nuclear reactor, spent nuclear fuels have to be stored in water pool of SNFS away from reactor to allow for radioactive to decay and removal of generated heat. To prevent corrosion damage of fuels and other equipments, the storage pool is filled with de-ionized water which serves as moderator, coolant and shielding. The de-ionized water will be provided from make-up water system. In this paper, design of a make-up water system for optimal water supply and its chemical properties in SNFS pool is presented. The main concern of design is to provide proper make-up water throughout the storage time. For design of make-up water system, characteristics of activated carbon purifier, anionic, cationic and mixed-bed ion-exchangers have been determined. Inlet water to make-up system provide from Tehran municipal water system. Regulatory Guide 1.13 of the and graver company manual that manufactured the Tehran research reactor (TRR) make-up water system have been used for make-up water system of TRR spent nuclear fuels storage pool design.

  8. Design of make-up water system for Tehran research reactor spent nuclear fuels storage pool

    International Nuclear Information System (INIS)

    Aghoyeh, Reza Gholizadeh; Khalafi, Hosein

    2010-01-01

    Spent nuclear fuels storage (SNFS) is an essential auxiliary system in nuclear facility. Following discharge from a nuclear reactor, spent nuclear fuels have to be stored in water pool of SNFS away from reactor to allow for radioactive to decay and removal of generated heat. To prevent corrosion damage of fuels and other equipments, the storage pool is filled with de-ionized water which serves as moderator, coolant and shielding. The de-ionized water will be provided from make-up water system. In this paper, design of a make-up water system for optimal water supply and its chemical properties in SNFS pool is presented. The main concern of design is to provide proper make-up water throughout the storage time. For design of make-up water system, characteristics of activated carbon purifier, anionic, cationic and mixed-bed ion-exchangers have been determined. Inlet water to make-up system provide from Tehran municipal water system. Regulatory Guide 1.13 of the and graver company manual that manufactured the Tehran research reactor (TRR) make-up water system have been used for make-up water system of TRR spent nuclear fuels storage pool design.

  9. Examination of Zircaloy-clad spent fuel after extended pool storage

    International Nuclear Information System (INIS)

    Bradley, E.R.; Bailey, W.J.; Johnson, A.B. Jr.; Lowry, L.M.

    1981-09-01

    This report presents the results from metallurgical examinations of Zircaloy-clad fuel rods from two bundles (0551 and 0074) of Shippingport PWR Core 1 blanket fuel after extended water storage. Both bundles were exposed to water in the reactor from late 1957 until discharge. The estimated average burnups were 346 GJ/kgU (4000 MWd/MTU) for bundle 0551 and 1550 GJ/kgU (18,000 MWd/MTU) for bundle 0074. Fuel rods from bundle 0551 were stored in deionized water for nearly 21 yr prior to examination in 1980, representing the world's oldest pool-stored Zircaloy-clad fuel. Bundle 0074 has been stored in deionized water since reactor discharge in 1964. Data from the current metallurgical examinations enable a direct assessment of extended pool storage effects because the metallurgical condition of similar fuel rods was investigated and documented soon after reactor discharge. Data from current and past examinations were compared, and no significant degradation of the Zircaloy cladding was indicated after almost 21 yr in water storage. The cladding dimensions and mechanical properties, fission gas release, hydrogen contents of the cladding, and external oxide film thicknesses that were measured during the current examinations were all within the range of measurements made on fuel bundles soon after reactor discharge. The appearance of the external surfaces and the microstructures of the fuel and cladding were also similar to those reported previously. In addition, no evidence of accelerated corrosion or hydride redistribution in the cladding was observed

  10. Criticality analysis of the CAREM-25 reactor irradiated fuel elements storage pool

    International Nuclear Information System (INIS)

    Albornoz, A.F.; Jatuff, F.E.; Gho, C.J.

    1993-01-01

    A criticality safety analysis of the irradiated fuel element pool storage of the CAREM-25 reactor was performed. The CAREM project is property of the Comision Nacional de Energia Atomica (CNEA) of Argentine, and it is being executed by INVAP S.E. difficult evaluation of the CAREM core (relatively high -3,4%- enriched U O 2 , Gd 2 O 3 burnable absorber in different densities, or criticality achievement with as few as 7 fuel elements is inherited by the pool storage. The lattice code CONDOR 1.1 was used for investigating the problem scene, and some results compared on the Monte Carlo codes MONK 5.0 and MONK 6.3. Circular and square tubes of 304-L stainless steel, borated steel and boral B 4 C in Al) were tested as suitable channels for fuel element containment, in square and hexagonal arrays; in addition, burnup, burnable absorber concentration, Sm and leakage credits were determined. It was found that the critical is strongly dependent on the separation of the fuel elements in the pool. Out-of-nominal conditions were investigated too, showing that the loss of coolant and the change in temperature and density conditions in the storage lead to an increase in reactivity, but the system's reactivity remains near the safety limits. (author)

  11. Sensitivity Analysis of Depletion Parameters for Heat Load Evaluation of PWR Spent Fuel Storage Pool

    International Nuclear Information System (INIS)

    Kim, In Young; Lee, Un Chul

    2011-01-01

    As necessity of safety re-evaluation for spent fuel storage facility has emphasized after the Fukushima accident, accuracy improvement of heat load evaluation has become more important to acquire reliable thermal-hydraulic evaluation results. As groundwork, parametric and sensitivity analyses of various storage conditions for Kori Unit 4 spent fuel storage pool and spent fuel depletion parameters such as axial burnup effect, operation history, and specific heat are conducted using ORIGEN2 code. According to heat load evaluation and parametric sensitivity analyses, decay heat of last discharged fuel comprises maximum 80.42% of total heat load of storage facility and there is a negative correlation between effect of depletion parameters and cooling period. It is determined that specific heat is most influential parameter and operation history is secondly influential parameter. And decay heat of just discharged fuel is varied from 0.34 to 1.66 times of average value and decay heat of 1 year cooled fuel is varied from 0.55 to 1.37 times of average value in accordance with change of specific power. Namely depletion parameters can cause large variation in decay heat calculation of short-term cooled fuel. Therefore application of real operation data instead of user selection value is needed to improve evaluation accuracy. It is expected that these results could be used to improve accuracy of heat load assessment and evaluate uncertainty of calculated heat load.

  12. Corrosion of aluminium alloy test coupons in water of spent fuel storage pool at RA reactor

    International Nuclear Information System (INIS)

    Pesic, M.; Maksin, T.; Jordanov, G.; Dobrijevic, R.

    2004-12-01

    Study on corrosion of aluminium cladding, of the TVR-S type of enriched uranium spent fuel elements of the research reactor RA in the storage water pool is examined in the framework nr the International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) 'Corrosion of Research Reactor Clad-Clad Spent Fuel in Water' since 2002. Standard racks with aluminium coupons are exposed to water in the spent fuel pools of the research reactor RA. After predetermined exposure times along with periodic monitoring of the water parameters, the coupons are examined according to the strategy and the protocol supplied by the IAEA. Description of the standard corrosion racks, experimental protocols, test procedures, water quality monitoring and compilation of results of visual examination of corrosion effects are present in this article. (author)

  13. Safety analysis of LWR irradiated fuel element pool storages before reprocessing

    International Nuclear Information System (INIS)

    Lefort, G.; Leclerc, J.; Hoffman, A.; Frejaville, C.; Domage, M.

    1984-01-01

    The protection of operators and environment requires imperatively that the safety must be taken into account as early as the design of the pools takes place and working conditions are defined. The analysis of criticality, irradiation, contamination, external or internal aggression hazards... allows to draw the main constraints which must be retained in the sizing of these pools: the criticality risk needs distances between fuel elements which results in a not very good utilization of the available area which leads to the utilization of neutron shieldings or requires a safe knowledge of the fuel elements burn up; the irradiation and contamination risks require a special quality of the pool water (temperature, activity, purity...) a good tightness of the basins to locate and to isolate the dubions fuel elements; the external or internal aggression risks such as earthquakes, missiles or loads drops, explosion, imply the civil engineering and involve the use of special technical devices. A brief presentation of the pool storages of the next UP2-800 and UP3 A reprocessing plants allows to show how the requirement drawn by safety analysis have been enforced, while carrying out civil engineering works without equivalent in the world, in this field. The foreseeable evolution of the uranium enrichment rate and burn-up of next PWR fuel elements have an effect upon the risk evaluations; a device apparatus, developed in CEA, for the measurement of burn up and cooling time is presented. At least, a short presentation of the mechanical structure durability studies of the reception and storage spent fuels installations are allowed to improve our knowledge in working conditions and in case of serious accidents

  14. Partition wall structure in spent fuel storage pool and construction method for the partition wall

    International Nuclear Information System (INIS)

    Izawa, Masaaki

    1998-01-01

    A partitioning wall for forming cask pits as radiation shielding regions by partitioning inside of a spent fuel storage pool is prepared by covering both surface of a concrete body by shielding metal plates. The metal plate comprises opposed plate units integrated by welding while sandwiching a metal frame as a reinforcing material for the concrete body, the lower end of the units is connected to a floor of a pool by fastening members, and concrete is set while using the metal plate of the units as a frame to form the concrete body. The shielding metal plate has a double walled structure formed by welding a lining plate disposed on the outer surface of the partition wall and a shield plate disposed to the inner side. Then the term for construction can be shortened, and the capacity for storing spent fuels can be increased. (N.H.)

  15. Quality of water from the pool, original containers and aluminum drums used for storage of spent nuclear fuel

    International Nuclear Information System (INIS)

    Idjakovic, Z.; Milonjic, S.; Cupic, S.

    2001-01-01

    Results of chemical analyses of water from the pool, including original containers and aluminium drums, for storage of spent nuclear fuel of the research reactor RA at the VINCA Institute and a short survey of the water properties from similar pools of other countries are presented in the paper. (author)

  16. Activity release from damaged fuel during the Paks-2 cleaning tank incident in the spent fuel storage pool

    International Nuclear Information System (INIS)

    Hozer, Zoltan; Szabo, Emese; Pinter, Tamas; Varju, Ilona Baracska; Bujtas, Tibor; Farkas, Gabor; Vajda, Nora

    2009-01-01

    During crud removal operations the integrity of 30 fuel assemblies was lost at high temperature at the unit No. 2 of the Paks NPP. Part of the fission products was released from the damaged fuel into the coolant of the spent fuel storage pool. The gaseous fission products escaped through the chimney from the reactor hall. The volatile and non-volatile materials remained mainly in the coolant and were collected on the filters of water purification system. The activity release from damaged fuel rods during the Paks-2 cleaning tank incident was estimated on the basis of coolant activity concentration measurements and chimney activity data. The typical release rate of noble gases, iodine and caesium was 1-3%. The release of non-volatile fission products and actinides was also detected.

  17. Activity release from damaged fuel during the Paks-2 cleaning tank incident in the spent fuel storage pool

    Energy Technology Data Exchange (ETDEWEB)

    Hozer, Zoltan, E-mail: hozer@aeki.kfki.h [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest 114, P.O. Box 49 (Hungary); Szabo, Emese [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest 114, P.O. Box 49 (Hungary); Pinter, Tamas; Varju, Ilona Baracska; Bujtas, Tibor; Farkas, Gabor [Nuclear Power Plant Paks, H-7031 Paks, P.O. Box 71 (Hungary); Vajda, Nora [Institute of Nuclear Techniques, Budapest University of Technology and Economics, H-1521 Budapest, Muegyetem rakpart 9 (Hungary)

    2009-07-01

    During crud removal operations the integrity of 30 fuel assemblies was lost at high temperature at the unit No. 2 of the Paks NPP. Part of the fission products was released from the damaged fuel into the coolant of the spent fuel storage pool. The gaseous fission products escaped through the chimney from the reactor hall. The volatile and non-volatile materials remained mainly in the coolant and were collected on the filters of water purification system. The activity release from damaged fuel rods during the Paks-2 cleaning tank incident was estimated on the basis of coolant activity concentration measurements and chimney activity data. The typical release rate of noble gases, iodine and caesium was 1-3%. The release of non-volatile fission products and actinides was also detected.

  18. Conception of a sub aquatic lighting system for nuclear fuels storage pools

    International Nuclear Information System (INIS)

    Bracco, P.; Rosenthal, E.

    1990-01-01

    Restrictions like contaminated water, irradiated fuel elements in racks located on the bottom of the pool and the impossibility of removing the water, require a non conventional design of pool lamps. The model developed is independent of the pool, permitting easily fabrication and maintenance. They are made of stainless steel tubes with borosilicate windows, where floodlight or light are located. The lamp assembly is fixed at the border of the pool. The system offers advantages over the conventional pool lighting systems in fabrication, operation and maintenance. (author)

  19. Safety aspects of the cleaning and conditioning of radioactive sludge from spent fuel storage pool on 'RA' Research reactor in the Vinca Institute

    International Nuclear Information System (INIS)

    Pavlovic, R; Pavlovic, S.; Plecas, I.

    1999-01-01

    Spent fuel elements from nuclear reactors in the Vinca Institute have been temporary stored in water filled storage pool. Due to the fact that the water in the spent fuel elements storage pool have not been purified for a long time, all metallic components submerged in the water have been hardly corroded and significant amount of the sludge has been settled on the bottom of the pool. As a first step in improving spent fuel elements storage conditions and slowing down corrosion in the storage spent fuel elements pool we have decided to remove the sludge from the bottom of the pool. Although not high, but slightly radioactive, this sludge had to be treated as radioactive waste material. Some safety aspects and radiation protection measures in the process of the spent fuel storage pool cleaning are presented in this paper

  20. ORIGAMI Automator Primer. Automated ORIGEN Source Terms and Spent Fuel Storage Pool Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Wieselquist, William A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Thompson, Adam B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bowman, Stephen M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Peterson, Joshua L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-04-01

    Source terms and spent nuclear fuel (SNF) storage pool decay heat load analyses for operating nuclear power plants require a large number of Oak Ridge Isotope Generation and Depletion (ORIGEN) calculations. SNF source term calculations also require a significant amount of bookkeeping to track quantities such as core and assembly operating histories, spent fuel pool (SFP) residence times, heavy metal masses, and enrichments. The ORIGEN Assembly Isotopics (ORIGAMI) module in the SCALE code system provides a simple scheme for entering these data. However, given the large scope of the analysis, extensive scripting is necessary to convert formats and process data to create thousands of ORIGAMI input files (one per assembly) and to process the results into formats readily usable by follow-on analysis tools. This primer describes a project within the SCALE Fulcrum graphical user interface (GUI) called ORIGAMI Automator that was developed to automate the scripting and bookkeeping in large-scale source term analyses. The ORIGAMI Automator enables the analyst to (1) easily create, view, and edit the reactor site and assembly information, (2) automatically create and run ORIGAMI inputs, and (3) analyze the results from ORIGAMI. ORIGAMI Automator uses the standard ORIGEN binary concentrations files produced by ORIGAMI, with concentrations available at all time points in each assembly’s life. The GUI plots results such as mass, concentration, activity, and decay heat using a powerful new ORIGEN Post-Processing Utility for SCALE (OPUS) GUI component. This document includes a description and user guide for the GUI, a step-by-step tutorial for a simplified scenario, and appendices that document the file structures used.

  1. Test on Similarity between the Flooded and Optimum Moderation Conditions of the Spent Fuel Storage Pool

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Gil Soo; Jang, Chang Sun; Woo, Sweng Woong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2009-05-15

    In the criticality safety analysis, uncertainty and bias should be considered. The final multiplication factor including uncertainty and bias in addition to calculated k-eff should be below the administrative limit. The administrative limit of spent fuel pool is 0.95 with flooded condition (filled with unborated water), and 0.98 with optimum moderation condition (filled with foggy unborated water, usually occurs near 0.1g/cc water density) for new fuel storage. The bias is determined by comparing the calculation results of the critical experiments ever performed. It is important to choose 'good' experiments which have 'similar' condition with application. To obtain realistic bias, many experiments with similar conditions should be chosen and considered. In previous approach, same critical experiment set are used to determine bias of the flooded and optimum moderation conditions. It would be correct way if two conditions are similar. The similarity test on this paper was performed by TSUNAMI code included in SCALE5.1 package. TSUNAMI code produces sensitivity data for each nuclear reaction by using first order perturbation theory. TSUNAMI code performs forward and adjoint multigroup Monte Carlo calculation. Sensitivity data are obtained by forward and adjoint results. TSUNAMI also produces uncertainty data with sensitivity data and cross section covariance data. In this paper, similarity is determined by comparing energy of average lethargy of fission (EALF), uncertainty data, sensitivity data, and correlation coefficient which is also output of the TSUNAMI code.

  2. Development of maintenance technology with underwater TIG welding for spent fuel storage pool

    International Nuclear Information System (INIS)

    Obana, Takeshi; Hamada, Yasumitsu; Ooeda, Kaoru; Katou, Masahide; Ootsuka, Toshihiro; Toyoda, Seiichi; Hosogane, Atsushi

    2007-01-01

    The core technology of underwater TIG welding process has been developed and welding equipment system has been manufactured, for application to the maintenance of the spent fuel storage pool of Rokkasho reprocessing plant. Basic experiments for understanding the conditions of dry area and the range of welding conditions was performed, and mock examination for simulation of real environment by using the developed welding equipment was also carried out to judge the applicability of the system. For the purpose that can be selected water removing method for different spatial conditions of the parts to be maintained in underwater, two kinds of welding equipment systems of Chamber type and Partition type were developed and manufactured. On the basis of fundamental experiments, the conditions of dry area formation and welding parameters range for high-reliability weld were discussed. Thus the proper condition in this process was able to be established. With the welding equipment systems of the Chamber type and Partition type, the practical use examination of underwater TIG welding process was executed by mock examination for simulating the real environment. As a result, it was confirmed that the underwater TIG welding could obtain the same reliability as a usual in-air TIG welding, and the operation and the control at remote distance were also possible. And the reliability of the patch-plate fillet weld could be evaluated by remote inspection with the expansion visual test. (author)

  3. Preliminary study on detection technology of the cladding weld of spent fuel storage pool

    Science.gov (United States)

    Qi, Pan; Cui, Hongyan; Feng, Meiming; Shao, Wenbin; Liao, Shusheng; Li, Wei

    2018-04-01

    As the first barrier of the Spent fuel storage pool, the steel cladding using different sizes (length×width) of 304L stainless steel with 3˜6mm thickness plate argon arc welded together which is direct contacted with boric acid water. Environmental humidity between the back of steel cladding and concrete, makes phosphate, chloride ion overflowed from the concrete that corroded on the weld zone with different mechanism. Part of the corrosion defects can penetrate leaded to leakage of boric acid water in penetration position accelerated crack propagation. In view of the above situation and combined with the actual needs of the power plant, the development of effective underwater nondestructive testing means of the weld area for periodic inspection and monitoring is necessary. A single method may lead to the missing of defects detection due to weld reinforcement unpolished. In this paper, eddy current array (ARRAY) and Alternating Current Field Measurement (ACFM) are adapted to test the limit sensitivity and resolution through by the specimens with artificial defects which make their detection abilities close to satisfy engineering requirements. The preliminary study found that Φ0.5mm through-wall hole and with 2mm length and 0.3mm width through-wall crack in the weld can be good inspected.

  4. Construction and cost experience regarding the 2nd pool house for spent fuel storage facility in the Atucha Power Station

    International Nuclear Information System (INIS)

    Barbosa, C.A.

    1980-01-01

    The Atucha I second pool house storage for spent fuel is designed as an extension of the Atucha I power station. The two are linked by civil structure, controlling circuits, electrical and compressed air and water supplies, low level wastes disposal, ventilation under pressure maintenance, and, most important, the ability to transfer spent and new fuel in both directions. Because the second pool house is, by location and design, an extension of the existing installation, and since there is no design departure, regarding storage and transfer of fuel from that of the original installation, the rules and regulations applied for its construction were the same as those valid for the Atucha I construction. The requirement not to exceed a four-year period for construction and commissioning was determined by the need to have storage room for the Atucha I fuel. Argentina will meet the 1982 target by having the installation available during the second half of 1981. The second pool house is a wet storage location with a capacity of 1000 tons metallic uranium. It was designed by the Kraftwerk Union of West Germany along the same lines as the 440-ton storage location originally built with the station. The Atomic Energy Commission of Argentina has managed the construction and participated in project and design stages. As in the original pool, the 6 m long assemblies are stacked in double tiers. The cost figures which are mentioned differ from previously released figures and are not the final ones. With civil construction almost finished and mechanical erection started, the present estimates should not differ by more than 10% from the final figures. The installation has an investment cost of 61 million dollars, (1980), and, depending on the amortization time span considered, a total yearly cost per kg of capacity of metallic uranium, ranging between 5.5 and 9.3 dollars per kg

  5. Fuel storage

    International Nuclear Information System (INIS)

    Palacios, C.; Alvarez-Miranda, A.

    2009-01-01

    ENSA is a well known manufacturer of multi-system primary components for the nuclear industry and is totally prepared to satisfy future market requirements in this industry. At the same time that ENSA has been gaining a reputation world wider for the supply of primary components, has been strengthening its commitment and experience in supplying spent fuel components, either pool racks or storage and transportation casks, and offers not only fabrication but also design capabilities for its products. ENSA has supplied Spent Fuel Pool Racks, in spain, Finland, Taiwan, Korea, China, and currently it is in the process of licensing its own rack design in the United States of America for the ESBWR along with Ge-Hitachi. ENSA has supplied racks for 20 pools and 22 different reactors and it has also manufactured racks under all available technologies and developed a design known as Interlock Cell Matrix whose main features are outlined in this article. Another ENSA achievement in rack technology is the use of remote control for re-racking activities instead of using divers, which improves the ALARA requirements. Regarding casks for storage and transportation, ENSA also has al leading worldwide position, with exports prevailing over the Spanish market where ENSA has supplied 16 storage and transportation casks to the Spanish nuclear power Trillo. In some cases, ENSA acts as subcontractor for other clients. Foreign markets are still a major challenge for ENSA. ENSA-is well known for its manufacturing capabilities in the nuclear industry, but has been always involved in design activities through its engineering division, which carries out different tasks: components Design; Tooling Design; Engineering and Documentation; Project Engineering; Calculations, Design and Development Engineering. (Author)

  6. Fuel storage tank

    International Nuclear Information System (INIS)

    Peehs, M.; Stehle, H.; Weidinger, H.

    1979-01-01

    The stationary fuel storage tank is immersed below the water level in the spent fuel storage pool. In it there is placed a fuel assembly within a cage. Moreover, the storage tank has got a water filling and a gas buffer. The water in the storage tank is connected with the pool water by means of a filter, a surge tank and a water purification facility, temperature and pressure monitoring being performed. In the buffer compartment there are arranged catalysts a glow plugs for recombination of radiolysis products into water. The supply of water into the storage tank is performed through the gas buffer compartment. (DG) [de

  7. Spent fuel storage rack

    International Nuclear Information System (INIS)

    Morikawa, Matsuo; Uchiyama, Yuichi.

    1983-01-01

    Purpose: To improve the safety and facilitate the design by limiting the relative displacement in a storage rack. Constitution: The outer wall of a storage rack disposed in water within a fuel pool, the pool wall opposing to the storage rack and the structure between the opposing storages racks are made as a space for confining the pool water or a structure formed with a slight gap, for example, a combination of a recessed structure and a protruded structure. In such a constitution, a space for confirming the pool water is established and the pool water thus confined forms a flow resistance when the storage rack vibrates upon earthquakes, serves as a damper and significantly reduces the responsivity. Furthermore, the relative displacement in the storage rack is limited to inhibit excess earthquake forces to exert on setting bolts and rack clamping bolts of the storage rack. (Sekiya, K.)

  8. A Critical Review of Practice of Equating the Reactivity of Spent Fuel to Fresh Fuel in Burnup Credit Criticality Safety Analyses for PWR Spent Fuel Pool Storage

    International Nuclear Information System (INIS)

    Wagner, J.C.; Parks, C.V.

    2000-01-01

    This research examines the practice of equating the reactivity of spent fuel to that of fresh fuel for the purpose of performing burnup credit criticality safety analyses for PWR spent fuel pool (SFP) storage conditions. The investigation consists of comparing k inf estimates based on reactivity equivalent fresh fuel enrichment (REFFE) to k inf estimates using the actual spent fuel isotopics. Analyses of selected storage configurations common in PWR SFPs show that this practice yields nonconservative results (on the order of a few tenths of a percent) in configurations in which the spent fuel is adjacent to higher-reactivity assemblies (e.g., fresh or lower-burned assemblies) and yields conservative results in configurations in which spent fuel is adjacent to lower-reactivity assemblies (e.g., higher-burned fuel or empty cells). When the REFFE is determined based on unborated water moderation, analyses for storage conditions with soluble boron present reveal significant nonconservative results associated with the use of the REFFE. This observation is considered to be important, especially considering the recent allowance of credit for soluble boron up to 5% in reactivity. Finally, it is shown that the practice of equating the reactivity of spent fuel to fresh fuel is acceptable, provided the conditions for which the REFFE was determined remain unchanged. Determination of the REFFE for a reference configuration and subsequent use of the REFFE for different configurations violates the basis used for the determination of the REFFE and, thus, may lead to inaccurate, and possibly, nonconservative estimates of reactivity. A significant concentration (approx. 2000 ppm) of soluble boron is typically (but not necessarily required to be) present in PWR SFPs, of which only a portion (le 500 ppm) may be credited in safety analyses. Thus, a large subcritical margin currently exists that more than accounts for errors or uncertainties associated with the use of the REFFE

  9. Isocrit: a burnup credit tool for spent fuel pool storage calculations - 333

    International Nuclear Information System (INIS)

    Kucukboyaci, V.N.; Marshall, W.J.

    2010-01-01

    In order to conservatively apply burnup credit in spent fuel pool criticality safety analyses, Westinghouse has developed a software tool, ISOCRIT, for generating depletion isotopics. This tool is used to create isotopics data based on specific reactor input parameters, such as design basis assembly type; bounding power/burnup profiles; reactor specific moderator temperature profiles; pellet percent theoretical density; burnable absorbers, axial blanket regions, and bounding ppm boron concentration. ISOCRIT generates burnup dependent isotopics using PARAGON; Westinghouse's state-of-the-art and licensed lattice physics code. Generation of isotopics and passing the data to the subsequent 3D KENO calculations are performed in an automated fashion, thus reducing the chance for human error. Furthermore, ISOCRIT provides the means for responding to any customer request regarding re-analysis due to changed parameters (e.g., power up-rate, exit temperature changes, etc) with a quick turnaround. (authors)

  10. Nuclear fuel storage facility

    International Nuclear Information System (INIS)

    Matsumoto, Takashi; Isaka, Shinji.

    1987-01-01

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

  11. Practical experience for liquid radioactive waste treatment from spent fuel storage pool on RA reactor in Vinca Institute

    International Nuclear Information System (INIS)

    Plecas, I.; Pavlovic, R.; Pavlovic, S.

    2002-01-01

    The present paper reports the results of the preliminary removal of sludge from the bottom of the spent fuel storage pool in the RA reactor, mechanical filtration of the pool water and sludge conditioning and storage. Yugoslavia is a country without a nuclear power plant (NPP) on its territory. The law which strictly forbids NPP construction is still valid, but, nevertheless we must handle and dispose radioactive waste. This is not only because of radwaste originating from the use of radioactive materials in medicine and industry, but also because of the waste generated by research in the Nuclear Sciences Institute Vinca. In the last forty years, in the Vinca Institute, as a result of two research reactors being operational, named RA and RB, and as a result of the application of radionuclides in medicine, industry and agriculture, radioactive waste materials of different levels of specific activity were generated. As a temporary solution, radioactive waste materials are stored in two interim storages. Radwaste materials that were immobilized in the inactive matrices are to be placed in concrete containers, for further manipulation and disposal.(author)

  12. Nuclear fuel storage

    International Nuclear Information System (INIS)

    Bevilacqua, F.

    1979-01-01

    A method and apparatus for the storage of fuel in a stainless steel egg crate structure within a storage pool are described. Fuel is initially stored in a checkerboard pattern or in each opening if the fuel is of low enrichment. Additional fuel (or fuel of higher enrichment) is later stored by adding stainless steel angled plates within each opening, thereby forming flux traps between the openings. Still higher enrichment fuel is later stored by adding poison plates either with or without the stainless steel angles. 8 claims

  13. Spent fuel element storage facility

    International Nuclear Information System (INIS)

    Ukaji, Hideo; Yamashita, Rikuo.

    1981-01-01

    Purpose: To always keep water level of a spent fuel cask pit equal with water level of spent fuel storage pool by means of syphon principle. Constitution: The pool water of a spent fuel storage pool is airtightly communicated through a pipe with the pool water of a spent fuel cask, and a gate is provided between the pool and the cask. Since cask is conveyed into the cask pit as the gate close while conveying, the pool water level is raised an amount corresponding to the volume of the cask, and water flow through scattering pipe and the communication pipe to the storage pool. When the fuel is conveyed out of the cask, the water level is lowered in the amount corresponding to the volume in the cask pit, and the water in the pool flow through the communication pipe to the cask pit. (Sekiya, K.)

  14. Method of performing shutdown reactivity measurements in spent nuclear fuel storage pools

    International Nuclear Information System (INIS)

    Levine, S.H.; Schultz, M.A.; Chang, D.

    1981-01-01

    The objective of this paper is to develop a device to measure the k/infinity/ of a spent fuel assembly used in light water reactors. A subcritical assembly having a cross configuration is designed to allow measurement of the k/sub //infinity/ of a spent fuel assembly by comparing the change in its multiplication with that of a fuel assembly of known k/infinity/. Calculations have been performed using nucleonic codes to develop polynomial equations that relate the k/infinity/ of the spent fuel assembly to measured data. The measurements involve taking count rates with the spent fuel assembly in the center position of the subcritical assembly, and the measured data are the count rate ratio of the spent fuel assembly over the count rate taken with a fuel assembly of known k/infinity/. The polynomial equations are easy to program on a microcomputer, which, together with the subcritical assembly, form the k/infinity/ meter. 9 refs

  15. Spent fuel storage requirements

    International Nuclear Information System (INIS)

    Fletcher, J.

    1982-06-01

    Spent fuel storage requirements, as projected through the year 2000 for U.S. LWRs, were calculated using information supplied by the utilities reflecting plant status as of December 31, 1981. Projections through the year 2000 combined fuel discharge projections of the utilities with the assumed discharges of typical reactors required to meet the nuclear capacity of 165 GWe projected by the Energy Information Administration (EIA) for the year 2000. Three cases were developed and are summarized. A reference case, or maximum at-reactor (AR) capacity case, assumes that all reactor storage pools are increased to their maximum capacities as estimated by the utilities for spent fuel storage utilizing currently licensed technologies. The reference case assumes no transshipments between pools except as currently licensed by the Nuclear Regulatory Commission (NRC). This case identifies an initial requirement for 13 MTU of additional storage in 1984, and a cumulative requirement for 14,490 MTU additional storage in the year 2000. The reference case is bounded by two alternative cases. One, a current capacity case, assumes that only those pool storage capacity increases currently planned by the operating utilities will occur. The second, or maximum capacity with transshipment case, assumes maximum development of pool storage capacity as described above and also assumes no constraints on transshipment of spent fuel among pools of reactors of like type (BWR, PWR) within a given utility. In all cases, a full core discharge capability (full core reserve or FCR) is assumed to be maintained for each reactor, except that only one FCR is maintained when two reactors share a common pool. For the current AR capacity case the indicated storage requirements in the year 2000 are indicated to be 18,190 MTU; for the maximum capacity with transshipment case they are 11,320 MTU

  16. Accident conditions analysis of spent fuel storage pool RA research reactor in Vinca; Analiza udesnih stanja u odlagalistu isluzenog goriva istrazivackog rektora RA u Vinci

    Energy Technology Data Exchange (ETDEWEB)

    Jovic, V; Jovic, L [Institute of Nuclear Sciences VINCA, Belgrade (Serbia and Montenegro)

    2000-07-01

    Based on Safety analysis of the spent fuel pool RA research reactor in Vinca, conditions and possibilities accident sequences in present configuration storage facility are considered (author) [Serbo-Croat] Na osnovu Analize sigurnosti odlagalista isluzenog goriva istrazivackog reaktora RA u Vinci razmatraju se uslovi i mogucnosti pojave udesnih stanja u postojecoj konfiguraciji odlagalista (author)

  17. Gravity-driven flow and heat transfer in a spent nuclear fuel storage pool

    International Nuclear Information System (INIS)

    Gay, R.R.

    1983-01-01

    The GFLOW code analyzes a three-dimensional rectangular porous medium by dividing the porous medium into a number of nodes or cells specified by the user. The finite difference form of the fluid conservation equations is solved for each node by application of a modified ''marker and cell'' numerical technique. The existence of spent nuclear fuel in any node is modeled by using a porosity value less than unity in that node and by including a surface heat transfer term in the fluid energy equation. In addition, local pressure losses due to grid spaces or other planar flow obstructions can be modeled by local loss coefficients. Heat conduction in the fuel is simulated by a fast running implicit finite difference model of the fuel, gap, and clad regions of the fuel rod

  18. Optimal cost design of base-isolated pool structures for the storage of nuclear spent fuel

    International Nuclear Information System (INIS)

    Ko, H. M.; Park, K. S.; Song, J. H.

    1999-01-01

    A method of cost-effectiveness evaluation for seismic isolated pool structures is presented. Input ground motion is modeled as spectral density function compatible with response spectrum for combination of acceleration coefficient and site coefficient. Interaction effects between flexible walls and contained fluid are considered in the form of added mass matrix. Wall thickness and isolator stiffness are adopted as design variables for optimization. Transfer function vector of the structure-isolator system is derived from the equation of motion. Spectral analysis method based on random vibration theories is used for the calculation of failure probability. The exemplifying designs and analyses show that cost-effectiveness of isolated pool structure is relatively high in low-moderate seismic region and stiff soil condition. Sensitiveness of optimal design variables to assumed damage scales is relatively low in such region

  19. High density fuel storage rack

    International Nuclear Information System (INIS)

    Zezza, L.J.

    1980-01-01

    High storage density for spent nuclear fuel assemblies in a pool achieved by positioning fuel storage cells of high thermal neutron absorption materials in an upright configuration in a rack. The rack holds the cells at required pitch. Each cell carries an internal fuel assembly support, and most cells are vertically movable in the rack so that they rest on the pool bottom. Pool water circulation through the cells and around the fuel assemblies is permitted by circulation openings at the top and bottom of the cells above and below the fuel assemblies

  20. Compact nuclear fuel storage

    International Nuclear Information System (INIS)

    Kiselev, V.V.; Churakov, Yu.A.; Danchenko, Yu.V.; Bylkin, B.K.; Tsvetkov, S.V.

    1983-01-01

    Different constructions of racks for compact storage of spent fuel assemblies (FA) in ''coolin''g pools (CP) of NPPs with the BWR and PWR type reactors are described. Problems concerning nuclear and radiation safety and provision of necessary thermal conditions arising in such rack design are discussed. It is concluded that the problem of prolonged fuel storage at NPPs became Very actual for many countries because of retapdation of the rates of fuel reprocessing centers building. Application of compact storage racks is a promising solution of the problem of intermediate FA storage at NPPs. Such racks of stainless boron steel and with neutron absorbers in the from of boron carbide panels enable to increase the capacity of the present CP 2-2.6 times, and the period of FA storage in them up to 5-10 years

  1. Extended storage of spent fuel

    International Nuclear Information System (INIS)

    1992-10-01

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

  2. Stress corrosion (Astm G30-90 standard) in 08x18H10T stainless steel of nuclear fuel storage pool in WWER reactors

    International Nuclear Information System (INIS)

    Herrera, V.; Zamora R, L.

    1997-01-01

    At the water storage of the irradiated nuclear fuel has been an important factor in its management. The actual pools have its walls covered with inoxidable steel and heat exchangers to dissipate the residual heat from fuel. It is essential to control the water purity to eliminate those conditions which aid to the corrosion process in fuel and at related components. The steel used in this research was obtained from an austenitic inoxidizable steel standardized with titanium 08x18H10T (Type 321) similar to one of the two steel coatings used to cover walls and the pools floor. the test consisted in the specimen deformation through an U ply according to the Astm G30-90 standard. The exposition of the deformed specimen it was realized in simulated conditions to the chemical regime used in pools. (Author)

  3. Nuclear fuel storage

    International Nuclear Information System (INIS)

    Bevilacqua, F.

    1981-01-01

    A nuclear fuel storage apparatus for use in a water-filled pool is fabricated of a material such as stainless steel in the form of an egg crate structure having vertically extending openings. Fuel may be stored in this basic structure in a checkerboard pattern with high enrichment fuel, or in all openings when the fuel is of low effective enrichment. Inserts of a material such as stainless steel are adapted to fit within these openings so that a water gap and, therefore, a flux trap is formed between adjacent fuel storage locations. These inserts may be added at a later time and fuel of a higher enrichment may be stored in each opening. When it is desired to store fuel of still greater enrichment, poison plates may be added to the water gap formed by the installed insert plates, or substituted for the insert plates. Alternately, or in addition, fuel may be installed in high neutron absorption poison boxes which surround the fuel assembly. The stainless steel inserts and the poison plates are each not required until the capacity of the basic egg crate structure is approached. Purchase of these items can, therefore, be deferred for many years. Should the fuel to be stored be of higher enrichment than initially forecast, the deferred decision on the poison plates makes it possible to obtain increased poison in the plates to satisfy the newly discovered requirement

  4. Casette for storage of spent fuel assemblies

    International Nuclear Information System (INIS)

    Ericsson, S.

    1992-01-01

    Describes a design of a casette for spent fuel storage in a fuelstorage pool. The new design, based on flexible spacers, allows the fuel assemblies to be packed more compact and the fuel storage pool used in a more economic way

  5. Expansion of capacity of spent fuel pools and associated problems

    International Nuclear Information System (INIS)

    Francisco, J.L. De; Lopez-Cotarelo, J.; Ramos, J.M.

    1978-01-01

    Expanding the spent fuel storage pool capacity is a good solution for utilities facing the current shortage in fuel reprocessing capacity. The problems more likely to be found when expanding a spent fuel storage facility by using high density storage racks are reviewed. Basically three types of problems arise: 1) Problems related with the characteristics of the new facility. 2) Problems related with the works of expansion. 3) Problems related with the long term storage of large quantities of spent fuel. (author)

  6. Spent fuel pool cleanup and stabilization

    International Nuclear Information System (INIS)

    Miller, R.L.

    1987-06-01

    Each of the plutonium production reactors at Hanford had a large water-filled spent fuel pool to provide interim storage of irradiated fuel while awaiting shipment to the separation facilities. After cessation of reactor operations the fuel was removed from the pools and the water levels were drawn down to a 5- to 10-foot depth. The pools were maintained with the water to provide shielding and radiological control. What appeared to be a straightforward project to process the water, remove the sediments from the basin, and stabilize the contamination on the floors and walls became a very complex and time consuming operation. The sediment characteristics varied from pool to pool, the ion exchange system required modification, areas of hard-pack sediments were discovered on the floors, special arrangements to handle and package high dose rate items for shipment were required, and contract problems ensued with the subcontractor. The original schedule to complete the project from preliminary engineering to final stabilization of the pools was 15 months. The actual time required was about 25 months. The original cost estimate to perform the work was $2,651,000. The actual cost of the project was $5,120,000, which included $150,000 for payment of claims to the subcontractor. This paper summarizes the experiences associated with the cleanup and radiological stabilization of the 100-B, -C, -D, and -DR spent fuel pools, and discusses a number of lessons learned items

  7. Equipment for nondestructive testing of the PWR and BWR spept fUel elements and assemblies in the NPP storage pools

    International Nuclear Information System (INIS)

    Gorskij, V.V.

    1983-01-01

    Design features are considered of units for nondestructive testing of spent fUel elements and fuel assemblies (FA) in the storage pools of NPP with the PWR and BWR reactors. Units for remote viewing control of fuel element cans and FA, for direct measurements of their geometrical dimensions, for FA leak-testing, fuel element can nondestructive testing and gamma scanning, for measuring gaseous fission product pressure and fuel element free volume are described along with units for complex checking of fuel element and FA parameters. The units for nondestructive testing of spent fuel elements and EA are shown to differ both in their designs and a number of checked parameters of fuel elements and FA. The remote viewing and those for measuring the basic FA parameters are most generally employed. Units for complex testing of multiple fuel element parameters, designed in the last few years, are intended for operation with FA disassembled partially or fully and are characteristic of a high degree of computer measuring automation both for the process control and data processing

  8. Nuclear reactors and fuel storage pools security in France and in Belgium, and the associated reinforcement measures. Press kit

    International Nuclear Information System (INIS)

    Besnard, Manon; Marignac, Yves; Boilley, David; MacKerron, Gordon; Becker, Oda; Lyman, Ed; Zerbib, Jean-Claude; Sotty, Meryl

    2017-10-01

    This report on the security of nuclear reactors and fuel storage pools in France and Belgium draws on the contributions of seven experts from France, Germany, the United Kingdom and the United States - specialists in nuclear safety, security, radioprotection and economics - to bring together the full range of expertise necessary to analyse the problem. Each contributor is only responsible for his/her contribution in the form it was commissioned and separately delivered to Greenpeace France. This report looks at an eminently sensitive subject in an extremely delicate context. Critical analysis and independent expertise must, in line with this, be approached in a way which conciliates two conflicting requirements. The first is that of democratic choice. The security of nuclear facilities in the face of external attacks must be open to public debate. There is no justifiable reason for this issue, a major factor in the assessment of risks related to different industrial and energy options, to be excluded from the democratic process. The public has a fundamental right to information about the risks associated with the operation of nuclear facilities, which includes assessing the risk of external attacks in all its dimensions. It is thus the responsibility of non-institutional experts to contribute to this debate. The second requirement, equally fundamental, is the preservation of public security. Contributing to this debate in a way that enhances the risk of an external attack on nuclear facilities, or, even worse, favours the success of a possible attack by revealing any flaws in the system, is of course out of the question. It is therefore also the responsibility of non-institutional experts to ensure that the protection of such facilities is not impaired by the information they collect or the analyses they produce and make available to the public. It is particularly difficult to strike this balance in the French context, since the authorities responsible for nuclear

  9. Investigation of the thermal performance of a vertical two-phase closed thermosyphon as a passive cooling system for a nuclear reactor spent fuel storage pool

    Energy Technology Data Exchange (ETDEWEB)

    Kusuma, Mukhsinun Hadi; Putra, Nandy; Imawan, Ficky Augusta [Heat Transfer Laboratory, Department of Mechanical Engineering Universitas Indonesia, Kampus (Indonesia); Antariksawan, Anhar Riza [Centre for Nuclear Reactor Safety and Technology, National Nuclear Energy Agency of Indonesia (BATAN), Kawasan Puspiptek Serpong (Indonesia)

    2017-04-15

    The decay heat that is produced by nuclear reactor spent fuel must be cooled in a spent fuel storage pool. A wickless heat pipe or a vertical two-phase closed thermosyphon (TPCT) is used to remove this decay heat. The objective of this research is to investigate the thermal performance of a prototype model for a large-scale vertical TPCT as a passive cooling system for a nuclear research reactor spent fuel storage pool. An experimental investigation and numerical simulation using RELAP5/MOD 3.2 were used to investigate the TPCT thermal performance. The effects of the initial pressure, filling ratio, and heat load were analyzed. Demineralized water was used as the TPCT working fluid. The cooled water was circulated in the water jacket as a cooling system. The experimental results show that the best thermal performance was obtained at a thermal resistance of 0.22°C/W, the lowest initial pressure, a filling ratio of 60%, and a high evaporator heat load. The simulation model that was experimentally validated showed a pattern and trend line similar to those of the experiment and can be used to predict the heat transfer phenomena of TPCT with varying inputs.

  10. Modular dry storage of spent fuel

    International Nuclear Information System (INIS)

    Baxter, J.W.

    1982-01-01

    Long term uncertainties in US spent fuel reprocessing and storage policies and programs are forcing the electric utilities to consider means of storing spent fuel at the reactor site in increasing quantitities and for protracted periods. Utilities have taken initial steps in increasing storage capacity. Existing wet storage pools have in many cases been reracked to optimize their capacity for storing spent fuel assemblies

  11. The management of the Spend Fuel Pool Water Quality (1996-2007)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Hwan; Lee, Eui Gyu; Choi, Ho Young; Choi, Mun Jo; Kim, Hyung Wook; Lee, Mun; Lee, Choong Sung; Hur, Soon Ock; Ahn, Guk Hun

    2008-12-15

    The water quality management of spent fuel storage pool water quality in HANARO is important to prevent the corrosion of nuclear fuel and reactor structure material. The condition of the spent fuel storage pool water has been monitored by measuring the electrical conductivity of the spent fuel storage pool purification system and pH periodically. The status of the spent fuel storage pool water quality management was investigated by using the measured data. taken from 1996 to 2007. In general, the electrical conductivity of the spent fuel storage pool water have been managed within 1 {mu}S/cm which is an operation target of HANARO.

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-09-01

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

  14. Anticipated corrosion in the Vermont Yankee spent fuel pool

    International Nuclear Information System (INIS)

    Weeks, J.R.

    1977-06-01

    The report provides additional information relating to a proposed modification to the spent fuel pool at the Vermont Yankee Nuclear Power Station (VYNPS) and addresses corrosion of spent fuel pool storage materials and zircaloy, and provides an analysis of the effectiveness of the Boral sealing

  15. Reracking to increase spent fuel storage capacity

    International Nuclear Information System (INIS)

    1980-05-01

    Many utilities have already increased their spent fuel pool storage capacity by replacing aluminum racks having storage densities as low as 0.2 MTU/ft 2 with stainless steel racks which can more than double storage densities. Use of boron-stainless steel racks or thin stainless steel cans containing reassembled fuel rods allows even higher fuel storage densities (up to approximately 1.25 MTU/ft 2 ). This report evaluates the economics of smaller storage gains that occur if pools, already converted to high density storage, are further reracked

  16. Dry spent fuel storage licensing

    International Nuclear Information System (INIS)

    Sturz, F.C.

    1995-01-01

    In the US, at-reactor-site dry spent fuel storage in independent spent fuel storage installations (ISFSI) has become the principal option for utilities needing storage capacity outside of the reactor spent fuel pools. Delays in the geologic repository operational date at or beyond 2010, and the increasing uncertainty of the US Department of Energy's (DOE) being able to site and license a Monitored Retrievable Storage (MRS) facility by 1998 make at-reactor-site dry storage of spent nuclear fuel increasingly desirable to utilities and DOE to meet the need for additional spent fuel storage capacity until disposal, in a repository, is available. The past year has been another busy year for dry spent fuel storage licensing. The licensing staff has been reviewing 7 applications and 12 amendment requests, as well as participating in inspection-related activities. The authors have licensed, on a site-specific basis, a variety of dry technologies (cask, module, and vault). By using certified designs, site-specific licensing is no longer required. Another new cask has been certified. They have received one new application for cask certification and two amendments to a certified cask design. As they stand on the brink of receiving multiple applications from DOE for the MPC, they are preparing to meet the needs of this national program. With the range of technical and licensing options available to utilities, the authors believe that utilities can meet their need for additional spent fuel storage capacity for essentially all reactor sites through the next decade

  17. Optimization of spent fuel pool weir gate driving mechanism

    Science.gov (United States)

    Liu, Chao; Du, Lin; Tao, Xinlei; Wang, Shijie; Shang, Ertao; Yu, Jianjiang

    2018-04-01

    Spent fuel pool is crucial facility for fuel storage and nuclear safety, and the spent fuel pool weir gate is the key related equipment. In order to achieve a goal of more efficient driving force transfer, loading during the opening/closing process is analyzed and an optimized calculation method for dimensions of driving mechanism is proposed. The result of optimizing example shows that the method can be applied to weir gates' design with similar driving mechanism.

  18. Spent fuel storage rack

    International Nuclear Information System (INIS)

    Kurokawa, Hideaki; Kumagaya, Naomi; Oda, Masashi; Matsuda, Masami; Maruyama, Hiromi; Yamanaka, Tsuneyasu.

    1997-01-01

    The structure of a spent fuel storage rack is determined by the material, thickness, size of square cylindrical tubes (the gap between spent fuel assemblies and the square cylindrical tubes) and pitch of the arrangement (the gap between each of the square cylindrical tubes). In the present invention, the thickness and the pitch of the arrangement of the square tubes are optimized while evaluating subcriticality. Namely, when the sum of the thickness of the water gap at the outer side (the pitch of arrangement of the cylindrical tubes) and the thickness of the cylindrical tubes is made constant, the storage rack is formed by determining the thickness of the cylindrical tubes which is smaller than the optimum value among the combination of the thickness of the water gap at the outer side and that of the cylindrical tube under the effective multiplication factor to be performed. Then, the weight of the rack can be reduced, and the burden of the load on the bottom of the pool can be reduced. Further, the amount of the constitutional materials of the rack itself can be reduced thereby capable of reducing the cost for the materials of the rack. (T.M.)

  19. Spent fuel storage criticality safety

    Energy Technology Data Exchange (ETDEWEB)

    Amin, E M; Elmessiry, A M [National center of nuclear safety and radiation control atomic energy authority, (Egypt)

    1995-10-01

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

  20. Spent fuel storage criticality safety

    International Nuclear Information System (INIS)

    Amin, E.M.; Elmessiry, A.M.

    1995-01-01

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

  1. Expertise on the Goesgen-Daeniken nuclear power plant on the granting of a licence for the construction and operation of a water storage pool for fuel assemblies at the site of the power plant

    International Nuclear Information System (INIS)

    2003-04-01

    On June 26, 2002, the Goesgen-Daeniken AG nuclear power plant (KKG) delivered a request to the Swiss Federal Council for the granting of a licence for the construction and operation of a water storage pool for the on-site storage of the power plant's fuel assemblies. The present report contains the results of the examination of the request by the Federal Agency for the Safety of Nuclear Installations (HSK), to check that the projected storage pool satisfies the legal requirements from the point of view of nuclear safety and protection against radioactivity. A water storage pool already exists in the reactor building of KKG. It was conceived for a fuel cycle based on the reprocessing of the spent fuel assemblies. Its capacity is not sufficient when the spent fuel assemblies are no longer reprocessed but have to be transferred and stored in the Central Intermediate Storage Facility (ZWILAG) in Wuerenlingen because their heat production is too high. The capacity of the actual water pool allows a maximum cooling time of 5-6 years, while 7-10 years are required before transfer to ZWILAG. The projected new water storage pool has to be aircraft crash and earthquake proof, in the same way that the reactor building itself has to be. It can store a maximum of 1008 fuel assemblies. The water in the pool as well as the pool walls shield the radiation from of the fuel assemblies almost completely. Each fuel assembly is put into a square steel channel. The channel walls are lined with 6.11 mg/cm 2 of the neutron absorbing nuclide B-10, which guaranties the subcriticality of the water pool even if the storage pool would be entirely filled with non-irradiated fuel assemblies with the maximal allowed enrichment or the maximal allowed content of Plutonium in case of MOX fuel assemblies, which is a very conservative assumption. The heat released by decay in the spent fuel assemblies is transferred to the pool water. Storage pool cooling is carried out by natural circulation through

  2. Total quality in spent fuel pool reracking

    International Nuclear Information System (INIS)

    Cranston, J.S.; Bradbury, R.B.; Cacciapouti, R.J.

    1993-01-01

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

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

    International Nuclear Information System (INIS)

    Mohd Fazli Zakaria; Tonny Lanyau; Ahmad Nabil Ab Rahim

    2010-01-01

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

  4. Horizontal above-rack pool storage

    International Nuclear Information System (INIS)

    Moscardini, R.L.

    1993-08-01

    This report describes a unique method for storing spent, six year out of core, fuel at a prototypical PWR nuclear power station. The study describes a conceptual design, with favorable structural, thermal and criticality technical evaluations. However, economic considerations and licensing risks are judged to be less favorable. The concept study prescribes a fuel over fuel arrangement in an existing Spent Fuel Pool (SFP) with full maintenance of ALARA principles. This concept study is specific to a prototypical pool design, but may easily be projected to other nuclear facilities with other SFP conditions. For the prototypical PWR, the conceptual fuel bridge design will store over 200 additional fuel assemblies without significant facility modifications and for an indefinite time period

  5. Method of storing the fuel storage pot in a fuel storage tank for away-from-reactor-storage

    International Nuclear Information System (INIS)

    Ishiguro, Jun-ichi.

    1980-01-01

    Purpose: To prevent the contact of sodium in the away-from-reactor-storage fuel storage tank with sodium in a fuel storage pool having radioactivity ana always retain clean state therein. Method: Sodium is filled in a container body of the away-from-reactor-storage fuel storage tank, and a conduit, a cycling pump, and cooling means are disposed to form a sodium coolant cycling loop. The fuel storage pool is so stored in the container body that the heat of the pool is projected from the liquid surface of the sodium in the container. Therefore, the sodium in the container is isolated from the sodium in the pool containing strong radioactivity to prevent contact of the former sodium from the latter sodium. (Sekiya, K.)

  6. Spent fuel storage

    International Nuclear Information System (INIS)

    Huppert

    1976-01-01

    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) [de

  7. Spent nuclear fuel storage

    International Nuclear Information System (INIS)

    Romanato, Luiz Sergio

    2005-01-01

    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)

  8. Economical evaluation on spent fuel storage technology away from reactor

    International Nuclear Information System (INIS)

    Itoh, Chihiro; Nagano, Koji; Saegusa, Toshiari

    2000-01-01

    Concerning the spent fuel storage away from reactor, economical comparison was carried out between metal cask and water pool storage technology. The economic index was defined by levelized cost (Unit storage cost) calculated on the assumption that the storage cost is paid at the receipt of the spent fuel at the storage facility. It is found that the cask storage is economical for small and large storage capacity. Unit storage cost of pool storage, however, is getting close to that of cask storage in case of storage capacity of 10,000 ton. Then, the unit storage cost is converted to power generation cost using data of the burn up of the fuel, etc. The cost is obtained as yen 0.09/kWh and yen 0. 15/kWh for cask storage and pool storage, respectively in case of the capacity of 5,000 tonU and the cooling time of 5 years. (author)

  9. Spent fuel storage requirements 1993--2040

    International Nuclear Information System (INIS)

    1994-09-01

    Historical inventories of spent fuel are combined with U.S. Department of Energy (DOE) projections of future discharges from commercial nuclear reactors in the United States to provide estimates of spent fuel storage requirements through the year 2040. The needs are estimated for storage capacity beyond that presently available in the reactor storage pools. These estimates incorporate the maximum capacities within current and planned in-pool storage facilities and any planned transshipments of spent fuel to other reactors or facilities. Existing and future dry storage facilities are also discussed. The nuclear utilities provide historical data through December 1992 on the end of reactor life are based on the DOE/Energy Information Administration (EIA) estimates of future nuclear capacity, generation, and spent fuel discharges

  10. Spent-fuel-storage alternatives

    International Nuclear Information System (INIS)

    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

  11. Spent fuel heatup following loss of water during storage

    International Nuclear Information System (INIS)

    Benjamin, A.S.; McCloskey, D.J.

    1978-01-01

    Spent fuel assemblies from light water reactors are typically stored for one year or more in the reactor spent fuel pool and then transported for long-term storage at an off-site location. Because of the design, construction, and operation features of spent fuel storage pools, an accident that might drain most of the water from a pool is assessed as being extremely improbable. As a limiting case, however, a hypothetical incident involving instantaneous draining of all the water from a storage pool has been postulated, and the subsequent heatup of the spent fuel elements has been evaluated. The model is analyzed, and results are summarized

  12. Equipment for the conditioning of core components in the fuel element storage pool with particular respect to the design required by the conditions for nuclear facilities in operation and the surveillance in accordance with atomic rules and regulations

    International Nuclear Information System (INIS)

    Dumpe, J.; Schwiertz, V.; Geiser, C.; Prucker, E.

    2001-01-01

    In nuclear power plants worn out and activated parts from the reactor core (core components) which are placed into the fuel element storage pool arise on a regular basis during the technical maintenance and the review. The disposal of these core components due to radiation protection aspects is only feasible within the fuel element storage pool during the operation of the NPP using techniques of the under water conditioning. Therefore, special GNS equipment is used for the conditioning, using under water conditioning equipment, such as UWS, BZ, and ZVA, a number of lifting and auxiliary equipment for mounting and dismantling purposes and the handling of the core components and the waste casks within the fuel element storage pool. These components must meet particular safety requirements with regard to their integrity and reliability. They are designed according to the requirements on nuclear components (KTA). The manipulating equipment must be partly redundant and the protection goals for nuclear accidents must be met. The Bavarian Ministry for Development and Environment tasked the TUeV Sueddeutschland with the surveillance and control. The conditioning equipment of GNS is therefore designed in co-ordination with the examiner of the Governmental Regulating Agency, in particular respect to all safety aspects. Furthermore the examiners perform reviews of the construction and the documentation during the design and construction phase. (orig.)

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

    International Nuclear Information System (INIS)

    Bailey, W.J.

    1990-02-01

    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

  14. Spent fuel storage requirements, 1991--2040

    International Nuclear Information System (INIS)

    1991-12-01

    Historical inventories of spent fuel are combined with US Department of Energy (DOE) projections of future discharges from commercial nuclear reactors in the United States to provide estimates of spent fuel storage requirements over the next 50 years, through the year 2040. The needs for storage capacity beyond that presently available in the pools are estimated. These estimates incorporate the maximum capacities within current and planned in-pool storage facilities and any planned transshipments of fuel to other reactors or facilities. Existing and future dry storage facilities are also discussed. Historical data through December 1990 are derived from the 1991 Form RW-859 data survey of nuclear utilities. Projected discharges through the end of reactor life are based on DOE estimates of future nuclear capacity, generation, and spent fuel discharges

  15. Spent fuel storage requirements, 1990--2040

    International Nuclear Information System (INIS)

    Walling, R.; Bierschbach, M.

    1990-11-01

    Historical inventories of spent fuel are combined with US Department of Energy (DOE) projections of future discharges from commercial nuclear reactors in the United States to provide estimates of spent fuel storage requirements over the next 51 years, through the year 2040. The needs for storage capacity beyond that presently available in the pools are estimated. These estimates incorporate the maximum capacities within current and planned in-pool storage facilities and any planned transshipments of fuel to other reactors or facilities. Existing and future dry storage facilities are also discussed. Historical data through December 1989 are derived from the 1990 Form RW-859 data survey of nuclear utilities. Projected discharges through the end of reactor life are based on DOE estimates of future nuclear capacity, generation, and spent fuel discharges. 15 refs., 3 figs., 11 tabs

  16. Robotic cleaning of a spent fuel pool

    International Nuclear Information System (INIS)

    Roman, H.T.; Marian, F.A.; Silverman, E.B.; Barkley, V.P.

    1987-01-01

    Spent fuel pools at nuclear power plants are not cleaned routinely, other than by purifying the water that they contain. Yet, debris can collect on the bottom of a pool and should be removed prior to fuel transfer. At Public Service Electric and Gas Company's Hope Creek Nuclear Power Plant, a submersible mobile robot - ARD Corporation's SCAVENGER - was used to clean the bottom of the spent fuel pool prior to initial fuel loading. The robotic device was operated remotely (as opposed to autonomously) with a simple forward/reverse control, and it cleaned 70-80% of the pool bottom. This paper reports that a simple cost-benefit analysis shows that the robotic device would be less expensive, on a per mission basis, than other cleaning alternatives, especially if it were used for other similar cleaning operations throughout the plant

  17. Evolution of spent fuel dry storage

    Energy Technology Data Exchange (ETDEWEB)

    Standring, Paul Nicholas [International Atomic Energy Agency, Vienna (Austria). Div. of Nuclear Fuel Cycle and Waste Technology; Takats, Ferenc [TS ENERCON KFT, Budapest (Hungary)

    2016-11-15

    Around 10,000 tHM of spent fuel is discharged per year from the nuclear power plants in operation. Whilst the bulk of spent fuel is still held in at reactor pools, 24 countries have developed storage facilities; either on the reactor site or away from the reactor site. Of the 146 operational AFR storage facilities about 80 % employ dry storage; the majority being deployed over the last 20 years. This reflects both the development of dry storage technology as well as changes in politics and trading relationships that have affected spent fuel management policies. The paper describes the various approaches to the back-end of the nuclear fuel cycle for power reactor fuels and provides data on deployed storage technologies.

  18. Reracking Possibilities of the NPP Krsko Spent Fuel Pool

    International Nuclear Information System (INIS)

    Bace, M.; Pevec, D.; Smuc, T.

    1998-01-01

    Using the SCALE-4 code package reracking possibilities of the NPP Krsko spent fuel pool were analyzed. Two cases were considered: the first case assuming the 40 years lifetime of the plant, and the second case assuming the 50 years lifetime of the plant. It was shown that it is possible to design the additional racks in free space of the spent fuel pool with the sufficient total capacity to store all the spent fuel generated during the 40 years lifetime of the plant. In the case of 50 years plant lifetime, completely new racks (capacity of 1890 spent fuel assemblies), containing 4mm boral in storage cell walls, were proposed for the NPP Krsko spent fuel pool. The effective multiplication factor of the spent fuel pool fully loaded with new racks containing spent fuel assemblies of initial enrichment 4.3 w/o, burned to 40 GWd/tU and cooled 2 years is lower than the value required by standard. It showed the possibility of the safe disposal of all spent fuel accumulated during more than 50 years lifetime of the plant. (author)

  19. Impacts of reactor. Induced cladding defects on spent fuel storage

    International Nuclear Information System (INIS)

    Johnson, A.B.

    1978-01-01

    Defects arise in the fuel cladding on a small fraction of fuel rods during irradiation in water-cooled power reactors. Defects from mechanical damage in fuel handling and shipping have been almost negligible. No commercial water reactor fuel has yet been observed to develop defects while stored in spent fuel pools. In some pools, defective fuel is placed in closed canisters as it is removed from the reactor. However, hundreds of defective fuel bundles are stored in numerous pools on the same basis as intact fuel. Radioactive species carried into the pool from the reactor coolant must be dealt with by the pool purification system. However, additional radiation releases from the defective fuel during storage appear tu be minimal, with the possible exception of fuel discharged while the reactor is operating (CANDU fuel). Over approximately two decades, defective commercial fuel has been handled, stored, shipped and reprocessed. (author)

  20. WWER spent fuel storage

    Energy Technology Data Exchange (ETDEWEB)

    Bower, C C; Lettington, C [GEC Alsthom Engineering Systems Ltd., Whetstone (United Kingdom)

    1994-12-31

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

  1. WWER spent fuel storage

    International Nuclear Information System (INIS)

    Bower, C.C.; Lettington, C.

    1994-01-01

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

  2. Guidebook on spent fuel storage

    International Nuclear Information System (INIS)

    1984-01-01

    The Guidebook summarizes the experience and information in various areas related to spent fuel storage: technological aspects, the transport of spent fuel, economical, regulatory and institutional aspects, international safeguards, evaluation criteria for the selection of a specific spent fuel storage concept, international cooperation on spent fuel storage. The last part of the Guidebook presents specific problems on the spent fuel storage in the United Kingdom, Sweden, USSR, USA, Federal Republic of Germany and Switzerland

  3. Spent fuel storage requirements 1989--2020

    International Nuclear Information System (INIS)

    1989-10-01

    Historical inventories of spent fuel are combined with Department of Energy (DOE) projections of future discharges from commercial nuclear reactors in the US to provide estimates of spent fuel storage requirements over the next 32 years, through the year 2020. The needs for storage capacity beyond that presently available in the pools are estimated. These estimates incorporate the maximum capacities within current and planned in-pool storage facilities and any planned transshipments of fuel to other reactors or facilities. Historical data through December 1988 are derived from the 1989 Form RW-859 data survey of nuclear utilities. Projected discharges through the end of reactor life are based on DOE estimates of future nuclear capacity, generation, and spent fuel discharges. 14 refs., 3 figs., 28 tabs

  4. Spent fuel and fuel pool component integrity. Annual report, FY 1979

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.; Bailey, W.J.; Schreiber, R.E.; Kustas, F.M.

    1980-05-01

    International meetings under the BEFAST program and under INFCE Working Group No. 6 during 1978 and 1979 continue to indicate that no cases of fuel cladding degradation have developed on pool-stored fuel from water reactors. A section from a spent fuel rack stand, exposed for 1.5 y in the Yankee Rowe (PWR) pool had 0.001- to 0.003-in.-deep (25- to 75-μm) intergranular corrosion in weld heat-affected zones but no evidence of stress corrosion cracking. A section of a 304 stainless steel spent fuel storage rack exposed 6.67 y in the Point Beach reactor (PWR) spent fuel pool showed no significant corrosion. A section of 304 stainless steel 8-in.-dia pipe from the Three Mile Island No. 1 (PWR) spent fuel pool heat exchanger plumbing developed a through-wall crack. The crack was intergranular, initiating from the inside surface in a weld heat-affected zone. The zone where the crack occurred was severely sensitized during field welding. The Kraftwerk Union (Erlangen, GFR) disassembled a stainless-steel fuel-handling machine that operated for 12 y in a PWR (boric acid) spent fuel pool. There was no evidence of deterioration, and the fuel-handling machine was reassembled for further use. A spent fuel pool at a Swedish PWR was decontaminated. The procedure is outlined in this report

  5. Horizontal modular dry irradiated fuel storage system

    Science.gov (United States)

    Fischer, Larry E.; McInnes, Ian D.; Massey, John V.

    1988-01-01

    A horizontal, modular, dry, irradiated fuel storage system (10) includes a thin-walled canister (12) for containing irradiated fuel assemblies (20), which canister (12) can be positioned in a transfer cask (14) and transported in a horizontal manner from a fuel storage pool (18), to an intermediate-term storage facility. The storage system (10) includes a plurality of dry storage modules (26) which accept the canister (12) from the transfer cask (14) and provide for appropriate shielding about the canister (12). Each module (26) also provides for air cooling of the canister (12) to remove the decay heat of the irradiated fuel assemblies (20). The modules (26) can be interlocked so that each module (26) gains additional shielding from the next adjacent module (26). Hydraulic rams (30) are provided for inserting and removing the canisters (12) from the modules (26).

  6. Baseline descriptions for LWR spent fuel storage, handling, and transportation

    International Nuclear Information System (INIS)

    Moyer, J.W.; Sonnier, C.S.

    1978-04-01

    Baseline descriptions for the storage, handling, and transportation of reactor spent fuel are provided. The storage modes described include light water reactor (LWR) pools, away-from-reactor basins, dry surface storage, reprocessing-facility interim storage pools, and deep geologic storage. Land and water transportation are also discussed. This work was sponsored by the Department of Energy/Office of Safeguards and Security as part of the Sandia Laboratories Fixed Facility Physical Protection Program. 45 figs, 4 tables

  7. Baseline descriptions for LWR spent fuel storage, handling, and transportation

    Energy Technology Data Exchange (ETDEWEB)

    Moyer, J.W.; Sonnier, C.S.

    1978-04-01

    Baseline descriptions for the storage, handling, and transportation of reactor spent fuel are provided. The storage modes described include light water reactor (LWR) pools, away-from-reactor basins, dry surface storage, reprocessing-facility interim storage pools, and deep geologic storage. Land and water transportation are also discussed. This work was sponsored by the Department of Energy/Office of Safeguards and Security as part of the Sandia Laboratories Fixed Facility Physical Protection Program. 45 figs, 4 tables.

  8. Fuel storage rack

    International Nuclear Information System (INIS)

    Mollon, L.

    1977-01-01

    Disclosed is a storage rack for spent nuclear fuel elements comprising a multiplicity of elongated hollow containers of uniform cross-section, preferably square,some of said containers having laterally extending continuous flanges extending between adjacent containers and defining continuous elongated chambers therebetween for the reception of neutron absorbing panels. 18 claims, 7 figures

  9. Fuel performance in water storage

    International Nuclear Information System (INIS)

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

    1993-11-01

    Westinghouse Idaho Nuclear Company operates the Idaho Chemical Processing Plant (ICPP) at the Idaho National Engineering Laboratory (INEL) for the Department of Energy (DOE). A variety of different types of fuels have been stored there since the 1950's prior to reprocessing for uranium recovery. In April of 1992, the DOE decided to end fuel reprocessing, changing the mission at ICPP. Fuel integrity in storage is now viewed as long term until final disposition is defined and implemented. Thus, the condition of fuel and storage equipment is being closely monitored and evaluated to ensure continued safe storage. There are four main areas of fuel storage at ICPP: an original underwater storage facility (CPP-603), a modern underwater storage facility (CPP-666), and two dry fuel storage facilities. The fuels in storage are from the US Navy, DOE (and its predecessors the Energy Research and Development Administration and the Atomic Energy Commission), and other research programs. Fuel matrices include uranium oxide, hydride, carbide, metal, and alloy fuels. In the underwater storage basins, fuels are clad with stainless steel, zirconium, and aluminum. Also included in the basin inventory is canned scrap material. The dry fuel storage contains primarily graphite and aluminum type fuels. A total of 55 different fuel types are currently stored at the Idaho Chemical Processing Plant. The corrosion resistance of the barrier material is of primary concern in evaluating the integrity of the fuel in long term water storage. The barrier material is either the fuel cladding (if not canned) or the can material

  10. 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)

  11. Spent fuel storage facility, Kalpakkam

    International Nuclear Information System (INIS)

    Shreekumar, B.; Anthony, S.

    2017-01-01

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

  12. Spent fuel storage and isolation

    International Nuclear Information System (INIS)

    Bensky, M.S.; Kurzeka, W.J.; Bauer, A.A.; Carr, J.A.; Matthews, S.C.

    1979-02-01

    The principal spent fuel activities conducted within the commercial waste and spent fuel within the Commercial Waste and Spent Fuel Packaging Program are: simulated near-surface (drywell) storage demonstrations at Hanford and the Nevada Test Site; surface (sealed storage cask) and drywell demonstrations at the Nevada Test Site; and spent fuel receiving and packaging facility conceptual design. These investigations are described

  13. Spent fuel storage options: a critical appraisal

    International Nuclear Information System (INIS)

    Singh, K.P.; Bale, M.G.

    1990-01-01

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

  14. Economics of spent LWR fuel storage

    International Nuclear Information System (INIS)

    Clark, H.J.

    1980-01-01

    A low cost option for spent fuel inventories would be to ship excess fuel from the overburdened reactor to another reactor in the utility's system that has available space. The only cost would be for cask leasing and shipping. Three other alternatives all require considerable capital expenditures: reracking, new at-reactor (AR) storage facilities, and away-from-reactor (AFR) storage facilities. Fuel storage requirements will be met best by transfer of fuel or by re-racking existing reactor basins whenever these options are available. These alternatives represent not only the lowest cost storage options but also the most timely. Fuel can be shipped to other storage pools for about $10/kg depending on the distance, while costs for reracking range from $18 to 25/kg depending on the approach. These alternatives are recognized to face environmental and regulatory obstacles. However, such obstacles should be less severe than similar issues that would be encountered with AR or AFR basin storage. When storage requirements cannot be met by the first two options, the next least costly alternative for most utilities will be use of a Federal AFR. Storage cost of about $137/kg at an AFR are less costly than charges of up to $350/kg that could be incurred by the use of AR basins. AR basins are practical only when a utility requires storage capacity to accommodate annual additions of 100 MT or more of spent fuel. The large reactor complexes discharging this much feul are not currently those that require relief from fuel storage problems. A recent development in Germany may offer an AR alternative of dry storage in transportation/storage casks at a cost of $200/kg; however, this method has not yet been accepted and licensed for use in the US

  15. Spent fuel storage - dry storage options and issues

    International Nuclear Information System (INIS)

    Akins, M.J.

    2007-01-01

    The increase in the number of nuclear energy power generation facilities will require the ability to store the spent nuclear fuel for a long period until the host countries develop reprocessing or disposal options. Plants have storage pools which are closely associated with the operating units. These are excellent for short term storage, but require active maintenance and operations support which are not desirable for the long term. Over the past 25 years, dry storage options have been developed and implemented throughout the world. In recent years, protection against terrorist attack has become an increasing source of design objectives for these facilities, as well as the main nuclear plant. This paper explores the current design options of dry storage cask systems and examines some of the current design issues for above ground , in-ground, or below-ground storage of spent fuel in dry casks. (author)

  16. Report by the committee assessing fuel storage

    International Nuclear Information System (INIS)

    Morgan, W.W.

    1977-11-01

    Various concepts for interim storage of spent nuclear fuel have been considered. Preliminary design studies and cost estimates have been prepared for the following concepts: two with water cooling - prolonged pool storage at a generating station and pool storage at a central site - , three with air cooling at a central site - ''canister'', ''convection vault'', and ''conduction vault'' - and one underground storage scheme in rock salt. Costs (1972 dollars) were estimated including transportation and a perpetual care fund for maintenance and periodical renewal of the storage facility. Part 2 provides details of the concepts and costing methods. All concepts gave moderate costs providing a contribution of about 0.1 m$/kWh to the total unit energy cost. Advantages and disadvantages of the respective schemes are compared. (author)

  17. ACRR fuel storage racks criticality safety analysis

    International Nuclear Information System (INIS)

    Bodette, D.E.; Naegeli, R.E.

    1997-10-01

    This document presents the criticality safety analysis for a new fuel storage rack to support modification of the Annular Core Research Reactor for production of molybdenum-99 at Sandia National Laboratories, Technical Area V facilities. Criticality calculations with the MCNP code investigated various contingencies for the criticality control parameters. Important contingencies included mix of fuel element types stored, water density due to air bubbles or water level for the over-moderated racks, interaction with existing fuel storage racks and fuel storage holsters in the fuel storage pool, neutron absorption of planned rack design and materials, and criticality changes due to manufacturing tolerances or damage. Some limitations or restrictions on use of the new fuel storage rack for storage operations were developed through the criticality analysis and are required to meet the double contingency requirements of criticality safety. As shown in the analysis, this system will remain subcritical under all credible upset conditions. Administrative controls are necessary for loading, moving, and handling the storage rack as well as for control of operations around it. 21 refs., 16 figs., 4 tabs

  18. Expanded spent fuel storage project at Yankee Atomic Electric Plant

    International Nuclear Information System (INIS)

    Chin, S.L.

    1980-01-01

    A detailed discussion on the project at the Yankee Rowe power reactor for expanding the capacity of the at-reactor storage pool by building double-tier storage racks. Various alternatives for providing additional capacity were examined by the operators. Away-from-reactor alternatives included shipment to existing privately owned facilities, a regional independent storage facility, and transshipments to other New England nuclear power plant pools. At-reactor alternatives evaluated included a new pool modification of the existing structure and finally, modification of the spent fuel pit. The establishment of a federal policy precluding transshipment of spent fuel prohibited the use of off-site alternatives. The addition of another pool was too expensive. The possibility of modifying an existing on-site structure required a new safety evaluation by the regulatory group with significant cost and time delays. Therefore, the final alternative - utilizing the existing spent fuel pool with some modification - was chosen due to cost, licensing possibility, no transport requirements, and the fact that the factors involved were mainly under the control of the operator. Modification of the pool was accomplished in phases. In the first phase, a dam was installed in the center of the pool (after the spent fuel was moved to one end). In the second phase, the empty end of the pool was drained and lined with stainless steel and the double-tier rack supports were added. In the third phase, the pool was refilled and the dam was removed. Then the spent fuel was moved into the completed end. In the fourth phase, the dam was replaced and the empty part of the pool was drained. The liner and double-tier rack supports were installed, the pool was refilled, and the dam was removed.The project demonstrated that the modification of existing spent fuel fuel pools for handling double-tier fuel racks is a viable solution for increasing the storage capacity at the reactor

  19. Spent fuel interim storage

    International Nuclear Information System (INIS)

    Bilegan, Iosif C.

    2003-01-01

    The official inauguration of the spent fuel interim storage took place on Monday July 28, 2003 at Cernavoda NNP. The inaugural event was attended by local and central public authority representatives, a Canadian Government delegation as well as newsmen from local and central mass media and numerous specialists from Cernavoda NPP compound. Mr Andrei Grigorescu, State Secretary with the Economy and Commerce Ministry, underlined in his talk the importance of this objective for the continuous development of nuclear power in Romania as well as for Romania's complying with the EU practice in this field. Also the excellent collaboration between the Canadian contractor AECL and the Romanian partners Nuclear Montaj, CITON, UTI, General Concret in the accomplishment of this unit at the planned terms and costs. On behalf of Canadian delegation, spoke Minister Don Boudria. He underlined the importance which the Canadian Government affords to the cooperation with Romania aiming at specific objectives in the field of nuclear power such as the Cernavoda NPP Unit 2 and spent fuel interim storage. After traditional cutting of the inaugural ribbon by the two Ministers the festivities continued on the Cernavoda NPP Compound with undersigning the documents regarding the project completion and a press conference

  20. 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.

  1. Problems of the Spent Nuclear Fuel Storage

    International Nuclear Information System (INIS)

    Negrivoda, G.

    1997-01-01

    Approximately 99% of the radioactivity in waste, produced in the process of operating a nuclear power plant, is contained in spent nuclear fuel. Safe handling and storage of the spent nuclear fuel is an important factor of a nuclear plant safety. Today at Ignalina NPP the spent fuel is stored in special water pools, located in the same buildings as the reactors. The volume of the pools is limited, for unit one the pool will be fully loaded in 1998, for unit 2 - in 2000. The further operation of the plant will only be possible if new storage is constructed. In 1994 contract with German company GNB was signed for the supply of 20 containers of the CASTOR type. Containers were delivered in accordance with agreed schedule. In the end of 1995 a new tender for new storage options was announced in order to minimize the storage costs. A proposal from Canadian company AECL now is being considered as one of the most suitable and negotiations to sign the contract started. (author)

  2. Advantages of dry hardened cask storage over wet storage for spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Romanato, Luiz Sergio, E-mail: romanato@ctmsp.mar.mil.b [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil). Dept. da Qualidade

    2011-07-01

    Pools are generally used to store and maintain spent nuclear fuel assemblies for cooling, after removed from reactors. After three to five years stored in the pools, spent fuel can be reprocessed or sent to a final disposition in a geological repository and handled as radioactive waste or sent to another site waiting for future solution. Spent fuel can be stored in dry or wet installations, depending on the method adopted by the nuclear plant. If this storage were exclusively wet, at the installation decommissioning in the future, another solution for storage will need to be found. Today, after a preliminary cooling, the spent fuel assemblies can be removed from the pool and sent to dry hardened storage installations. This kind of storage does not need complex radiation monitoring and it is safer than wet storage. Brazil has two nuclear reactors in operation, a third reactor is under construction and they use wet spent fuel storage . Dry hardened casks use metal or both metal and concrete for radiation shielding and they are safe, especially during an earthquake. An earthquake struck Japan on March 11, 2011 damaging Fukushima Daiichi nuclear power plant. The occurrence of earthquakes in Brazil is very small but dry casks can resist to other events, including terrorist acts, better than pools. This paper shows the advantages of dry hardened cask storage in comparison with the wet storage (water pools) for spent nuclear fuel. (author)

  3. Advantages of dry hardened cask storage over wet storage for spent nuclear fuel

    International Nuclear Information System (INIS)

    Romanato, Luiz Sergio

    2011-01-01

    Pools are generally used to store and maintain spent nuclear fuel assemblies for cooling, after removed from reactors. After three to five years stored in the pools, spent fuel can be reprocessed or sent to a final disposition in a geological repository and handled as radioactive waste or sent to another site waiting for future solution. Spent fuel can be stored in dry or wet installations, depending on the method adopted by the nuclear plant. If this storage were exclusively wet, at the installation decommissioning in the future, another solution for storage will need to be found. Today, after a preliminary cooling, the spent fuel assemblies can be removed from the pool and sent to dry hardened storage installations. This kind of storage does not need complex radiation monitoring and it is safer than wet storage. Brazil has two nuclear reactors in operation, a third reactor is under construction and they use wet spent fuel storage . Dry hardened casks use metal or both metal and concrete for radiation shielding and they are safe, especially during an earthquake. An earthquake struck Japan on March 11, 2011 damaging Fukushima Daiichi nuclear power plant. The occurrence of earthquakes in Brazil is very small but dry casks can resist to other events, including terrorist acts, better than pools. This paper shows the advantages of dry hardened cask storage in comparison with the wet storage (water pools) for spent nuclear fuel. (author)

  4. Spent fuel heatup following loss of water during storage

    International Nuclear Information System (INIS)

    Benjamin, A.S.; McCloskey, D.J.; Powers, D.A.; Dupree, S.A.

    1979-03-01

    An analysis of spent fuel heatup following a hypothetical accident involving drainage of the storage pool is presented. Computations based upon a new computer code called SFUEL have been performed to assess the effect of decay time, fuel element design, storage rack design, packing density, room ventilation, drainage level, and other variables on the heatup characteristics of the spent fuel and to predict the conditions under which clad failure will occur. Possible storage pool design modifications and/or onsite emergency action have also been considered

  5. Materials in the environment of the fuel in dry storage

    Energy Technology Data Exchange (ETDEWEB)

    Issard, H [TN International (Cogema Logistics) (France)

    2012-07-01

    Spent nuclear fuel has been stored safely in pools or dry systems in over 30 countries. The majority of IAEA Member States have not yet decided upon the ultimate disposition of their spent nuclear fuel: reprocessing or direct disposal. Interim storage is the current solution for these countries. For developing the technological knowledge data base, a continuation of the IAEA's spent fuel storage performance assessment was achieved. The objectives are: Investigate the dry storage systems and gather basic fuel behaviour assessment; Gather data on dry storage environment and cask materials; Evaluate long term behaviour of cask materials.

  6. Used fuel extended storage security and safeguards by design roadmap

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, Samuel G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lindgren, Eric Richard [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Jones, Robert [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Ketusky, Edward [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); England, Jeffrey [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Scherer, Carolynn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Sprinkle, James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Miller, Michael. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rauch, Eric [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Scaglione, John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Dunn, T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-05-01

    In the United States, spent nuclear fuel (SNF) is safely and securely stored in spent fuel pools and dry storage casks. The available capacity in spent fuel pools across the nuclear fleet has nearly reached a steady state value. The excess SNF continues to be loaded in dry storage casks. Fuel is expected to remain in dry storage for periods beyond the initial dry cask certification period of 20 years. Recent licensing renewals have approved an additional 40 years. This report identifies the current requirements and evaluation techniques associated with the safeguards and security of SNF dry cask storage. A set of knowledge gaps is identified in the current approaches. Finally, this roadmap identifies known knowledge gaps and provides a research path to deliver the tools and models needed to close the gaps and allow the optimization of the security and safeguards approaches for an interim spent fuel facility over the lifetime of the storage site.

  7. Spent fuel behaviour during dry storage - a review

    International Nuclear Information System (INIS)

    Shivakumar, V.; Anantharaman, K.

    1997-09-01

    One of the strategies employed for management of spent fuel prior to their final disposal/reprocessing is their dry storage in casks, after they have been sufficiently cooled in spent fuel pools. In this interim storage, one of the main consideration is that the fuel should retain its integrity to ensure (a) radiological health hazard remains minimal and (b) the fuel is retrievable for down steam fuel management processes such as geological disposal or reprocessing. For dry storage of spent fuel in air, oxidation of the exposed UO 2 is the most severe of phenomena affecting the integrity of fuel. This is kept within acceptable limits for desired storage time by limiting the fuel temperature in the storage cask. The limit on the fuel temperature is met by having suitable limits on maximum burn-up of fuel, minimum cooling period in storage pool and optimum arrangement of fuel bundles in the storage cask from heat removal considerations. The oxidation of UO 2 by moist air has more deleterious effects on the integrity of fuel than that by dry air. The removal of moisture from the storage cask is therefore a very important aspect in dry storage practice. The kinetics of the oxidation phenomena at temperatures expected during dry storage in air is very slow and therefore the majority of the existing data is based on extrapolation of data obtained at higher fuel temperatures. This and the complex effects of factors like fission products in fuel, radiolysis of storage medium etc. has necessitated in having a conservative limiting criteria. The data generated by various experimental programmes and results from the on going programmes have shown that dry storage is a safe and economical practice. (author)

  8. On-site concrete cask storage system for spent nuclear fuel

    International Nuclear Information System (INIS)

    Craig, P.A.; Haelsig, R.T.; Kent, J.D.; Schmoker, D.S.

    1989-01-01

    A method is described of storing spent nuclear fuel assemblies including the steps of: transferring the fuel assemblies from a spent-fuel pool to a moveable concrete storage cask located outside the spent-fuel pool; maintaining a barrier between the fuel and the concrete in the cask to prevent contamination of the concrete by the fuel; maintaining the concrete storage cask containing the spent-fuel on site at the reactor complex for some predetermined period; transferring the fuel assemblies from the concrete storage cask to a shipping container; and, recycling the concrete storage cask

  9. Study on increasing spent fuel storage capacity at Juragua NPP

    International Nuclear Information System (INIS)

    Guerra Valdes, R.; Lopez Aldama, D.; Rodriguez Gual, M.; Garcia Yip, F.

    1999-01-01

    The delay in decision about the final disposal of the spent fuel, led to longer interim storage. The reracking og the storage pools was an economical and feasible option to increase the storage capacity on the site. Reracking of the storage facility led to the analysis of the new conditions for criticality, shielding, residual heat removal and mechanical loads over the structures. This paper includes a summary of the studies on criticality and dose rate changes in the vicinity of the storage pool of Juragua NPP

  10. Evaluation of economics of spent fuel storage techniques

    International Nuclear Information System (INIS)

    Yamaji, Kenji; Nagano, Koji

    1988-01-01

    Various spent fuel storage techniques are evaluated in terms of required costs. The unit storage cost for each spent fuel storage scenario is calculated based on the total cost required for the scenario including capital expenditure, operation cost, maintenance cost and transport cost. Intermediate storage may be performed in relatively small facilities in the plant or in independent large-scale facilities installed away from the plant. Dry casks or water pools are assumed to be used in in-plant storage facilities while vaults may also be employed in independent facilities. Evaluation is made for these different cases. In in-plant facilities, dry cask storage is found to be more economical in all cases than water pool storage, especially when large-sized casks are employed. In independent facilities, on the other hand, the use of vaults is the most desirable because the required capital expenditure is the lowest due to the effect of scale economics. Dry cask storage is less expensive than water pool storage also in independent facilities. The annual discount rate has relatively small influence on the unit cost for storage. An estimated unit cost for storage in independent storage facilities is shown separately for facilities with a capacity of 1,000 tons, 3,000 tons or 5,000 tons. The report also outlines the economics of spent fuel storage in overseas facilities (Finland, Sweden and U.S.A.). (Nogami, K.)

  11. Spent-fuel pool thermal hydraulics: The evaporation question

    International Nuclear Information System (INIS)

    Yilmaz, T.P.; Lai, J.C.

    1996-01-01

    Many nuclear power plants are currently using dense fuel arrangements that increase the number of spent fuel elements stored in their spent-fuel pools (SFPs). The denser spent-fuel storage results in higher water temperatures, especially when certain event scenarios are analyzed. In some of these event scenarios, it is conservative to maximize the evaporation rate, while in other circumstances it is required to minimize the evaporation rates for conservatism. Evaporation is such a fundamental phenomenon that many branches of engineering developed various equations based on theory and experiments. The evaporation rates predicted by existing equations present a wide range of variation, especially at water temperatures >40 degrees C. Furthermore, a study on which equations provide the highest and lowest evaporation rates has not been done until now. This study explores the sensitivity of existing evaporation equations to various parameters and recommends the limiting evaporation equations for use in the solution of SFP thermal problems. Note that the results of this study may be applicable to a much wider range of applications from irrigation ponds, cooling lakes, and liquid-waste management to calculating adequate air exchange rate for swimming pools and health spas

  12. HTGR spent fuel storage study

    International Nuclear Information System (INIS)

    Burgoyne, R.M.; Holder, N.D.

    1979-04-01

    This report documents a study of alternate methods of storing high-temperature gas-cooled reactor (HTGR) spent fuel. General requirements and design considerations are defined for a storage facility integral to a fuel recycle plant. Requirements for stand-alone storage are briefly considered. Three alternate water-cooled storage conceptual designs (plug well, portable well, and monolith) are considered and compared to a previous air-cooled design. A concept using portable storage wells in racks appears to be the most favorable, subject to seismic analysis and economic evaluation verification

  13. Dry storage of Magnox fuel

    International Nuclear Information System (INIS)

    1986-09-01

    This work, commissioned by the CEGB, studies the feasibility of a combination of short-term pond storage and long-term dry storage of Magnox spent fuel as a cheaper alternative to reprocessing. Storage would be either at the reactor site or a central site. Two designs are considered, based on existing design work done by GEC-ESL and NNC; the capsule design developed by NNC and with storage in passive vaults for up to 100 yrs and the GEC-ESL tube design developed at Wylfa for the interim storage of LWR. For the long-term storage of Magnox spent fuel the GEC-ESL tubed vault all-dry storage method is recommended and specifications for this method are given. (U.K.)

  14. CHARACTERISTICS OFBENTHIC FISH COMMUNITY OF DNIEPER STORAGE POOL LITTORAL ZONE

    Directory of Open Access Journals (Sweden)

    Novitskiy R. A.

    2011-11-01

    Full Text Available Specific composition of bull-calves is analyzed together with their quantitative and quality parameters in the littoral zone of the Dnepr storage pool. The structural-functional features of organization of littoral communities of bull-calves were studied. The patterns of spatial distribution of Bull-calf (Gobiidae representatives were analyzed for the storage pool; their role in the littoral fish communities was clarified.

  15. Projection of US LWR spent fuel storage requirements

    International Nuclear Information System (INIS)

    Fletcher, J.F.; Cole, B.M.; Purcell, W.L.; Rau, R.G.

    1982-11-01

    The spent fuel storage requirements projection is based on data supplied for each operating or planned nuclear power power plant by the operting utilities. The data supplied by the utilities encompassed details of plant operating history, past records of fuel discharges, current inventories in reactor spent fuel storage pools, and projections of future discharge patterns. Data on storage capacity of storage pools and on characterization of the discharged fuel are also included. The data supplied by the utilities, plus additional data from other appropriate sources, are maintained on a computerized data base by Pacific Northwest Laboratory. The spent fuel requirements projection was based on utility data updated and verified as of December 31, 1981

  16. Canadian experience with wet and dry fuel storage concepts

    International Nuclear Information System (INIS)

    Mayman, S.A.

    1978-07-01

    Canada has been storing fuel in water-filled pools for 30 years. There have been no significant problems, but until recently little effort has been invested in quantitative assessment of fuel performance under storage conditions. Work is now in progress to provide such information. Storage pools at nuclear generating stations have operated satisfactorily. The Canadian nuclear industry has nevertheless been studying methods for reducing storage costs and/or increasing reliability. Various concepts, using both water and air cooling, have been suggested. One such concept - the air-cooled concrete canister - is presently under test at the Whiteshell Nuclear Research Establishment. (author)

  17. Utilization of the NFS West Valley Installation for spent fuel storage

    International Nuclear Information System (INIS)

    MacDonald, R.W.

    1978-04-01

    Several thousand MT of capacity of AFR storage will be required in the 1980's. The pool at NFS has capacity for an additional 60 MT of BWR fuel or 150 MT of PWR assemblies. Zircaloy-clad LWR fuel can be stored in pools for up to 100 years. Environmental effects are discussed. Expansion of the pool capacity for as much as 1000 MT more, either by using more compact storage racks or constructing a new pool or an independent pool, is considered. Some indication of the environmental impacts of expanded fuel storage capacity at West Valley is offered by experience at Barnwell

  18. Cask operation and maintenance for spent fuel storage

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J.S. [International Atomic Energy Agency, Vienna (Austria)

    2004-07-01

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

  19. Cask operation and maintenance for spent fuel storage

    International Nuclear Information System (INIS)

    Lee, J.S.

    2004-01-01

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

  20. Spent fuel pool thermal-hydraulic analysis using RELAP5-3D

    Energy Technology Data Exchange (ETDEWEB)

    Ramos, M. C.; Fernandes, G.H.N.; Costa, A.L.; Pereira, F.; Pereira, C., E-mail: marc5663@gmail.com, E-mail: ghnfernandes@pq.cnpq.br, E-mail: claubia@nuclear.ufmg.br, E-mail: antonella@nuclear.ufmg.br [Universidade Federal de Minas Gerais, Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-07-01

    In order to analyze the thermo-hydraulic behavior of spent fuel pools, and taking as reference a hypothetic PWR nuclear plant, a model of RELAP-3D for a spent fuel pool has been built. This model has been used to simulate a loss of coolant in SPF. This study focuses on the loss of coolant flow accident in spent fuel storage pool which is modelled by using RELAP5-3D code to observe the coolant level reduction and fuel uncovery because of decay heat generation of the spent fuel in the pool. The results have been compared with the available data. The developed model demonstrated that the RELAP5-3D is capable of reproduce the thermal behavior of SPF in a transient scenario. (author)

  1. Studies and research concerning BNFP: spent fuel dry storage studies at the Barnwell Nuclear Fuel Plant

    International Nuclear Information System (INIS)

    Anderson, K.J.

    1980-09-01

    Conceptual designs are presented utilizing the Barnwell Nuclear Fuel Plant for the dry interim storage of spent light water reactor fuel. Studies were conducted to determine feasible approaches to storing spent fuel by methods other than wet pool storage. Fuel that has had an opportunity to cool for several years, or more, after discharge from a reactor is especially adaptable to dry storage since its thermal load is greatly reduced compared to the thermal load immediately following discharge. A thermal analysis was performed to help in determining the feasibility of various spent fuel dry storage concepts. Methods to reject the heat from dry storage are briefly discussed, which include both active and passive cooling systems. The storage modes reviewed include above and below ground caisson-type storage facilities and numerous variations of vault, or hot cell-type, storage facilities

  2. Behaviour of Spent WWER fuel under long term storage conditions

    Energy Technology Data Exchange (ETDEWEB)

    Kadarmetov, I M [A.A.Bochvar All-Russia Research Institute of Inorganic Materials, Moscow (Russian Federation)

    1999-07-02

    Results of experimental investigation into thermomechanical properties of pre-irradiated Zr-1%Nb alloy over a range temperatures 500-570 grad C are presented. Safety examination of the Ventilation Storage Casks dry storage system has been carried out. Preliminary safety criteria under dry storage conditions in an environment of inert gas are follows: maximum cladding temperature under normal conditions of dry storage should not exceed 330 grad C after 5-year cooling in water-filled pools; maximum allowable temperature of spent fuel rod cladding under operational mode with infringement of heat removal should not exceed 440 grad C over 8 hours. As each SFA dry storage project comprises its individual technology of spent fuel management, it is necessary to evaluate allowable parameters (terms of storage, maximum temperatures of fuel) for each project respectively. The programme of experimental investigations for the justification of safety criteria for WWER-1000 dry spent fuel storage systems is underway. (author)

  3. Storage arrangements for nuclear fuel

    International Nuclear Information System (INIS)

    Ealing, C.J.

    1985-01-01

    A storage arrangement for nuclear fuel has a plurality of storage tubes connected by individual pipes to manifolds which are connected, in turn, to an exhaust system for maintaining the tubes at sub-atmospheric pressure, and means for producing a flow of a cooling fluid, such as air, over the exterior surfaces of the tubes. (author)

  4. Preliminary Calculation on a Spent Fuel Pool Accident using GOTHIC

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jaehwan; Choi, Yu Jung; Hong, Tae Hyub; Kim, Hyeong-Taek [KHNP-CRI, Daejeon (Korea, Republic of)

    2015-10-15

    The probability of an accident happening at the spent fuel pool was believed to be quite low until the 2011 Fukushima accident occurred. Notably, large amount of spent fuel are normally stored in the spent fuel pool for a long time compared to the amount of fuel in the reactor core and the total heat released from the spent fuel is high enough to boil the water of the spent fuel pool when the cooling system does not operate. In addition, the enrichment and the burnup of the fuel have both increased in the past decade and heat generation from the spent fuel thereby has also increased. The failure of the cooling system at the spent fuel pool (hereafter, a loss-of-cooling accident) is one of the principal hypothetical causes of an accident that could occur at the spent fuel pool. In this paper, the preliminary calculation of a loss-of-cooling accident was performed. In this paper, the preliminary calculation of a loss-of cooling accident was performed with GOTHIC. The calculation results show boiling away of water in the spent fuel pool due to the loss-of-cooling accident and similar thermal performance of the spent fuel pool with previous research results.

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

    International Nuclear Information System (INIS)

    1990-06-01

    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

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

    International Nuclear Information System (INIS)

    Khattab, M.; Mariy, Ahmed

    1999-01-01

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

  7. Rethinking the economics of centralized spent fuel storage

    International Nuclear Information System (INIS)

    Wood, T.W.; Short, S.M.; Dippold, D.G.; Rod, S.R.; Williams, J.W.

    1991-01-01

    The technology for extended storage of spent nuclear fuel (SNF), either at-reactor or in a centralized facility such as a monitored retrievable storage (MRS) facility, is well-developed and proven from an engineering and safety perspective. The question of whether spent fuel should await its final geologic disposal while at a reactor site or in an MRS facility is essentially an economic one. While intuition and previous results suggest that centralized storage will be more economical than at-reactor storage beyond some break-even quantity of SNF, the incremental costs of pool storage at-reactor are close to zero as long as pool capacity is generally available. Thus, if economics is the prime motivator, the quantity of spent fuel required to warrant centralized storage could be quite large. The economics of centralizing the storage of spent fuel at a single site, as opposed to continued storage at over 100 reactor sites, has been the subject of several recent analyses. Most of these analyses involved calculating the benefits of an MRS facility (in terms of avoided utility costs) with a pre-defined MRS operating scenario (e.g., spent fuel acceptance schedule, storage capacity, and typical storage cycle). While these analyses provided some insight into the economic justification for an MRS facility, even the most favorable scenarios resulted in net costs of hundreds of millions of dollars when evaluated on a discounted cash flow basis

  8. Special storage of leaking fuel at Paks NPP

    International Nuclear Information System (INIS)

    Biro, Janos; Szőke, L.; Burján, T.; Lukács, R.; Hózer, H.

    2015-01-01

    In this paper the activities related with spent, hermetic as well as leaking fuel handling and storage, including: Spent fuel pool; Transportation criteria for the spent fuel assemblies and Interim spent fuel dry storage; Short-term storage in the spent fuel pool; Identification of the leaking assemblies by the TS-device; Present conception of Identification, handling of the leaking FAs; Modified transport procedure for the leaking FAs; Calculation of solved activity inside the leaking fuel rod; Solved activity limit values for the leaking FAs; Long-term storage in the interim spent Fuel dry storage are presented. At the end authors’ concluded that: 1) The leaking FA can be transported to the interim dry storage together with the other spent fuel assemblies in the transport container. 2) The transport-documentation of the leaking FA has to contain: isotope inventory, calculated solved activity values of the failed FA and the quantity of failed fuel rods. 3) Performing three leakage tests of the identified leaking FA before the transportation in the 5FP. it is useful to decrease the solved activity concentration inside the leaking FA and give additional information about the extent of the leakage. 4) We can calculate simply the solved activity of the leaking FA. 5) The modified transport procedure will have to be authorized. 6) The radiological effects of the leaking FA are negligible relative to the natural background radiation

  9. Seismic analysis and design of spent subassembly storage bay (SSSB) pool

    International Nuclear Information System (INIS)

    Abdul Gani, H.I.; Ramanjaneyulu, K.V.S.; Pillai, C.S.; Chetal, S.C.

    2003-01-01

    Fuel bundles, after their specified stay in reactor core, are replaced by fresh fuel for sustaining power generation at rated levels. The irradiated fuel subassembly, removed fresh from core, known as spent fuel sub assembly, is radioactive and decay heat generating. It needs to be cooled before it becomes amenable for handling, either for reprocessing or for immobilisation. For this purpose, it is immersed in a pool of water, retained in a concrete structure referred as Spent Subassembly Storage Bay (SSSB) pool. The height of water column above fuel bundles is arrived from shielding considerations. SSSB pool is one of the nuclear safety related structures and warrants rigorous analysis and design. The SSSB pool, in case of PFBR 500 MW(e) is located in fuel building. It is a stainless steel lined. water retaining rectangular R.C.C. open tank of size 7.5 X 29.0 m, with a height of 11.0 m. This structure is analysed for two levels of site specific earthquakes taking in to account liquid structure interactions as per ASCE-4, 1998. The design of walls and bottom slab is carried out satisfying the AERB code for nuclear safety related structures. Analysis and design of SSSB pool of PFBR is presented in the following paper. (author)

  10. Rethinking the economics of centralized spent fuel storage

    International Nuclear Information System (INIS)

    Wood, T.W.; Short, S.M.; Dippold, D.G.; Rod, S.R.; Williams, J.W.

    1991-04-01

    The technology for extended storage of spent nuclear fuel (SNF), either at-reactor or in a centralized facility such as a monitored retrievable storage (MRS) facility, is well-developed and proven from an engineering and safety perspective. The question of whether spent fuel should await its final geologic disposal while at a reactor site or in an MRS facility is essentially an economic one. While intuition and previous results suggest that centralized storage will be more economical than at-reactor storage beyond some break-even quantity of SNF, the incremental costs of pool storage at-reactor are close to zero as long as pool capacity is generally available. Thus, if economics is the prime motivator, the quantity of spent fuel required to warrant centralized storage could be quite large. The economics of centralizing the storage of spent fuel at a single site, as opposed to continued storage at over 100 reactor sites, has been the subject of several recent analyses. Most of these analyses involved calculating the benefits of an MRS facility with a pre-defined MRS operating scenario. This paper reverses this approach to economic analysis of the MRS by seeking the optimal MRS operating scenario (in terms of the parameters listed above) implied by the economic incentives arising from the relative costs of at-reactor storage and centralized storage. This approach treats an MRS as a possible storage location that will be used according to its economic value in system operation. 5 refs., 5 figs

  11. Stress corrosion (Astm G30-90 standard) in 08x18H10T stainless steel of nuclear fuel storage pool in WWER reactors; Corrosion bajo esfuerzo (Norma ASTM G30-90) en acero inoxidable 08x18H10T de piscinas de almacenamiento de combustible nuclear en reactores V.V.E.R

    Energy Technology Data Exchange (ETDEWEB)

    Herrera, V.; Zamora R, L. [Centro de Estudios Aplicados al Desarrollo Nuclear (Cuba)

    1997-07-01

    At the water storage of the irradiated nuclear fuel has been an important factor in its management. The actual pools have its walls covered with inoxidable steel and heat exchangers to dissipate the residual heat from fuel. It is essential to control the water purity to eliminate those conditions which aid to the corrosion process in fuel and at related components. The steel used in this research was obtained from an austenitic inoxidizable steel standardized with titanium 08x18H10T (Type 321) similar to one of the two steel coatings used to cover walls and the pools floor. the test consisted in the specimen deformation through an U ply according to the Astm G30-90 standard. The exposition of the deformed specimen it was realized in simulated conditions to the chemical regime used in pools. (Author)

  12. Spent nuclear fuel storage. (Latest citations from the NTIS bibliographic database). Published Search

    International Nuclear Information System (INIS)

    1997-07-01

    The bibliography contains citations concerning spent nuclear fuel storage technologies, facilities, sites, and assessment. References review wet and dry storage, spent fuel casks and pools, underground storage, monitored and retrievable storage systems, and aluminum-clad spent fuels. Environmental impact, siting criteria, regulations, and risk assessment are also discussed. Computer codes and models for storage safety are covered. (Contains 50-250 citations and includes a subject term index and title list.) (Copyright NERAC, Inc. 1995)

  13. Assessment of spent WWER-440 fuel performance under long-term storage conditions

    Energy Technology Data Exchange (ETDEWEB)

    Takats, F [TS Enercon Kft. (Hungary)

    2012-07-01

    Paks Nuclear Power Plant is the only NPP in Hungary. It has four WWER-440 type reactor units. The fresh fuel is imported from Russia so far. The spent fuel assemblies were shipped back to Russia until 1997 after about 6 years cooling at the plant. A dry storage facility (MVDS type) has been constructed and is operational since then. By 1 January 2008, there were 5107 assemblies in dry storage. The objectives are: 1) Wet AR storage of spent fuel from the NPP Paks: Measurements of conditions for spent fuel storage in the at-reactor (AR) storage pools of Paks NPP (physical and chemical characteristics of pool water, corrosion product data); Measurements and visual control of storage pool component characteristics; Evaluation of storage characteristics and conditions with respect to long-term stability (corrosion of fuel cladding, construction materials); 2) Dry AFR storage at Paks NPP: Calculation and measurement of spent fuel conditions during the transfer from the storage pool to the modular vault dry storage (MVDS) on the site; Calculation and measurement of spent fuel conditions during the preparation of fuel for dry storage (drying process), such as crud release, activity build-up; Measurement of spent fuel conditions during the long-term dry storage, activity data in the storage tubes and amount of crud.

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

    International Nuclear Information System (INIS)

    Bailey, W.J.; Pankaskie, P.J.; Langstaff, D.C.; Gilbert, E.R.; Rising, K.H.; Schreiber, R.E.

    1982-11-01

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

  15. Reracking of fuel pools, experience with improved codes and design for reactor sites with high seismic loads

    International Nuclear Information System (INIS)

    Banck, J.; Wirtz, K.

    1998-01-01

    Reracking of existing pools to the maximum extent is desirable from the economical point of view. Although the load onto the storage rack structure and the fuel pool bottom will be increased, new improved codes, optimized structural qualification procedures and advanced design enable to demonstrate the structural integrity for all normal and accident conditions so that the design provides a safe compact storage of spent fuel under any condition.(author)

  16. Nuclear reactor spent fuel storage rack

    International Nuclear Information System (INIS)

    Machado, O.J.; Flynn, W.M.; Flanders, H.E. Jr.; Booker, L.W.

    1989-01-01

    A fuel rack is described for use in storing nuclear fuel assemblies in a nuclear fuel storage pool having a floor on which an upwardly projecting stud is mounted; the fuel rack comprising: a base structure at the lower end of the fuel rack including base-plate means having flow openings therein, the base-plate means supporting a first network of interlaced beams which form a multiplicity of polygonal openings; a second network of interlaced beams forming polygonal openings positioned in spaced vertical alignment with corresponding polygonal openings in the first network of beams; a plurality of cells, each cell having sides bounded by inner and outer surfaces and being of a size and configuration designed to hold therein a fuel assembly, each cell positioned in a corresponding pair of the aligned polygonal openings, each cell being open at both ends with a guiding funnel at the upper end, and the cells being positioned over the flow openings in the base-plate to permit flow of coolant through the cells; spaced, outwardly directed, projections on the outer surfaces of the sides of the cells near the tops and bottoms of the sides thereof, each cell being sized to be received within a corresponding of the pair of aligned polygonal openings in which the cells are respectively positioned; and means fixedly securing the projections to the beams in the first and second networks of beams thereby to provide a substantially rigid fuel rack of modular design

  17. Conceptual design of reactor TRIGA PUSPATI (RTP) spent fuel pool cooling system

    International Nuclear Information System (INIS)

    Tonny Lanyau; Mazleha Maskin; Mohd Fazli Zakaria; Mohmammad Suhaimi Kassim; Ahmad Nabil Abdul Rahim; Phongsakorn Prak Tom; Mohd Fairus Abdul Farid; Mohd Huzair Hussain

    2012-01-01

    After undergo about 30 years of safe operation, Reactor TRIGA PUSPATI (RTP) was planned to be upgraded to ensure continuous operation at optimum safety condition. In the meantime, upgrading is essential to get higher flux to diversify the reactor utilization. Spent fuel pool is needed for temporary storage of the irradiated fuel before sending it back to original country for reprocessing, reuse after the upgrading accomplished or final disposal. The irradiated fuel elements need to be secure physically with continuous cooling to ensure the safety of the fuels itself. The decay heat probably still exist even though the fuel elements not in the reactor core. Therefore, appropriate cooling is required to remove the heat produced by decay of the fission product in the irradiated fuel element. The design of spent fuel pool cooling system (SFPCS) was come to mind in order to provide the sufficient cooling to the irradiated fuel elements and also as a shielding. The spent fuel pool cooling system generally equipped with pumps, heat exchanger, water storage tank, valve and piping. The design of the system is based on criteria of the primary cooling system. This paper provides the conceptual design of the spent fuel cooling system. (author)

  18. Safety aspects of dry spent fuel storage and spent fuel management

    International Nuclear Information System (INIS)

    Botsch, W.; Smalian, S.; Hinterding, P.; Voelzke, H.; Wolff, D.; Kasparek, E.

    2014-01-01

    The storage of spent nuclear fuel (SF) and high-level radioactive waste (HLW) must conform to safety requirements. Safety aspects like safe enclosure of radioactive materials, safe removal of decay heat, nuclear criticality safety and avoidance of unnecessary radiation exposure must be achieved throughout the storage period. 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. In Germany dual purpose casks for SF or HLW are used for safe transportation and interim storage. TUV and BAM, who work as independent experts for the competent authorities, present the storage licensing process including sites and casks and inform about spent nuclear fuel management and issues concerning dry storage of spent nuclear fuel, based on their long experience in these fields (authors)

  19. Spent fuel storage at Prairie Island: January 1995 status

    International Nuclear Information System (INIS)

    Closs, J.; Kress, L.

    1995-01-01

    The disposal of spent nuclear fuel has been an issue for the US since the inception of the commercial nuclear power industry. In the past decade, it has become a critical factor in the continued operation of some nuclear power plants, including the two units at Prairie Island. As the struggles and litigation over storage alternatives wage on, spent fuel pools continue to fill and plants edge closer to premature shutdown. Due to the delays in the construction of a federal repository, many nuclear power plants have had to seek interim storage alternatives. In the case of Prairie Island, the safest and most feasible option is dry cask storage. This paper discusses the current status of the Independent Spent Fuel Storage Installation (ISFSI) Project at Prairie Island. It provides a historical background to the project, discusses the notable developments over the past year, and presents the projected plans of the Northern States Power Company (NSP) in regards to spent fuel storage

  20. Licensing of spent fuel dry storage and consolidated rod storage: A Review of Issues and Experiences

    Energy Technology Data Exchange (ETDEWEB)

    Bailey, W.J.

    1990-02-01

    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.

  1. Report on the possibilities of long-term storage of irradiated nuclear fuels

    International Nuclear Information System (INIS)

    2001-01-01

    This report aims at giving a legislative aspect to the many rules that govern the activities of the back-end of the fuel cycle in France. These activities concern the unloading of spent nuclear fuels, their reprocessing, storage, recycling and definitive disposal. The following points are reviewed and commented: the management of non-immediately reprocessed fuels (historical reasons of the 'all wastes reprocessing' initial choice, evolution of the economic and political context, the future reprocessing or the definitive disposal of spent fuels in excess); the inevitable long-term storage of part of the spent fuels (quantities and required properties of long-term stored fuels, the eventuality of a definitive disposal of spent fuels); the criteria that long-term storage facilities must fulfill (confinement measures, reversibility, surveillance and control during the whole duration of the storage); storage concept to be retained (increase of storage pools capacity, long-term storage in pools of reprocessing plants, centralized storage in pools, surface dry-storage on power plant sites, reversible underground storage, subsurface storage and storage/disposal in galleries, surface dry-storage facilities); the preliminary studies for the creation of long-term storage facilities (public information, management by a public French organization, clarifying of the conditions of international circulation of spent fuels); problems linked with the presence of foreign spent fuels in France (downstream of the reprocessing cycle, foreign plutonium and wastes re-shipment); conclusions and recommendations. (J.S.)

  2. Development of spent fuel dry storage technology

    International Nuclear Information System (INIS)

    Maruoka, Kunio; Matsunaga, Kenichi; Kunishima, Shigeru

    2000-01-01

    The spent fuels are the recycle fuel resources, and it is very important to store the spent fuels in safety. There are two types of the spent fuel interim storage system. One is wet storage system and another is dry storage system. In this study, the dry storage technology, dual purpose metal cask storage and canister storage, has been developed. For the dual purpose metal cask storage, boronated aluminum basket cell, rational cask body shape and shaping process have been developed, and new type dual purpose metal cask has been designed. For the canister storage, new type concrete cask and high density vault storage technology have been developed. The results of this study will be useful for the spent fuel interim storage. Safety and economical spent fuel interim storage will be realized in the near future. (author)

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

    International Nuclear Information System (INIS)

    Siu, N.; Khericha, S.; Conroy, S.; Beck, S.; Blackman, H.

    1996-09-01

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

  4. Sustainable Solutions for Nuclear used Fuels Interim Storage

    International Nuclear Information System (INIS)

    Arslan, Marc; Favet, Dominique; Issard, Herve; Le Jemtel, Amaury; Drevon, Caroline

    2014-01-01

    AREVA has a unique experience in providing sustainable solutions for used fuel management, fitted with the needs of different customers in the world and with regulation in different countries. These solutions entail both recycling and interim storage technologies. In a first part, we will describe the various types of solutions for Interim Storage of UNF that have been implemented around the world for interim storage at reactor or centralized Pad solution in canisters dry storage, vault type storages for dry storage, dry storage of transportation casks (dual purpose) pools for wet storage, The experience for all these different families of interim storages in which AREVA is involved is extensive and will be discussed with respect to the new challenges: increase of the duration of the interim storage (long term interim storage) increase of burn up of the fuels In a second part of the presentation, special recycling features will be presented. In that case, interim storage of the used fuels is ensured in pools. This provides in the long term good conditions for the behaviour of the fuel and its retrievability. With recycling, the final waste (Universal Canister of vitrified fission products and compacted hulls and end pieces): is stable and licensed in many countries for the final disposal (France, UK, Belgium, NL, Switzerland, Germany, Japan, upcoming: Spain, Australia, Italy). Presents neither safety criticality risks nor proliferation risks (AREVA conditioned HLW and LL-ILW are free of IAEA safeguard constraints thanks to AREVA process high recovery and purification yields). It can therefore be safely stored in interim storage for more than 100 years before final disposal. Some economic considerations will also be discussed. In particular, in the case of long term interim storage of used fuels, there are growing uncertainties regarding the future needs of repackaging and transportation, which can result in future cost overruns. Meanwhile, in the recycling policy

  5. Nuclear spent fuel dry storage in the EWA reactor shaft

    International Nuclear Information System (INIS)

    Mieleszczenko, W.; Moldysz, A.; Hryczuk, A.; Matysiak, T.

    2001-01-01

    The EWA reactor was in operation from 1958 until February 1995. Then it was subjected to the decommissioning procedure. Resulting from a prolonged operation of Polish research reactors a substantial amount of nuclear spent fuel of various types, enrichment and degree of burnup have been accumulated. The technology of storage of spent nuclear fuel foresees the two stages of wet storing in a water pool (deferral period from tens to several dozens years) and dry storing (deferral period from 50 to 80 years). In our case the deferral time in the water environment is pretty significant (the oldest fuel elements have been stored in water for more than 40 years). Though the state of stored fuel elements is satisfactory, there is a real need for changing the storage conditions of spent fuel. The paper is covering the description of philosophy and conceptual design for construction of the spent fuel dry storage in the decommissioned EWA reactor shaft. (author)

  6. Criticality impacts on LWR fuel storage efficiency

    International Nuclear Information System (INIS)

    Napolitano, D.

    1992-01-01

    This presentation discusses the criticality impacts throughout storage of fuel onsite including new fuel storage, spent fuel storage, consolidation, and dry storage. The general principles for criticality safety are also be discussed. There is first an introduction which explains today's situation for criticality safety concerns. This is followed by a discussion of criticality safety Regulatory Guides, safety limits and fundamental principles. Design objectives for criticality safety in the 1990's include higher burnups, longer cycles, and higher enrichments which impact the criticality safety design. Criticality safety for new fuel storage, spent fuel storage, fuel consolidation, and dry storage are followed by conclusions. Today's situation is one in which the US does not reprocess, and does not have an operating MRS facility or repository. High density fuel storage rack designs of the 1980s, are filling up. Dry cask storage systems for spent fuel storage are being utilized. Enrichments continue to increase PWR fuel assemblies with enrichments of 4.5 to 5.0 weight percent U-235 and BWR fuel assemblies with enrichments of 3.25 to 3.5 weight percent U-235 are common. Criticality concerns affect the capacity and the economics of light water reactor (LWR) fuel storage arrays by dictating the spacing of fuel assemblies in a storage system, or the use of poisons or exotic materials in the storage system design

  7. Storage device of reactor fuel

    International Nuclear Information System (INIS)

    Nakamura, Masaaki.

    1997-01-01

    The present invention concerns storage of spent fuels and provides a storage device capable of securing container-cells in shielding water by remote handling and moving and securing the container-cells easily. Namely, a horizontal support plate has a plurality of openings formed in a lattice like form and is disposed in a pit filled with water. The container-cell has a rectangular cross section, and is inserted and disposed vertically in the openings. Securing members are put between the container-cells above the horizontal support plate, and constituted so as to be expandable from above by remote handling. The securing member is preferably comprised of a vertical screw member and an expandable urging member. Since securing members for securing the container-cells for incorporating reactor fuels are disposed to the horizontal support plate controllable from above by the remote handling, fuel storage device can be disposed without entering into a radiation atmosphere. The container-cells can be settled and exchanged easily after starting of the use of a fuel pit. (I.S.)

  8. Storage arrangements for nuclear fuel

    International Nuclear Information System (INIS)

    Deacon, D.

    1982-01-01

    A storage arrangement for spent nuclear fuel either irradiated or pre-irradiated or for vitrified waste after spent fuel reprocessing, comprises a plenum chamber which has a base pierced by a plurality of openings each of which has sealed to it an open topped tube extending downwards and closed at its lower end. The plenum chamber, with the tubes, forms an air-filled enclosure associated with an exhaust system for exhausting air from the system through filters to maintain the interior of the enclosure at sub-atmospheric pressure. The tubes are arranged to accommodate the stored fuel and the arrangement includes a means for producing a flow of cooling air over the exterior of the tubes so that the latter effectively form a plurality of heat exchangers in close proximity to the fuel. The air may be caused to flow over the tube surfaces by a natural thermosyphon process. (author)

  9. Spent fuel storage requirements, 1988

    International Nuclear Information System (INIS)

    1988-10-01

    Historical inventories of spent fuel and Department of Energy (DOE) estimates of future discharges from US commercial nuclear reactors are presented for the next 20 years, through the year 2007. The eventual needs for additional spent fuel storage capacity are estimated. These estimates are based on the maximum capacities within current and planned at-reactor facilities and on any planned transshipments of fuel to other reactors or facilities. Historical data through December 1987 and projected discharges through the end of reactor life are used in this analysis. The source data was supplied by the utilities to DOE through the 1988 RW-859 data survey and by DOE estimates of future nuclear capacity, generation, and spent fuel discharges. 12 refs., 3 figs., 28 tabs

  10. Underwater Nuclear Fuel Disassembly and Rod Storage Process and Equipment Description. Volume II

    International Nuclear Information System (INIS)

    Viebrock, J.M.

    1981-09-01

    The process, equipment, and the demonstration of the Underwater Nuclear Fuel Disassembly and Rod Storage System are presented. The process was shown to be a viable means of increasing spent fuel pool storage density by taking apart fuel assemblies and storing the fuel rods in a denser fashion than in the original storage racks. The assembly's nonfuel-bearing waste is compacted and containerized. The report documents design criteria and analysis, fabrication, demonstration program results, and proposed enhancements to the system

  11. Dry spent fuel storage in the 1990's

    International Nuclear Information System (INIS)

    Roberts, J.P.

    1991-01-01

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

  12. Nuclear criticality assessment of Oak Ridge research fuel element storage

    International Nuclear Information System (INIS)

    Thomas, J.T.

    1978-06-01

    Spent and partially spent Oak Ridge Research Reactor (ORR) fuel elements are retained in the storage section of the ORR pool facility. Determination of a maximum expected neutron multiplication factor for the storage area is accomplished by a validated calculational method. The KENO Monte Carlo code and the Hansen-Roach 16-group neutron cross section sets were validated by calculations of critical experiments performed with early ORR fuel elements and with SPERT-D fuel elements. Calculations of various fuel element arrangements are presented which confirm the subcriticality previously inferred from critical experiments and indicate the k/sub eff/ would not exceed 0.85, were the storage area to be filled to capacity with storage racks containing elements with the fissionable material loading increased to 350 g of 235 U

  13. Spent fuel storage rack for BWR fuel assemblies

    International Nuclear Information System (INIS)

    Machado, O.; Henry, C.W.; Congleton, R.L.; Flynn, W.M.

    1990-01-01

    This patent describes for the use in storing nuclear fuel assemblies in a storage pool containing a coolant and having a pool floor, a fuel rack module. It comprises: a base plate to be disposed generally horizontally on the floor and having a horizontal surface area sufficient to support a fuel assemblies; uniformly spaced openings in the base plate, disposed in rows and columns throughout the surface area; fabricated cells of rectangular cross section extending over alternate openings along each row of the openings, the fabricated cells of each row being uniformly staggered by one opening with respect to the cells of its just adjacent rows so that the fabricated cells form a checkerboard like array; each of the fabricated cells having elongated walls mounted generally vertically on the base plate; each of the corners formed by the walls of each fabricated cell, which corners are internal of the periphery of the array, being disposed as closely adjacent as practicable to and face-to-face with a corner of an adjacent fabricated cell and joined by weld means so that substantially no space exists between adjacent cells. The cells being welded to their bottom ends to the base plate so that a strong compact modular structure is produced; neutron-absorbing means on the external surface of the fabricated cell walls except on the coextensive sections of the outer wall around the periphery of the array; and leveling pads are mounted under the base plate near the periphery thereof and adjustably engage the pool floor and intermediate leveling pads are mounted under cells within the fuel-rack module, the intermediate pads being uniformly disposed

  14. Spent fuel storage requirements 1987

    International Nuclear Information System (INIS)

    1987-09-01

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

  15. Compressed gas fuel storage system

    Science.gov (United States)

    Wozniak, John J.; Tiller, Dale B.; Wienhold, Paul D.; Hildebrand, Richard J.

    2001-01-01

    A compressed gas vehicle fuel storage system comprised of a plurality of compressed gas pressure cells supported by shock-absorbing foam positioned within a shape-conforming container. The container is dimensioned relative to the compressed gas pressure cells whereby a radial air gap surrounds each compressed gas pressure cell. The radial air gap allows pressure-induced expansion of the pressure cells without resulting in the application of pressure to adjacent pressure cells or physical pressure to the container. The pressure cells are interconnected by a gas control assembly including a thermally activated pressure relief device, a manual safety shut-off valve, and means for connecting the fuel storage system to a vehicle power source and a refueling adapter. The gas control assembly is enclosed by a protective cover attached to the container. The system is attached to the vehicle with straps to enable the chassis to deform as intended in a high-speed collision.

  16. High density fuel storage racks

    International Nuclear Information System (INIS)

    Groves, M.D.

    1978-01-01

    An apparatus is described for the safe and compact storage of nuclear fuel assemblies in an array of discrete open-ended neutron absorbing shields for which the theoretical minimum safe separation distance and cell pitch are known. Open-ended stainless steel end fittings are welded to each end of each shield and the end fittings are welded to each other in side-by-side relation, thereby reducing the cell pitch tolerance due to fabrication uncertainties. In addition, a multiplicity of ridges on the sides of each shield having a height equal to one half the theoretical minimum safe separation distance further reduce shield bowing tolerances. The net tolerance reduction permits a significant increase in the number of fuel assemblies that can be safely contained in a storage area of fixed size

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

    International Nuclear Information System (INIS)

    Viebrock, J.M.; Carlson, R.W.

    1979-09-01

    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/ft 2 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/ft 2 . Even this is a substantial increase over the 0.39 MTU/ft 2 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

  18. PWR Core II blanket fuel disposition recommendation of storage option study

    International Nuclear Information System (INIS)

    Dana, C.M.

    1995-01-01

    After review of the options available for current storage of T Plant Fuel the recommended option is wet storage without the use of chillers. A test has been completed that verifies the maximum temperature reached is below the industrial standard for storage of spent fuel. This option will be the least costly and still maintain the fuel in a safe environment. The options that were evaluated included dry storage with and without chillers, and wet storage with and without chillers. Due to the low decay heat of the Shippingport Core II Blanket fuel assemblies the fuel pool temperature will not exceed 100 deg. F

  19. Alternatives for water basin spent fuel storage: executive summary and comparative evaluation

    International Nuclear Information System (INIS)

    Viebrock, J.M.

    1979-09-01

    A five part report identifies and evaluates alternatives to conventional methods for water basin storage of irradiated light water reactor fuel assemblies (spent fuel). A recommendation is made for development or further evaluation of one attractive alternative: Proceed to develop fuel disassembly with subsequent high density storage of fuel pins (pin storage). The storage alternatives were evaluated for emplacement at reactor, in existing away-from-reactor storage facilities and in new away-from-reactor facilities. In the course of the study, the work effort necessarily extended beyond the pool wall in scope to properly assess the affects of storage alternatives on AFT systems

  20. Interim spent-fuel storage options at commercial nuclear power plants

    International Nuclear Information System (INIS)

    Thakkar, A.R.; Hylko, J.M.

    1991-01-01

    Although spent fuel can be stored safely in waterfilled pools at reactor sites, some utilities may not possess sufficient space for life-of-plant storage capability. In-pool storage capability may be increased by reracking assemblies, rod consolidation, double tiering spent-fuel racks, and by shipping spent fuel to other utility-owned facilities. Long-term on-site storage capability for spent fuel may be provided by installing (dry-type) metal casks, storage and transportation casks, concrete casks, horizontal concrete modules, modular concrete vaults, or by constructing additional (pool-type) storage installations. Experience to date has provided valuable information regarding dry-type or pool-type installations, cask handling and staffing requirements, security features, decommissioning activities, and radiological issues

  1. Storage of spent nuclear fuel

    International Nuclear Information System (INIS)

    Machado, O.J.; Moore, J.T.; Cooney, B.F.

    1989-01-01

    This patent describes a rack for storing nuclear fuel assemblies. The rack including a base, an array of side-by-side fuel-storage locations, each location being a hollow body of rectangular transverse cross section formed of metallic sheet means which is readily bent, each body having a volume therein dimensioned to receive a fuel assembly. The bodies being mounted on the base with each body secured to bodies adjacent each body along welded joints, each joint joining directly the respective contiguous corners of each body and of bodies adjacent to each body and being formed by a series of separate welds spaced longitudinally between the tops and bottoms of the secured bodies along each joint. The spacings of the separate welds being such that the response of the rack when it is subjected to the anticipated seismic acceleration of the rack, characteristic of the geographical regions where the rack is installed, is minimized

  2. Hydrogen storage and fuel cells

    Science.gov (United States)

    Liu, Di-Jia

    2018-01-01

    Global warming and future energy supply are two major challenges facing American public today. To overcome such challenges, it is imperative to maximize the existing fuel utilization with new conversion technologies while exploring alternative energy sources with minimal environmental impact. Hydrogen fuel cell represents a next-generation energy-efficient technology in transportation and stationary power productions. In this presentation, a brief overview of the current technology status of on-board hydrogen storage and polymer electrolyte membrane fuel cell in transportation will be provided. The directions of the future researches in these technological fields, including a recent "big idea" of "H2@Scale" currently developed at the U. S. Department of Energy, will also be discussed.

  3. Preliminary assessment of alternative dry storage methods for the storage of commercial spent nuclear fuel

    International Nuclear Information System (INIS)

    1981-11-01

    This report presents the results of an assessment of the (1) state of technology, (2) licensability, (3) implementation schedule, and (4) costs of alternative dry methods for storage of spent fuel at a reactor location when used to supplement reactor pool storage facilities. The methods of storage that were considered included storage in casks, drywells, concrete silos and air-cooled vaults. The impact of disassembly of spent fuel and storage of consolidated fuel rods was also determined. The economic assessments were made based on the current projected storage requirements of Virginia Electric and Power Company's Surry Station for the period 1985 to 2009, which has two operating pressurized water reactors (824 MWe each). It was estimated that the unit cost for storage of spent fuel in casks would amount to $117/kgU and that such costs for storage in drywells would amount to $137/kgU. However, based on the overall assessment it was concluded both storage methods were equal in merit. Modular methods of storage were generally found to be more economic than those requiring all or most of the facilities to be constructed prior to commencement of storage operations

  4. A Probabilistic Analysis Methodology and Its Application to A Spent Fuel Pool System

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyowon; Jae, Moosung [Hanyang Univ., Seoul (Korea, Republic of); Ryu, Ho G. [Daedeok R and D Center, Daejeon (Korea, Republic of)

    2013-05-15

    There was a similar accident occurring at the 2{sup nd} unit of PAKS nuclear power station in Hungary on the 10{sup th} April 2003. Insufficient cooling of spent fuel caused the spent fuel burn up or partly melting. There were many previous studies performed for analyzing and measuring the risk of spent fuel damage. In the 1980s, there are changes in conditions such as development of high density storage racks and new information concerning the possibility of cladding fires in the drained spent fuel pools. The US NRC assessed the spent fuel pool risk under the Generic Issue 82. In the 1990s, under the US NRC sponsorship, the risk assessment about the spent fuel pool at Susquehanna Steam Electric Station (SSES) has been performed and Analysis Evaluation of Operational Data (AEOD) has been organized for accumulating the reliability data. A methodology for assessing the risk associated with the spent fuel pool facility has been developed and is applied to the reference plant. It is shown that the methodology developed in this study might contribute to assessing these kinds of the SFP facilities. In this probabilistic risk analysis, the LINV Initial event results in the high frequent occurrence. The most dominant cut-sets include the human errors. The result of this analysis might contribute to identifying the weakness of the preventive and mitigating system in the SFP facility.

  5. 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

  6. Long term wet spent nuclear fuel storage

    International Nuclear Information System (INIS)

    1987-04-01

    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

  7. Near surface spent fuel storage: environmental issues

    International Nuclear Information System (INIS)

    Nelson, I.C.; Shipler, D.B.; McKee, R.W.; Glenn, R.D.

    1979-01-01

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

  8. Decontamination of transport casks and of spent fuel storage facilities

    International Nuclear Information System (INIS)

    1990-06-01

    The present document provides an analysis of the technical papers presented at the meeting as well as a summary of the panel discussion. Conclusions and Recommendations: The meeting agreed that the primary source of contamination of transport casks is the production of radioactive isotopes in nuclear fuel and activation products of fuel components in nuclear reactors. The type, amount of mechanism for the release of these isotopes depend on the reactor type and fuel handling process. The widespread use of pools for the storage and handling of fuel provides an easy path for the transfer of contamination. Control of pool water conditions is essential for limiting the spread of contamination. For plants where casks are immersed in pools for loading, the immersion times should be minimised. Casks should be designed for ease of decontamination. The meeting discussed the use of stainless steel and suitable paints for coating casks. Designers should consider the appropriate coating for specific applications. The use of pressurized water for decontamination is recommended whenever possible. A number of commercially available reagents exist for decontaminating cask external surfaces. More work, however, is needed to cope with Pressurized Water Reactor crud within casks. Leaking fuel should be identified and isolated before storage in pools. Basic studies of the uptake and release of contamination from cask surfaces should be initiated. Standardization of methods of contamination measurement and instrumentation should be instituted. Refs, figs and tabs

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

    International Nuclear Information System (INIS)

    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 2 oxidation in breached fuel rods would not become a problem. Scoping tests at PNL with nonirradiated UO 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

  10. Dry storage of spent fuel

    International Nuclear Information System (INIS)

    Jeffrey, R.

    1993-01-01

    Scottish Nuclear's plans to build and operate dry storage facilities at each of its two nuclear power station sites in Scotland are explained. An outline of where waste materials arise as part of the operation and decommissioning of nuclear power stations, the volumes for each category of high-, intermediate-and low-level wastes and the costs involved are given. The present procedure for the spent fuels from Hunterston-B and Torness stations is described and Scottish Nuclear's aims of driving output up and costs down are studied. (UK)

  11. In-pool damaged fuel bundle recovery

    International Nuclear Information System (INIS)

    Piascik, T.G.; Patenaude, R.S.

    1988-01-01

    While preparing to rerack the Oyster Creek Nuclear Generating Station, GPU Nuclear had need to move a damaged fuel bundle. This bundle had no upper tie plate and could not be moved in the normal manner. GPU Nuclear formed a small, dedicated project team to disassemble, package and move this damaged bundle. The team was composed of key personnel from GPU Nuclear Fuels Projects, OCNGS Operations and Proto-Power / Bisco, a specialty contractor who has fuel bundle reconstitution and rod consolidation experience, remote tooling, underwater video systems and experienced technicians. Proven tooling, clear procedures and a simple approach were important, but the key element was the spirit of teamwork and leadership exhibited by the people involved

  12. In-pool damaged fuel bundle recovery

    International Nuclear Information System (INIS)

    Piascik, T.G.; Patenaude, R.S.

    1988-01-01

    While preparing to rerack the Oyster Creek Nuclear Generating Station, GPU Nuclear had need to move a damaged fuel bundle. This bundle had no upper tie plate and could not be moved in the normal manner. GPU Nuclear formed a small, dedicated project team to disassemble, package, and move this damaged bundle. The team was composed of key personnel from GPU Nuclear Fuels Projects, OCNGS Operations and Proto-Power/Bisco, a specialty contractor who has fuel bundle reconstitution and rod consolidation experience, remote tooling, underwater video systems and experienced technicians. Proven tooling, clear procedures and a simple approach were important, but the key element was the spirit of teamwork and leadership exhibited by the people involved. In spite of several emergent problems which a task of this nature presents, this small, close knit utility/vendor team completed the work on schedule and within the exposure and cost budgets

  13. Costing of spent nuclear fuel storage

    International Nuclear Information System (INIS)

    2009-01-01

    This report deals with economic analysis and cost estimation, based on exploration of relevant issues, including a survey of analytical tools for assessment and updated information on the market and financial issues associated with spent fuel storage. The development of new storage technologies and changes in some of the circumstances affecting the costs of spent fuel storage are also incorporated. This report aims to provide comprehensive information on spent fuel storage costs to engineers and nuclear professionals as well as other stakeholders in the nuclear industry. This report is meant to provide informative guidance on economic aspects involved in selecting a spent fuel storage system, including basic methods of analysis and cost data for project evaluation and comparison of storage options, together with financial and business aspects associated with spent fuel storage. After the review of technical options for spent fuel storage in Section 2, cost categories and components involved in the lifecycle of a storage facility are identified in Section 3 and factors affecting costs of spent fuel storage are then reviewed in the Section 4. Methods for cost estimation and analysis are introduced in Section 5, and other financial and business aspects associated with spent fuel storage are discussed in Section 6.

  14. Rock cavern storage of spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Won Jin; Kim, Kyung Soo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kwon, Sang Ki [Inha University, Incheon (Korea, Republic of)

    2015-12-15

    The rock cavern storage for spent fuel has been assessed to apply in Korea with reviewing the state of the art of the technologies for surface storage and rock cavern storage of spent fuel. The technical feasibility and economic aspects of the rock cavern storage of spent fuel were also analyzed. A considerable area of flat land isolated from the exterior are needed to meet the requirement for the site of the surface storage facilities. It may, however, not be easy to secure such areas in the mountainous region of Korea. Instead, the spent fuel storage facilities constructed in the rock cavern moderate their demands for the suitable site. As a result, the rock cavern storage is a promising alternative for the storage of spent fuel in the aspect of natural and social environments. The rock cavern storage of spent fuel has several advantages compared with the surface storage, and there is no significant difference on the viewpoint of economy between the two alternatives. In addition, no great technical difficulties are present to apply the rock cavern storage technologies to the storage of domestic spent fuel.

  15. Analysis of radiation shields of BNPP spent fuel pool

    International Nuclear Information System (INIS)

    Ayoobian, N.; Hadad, K.; Nematollahi, M. R.

    2007-01-01

    Radioactive protection is one of the most important subjects in nuclear power plants safety. Analysis of BNPP spent fuel pool shielding , as a main source of energetic γ-rays was the main goal of this project. Firstly, we simulated the reactor core using WIMSD-4 neutronic code and the amount of fission product in the fuel assembly (FA) was calculated during the reactor operation. Then, by obtaining the results from the previous calculation and by using MCNP4C nuclear code , the intensity of γ-rays was obtained in layers of spent fuel pool shields. The results have shown that no significant γ-rays passed through these shields. Finally, an accident and resulting exposure dose above the pool was analyzed

  16. Storage method for spent fuel assembly

    International Nuclear Information System (INIS)

    Tajiri, Hiroshi.

    1992-01-01

    In the present invention, spent fuel assemblies are arranged at a dense pitch in a storage rack by suppressing the reactivity of the assemblies, to increase storage capacity for the spent fuel assemblies. That is, neutron absorbers are filled in the cladding tube of an absorbing rod, and the diameter thereof is substantially equal with that of a fuel rod. A great amount of the absorbing rods are arranged at the outer circumference of the fuel assembly. Then, they are fixed integrally to the fuel assembly and stored in a storage rack. In this case, the storage rack may be constituted only with angle materials which are inexpensive and installed simply. With such a constitution, in the fuel assembly having absorbing rods wound therearound, neutrons are absorbed by absorbing rods and the reactivity is lowered. Accordingly, the assembly arrangement pitch in the storage rack can be made dense. As a result, the storage capacity for the assemblies is increased. (I.S.)

  17. Seismic analysis of spent nuclear fuel storage racks

    International Nuclear Information System (INIS)

    Shah, S.J.; Biddle, J.R.; Bennett, S.M.; Schechter, C.B.; Harstead, G.A.; Marquet, F.

    1996-01-01

    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

  18. Interim dry fuel storage for magnox reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bradley, N [National Nuclear Corporation, Risley, Warrington (United Kingdom); Ealing, C [GEC Energy Systems Ltd, Whetstone, Leicester (United Kingdom)

    1985-07-01

    In the UK the practice of short term buffer storage in water ponds prior to chemical reprocessing had already been established on the early gas cooled reactors in Calder Hall. Thus the choice of water pond buffer storage for MGR power plants logically followed the national policy decision to reprocess. The majority of the buffer storage period would take place at the reprocessing plant with only a nominal of 100 days targeted at the station. Since Magnox clad fuel is not suitable for long term pond storage, alternative methods of storage on future stations was considered desirable. In addition to safeguards considerations the economic aspects of the fuel cycle has influenced the conclusion that today the purchase of a MGR power plant with dry spent fuel storage and without commitment to reprocess would be a rational decision for a country initiating a nuclear programme. Dry storage requirements are discussed and two designs of dry storage facilities presented together with a fuel preparation facility.

  19. Interim dry fuel storage for magnox reactors

    International Nuclear Information System (INIS)

    Bradley, N.; Ealing, C.

    1985-01-01

    In the UK the practice of short term buffer storage in water ponds prior to chemical reprocessing had already been established on the early gas cooled reactors in Calder Hall. Thus the choice of water pond buffer storage for MGR power plants logically followed the national policy decision to reprocess. The majority of the buffer storage period would take place at the reprocessing plant with only a nominal of 100 days targeted at the station. Since Magnox clad fuel is not suitable for long term pond storage, alternative methods of storage on future stations was considered desirable. In addition to safeguards considerations the economic aspects of the fuel cycle has influenced the conclusion that today the purchase of a MGR power plant with dry spent fuel storage and without commitment to reprocess would be a rational decision for a country initiating a nuclear programme. Dry storage requirements are discussed and two designs of dry storage facilities presented together with a fuel preparation facility

  20. Integrated project for increasing the capacity of spent fuel pools at Cofrentes NPP

    International Nuclear Information System (INIS)

    Rebollo Garcia, C.; Arana, S.

    1996-01-01

    The current storage capacity of the Cofrentes NPP will have reached its limit by the end of its 15th cycle, in the year 2005. The works performed by Empresarios Agrupados for IBERDROLA show that it is possible to increase this capacity in successive phases, so as to make the Power Plant self-sufficient for 16 more years (up to 2021) in the case of compact storage, or for 50 more years (2055) in the case of consolidated storage or second level storage. Optimisation of the management of high-activity wastes goes with a series of tasks which come under the group referred to as Integrated Project for Increasing the Capacity of Spent Fuel Pools. The main activities of the project can be summarised under the following three items: increase of storage capacity (feasibility study, specification for the purchase of racks, manufacture, assembly and tests), improvement of the capacity of the pool cooling system and modification of the components and accessories located inside the pools which interfere with the new racks. Another series of activities with less technical and economic impact are: modification of fuel handling machines, management of generated radwaste, licensing and modification of plant documentation (seismic analysis, radiation areas, as-built drawings and verification of the validation of purification and HVAC systems). (Author)

  1. Project management for the Virginia power spent fuel storage project

    International Nuclear Information System (INIS)

    Smith, M.

    1992-01-01

    Like Duke Power, Virginia Power has been involved in spent fuel storage expansion studies for a long time - possibly a little longer than Duke Power. Virginia Power's initial studies date back to the late 70s and into the early 80s. Large variety of storage techniques are reviewed including reracking and transshipment. Virginia Power also considered construction a new spent fuel pool. This was one of the options that was considered early on since Virginia Power started this process before any dry storage techniques had been proven. Consolidation of spent fuel is something that was also studied. Finally, construction of dry storage facility was determined to be the technology of choice. They looked a large variety of dry storage technologies and eventually selected dry storage in metal casks at Surry. There are many of reasons why a utility may choose one technology over another. In Virginia Power's situation, additional storage was needed at Surry much earlier than at other utilities. Virginia Power was confronted with selecting a storage technique and having to be a leader in that it was the first U.S. utility to implement a dry storage system

  2. Sogin enriched uranium extraction (EUREX) plant spent fuel pool cleaning and decontamination utilizing the Smart Safe solution

    International Nuclear Information System (INIS)

    Denton, M.S.; Gili, M.; Nasta, M.; Quintiliani, R.; Caccia, G.; Botzen, W.; Forrester, K.

    2009-01-01

    SOGIN's EUREX facility in Italy was developed as a pilot plant functional testing laboratory for spent fuel reprocessing. This facility was operated successfully for many years since 1970 and was eventually shutdown consistent with Italy's suspension of all nuclear operations. At the time of suspension, the EUREX facility still had spent nuclear fuel assemblies in storage from a nearby PWR. Other fuel assemblies from an Italian AGR had remained stored in the spent fuel pool for the 20 years or so waiting for removal and reprocessing abroad. Being Magnox fuel elements, their recovery for the transport produced a huge amount of sludge in the pool. During this time, sediment, dirt, corrosion products, fuel cladding, etc. has collected within the fuel pool as a crud layer dispersed throughout. Most of this crud has accumulated on the horizontal surfaces of the pool and fuel element assemblies, while some remains as a suspended colloidal material. Furthermore many other contaminated metal components, used during the operation years, where still inside the pool. Due to a pool leak discovered in 2006, SOGIN speeded up its pool decommissioning program, making also available the transfer of the spent fuel to a nearby interim repository, with the goal to completely drain the pool in the shortest period of time. In order for SOGIN to successfully transfer the fuel assemblies from their current storage basket locations to the spent fuel transfer cask, the bulk of the crud needed to be removed. This cleanup operation was deemed necessary to minimize the suspension of contamination in the water during underwater handling operations. This would reduce the decontamination efforts on the transfer cask upon removal, once loaded with the spent fuel, and enhance safety by reducing potential underwater visibility issues. The operations were completed in July 2008 with the release to the environment of the pool water, thoroughly purified and without any relevant radiological impact. The

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

    International Nuclear Information System (INIS)

    Bailey, W.J.

    1990-11-01

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

  4. Spent nuclear fuel storage - Basic concept

    International Nuclear Information System (INIS)

    Krempel, Ascanio; Santos, Cicero D. Pacifici dos; Sato, Heitor Hitoshi; Magalhaes, Leonardo de

    2009-01-01

    According to the procedures adopted in others countries in the world, the spent nuclear fuel elements burned to produce electrical energy in the Brazilian Nuclear Power Plant of Angra do Reis, Central Nuclear Almirante Alvaro Alberto - CNAAA will be stored for a long time. Such procedure will allow the next generation to decide how they will handle those materials. In the future, the reprocessing of the nuclear fuel assemblies could be a good solution in order to have additional energy resource and also to decrease the volume of discarded materials. This decision will be done in the future according to the new studies and investigations that are being studied around the world. The present proposal to handle the nuclear spent fuel is to storage it for a long period of time, under institutional control. Therefore, the aim of this paper is to introduce a proposal of a basic concept of spent fuel storage, which involves the construction of a new storage building at site, in order to increase the present storage capacity of spent fuel assemblies in CNAAA installation; the concept of the spent fuel transportation casks that will transfer the spent fuel assemblies from the power plants to the Spent Fuel Complementary Storage Building and later on from this building to the Long Term Intermediate Storage of Spent Fuel; the concept of the spent fuel canister and finally the basic concept of the spent fuel long term storage. (author)

  5. Spent fuel storage for ISER plant

    International Nuclear Information System (INIS)

    Nakajima, Takasuke; Kimura, Yuzi

    1987-01-01

    ISER is an intrinsically safe reactor basing its safety only on physical laws, and uses a steel reactor vessel in order to be economical. For such a new type reactor, it is essentially important to be accepted by the society by showing that the reactor is more profitable than conventional reactors to the public in both technical and economic viewpoint. It is also important that the reactor raises no serious problem in the total fuel cycle. Reprocessing seems one of the major worldwide fuel cycle issues. Spent fuel storage is also one of the key technologies for fuel cycle back end. Various systems for ISER spent fuel storages are examined in the present report. Spent fuel specifications of ISER are similar to those of LWR and therefore, most of LWR spent fuel technologies are basically applicable to ISER spent fuel. Design requirements and examples of storage facilities are also discussed. Dry storage seems to be preferable for the relatively long cooling time spent fuel like ISER's one from economical viewpoint. Vault storage will possibly be the most advantageous for large storage capacity. Another point for discussion is the location and international collaboration for spent fuel storages: ISER expected to be a worldwide energy source and therefore, international spent fuel management seems to be fairly attractive way for an energy recipient country. (Nogami, K.)

  6. Advanced compressed hydrogen fuel storage systems

    International Nuclear Information System (INIS)

    Jeary, B.

    2000-01-01

    Dynetek was established in 1991 by a group of private investors, and since that time efforts have been focused on designing, improving, manufacturing and marketing advanced compressed fuel storage systems. The primary market for Dynetek fuel systems has been Natural Gas, however as the automotive industry investigates the possibility of using hydrogen as the fuel source solution in Alternative Energy Vehicles, there is a growing demand for hydrogen storage on -board. Dynetek is striving to meet the needs of the industry, by working towards developing a fuel storage system that will be efficient, economical, lightweight and eventually capable of storing enough hydrogen to match the driving range of the current gasoline fueled vehicles

  7. Analysis of the Processes in Spent Fuel Pools in Case of Loss of Heat Removal due to Water Leakage

    Directory of Open Access Journals (Sweden)

    Algirdas Kaliatka

    2013-01-01

    Full Text Available The safe storage of spent fuel assemblies in the spent fuel pools is very important. These facilities are not covered by leaktight containment; thus, the consequences of overheating and melting of fuel in the spent fuel pools can be very severe. On the other hand, due to low decay heat of fuel assemblies, the processes in pools are slow in comparison with processes in reactor core during LOCA accident. Thus, the accident management measures play a very important role in case of some accidents in spent fuel pools. This paper presents the analysis of possible consequences of fuel overheating due to leakage of water from spent fuel pool. Also, the accident mitigation measure, the late injection of water was evaluated. The analysis was performed for the Ignalina NPP Unit 2 spent fuel pool, using system thermal hydraulic code for severe accident analysis ATHLET-CD. The phenomena, taking place during such accident, are discussed. Also, benchmarking of results of the same accident calculation using ASTEC and RELAP/SCDAPSIM codes is presented here.

  8. Thermal Analysis Evaluation of Spent Fuel Storage Rack for Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sangjin; Oh, Jinho; Kwak, Jinsung; Lee, Jongmin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Spent fuel storage rack is to store spent fuel assemblies. The spent fuel storage rack is submerged in the designated pool for cooling. Due to the condition change of the pool water, the effect of thermal load on spent fuel storage rack must be analyzed and evaluated. In this paper, thermal stress analysis is performed and evaluated on a spent fuel storage rack. For thermal stress evaluation of the spent fuel storage rack, load combinations and allowable criteria in ASME Sec. III NB-3220 are applied. In cases of A-1 and B-1, the same temperature applied on the whole model, thermal stress doesn't occur because there is no constraint about the thermal expansion. The support frame is located on the pool bottom in free standing type and the racks are located in the support frame with enough space. Thermal expansion was considered and reflected in the design of spent fuel storage rack in advance. Thermal stress analysis is performed and evaluated on a spent fuel storage rack with consideration of pool water temperature variation. The thermal analysis including a linear heat transfer and the thermal stress analysis is performed for the racks and support frame and resulted stresses are within allowable criteria.

  9. Strontium and cesium radionuclide leak detection alternatives in a capsule storage pool

    International Nuclear Information System (INIS)

    Larson, D.E.; Crawford, T.W.; Joyce, S.M.

    1981-08-01

    A study was performed to assess radionuclide leak-detection systems for use in locating a capsule leaking strontium-90 or cesium-137 into a water-filled pool. Each storage pool contains about 35,000 L of water and up to 715 capsules, each of which contains up to 150 kCi strontium-90 or 80 kCi cesium-137. Potential systems assessed included instrumental chemical analyses, radionuclide detection, visual examination, and other nondestructive nuclear-fuel examination techniques. Factors considered in the assessment include: cost, simplicity of maintenance and operation, technology availability, reliability, remote operation, sensitivity, and ability to locate an individual leaking capsule in its storage location. The study concluded that an adaption of the spent nuclear-fuel examination technique of wet sipping be considered for adaption. In the suggested approoch, samples would be taken continuously from pool water adjacent to the capsule(s) being examined for remote radiation detection. In-place capsule isolation and subsequent water sampling would confirm that a capsule was leaking radionuclides. Additional studies are needed before implementing this option. Two other techniques that show promise are ultrasonic testing and eddy-current testing

  10. Demonstration of a transportable storage system for spent nuclear fuel

    International Nuclear Information System (INIS)

    Shetler, J.R.; Miller, K.R.; Jones, R.E.

    1993-01-01

    The purpose of this paper is to discuss the joint demonstration project between the Sacramento Municipal Utility District (SMUD) and the US Department of Energy (DOE) regarding the use of a transportable storage system for the long-term storage and subsequent transport of spent nuclear fuel. SMUD's Rancho Seco nuclear generating station was shut down permanently in June 1989. After the shutdown, SMUD began planning the decommissioning process, including the disposition of the spent nuclear fuel. Concurrently, Congress had directed the Secretary of Energy to develop a plan for the use of dual-purpose casks. Licensing and demonstrating a dual-purpose cask, or transportable storage system, would be a step toward achieving Congress's goal of demonstrating a technology that can be used to minimize the handling of spent nuclear fuel from the time the fuel is permanently removed from the reactor through to its ultimate disposal at a DOE facility. For SMUD, using a transportable storage system at the Rancho Seco Independent Spent-Fuel Storage Installation supports the goal of abandoning Rancho Seco's spent-fuel pool as decommissioning proceeds

  11. Spent fuel and high-level radioactive waste storage

    International Nuclear Information System (INIS)

    Trigerman, S.

    1988-06-01

    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)

  12. Nuclear fuel storage apparatus for seismic areas

    International Nuclear Information System (INIS)

    Anthony, A.J.

    1982-01-01

    A structural grid for supporting spent fuel is located underwater in a pool. The grid is spaced from the walls of the pool and supported by cables from above. Horizontal acceleration due to seismic forces results in a movement of the support members and of the pool walls. The cables, being flexible, continue to support the grid but do not contribute to the horizontal movement of the grid. Accordingly, no significant earthquake forces are transmitted from the supporing structure

  13. A complete NUHOMS {sup registered} solution for storage and transport of high burnup spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bondre, J. [Transnuclear, Inc. (AREVA Group), Fremont, CA (United States)

    2004-07-01

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

  14. Fuel performance of DOE fuels in water storage

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    International Nuclear Information System (INIS)

    Dey, P.K.

    2005-01-01

    The spent fuel discharged from the reactors are temporarily stored at the reactor pool. After a certain cooling time, the spent fuel is moved to the storage locations either on or off reactor site depending on the spent fuel management strategy. As India has opted for a closed fuel cycle for its nuclear energy development, reprocessing of the spent fuel, recycling of the reprocessed plutonium and uranium and disposal of the wastes from the reprocessing operations forms the spent fuel management strategy. Since the reprocessing operations are planned to match the nuclear energy programme, storage of the spent fuel in ponds are adopted prior to reprocessing. Transport of the spent fuel to the storage locations are carried out adhering to international and national guide lines. India is having 14 operating power reactors and three research reactors. The spent fuel from the two safeguarded BWRs are stored at-reactor (AR) storage pond. A separate wet storage facility away-from-reactor (AFR) has been designed, constructed and made operational since 1991 for additional fuel storage. Storage facilities are provided in ARs at other reactor locations to cater to 10 reactor-years of operation. A much lower capacity spent fuel storage is provided in reprocessing plants on the same lines of AR fuel storage design. Since the reprocessing operations are carried out on a need basis, to cater to the increased storage needs two new spent fuel storage facilities (SFSF) are being designed and constructed near the existing nuclear plant sites. India has mastered the technology for design, construction and operation of wet spent fuel storage facility meeting all the international standards Wet storage of the spent fuel is the most commonly adopted mode all over the world. Recently an alternate mode viz. dry storage has also been considered. India has designed, constructed and operated lead shielded dry storage casks and is operational at one site. A dry storage cask made of concrete

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

    International Nuclear Information System (INIS)

    Oh, Jinho

    2013-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Jinho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-07-01

    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.

  18. Fuel removal, transport, and storage

    International Nuclear Information System (INIS)

    Reno, H.W.

    1986-01-01

    The March 1979 accident at Unit 2 of the Three Mile Island Nuclear Power Station (TMI-2) which damaged the core of the reactor resulted in numerous scientific and technical challenges. Some of those challenges involve removing the core debris from the reactor, packaging it into canisters, loading canisters into a rail cask, and transporting the debris to the Idaho National Engineering Laboratory (INEL) for storage, examination, and preparation for final disposal. This paper highlights how some challenges were resolved, including lessons learned and benefits derived therefrom. Key to some success at TMI was designing, testing, fabricating, and licensing two rail casks, which each provide double containment of the damaged fuel. 10 refs., 12 figs

  19. Behaviour of power and research reactor fuel in wet and dry storage

    Energy Technology Data Exchange (ETDEWEB)

    Freire-Canosa, J [Nuclear Waste Management Organization (Canada)

    2012-07-01

    Canada has developed extensive experience in both wet and dry storage of CANDU fuel. Fuel has been stored in water pools at CANDU reactor sites for approximately 45 years, and in dry storage facilities for a large part of the past decade. Currently, Canada has 38 450 t U of spent fuel in storage, of which 8850 t U are in dry storage. In June 2007, the Government of Canada selected the Adaptive Phased Management (APM) approach, recommended by the Nuclear Waste Management Organization (NWMO), for the long-term management of Canada's nuclear-fuel waste. The Canadian utilities and AECL are conducting development work in extended storage systems as well as research on fuel behaviour under storage conditions. Both activities have as ultimate objective to establish a technical basis for assuring the safety of long-term fuel storage.

  20. Achieving increased spent fuel storage capacity at the High Flux Isotope Reactor (HFIR)

    International Nuclear Information System (INIS)

    Cook, D.H.; Chang, S.J.; Dabs, R.D.; Freels, J.D.; Morgan, K.A.; Rothrock, R.B.; Griess, J.C.

    1994-01-01

    The HFIR facility was originally designed to store approximately 25 spent cores, sufficient to allow for operational contingencies and for cooling prior to off-site shipment for reprocessing. The original capacity has now been increased to 60 positions, of which 53 are currently filled (September 1994). Additional spent cores are produced at a rate of about 10 or 11 per year. Continued HFIR operation, therefore, depends on a significant near-term expansion of the pool storage capacity, as well as on a future capability of reprocessing or other storage alternatives once the practical capacity of the pool is reached. To store the much larger inventory of spent fuel that may remain on-site under various future scenarios, the pool capacity is being increased in a phased manner through installation of a new multi-tier spent fuel rack design for higher density storage. A total of 143 positions was used for this paper as the maximum practical pool capacity without impacting operations; however, greater ultimate capacities were addressed in the supporting analyses and approval documents. This paper addresses issues related to the pool storage expansion including (1) seismic effects on the three-tier storage arrays, (2) thermal performance of the new arrays, (3) spent fuel cladding corrosion concerns related to the longer period of pool storage, and (4) impacts of increased spent fuel inventory on the pool water quality, water treatment systems, and LLLW volume

  1. Review of Current Criteria of Spent Fuel Rod Integrity during Dry Storage

    International Nuclear Information System (INIS)

    Yang, Yong Sik; Kim, Sun Ki; Bang, Je Geon; Song, Kun Woo

    2006-01-01

    A PWR spent fuel has been stored in a wet storage pool in Korea. However, the amount of spent fuel is expected to exceed the capacity of a wet storage pool within 10∼15 years. From the early 1970's, a research on the PWR spent fuel dry storage started because the dry storage system has been economical compared with the wet storage system. The dry storage technology for Zircaloy-clad fuel was assessed and licensed in many countries such as USA, Canada, FRG and Switzerland. In the dry storage system, a clad temperature may be higher than in the wet storage system and can reach up to 400 .deg.. A higher clad temperature can cause cladding failures during the period of dry storage, and thus a dry storage related research has essentially dealt with the prevention of clad degradation. It is temperature and rod internal pressure that cause cladding failures through the mechanisms such as clad creep rupture, hydride re-orientation, and stress-corrosion cracking etc.. In this paper, the current licensing criteria are summarized for the PWR spent fuel dry storage system, especially on spent fuel rod integrity. And it is investigated that an application propriety of existing criteria to Korea spent fuel dry storage system

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

    International Nuclear Information System (INIS)

    Maaranen, Paeivi

    2014-01-01

    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

  3. Nonlinear dynamic response of whole pool multiple spent fuel racks subject to three-dimensional excitations

    International Nuclear Information System (INIS)

    Zhao, Y.; Wilson, P.R.; Stevenson, J.D.

    1995-01-01

    The seismic evaluation of submerged free standing spent fuel storage racks is more complicated than most other nuclear structural systems. When subjected to three dimensional (3-D) floor seismic excitations the dynamic responses of racks in a pool are hydro dynamically coupled with each other, with the fuel assemblies water in gaps. The motion behavior of the racks is significantly different from that observed using a 3D single rack mode. Few seismic analyses using 3-D whole pool multiple rack models are available in the literature. I this paper an analysis was performed for twelve racks using potential theory for the fluid-structure interaction, and using a 3-D whole pool multi-rack finite element model developed herein. The analysis includes the potential nonlinear dynamic behavior of the impact of fuel-rack, rack-rack and rack-pool wall, the tilting or uplift and the frictional sliding of rack supports, and the impact of the rack supports to the pool floor. (author). 12 refs., 7 figs., 1 tab

  4. Probabilistic risk analysis for Test Area North Hot Shop Storage Pool Facility

    International Nuclear Information System (INIS)

    Meale, B.M.; Satterwhite, D.G.

    1990-01-01

    A storage pool facility used for storing spent fuel and radioactive debris from the Three Mile Island (TMI) accident was evaluated to determine the risk associated with its normal operations. Several hazards were identified and examined to determine if any any credible accident scenarios existed. Expected annual occurrence frequencies were calculated for hazards for which accident scenarios were identified through use of fault trees modeling techniques. Fault tree models were developed for two hazards: (1) increased radiation field and (2) spread of contamination. The models incorporated facets of the operations within the facility as well as the facility itself. 6 refs

  5. A safety study on the wet storage of spent fuel

    International Nuclear Information System (INIS)

    Chun, Kwan Sik; Whang, Joo Ho; Lee, Hoo Kun; Choi, Jong Won; Lee, Jong Geun

    1989-02-01

    This study is to provide data related with a basic design of the spent fuel storage facility in the field of radiation and to establish the safety assessment methodology of away from reactor spent fuel storage facility. This is in progress and continue upto the year of 1991. The mathematical model which predict the quantity of environmental release of fission and corrosion products from spent fuel received and stored in wet storage facility operated in normal conditions was prepared. The decay characteristic of domestic spent fuels are analysed and then the coefficients for the prediction of the decay heat by simple formular was determined. This correlations could predict decay heat of spent fuel with ±10% difference from ORIGEN2 results. The release factor of cobalt out of PWR spent fuel in PIE pool is 7.97 x 10-12∼8.49 x 10-11 Ci/ sec-rod, which appears to be linear without being connected with the types of fuel defects, but that of cesium varies with the defect type and the exposure time in water. In water condition, release factor of uranium out of CANDU fuel pellets appears to be about 5 x 10-8/day, whose tendency is similar to that of cesium of the latter half of the exposure time of water. (Author)

  6. Assessment of spent WWER-440 fuel performance under long-term storage conditions

    Energy Technology Data Exchange (ETDEWEB)

    Kupca, L [VUJE Inc. (Slovakia)

    2012-07-01

    In the Slovak Republic are under operation 6 units (4 in the Jaslovske Bohunice site, and 2 in the Mochovce), 2 units are under construction in Mochovce site. All units are WWER-440 type. The fresh fuel is imported from the Russian Federation. The spent fuel assemblies are stored in wet conditions in Bohunice Interim Storage Spent Fuel Facility (SFIS). By 15 July 2008, there were 8413 assemblies in SFIS. The objectives are: 1) Wet AR storage of spent fuel from the NPP Bohunice and Mochovce: Surveillance of conditions for spent fuel storage in the at-reactor (AR) storage pools of both NPP's (characteristics of pool water, corrosion product data); Visual control of storage pool components; Evaluation of storage conditions with respect to long-term stability (corrosion of fuel cladding, structural materials); 2) Wet SFIS storage at Bohunice: Measurement of spent fuel conditions during the long-term wet storage, activity data in the storage casks and amount of crud; Surveillance program for SFIS structural materials.

  7. Conceptual design of an interim dry storage system for the Atucha nuclear power plant spent fuels

    International Nuclear Information System (INIS)

    Nassini, Horacio E.P.; Fuenzalida Troyano, C.S.; Bevilacqua, Arturo M.; Bergallo, Juan E.

    2005-01-01

    The Atucha I nuclear power station, after completing the rearrangement and consolidation of the spent fuels in the two existing interim wet storage pools, will have enough room for the storage of spent fuel from the operation of the reactor till December 2014. If the operation is extended beyond 2014, or if the reactor is decommissioned, it will be necessary to empty both pools and to transfer the spent fuels to a dry storage facility. This paper shows the progress achieved in the conceptual design of a dry storage system for Atucha I spent fuels, which also has to be adequate, without modifications, for the storage of fuels from the second unity of the nuclear power station, Atucha II, that is now under construction. (author) [es

  8. Spent fuel handling and storage facility for an LWR fuel reprocessing plant

    International Nuclear Information System (INIS)

    Baker, W.H.; King, F.D.

    1979-01-01

    The facility will have the capability to handle spent fuel assemblies containing 10 MTHM/day, with 30% if the fuel received in legal weight truck (LWT) casks and the remaining fuel received in rail casks. The storage capacity will be about 30% of the annual throughput of the reprocessing plant. This size will provide space for a working inventory of about 50 days plant throughput and empty storage space to receive any fuel that might be in transit of the reprocessing plant should have an outage. Spent LWR fuel assemblies outside the confines of the shipping cask will be handled and stored underwater. To permit drainage, each water pool will be designed so that it can be isolated from the remaining pools. Pool water quality will be controlled by a filter-deionizer system. Radioactivity in the water will be maintained at less than or equal to 2 x 10 -4 Ci/m 3 ; conductivity will be maintained at 1 to 2 μmho/cm. The temperature of the pool water will be maintained at less than or equal to 40 0 C to retard algae growth and reduce evaporation. Decay heat will be transferred to the environment via a heat exchanger-cooling tower system

  9. The prospects for dry fuel storage

    International Nuclear Information System (INIS)

    Harris, G.G.; Elliott, D.

    1994-01-01

    Dry storage of spent nuclear fuels is one method of dealing with radioactive waste. This article reports from a one day seminar on future prospects for dry fuel storage held in November 1993. Dry storage in an inert gas or air environment in vaults or casks, is an alternative to wet storage in water-filled ponds. Both wet and dry storage form part of the Interim Storage option for radioactive waste materials, and form alternatives to reprocessing or direct disposal in a deep repository. It has become clear that a large market for dry fuel storage will exist in the future. It will therefore be necessary to ensure that the various technical, safety, commercial, legislative and political constraints associated with it can be met effectively. (UK)

  10. Foreign programs for the storage of spent nuclear power plant fuels, high-level waste canisters and transuranic wastes

    International Nuclear Information System (INIS)

    Harmon, K.M.; Johnson, A.B. Jr.

    1984-04-01

    The various national programs for developing and applying technology for the interim storage of spent fuel, high-level radioactive waste, and TRU wastes are summarized. Primary emphasis of the report is on dry storage techniques for uranium dioxide fuels, but data are also provided concerning pool storage

  11. Safety Aspects of Long Term Spent Fuel Dry Storage

    International Nuclear Information System (INIS)

    Botsch, Wolfgang; Smalian, S.; Hinterding, P.; Drotleff, H.; Voelzke, H.; Wolff, D.; Kasparek, E.

    2014-01-01

    As a consequence of the lack of a final repository for spent nuclear fuel (SF) and high level waste (HLW), long term interim storage of SF and HLW will be necessary. As with the storage of all radioactive materials, the long term storage of SF and HLW must conform to safety requirements. Safety aspects such as safe enclosure of radioactive materials, safe removal of decay heat, sub-criticality and avoidance of unnecessary radiation exposure must be achieved throughout the complete storage period. 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. After the events of Fukushima, the advantages of passively and inherently safe dry storage systems have become more obvious. In Germany, dry storage of SF in casks fulfils both transport and storage requirements. Mostly, storage facilities are designed as concrete buildings above the ground; one storage facility has also been built as a rock tunnel. In all these facilities the safe enclosure of radioactive materials in dry storage casks is achieved by a double-lid sealing system with surveillance of the sealing system. The safe removal of decay heat is ensured by the design of the storage containers and the storage facility, which also secures to reduce the radiation exposure to acceptable levels. TUV and BAM, who work as independent experts for the competent authorities, inform about spent fuel management and issues concerning dry storage of spent nuclear fuel, based on their long experience in these fields. All relevant safety issues such as safe enclosure, shielding, removal of decay heat and sub-criticality are checked and validated with state-of-the-art methods and computer codes before the license approval. In our presentation we discuss which of these aspects need to be examined closer for a long term interim storage. It is shown

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  13. A central spent fuel storage in Sweden

    International Nuclear Information System (INIS)

    Gustafsson, B.; Hagberth, R.

    1978-01-01

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

  14. Management and storage of spent fuel from CEA research reactors

    International Nuclear Information System (INIS)

    Merchie, F.

    1996-01-01

    CEA research reactors and their interim spent fuel storage facilities are described. Long-term solutions for spent fuel storage problems, involving wet storage at PEGASE or dry storage at CASCAD, are outlined in some detail. (author)

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

    International Nuclear Information System (INIS)

    Takats, F.

    1999-01-01

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

  16. Concrete storage cask for interim storage of spent nuclear fuel

    International Nuclear Information System (INIS)

    Nabemoto, Toyonobu; Fujiwara, Hiroaki; Kobayashi, Shunji; Shionaga, Ryosuke

    2004-01-01

    Experiments and analytical evaluation of the fabrication, non-destructive inspection and structural integrity of reinforced concrete body for storage casks were carried out to demonstrate the concrete storage cask for spent fuel generated from nuclear power plants. Analytical survey on the type of concrete material and fabrication method of the storage cask was performed and the most suitable fabrication method for the concrete body was identified to reduce concrete cracking. The structural integrity of the concrete body of the storage cask under load conditions during storage was confirmed and the long term integrity of concrete body against degradation dependent on environmental factors was evaluated. (author)

  17. Hydrogen storage and integrated fuel cell assembly

    Science.gov (United States)

    Gross, Karl J.

    2010-08-24

    Hydrogen is stored in materials that absorb and desorb hydrogen with temperature dependent rates. A housing is provided that allows for the storage of one or more types of hydrogen-storage materials in close thermal proximity to a fuel cell stack. This arrangement, which includes alternating fuel cell stack and hydrogen-storage units, allows for close thermal matching of the hydrogen storage material and the fuel cell stack. Also, the present invention allows for tailoring of the hydrogen delivery by mixing different materials in one unit. Thermal insulation alternatively allows for a highly efficient unit. Individual power modules including one fuel cell stack surrounded by a pair of hydrogen-storage units allows for distribution of power throughout a vehicle or other electric power consuming devices.

  18. Initial evaluation of dry storage issues for spent nuclear fuels in wet storage at the Idaho Chemical Processing Plant

    Energy Technology Data Exchange (ETDEWEB)

    Guenther, R J; Johnson, Jr, A B; Lund, A L; Gilbert, E R [and others

    1996-07-01

    The Pacific Northwest Laboratory has evaluated the basis for moving selected spent nuclear fuels in the CPP-603 and CPP-666 storage pools at the Idaho Chemical Processing Plant from wet to dry interim storage. This work is being conducted for the Lockheed Idaho Technologies Company as part of the effort to determine appropriate conditioning and dry storage requirements for these fuels. These spent fuels are from 22 test reactors and include elements clad with aluminum or stainless steel and a wide variety of fuel materials: UAl{sub x}, UAl{sub x}-Al and U{sub 3}O{sub 8}-Al cermets, U-5% fissium, UMo, UZrH{sub x}, UErZrH, UO{sub 2}-stainless steel cermet, and U{sub 3}O{sub 8}-stainless steel cermet. The study also included declad uranium-zirconium hydride spent fuel stored in the CPP-603 storage pools. The current condition and potential failure mechanisms for these spent fuels were evaluated to determine the impact on conditioning and dry storage requirements. Initial recommendations for conditioning and dry storage requirements are made based on the potential degradation mechanisms and their impacts on moving the spent fuel from wet to dry storage. Areas needing further evaluation are identified.

  19. Initial evaluation of dry storage issues for spent nuclear fuels in wet storage at the Idaho Chemical Processing Plant

    International Nuclear Information System (INIS)

    Guenther, R.J.; Johnson, A.B. Jr.; Lund, A.L.; Gilbert, E.R.

    1994-11-01

    The Pacific Northwest Laboratory has evaluated the basis for moving selected spent nuclear fuels in the CPP-603 and CPP-666 storage pools at the Idaho Chemical Processing Plant from wet to dry interim storage. This work is being conducted for the Lockheed Idaho Technologies Company as part of the effort to determine appropriate conditioning and dry storage requirements for these fuels. These spent fuels are from 22 test reactors and include elements clad with aluminum or stainless steel and a wide variety of fuel materials: UAl x , UAl x -Al and U 3 O 8 -Al cermets, U-5% fissium, UMo, UZrH x , UErZrH, UO 2 -stainless steel cermet, and U 3 O 8 -stainless steel cermet. The study also included declad uranium-zirconium hydride spent fuel stored in the CPP-603 storage pools. The current condition and potential failure mechanisms for these spent fuels were evaluated to determine the impact on conditioning and dry storage requirements. Initial recommendations for conditioning and dry storage requirements are made based on the potential degradation mechanisms and their impacts on moving the spent fuel from wet to dry storage. Areas needing further evaluation are identified

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

  1. Criticality safety assessment of a TRIGA reactor spent-fuel pool under accident conditions

    International Nuclear Information System (INIS)

    Glumac, B.; Ravnik, M.; Logar, M.

    1997-01-01

    Additional criticality safety analysis of a pool-type storage for TRIGA spent fuel at the Jozef Stefan Institute in Ljubljana, Slovenia, is presented. Previous results have shown that subcriticality is not guaranteed for some postulated accidents (earthquake with subsequent fuel rack disintegration resulting in contact fuel pitch) under the assumption that the fuel rack is loaded with fresh 12 wt% standard fuel. To mitigate this deficiency, a study was done on replacing a certain number of fuel elements in the rack with cadmium-loaded absorber rods. The Monte Carlo computer code MCNP4A with an ENDF/B-V library and detailed three-dimensional geometrical model of the spent-fuel rack was used for this purpose. First, a minimum critical number of fuel elements was determined for contact pitch, and two possible geometries of rack disintegration were considered. Next, it was shown that subcriticality can be ensured when pitch is decreased from a rack design pitch of 8 cm to contact, if a certain number of fuel elements (8 to 20 out of 70) are replaced by absorber rods, which are uniformly mixed into the lattice. To account for the possibility that random mixing of fuel elements and absorber rods can occur during rack disintegration and result in a supercritical configuration, a probabilistic study was made to sample the probability density functions for random absorber rod lattice loadings. Results of the calculations show that reasonably low probabilities for supercriticality can be achieved (down to 10 -6 per severe earthquake, which would result in rack disintegration and subsequent maximum possible pitch decrease) even in the case where fresh 12 wt% standard TRIGA fuel would be stored in the spent-fuel pool

  2. Storage of spent fuel from power reactors in India management and experience

    International Nuclear Information System (INIS)

    Changrani, R.D.; Bajpai, D.D.; Kodilkar, S.S.

    1999-01-01

    The spent fuel management programme in India is based on closing the nuclear fuel cycle with reprocessing option. This will enable the country to enhance energy security through maximizing utilization of available limited uranium resources while pursuing its Three Stage Nuclear Power Programme. Storage of spent fuel in water pools remains as prevailing mode in the near term. In view of inventory build up of spent fuel, an Away-From-Reactor (AFR) On-Site (OS) spent fuel storage facility has been made operational at Tarapur. Dry storage casks also have been developed as 'add on' system for additional storage of spent fuels. The paper describes the status and experience pertaining to spent fuel storage practices in India. (author)

  3. Past experience and future needs for the use of burnup credit in LWR fuel storage

    International Nuclear Information System (INIS)

    Boyd, W.A.; Wrights, G.N.

    1987-01-01

    To achieve improved fuel economics and reduce the amount of fuel discharged annually, utilities are engaging in fuel management strategies that will achieve higher discharge burnups for their fuel assemblies. Although burnup credit methodologies have been developed and spent-fuel racks have been licensed, burnup credit fuel storage racks are not the answer for all utilities. Off-site and out-of-pool spent-fuel storage may be more appropriate. This is leading to the development of dry spent-fuel storage and shipping casks. Cask designs with spent-fuel storage capability between 20 and 32 assemblies are being developed by several vendors. The US Dept. of Energy is also funding work by VEPCO. Westinghouse is currently licensing its dry storage cask, developing a shipping cask for the domestic market, and is involved in a joint venture to develop a cask for the international market. Although methods of taking credit for fuel burnup in spent-fuel storage racks have been developed and licensed, use of these methods on dry spent-fuel storage and shipping casks can lead to new issues. These issues arise because the excess reactivity margin that is inherent in a burnup credit spent-fuel storage rack criticality analysis will not be available in a dry cask analysis

  4. Economics of spent LWR fuel storage

    International Nuclear Information System (INIS)

    Clark, H.J.; O'Neill, G.F.

    1980-01-01

    A power reactor operator, confronted with rising spent fuel inventories that would soon exceed his storage capacity, has to decide what to do with this fuel if he wants to continue reactor operations. A low cost option would be to ship excess fuel from the overburdened reactor to another reactor in the utility's system that has available space. The only cost would be for cask leasing and shipping. Three other alternatives all require considerable capital expenditures: reracking, new at-reactor (AR) basins for storage, and away-from-reactor (AFR) basins for storage. Economic considerations for each of the alternatives are compared

  5. Storage arrangement for nuclear reactor fuel assemblies

    International Nuclear Information System (INIS)

    Wade, E.E.

    1977-01-01

    Said invention is intended for providing an arrangement of spent fuel assembly storage inside which the space is efficiently used without accumulating a critical mass. The storage is provided for long fuel assemblies having along their longitudinal axis an active part containing the fuel and an inactive part empty of fuel. Said storage arrangement comprises a framework constituting some long-shaped cells designed so as each of them can receive a fuel assembly. Means of axial positioning of said assembly in a cell make it possible to support the fuel assemblies inside the framework according to a spacing ratio, along the cell axis, such as the active part of an assembly is adjacent to the inactive part of the adjacent assemblies [fr

  6. Calculation of spent fuel pool severe accident with MELCOR

    International Nuclear Information System (INIS)

    Deng Jian; Xiang Qing'an; Zhou Kefeng

    2014-01-01

    A calculation model was established for spent fuel pool (SFP) using MELCOR code to study the severe accident phenomena caused by the long term station black-out (SBO), including spent fuel heatup, zirconium cladding oxidation, and the injection into SFP to mitigate the severe accident. The results show that the severe accident progression is slow and relates directly with the initial water level in SFP. It is illustrated that the injection into SFP is one of the best mitigated measures for the SFP severe accident. (authors)

  7. Design considerations and operating experience with wet storage of Ontario Hydro's irradiated fuel

    International Nuclear Information System (INIS)

    Frost, C.R.; Naqvi, S.J.; McEachran, R.A.

    1987-01-01

    The characteristics of Ontario Hydro's fuel and at-reactor irradiated fuel storage water pools (or irradiated fuel bays) are described. There are two types of bay known respectively as primary bays and auxiliary bays, used for at-reactor irradiated fuel storage. Irradiated fuel is discharged remotely from Ontario Hydro's reactors to the primary bays for initial storage and cooling. The auxiliary bays are used to receive and store fuel after its initial cooling in the primary bay, and provide additional storage capacity as needed. The major considerations in irradiated fuel bay design, including site-specific requirements, reliability and quality assurance, are discussed. The monitoring of critical fuel bay components, such as bay liners, the development of high storage density fuel containers, and the use of several irradiated fuel bays at each reactor site have all contributed to the safe handling of the large quantities of irradiated fuel over a period of about 25 years. Routine operation of the irradiated fuel bays and some unusual bay operational events are described. For safety considerations, the irradiated fuel in storage must retain its integrity. Also, as fuel storage is an interim process, likely for 50 years or more, the irradiated fuel should be retrievable for downstream fuel management phases such as reprocessing or disposal. A long-term experimental program is being used to monitor the integrity of irradiated fuel in long-term wet storage. The well characterized fuel, some of which has been in wet storage since 1962 is periodically examined for possible deterioration. The evidence from this program indicates that there will be no significant change in irradiated fuel integrity (and retrievability) over a 50 year wet storage period

  8. Activity release during the dry storage of fuel assemblies

    International Nuclear Information System (INIS)

    Valentine, M.K.; Fettel, W.; Gunther, H.

    1991-01-01

    This paper reports that wet storage is the predominant storage method in the USA for spent fuel assemblies. Nevertheless, most utilities have stretched their storage capacities and several reactors will lose their full-core reserve in the 90's. A great variety of out-of-pool storage methods already exist, including the FUELSTOR vault-type dry storage concept. A FUELSTOR vault relies on double containment of the spent fuel (intact cladding as the primary containment and sealing of assemblies in canisters filled with an inert gas as the secondary containment) to reduce radiation levels at the outside wall of the vault to less than site boundary levels. Investigation of accident scenarios reveals that radiation release limits are only exceeded following complete failure of all canisters and simultaneous cladding breach for more than 40% of the rods (or for more than 1% of failed rods if massive fuel oxidation occurs following cladding failure). Such failures are considered highly improbable. Thus, it can be concluded that this type of dry storage is safe and individual canister monitoring is not required in the facility

  9. 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)

  10. Benchmarking criticality analysis of TRIGA fuel storage racks.

    Science.gov (United States)

    Robinson, Matthew Loren; DeBey, Timothy M; Higginbotham, Jack F

    2017-01-01

    A criticality analysis was benchmarked to sub-criticality measurements of the hexagonal fuel storage racks at the United States Geological Survey TRIGA MARK I reactor in Denver. These racks, which hold up to 19 fuel elements each, are arranged at 0.61m (2 feet) spacings around the outer edge of the reactor. A 3-dimensional model was created of the racks using MCNP5, and the model was verified experimentally by comparison to measured subcritical multiplication data collected in an approach to critical loading of two of the racks. The validated model was then used to show that in the extreme condition where the entire circumference of the pool was lined with racks loaded with used fuel the storage array is subcritical with a k value of about 0.71; well below the regulatory limit of 0.8. A model was also constructed of the rectangular 2×10 fuel storage array used in many other TRIGA reactors to validate the technique against the original TRIGA licensing sub-critical analysis performed in 1966. The fuel used in this study was standard 20% enriched (LEU) aluminum or stainless steel clad TRIGA fuel. Copyright © 2016. Published by Elsevier Ltd.

  11. Storage and Reprocessing of Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-02-02

    Addressing the problem of waste, especially high-level waste (HLW), is a requirement of the nuclear fuel cycle that cannot be ignored. We explore the two options employed currently, long-term storage and reprocessing.

  12. Choosing a spent fuel interim storage system

    International Nuclear Information System (INIS)

    Roland, V.; Hunter, I.

    2001-01-01

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

  13. Experience on wet storage spent fuel sipping at IEA-R1 Brazilian research reactor

    International Nuclear Information System (INIS)

    Perrotta, J.A.; Terremoto, L.A.A.; Zeituni, C.A.

    1998-01-01

    The IEA-R1 research reactor of the Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP) is a pool type reactor of B and W design, that has been operating since 1957 at a power of 2 MW. Irradiated (spent) fuels have been stored at the facility during the various years of operation. At present there are 40 spent fuel assemblies at dry storage, 79 spent fuel assemblies at wet storage and 30 fuel assemblies in the core. The oldest fuels are of United States origin, made with U-Al alloy, both of LEU and HEU MTR fuel type. Many of these fuel assemblies have corrosion pits along their lateral fuel plates. These pits originate by galvanic corrosion between the fuel plate and the stainless steel storage racks. As a consequence of the possibility of sending the irradiated old fuels back the U.S.A., sipping tests were performed with the spent fuel assemblies. The reason for this was to evaluate their 137 Cs leaking rate, if any. This work describes the procedure and methodology used to perform the sipping tests with the fuel assemblies at the storage pool, and presents the results obtained for the 137 Cs sipping water activity for each fuel assembly. A correlation is made between the corrosion pits and the activity values measured. A 137 Cs leaking rate is determined and compared to the criteria established for canning spent fuel assemblies before shipment

  14. Spent fuel storage rack for atomic power plant

    International Nuclear Information System (INIS)

    Kodama, Tatemitsu.

    1981-01-01

    Purpose: To flexibly cope with the changes in the size and shape of spent fuel storage containers by placing a number of independently-constructed rack cells in a rack frame in such a manner that the guide support members of the storage rack, mounted on each rack cell may be replaced. Constitution: Independently-constructed rack cells are inserted from above into a rack frame rigidly installed on the bottom of a water pool. Each cell is produced by welding, has a handling head mounted at the top, and guide support members made of three replaceable guide tubes are mounted with bolts. If the size and the shape of the containers are altered, this configuration can easily cope with the new container shape by merely having the guide tubes replaced, without adversely affecting other cells and without necessitating draining of the water in the pool. (Yoshino, Y.)

  15. Dry storage of spent nuclear fuel in UAE – Economic aspect

    International Nuclear Information System (INIS)

    Al Saadi, Sara; Yi, Yongsun

    2015-01-01

    Highlights: • Cost analysis of interim storage of spent nuclear fuel in the UAE was performed. • Two scenarios were considered: accelerated transfer of SNF and max. use of fuel pool. • Additional cost by accelerated transfer of SNF to dry storage was not significant. • Multiple regression analysis was applied to the resulting dry storage costs. • Dry storage costs for different cases could be expressed by single equations. - Abstract: Cost analysis of dry storage of spent nuclear fuel (SNF) discharged from Barakah nuclear power plants in the UAE was performed using three variables: average fuel discharge rate (FD), discount rate (d), and cooling time in a spent fuel pool (T cool ). The costs of dry storage as an interim spent fuel storage option in the UAE were estimated and compared between the following two scenarios: Scenario 1 is ‘accelerated transfer of spent fuel to dry storage’ that SNF will be transferred to dry storage facilities as soon as spent fuel has been sufficiently cooled down in a pool for the dry storage; Scenario 2 is defined as ‘maximum use of spent fuel pool’ that SNF will be stored in a pool as long as possible till the amount of stored SNF in the pool reaches the capacity of the pools and, then, to be moved to dry storage. A sensitivity analysis on the costs was performed and multiple regression analysis was applied to the resulting net present values (NPVs) for Scenarios 1 and 2 and ΔNPV that is difference in the net present values between the two scenarios. The results showed that NPVs and ΔNPV could be approximately expressed by single equations with the three variables. Among the three variables, the discount rate had the largest effect on the NPVs of the dry storage costs. However, ΔNPV was turned out to be equally sensitive to the discount rate and cooling period. Over the ranges of the variables, the additional cost for accelerated fuel transfer (Scenario 1) ranged from 86.4 to 212.9 million $. Calculated using

  16. Dry well storage of spent LWBR fuel

    International Nuclear Information System (INIS)

    Christensen, A.B.; Fielding, K.D.

    1985-01-01

    Recently, 50 dry wells were constructed at the Idaho Chemical Processing Plant (ICPP) to temporarily store the Light Water Breeder Reactor (LWBR) fuel. Over 400 dry wells of the same design are projected to be constructed in the next 5 yr at the ICPP to store unreprocessible fuels until a permanent repository becomes available. This summary describes the LWBR fuel storage dry wells and the enhancements made over the Peach Bottom fuel and Fermi blanket dry wells that have been in use for up to 4 yr. Dry well storage at the ICPP has historically been found to be a safe and efficient method of temporary fuel storage. The LWBR dry wells should be more reliable than the original dry wells and provide data not previously available

  17. Interim licensing criteria for physical protection of certain storage of spent fuel

    International Nuclear Information System (INIS)

    Dwyer, P.A.

    1994-11-01

    This document presents interim criteria to be used in the physical protection licensing of certain spent fuel storage installations. Installations that will be reviewed under this criteria are those that store power reactor spent fuel at decommissioned power reactor sites; independent spent fuel storage installations located outside of the owner controlled area of operating nuclear power reactors; monitored retrievable storage installations owned by the Department of Energy, designed and constructed specifically for the storage, of spent fuel; the proposed geologic repository operations area; or permanently shutdown power reactors still holding a Part 50 license. This criteria applies to both dry cask and pool storage. However, the criteria in this document does not apply to the storage of spent fuel within the owner-controlled area of operating nuclear power reactors

  18. Design ampersand operational experience of the NUHOMS reg-sign-24P spent fuel storage system

    International Nuclear Information System (INIS)

    McConaghy, W.J.; Lehnert, R.A.; Rasmussen, R.W.

    1991-01-01

    The NUHOMS reg-sign Spent Fuel Storage System provides a safe and economical method for the dry storage of spent fuel assemblies either at an at-reactor Independent Spent Fuel Storage Installation (ISFSI) or at a centralized away-from-reactor (AFR) storage facility. The system consists of three major safety related components: a dry shielded canister (DSC) which provides a high integrity containment boundary and a controlled storage environment for the fuel; a reinforced concrete horizontal storage module (HSM) which houses the stored DSC and provides radiation shielding, protection against natural phenomena, and an efficient means for decay heat removal; and a transfer cask which provides for the safe shielded transfer of the DSC from the plant spent fuel pool to the HSM. The NUHOMS reg-sign system is designed and licensed to the requirements of 10 CFR 72 and ANS/ANSI 57.9 for ISFSIs

  19. Method of assembling spent nuclear fuel storage rack

    International Nuclear Information System (INIS)

    Igarashi, Ryokichi; Hasegawa, Hidenobu.

    1982-01-01

    Purpose: To improve the safety of a spent fuel storage rack by stably installing the spent fuel in a pool without using supporting beams. Constitution: A restricted unit is composed of a plurality of spuare cylinders. A plurality of such restricted units are aligned in a direction perpendicularly to the arraying direction of the cylinders in the respective restricted units, are coupled with long connecting plates, and are fixed by welding on a common small base, thereby forming a restricted body. According to such assembling method, a plurality of restricted bodies are connected in a direction that the respective restricted bodies are readily overturned, and are secured to the common base. Accordingly, the restricted bodies can be stably installed in a pool without using supporting beams as the conventional one. (Sekiya, K.)

  20. Corrosion in ICPP fuel storage basins

    International Nuclear Information System (INIS)

    Dirk, W.J.

    1993-09-01

    The Idaho Chemical Processing Plant currently stores irradiated nuclear fuel in fuel storage basins. Historically, fuel has been stored for over 30 years. During the 1970's, an algae problem occurred which required higher levels of chemical treatment of the basin water to maintain visibility for fuel storage operations. This treatment led to higher levels of chlorides than seen previously which cause increased corrosion of aluminum and carbon steel, but has had little effect on the stainless steel in the basin. Corrosion measurements of select aluminum fuel storage cans, aluminum fuel storage buckets, and operational support equipment have been completed. Aluminum has exhibited good general corrosion rates, but has shown accelerated preferential attack in the form of pitting. Hot dipped zinc coated carbon steel, which has been in the basin for approximately 40 years, has shown a general corrosion rate of 4 mpy, and there is evidence of large shallow pits on the surface. A welded Type 304 stainless steel corrosion coupon has shown no attack after 13 years exposure. Galvanic couples between carbon steel welded to Type 304 stainless steel occur in fuel storage yokes exposed to the basin water. These welded couples have shown galvanic attack as well as hot weld cracking and intergranular cracking. The intergranular stress corrosion cracking is attributed to crevices formed during fabrication which allowed chlorides to concentrate

  1. Comprehensive studies on regulatory issues of spent fuel pools

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    An existence of safety issues in the spent fuel pool (SFP) was recognized by the nuclear accident at the Fukushima Daiichi Nuclear Power Station, and many reports on the accident describe needs of countermeasures for SFP under sever accidents. For research planning, thermal hydraulic behaviors of SFP and possibility of occurrence of re-criticality conditions in SFP were studied by computational approaches. In the studies on thermal hydraulic behaviors, possibilities of adiabatic conditions in a spent fuel bundle were identified because natural circulation cooling of air could be terminated due to flow path blockage by pool water and steam cooling could be terminated due to reduction of pool water evaporation originated from cold water injection by emergency water supply. In the re-criticality study, in the case of the un-borated lack, it was shown that the neutron multiplication factor became larger than unity when the difference of water levels inside and outside the channel box larger than some values. (author)

  2. Integrated spent fuel storage and transportation system using NUHOMS

    International Nuclear Information System (INIS)

    Lehnert, R.; McConaghy, W.; Rosa, J.

    1990-01-01

    As utilities with nuclear power plants face increasing near term spent fuel store needs, various systems for dry storage such as the NUTECH Horizontal Modular Storage (NUHOMS) system are being implemented to augment existing spent fuel pool storage capacities. These decisions are based on a number of generic and utility specific considerations including both short term and long term economics. Since the US Department of Energy (DOE) is tasked by the Nuclear Waste Policy Act with the future responsibility of transporting spent fuel from commercial nuclear power plants to a Monitored Retrievable Storage (MRS) facility anchor a permanent geologic repository, the interfaces between the utilities at-reactor dry storage system and the DOE's away-from-reactor transportation system become important. This paper presents a study of the interfaces between the current at-reactor NUHOMS system and the future away-from-reactor DOE transportation system being developed under the Office of Civilian Radioactive Waste Management (OCRWM) program. 7 refs., 9 figs., 1 tab

  3. The shutdown reactor: Optimizing spent fuel storage cost

    International Nuclear Information System (INIS)

    Pennington, C.W.

    1995-01-01

    Several studies have indicated that the most prudent way to store fuel at a shutdown reactor site safely and economically is through the use of a dry storage facility licensed under 10CFR72. While such storage is certainly safe, is it true that the dry ISFSI represents the safest and most economical approach for the utility? While no one is really able to answer that question definitely, as yet, Holtec has studied this issue for some time and believes that both an economic and safety case can be made for an optimization strategy that calls for the use of both wet and dry ISFSI storage of spent fuel at some plants. For the sake of brevity, this paper summarizes some of Holtec's findings with respect to the economics of maintaining some fuel in wet storage at a shutdown reactor. The safety issue, or more importantly the perception of safety of spent fuel in wet storage, still varies too much with the eye of the beholder, and until a more rigorous presentation of safety analyses can be made in a regulatory setting, it is not practically useful to argue about how many angels can sit on the head of a safety-related pin. Holtec is prepared to present such analyses, but this does not appear to be the proper venue. Thus, this paper simply looks at certain economic elements of a wet ISFSI at a shutdown reactor to make a prima facie case that wet storage has some attractiveness at a shutdown reactor and should not be rejected out of hand. Indeed, an optimization study at certain plants may well show the economic vitality of keeping some fuel in the pool and converting the NRC licensing coverage from 10CFR50 to 10CFR72. If the economics look attractive, then the safety issue may be confronted with a compelling interest

  4. Biofouling on austenitic stainless steels in spent nuclear fuel pools

    Energy Technology Data Exchange (ETDEWEB)

    Sarro, M I; Moreno, D A; Chicote, E; Lorenzo, P I; Garcia, A M [Universidad Politecnica de Madrid, Departamento de Ingenieria y Ciencia de los Materiales, Escuela Tecnica Superior de Ingenieros Industriales, Jose Gutierrez Abascal, 2, E-28006 Madrid (Spain); Montero, F [Iberdrola Generacion, S.A., y C.M.D.S., Centro de Tecnologia de Materiales, Paseo de la Virgen del Puerto, 53, E-28005 Madrid (Spain)

    2003-07-01

    The objective of this study was to investigate the biofilm formation on three different types of austenitic stainless steel (UNS S30400, S30466 and S31600) submerged in a spent nuclear fuel pool. The presence of microorganisms in coupons was characterised using standard culture microbiological methods, microscopic techniques (epifluorescence microscopy and scanning electron microscopy), and molecular biology techniques (denaturing gradient gel electrophoresis and sequencing fragments of 16S rDNA). The microscopy techniques showed signs of colonisation of stainless steels in spite of these extreme conditions. Based on sequencing of cultured microorganisms, different bacteria belonging to {alpha}, {beta}, {gamma}-Proteobacteria, Bacilli, and Actinobacteria classes have been identified. The biofilm radioactivity was measured using gamma-ray spectrometry and, according to the data gathered, the radionuclides present in the water pool were entrapped in the biofilm increasing the amount of radiation at the surface of the different materials. (Abstract Copyright [2003], Wiley Periodicals, Inc.)

  5. Biofouling on austenitic stainless steels in spent nuclear fuel pools

    International Nuclear Information System (INIS)

    Sarro, M.I.; Moreno, D.A.; Chicote, E.; Lorenzo, P.I.; Garcia, A.M.; Montero, F.

    2003-01-01

    The objective of this study was to investigate the biofilm formation on three different types of austenitic stainless steel (UNS S30400, S30466 and S31600) submerged in a spent nuclear fuel pool. The presence of microorganisms in coupons was characterised using standard culture microbiological methods, microscopic techniques (epifluorescence microscopy and scanning electron microscopy), and molecular biology techniques (denaturing gradient gel electrophoresis and sequencing fragments of 16S rDNA). The microscopy techniques showed signs of colonisation of stainless steels in spite of these extreme conditions. Based on sequencing of cultured microorganisms, different bacteria belonging to α, β, γ-Proteobacteria, Bacilli, and Actinobacteria classes have been identified. The biofilm radioactivity was measured using gamma-ray spectrometry and, according to the data gathered, the radionuclides present in the water pool were entrapped in the biofilm increasing the amount of radiation at the surface of the different materials. (Abstract Copyright [2003], Wiley Periodicals, Inc.)

  6. The cascad spent fuel dry storage facility

    International Nuclear Information System (INIS)

    Guay, P.; Bonnet, C.

    1991-01-01

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

  7. Design of spent fuel storage facilities

    International Nuclear Information System (INIS)

    1994-01-01

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

  8. Operational experience in the spent fuel receipt and storage facility at the Tokai Reprocessing Plant

    International Nuclear Information System (INIS)

    Nakashima, S.; Yamaguchi, Y.; Iimura, I.; Yamamura, O.; Ogata, Y.

    1992-01-01

    The development of the double containment system led to the reduction of labor time for the cask decontamination to one-tenth compared to the original manner. And also it led to the great decrease of floor contamination in the receipt and storage facility. The decrease permitted as many as about 20,000 visitors to take tours in the fuel receipt and storage facility in the past three years without contamination trouble with the visitors. Different types of spent fuels can be easily handled and stored by the specially designed tools in the pool water. The exchange of the cooling water in the transport cask before unloading and the use of the storage container keep contamination of the pool water to a minimum. The pool water treatment system has been successfully operated. As result, the pool water condition has been well-controlled

  9. Gasket structure improvement for the spent fuel pool cooler

    International Nuclear Information System (INIS)

    Li Yun; He Shaohua; Qi Hongchang; Wang Cong; Wang Chenglin; Zhong Boling

    2014-01-01

    The two spent fuel pool coolers for the 320 MW unit in CNNC Nuclear Power Operation Management Co., Ltd. have operated for more than 20 years. In accordance with the preventive maintenance programs, they must be overhauled. It is decided to improve the original gasket structure of the component and adopt the method of a short-length U-tubes pulling after analysis and study. There are no leakages and other abnormal situations after the equipment being put into operation. The unit is kept safe and stable. At the same time, thought and method for the maintenance of other similar equipment are provided. (authors)

  10. Environmental Assessment: Relocation and storage of TRIGA reg-sign reactor fuel, Hanford Site, Richland, Washington

    International Nuclear Information System (INIS)

    1995-08-01

    In order to allow the shutdown of the Hanford 308 Building in the 300 Area, it is proposed to relocate fuel assemblies (101 irradiated, three unirradiated) from the Mark I TRIGA Reactor storage pool. The irradiated fuel assemblies would be stored in casks in the Interim Storage Area in the Hanford 400 Area; the three unirradiated ones would be transferred to another TRIGA reactor. The relocation is not expected to change the offsite exposure from all Hanford Site 300 and 400 Area operations

  11. Transport and storage of spent nuclear fuel

    International Nuclear Information System (INIS)

    Lung, M.; Lenail, B.

    1987-01-01

    From a safety standpoint, spent fuel is clearly not ideal for permanent disposal and reprocessing is the best method of preparing wastes for long-term storage in a repository. Furthermore, the future may demonstrate that some fission products recovered in reprocessing have economic applications. Many countries have in fact reached the point at which the recycling of plutonium and uranium from spent fuel is economical in LWR's. Even in countries where this is not yet evident, (i.e., the United States), the French example shows that the day will come when spent fuel will be retrieved for reprocessing and recycle. It is highly questionable whether spent fuel will ever be considered and treated as waste in the same sense as fission products and processed as such, i.e., packaged in a waste form for permanent disposal. Even when recycled fuel material can no longer be reused in LWR's because of poor reactivity, it will be usable in FBR's. Based on the considerable experience gained by SGN and Cogema, this paper has provided practical discussion and illustrations of spent fuel transport and storage of a very important step in the nuclear fuel management process. The best of spent fuel storage depends on technical, economic and policy considerations. Each design has a role to play and we hope that the above discussion will help clarify certain issues

  12. Spent fuel storage requirements for nuclear utilities and OCRWM [Office of Civilian Radioactive Waste Management

    International Nuclear Information System (INIS)

    Wood, T.W.

    1990-03-01

    Projected spent fuel generation at US power reactors exceeds estimated aggregate pool storage capacity by approximately 30,000 metric tons of uranium (MTU). Based on the current repository schedule, little of the spent fuel inventory will be disposed of prior to shutdown of existing reactors, and a large additional capacity for surface storage of spent fuel will be required, either at reactors or at a centralized DOE storage site. Allocation of this storage requirement across the utility-DOE interface, and the resulting implications for reactor sites and the performance of the federal waste management system, were studied during the DOE MRS System Study and again subsequent to the reassessment of the repository schedule. Spent fuel logistics and cost results from these analyses will be used in definition of spent fuel storage capacity requirements for the federal system. 9 refs., 8 figs., 1 tab

  13. Role of transportation in the utilities' management of spent fuel storage

    International Nuclear Information System (INIS)

    Newman, D.F.

    1985-01-01

    Additional spent fuel storage can be provided by using a combination of wet and dry storage technologies, with the technology or technologies used in any specific instance being determined by the particular circumstances involved. The capability for spent fuel storage at a reactor site can be enhanced using any one or a combination of the following: expansion of existing pool storage capacity; more efficient use of available capacity; and addition of an independent spent fuel storage installation (ISFSI). Each of these methods, which are described more fully below, have characteristics that may make them more or less suitable for use, depending on the nuclear power plant where they will be deployed, the magnitude of the need for additional storage, the utility's overall spent fuel management strategy, and other factors. 15 refs., 2 figs., 2 tabs

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

    International Nuclear Information System (INIS)

    2003-01-01

    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

  15. Criticality and Its Uncertainty Analysis of Spent Fuel Storage Rack for Research Reactor

    International Nuclear Information System (INIS)

    Han, Tae Young; Park, Chang Je; Lee, Byung Chul

    2011-01-01

    For evaluating the criticality safety of spent fuel storage rack in an open pool type research reactor, a permissible upper limit of criticality should be determined. It can be estimated from the criticality upper limit presented by the regulatory guide and an uncertainty of criticality calculation. In this paper, criticalities for spent fuel storage rack are carried out at various conditions. The calculation uncertainty of MCNP system is evaluated from the calculation results for the benchmark experiments. Then, the upper limit of criticality is determined from the uncertainties and the calculated criticality of the spent fuel storage rack is evaluated

  16. Decommissioning the Dresden Unit 1 Spent Fuel Pool

    International Nuclear Information System (INIS)

    Demmer, R.L.; Bargelt, R.J.; Panozzo, J.B.; Christensen, R.J.

    2006-01-01

    The Dresden Nuclear Power Station, Unit 1 Spent Fuel Pool (SFP) (Exelon Generation Co.) was decommissioned using a new underwater coating strategy developed in cooperation with the Idaho National Laboratory (INL). This was the first time that a commercial nuclear power plant (NPP) SFP was decommissioned using this underwater coating approach. This approach has advantages in many aspects, particularly in reducing airborne contamination and in safer, more cost effective deactivation. The process was pioneered at the INL and used to decommission three SFPs with a total combined pool volume of over 900,000 gallons. The INL provided engineering support and shared project plans to successfully initiate the Dresden project. This report outlines the steps taken by the INL and Exelon on the pathway for this activity. The rationale used to select the underwater coating option and the advantages and disadvantages are shown. Special circumstances, such as the use of a remotely operated underwater vehicle to map (visually and radiologically) the pool areas that were not readily accessible, are discussed. Several specific areas where special equipment was employed are given and a lessons learned evaluation is included. (authors)

  17. Spent fuel pool cooling system upgrade for Kori Unit 1

    International Nuclear Information System (INIS)

    Sun Park, Jong; In Shin, Kyung

    2014-01-01

    Following Fukushima nuclear power plant accident, the needs for reliable performance of its own safety functions of Spent Fuel Pool Cooling System (SFPCS) has risen significantly to maintain the plant in a safe condition. Regulatory Guide 1.13 of United States Nuclear Regulatory Commission (USNRC) requires the SFPCS shall be designed safety related as Quality Group C and Seismic Category 1. However, the existing Spent Fuel Pool (SFP) of KORI Unit 1 was not designed as a safety system. In order to comply with the above licensing requirement for the extended operational life of KORI Unit 1, it has been decided to add a safety related Seismic Category 1 Makeup System to KORI Unit 1 and the existing SFPCS to be modified in dedicated channels with safety related equipment to enhance system's reliability as a means of providing diversity. This paper focuses on describing the relevant design requirements, applications, and supplemental facilities to the SFPCS of KORI Unit 1. (authors)

  18. Spent fuel dry storage in Hungary

    International Nuclear Information System (INIS)

    Buday, G.; Szabo, B.; Oerdoegh, M.; Takats, F.

    1999-01-01

    Paks Nuclear Power Plant is the only NPP in Hungary. It has four WWER-440 type reactor units. Since 1989, approximately 40-50% of the total annual electricity generation of the country has been supplied by this plant. The fresh fuel is imported from Russia. Most of the spent fuel assemblies have been shipped back to Russia. Difficulties with spent fuel transportation to Russia have begun in 1992. Since that time, some of the shipments were delayed, some of them were completely cancelled, thus creating a backlog of spent fuel filling all storage positions of the plant. To provide assurance of the continued operation, Paks NPPs management decided to implement an independent spent fuel storage facility and chose GEC-Althom's MVDS design. The construction of the facility started in February 1995 and the first spent fuel assembly was placed in the store in September 1997. The paper gives an overview of the situation, describing the conditions leading to the construction of the dry storage facility at Paks and its implementation. Finally, some information is given about the new Public Agency for Radioactive Waste Management established this year and responsible for managing the issues related to spent fuel management. (author)

  19. Leak testing fuel stored in the ICPP fuel storage basin

    International Nuclear Information System (INIS)

    Lee, J.L.; Rhodes, D.W.

    1977-06-01

    Irradiated fuel to be processed at the Idaho Chemical Processing Plant is stored under water at the CPP-603 Fuel Storage Facility. Leakage of radionuclides through breaks in the cladding of some of the stored fuels contaminates the water with radionuclides resulting in radiation exposure to personnel during fuel handling operations and contamination of the shipping casks. A leak test vessel was fabricated to test individual fuel assemblies which were suspected to be leaking. The test equipment and procedures are described. Test results demonstrated that a leaking fuel element could be identified by this method; of the eleven fuel assemblies tested, six were estimated to be releasing greater than 0.5 Ci total radionuclides/day to the basin water

  20. Spent-fuel-storage requirements: an update of DOE/RL-82-1

    International Nuclear Information System (INIS)

    1983-01-01

    Spent fuel storage requirements as projected through the year 2000 for US light water reactor (LWR) nuclear power plants were calculated using information supplied by the utilities reflecting plant status as of September 30, 1982. Projections through the year 2000 combined fuel discharge projections of the utilities with the assumed discharges of typical reactors required to meet the nuclear capacity of 132 gigawatts electrical (GWe) projected by the Energy Information Administration (EIA) for the year 2000. Three cases were developed and are summarized. A reference case, or maximum at-reactor (AR) capacity case, assumes that all reactor storage pools are increased to their maximum capacities, as estimated by the utilities, for spent fuel storage utilizing currently licensed technologies. Rod consolidation and dry storage technologies were not considered. The reference case assumes no transshipments between pools except as currently licensed by the Nuclear Regulatory Commission (NRC). This case identifies an initial requirement for 13 metric tons uranium (MTU) of additional storage in 1984, and a cumulative requirement for 13,090 MTU additional storage in the year 2000. The reference case is bounded by two alternative cases. One, a current capacity case, assumes that only those pool storage capacity increases currently planned by the operating utilities will occur. The second, or maximum capacity with transshipment case, assumes maximum development of pool storage capacity as described above and also assumes no constraints on transshipment of spent fuel among pools of reactors of like type (BWR) within a given utility. In all cases, a full core discharge capability is assumed to be maintained for each reactor, except that only one FCR is maintained when two reactors share a common pool. 1 figure, 12 tables

  1. HLM fuel pin bundle experiments in the CIRCE pool facility

    Energy Technology Data Exchange (ETDEWEB)

    Martelli, Daniele, E-mail: daniele.martelli@ing.unipi.it [University of Pisa, Department of Civil and Industrial Engineering, Pisa (Italy); Forgione, Nicola [University of Pisa, Department of Civil and Industrial Engineering, Pisa (Italy); Di Piazza, Ivan; Tarantino, Mariano [Italian National Agency for New Technologies, Energy and Sustainable Economic Development, C.R. ENEA Brasimone (Italy)

    2015-10-15

    Highlights: • The experimental results represent the first set of values for LBE pool facility. • Heat transfer is investigated for a 37-pin electrical bundle cooled by LBE. • Experimental data are presented together with a detailed error analysis. • Nu is computed as a function of the Pe and compared with correlations. • Experimental Nu is about 25% lower than Nu derived from correlations. - Abstract: Since Lead-cooled Fast Reactors (LFR) have been conceptualized in the frame of GEN IV International Forum (GIF), great interest has focused on the development and testing of new technologies related to HLM nuclear reactors. In this frame the Integral Circulation Experiment (ICE) test section has been installed into the CIRCE pool facility and suitable experiments have been carried out aiming to fully investigate the heat transfer phenomena in grid spaced fuel pin bundles providing experimental data in support of European fast reactor development. In particular, the fuel pin bundle simulator (FPS) cooled by lead bismuth eutectic (LBE), has been conceived with a thermal power of about 1 MW and a uniform linear power up to 25 kW/m, relevant values for a LFR. It consists of 37 fuel pins (electrically simulated) placed on a hexagonal lattice with a pitch to diameter ratio of 1.8. The FPS was deeply instrumented by several thermocouples. In particular, two sections of the FPS were instrumented in order to evaluate the heat transfer coefficient along the bundle as well as the cladding temperature in different ranks of sub-channels. Nusselt number in the central sub-channel was therefore calculated as a function of the Peclet number and the obtained results were compared to Nusselt numbers obtained from convective heat transfer correlations available in literature on Heavy Liquid Metals (HLM). Results reported in the present work, represent the first set of experimental data concerning fuel pin bundle behaviour in a heavy liquid metal pool, both in forced and

  2. Cost analysis methodology of spent fuel storage

    International Nuclear Information System (INIS)

    1994-01-01

    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

  3. Spent LWR fuel-storage costs

    International Nuclear Information System (INIS)

    Clark, H.J.

    1981-01-01

    Expanded use of existing storage basins is clearly the most economic solution to the spent fuel storage problem. The use of high-density racks followed by fuel disassembly and rod storage is an order of magnitude cheaper than building new facilities adjacent to the reactor. The choice of a new storage facility is not as obvious; however, if the timing of expenditures and risk allowance are to be considered, then modular concepts such as silos, drywells, and storage casks may cost less than water basins and air-cooled vaults. A comparison of the costs of the various storage techniques without allowances for timing or risk is shown. The impact of allowances for discounting and early resumption of reprocessing is also shown. Economics is not the only issue to be considered in selecting a storage facility. The licensing, environmental impact, timing, and social responses must also be considered. Each utility must assess all of these issues for their particular reactors before the best storage solution can be selected

  4. Risk assessment in spent fuel storage and transportation

    International Nuclear Information System (INIS)

    Pandimani, S.

    1989-01-01

    Risk assessment in various stages of nuclear fuel cycle is still an active area of Nuclear safety studies. From the results of risk assessment available in literature, it can be determined that the risk resulting from shipments of plutonium and spent-fuel are much greater than that resulting from the transport of other materials within the nuclear fuel cycle. In India spent fuels are kept in Spent Fuel Storage Pool (SFSP) for about 240-400 days, which is relatively a longer period compared to the usual 120 days as recommended by regulatory authorities. After cooling spent fuels are transported to the reprocessing sites which are mostly situated close to the plants. India has two high level waste treatment facilities, one PREFRE (Plutonium Reprocessing and Fuel Recycling) at Tarapur and the other one, a unit of Nuclear Fuel Complex at Hyderabad. This paper presents the risk associated with spent fuel storage and transportation for the Indian conditions. All calculations are based on a typical CANDU reactor system. Simple fault tree models are evolved for SFSP and for Transportation Accident Mode (TAM) for both road and rail. Fault tree quantification and risk assessment are done to each of these models. All necessary data for SFSP are taken mostly from Reactor Safety Study, (1975). Similarly, the data for rail TAM are taken from Annual Statistical Statements, (1987-8) and that for road TAM from Special Issue on Motor Vehicle Accident Statistics in India, (1986). Simulation method is used wherever necessary. Risk is also estimated for normal/accident free transport

  5. Operation of spent fuel storage facilities

    International Nuclear Information System (INIS)

    1994-01-01

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

  6. Heat transfer in a spent fuel pool concept containing PWR, Hybrid ADS-Fission, and VHTR spent fuels

    Energy Technology Data Exchange (ETDEWEB)

    Faria, Fernando P.; Cardoso, Fabiano; Salomé, Jean A.D.; Velasquez, Carlos E.; Pereira, Claubia, E-mail: fernandopereirabh@gmail.com, E-mail: fabinuclear@yahoo.com.br, E-mail: jadsalome@yahoo.com.br, E-mail: carlosvelcab@hotmail.com, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais, Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-07-01

    Thermal evaluation under wet storage conditions of spent fuels (SF) of the types UO{sub 2} discharged from Pressurized Water Reactor (PWR) and Very High-temperature Reactor (VHTR), and (Th,TRU)O{sub 2} from Accelerator-Driven Subcritical Reactor System (ADS) and VHTR are presented. The analyzes are in the absence of an external cooling system of the pool, and the goal is to compare the water boiling time of the pool storing these different types of SF, at time t=0 year after reactor discharge. Two techniques were implemented. In the first one, all the materials of the fuel elements are considered. In the second, the SF is treated as holes inside the pool, assuming the heat transfer directly from the SF to the water. Results from first technique show that the boiling time (T{sub b}) ranged from 23 minutes for (Th,TRU)O{sub 2} from VHTR to 3 hours for UO{sub 2} from VHTR, while for the second technique, T{sub b} ranged from 10 minutes for (Th,TRU)O{sub 2} from VHTR to 2.7 hours for UO{sub 2} from VHTR. The discrepancies between Tb from both techniques reveal that the pathways considered for the heat transfer are crucial to the results. The thermal studies used the module CFX of the ANSYS Workbench 16.2 - student version. (author)

  7. Extending Spent Fuel Storage until Transport for Reprocessing or Disposal

    Energy Technology Data Exchange (ETDEWEB)

    Carlsen, Brett; Chiguer, Mustapha; Grahn, Per; Sampson, Michele; Wolff, Dietmar; Bevilaqua, Arturo; Wasinger, Karl; Saegusa, Toshiari; Seelev, Igor

    2016-09-01

    Spent fuel (SF) must be stored until an end point such as reprocessing or geologic disposal is imple-mented. Selection and implementation of an end point for SF depends upon future funding, legisla-tion, licensing and other factors that cannot be predicted with certainty. Past presumptions related to the availability of an end point have often been wrong and resulted in missed opportunities for properly informing spent fuel management policies and strategies. For example, dry cask storage systems were originally conceived to free up needed space in reactor spent fuel pools and also to provide SFS of up to 20 years until reprocessing and/or deep geological disposal became available. Hundreds of dry cask storage systems are now employed throughout the world and will be relied upon well beyond the originally envisioned design life. Given present and projected rates for the use of nuclear power coupled with projections for SF repro-cessing and disposal capacities, one concludes that SF storage will be prolonged, potentially for several decades. The US Nuclear Regulatory Commission has recently considered 300 years of storage to be appropriate for the characterization and prediction of ageing effects and ageing management issues associated with extending SF storage and subsequent transport. This paper encourages addressing the uncertainty associated with the duration of SF storage by de-sign – rather than by default. It suggests ways that this uncertainty may be considered in design, li-censing, policy, and strategy decisions and proposes a framework for safely extending spent fuel storage until SF can be transported for reprocessing or disposal – regardless of how long that may be. The paper however is not intended to either encourage or facilitate needlessly extending spent fuel storage durations. Its intent is to ensure a design and safety basis with sufficient margin to accommodate the full range of potential future scenarios. Although the focus is primarily on

  8. Taxing fossil fuels under speculative storage

    International Nuclear Information System (INIS)

    Tumen, Semih; Unalmis, Deren; Unalmis, Ibrahim; Unsal, D. Filiz

    2016-01-01

    Long-term environmental consequences of taxing fossil fuel usage have been extensively studied in the literature. However, these taxes may also impose several short-run macroeconomic policy challenges, the nature of which remains underexplored. This paper investigates the mechanisms through which environmental taxes on fossil fuel usage can affect the main macroeconomic variables in the short-run. We concentrate on a particular mechanism: speculative storage. Formulating and using a dynamic stochastic general equilibrium (DSGE) model, calibrated for the United States, with an explicit storage facility and nominal rigidities, we show that in designing environmental tax policies it is crucial to account for the fact that fossil fuel prices are subject to speculation. The existence of forward-looking speculators in the model improves the effectiveness of tax policies in reducing fossil fuel usage. Improved policy effectiveness, however, is costly: it drives inflation and interest rates up, while impeding output. Based on this tradeoff, we seek an answer to the question how monetary policy should interact with environmental tax policies in our DSGE model of fossil fuel storage. We show that, in an environment with no speculative storers, monetary policy should respond to output along with CPI inflation in order to minimize the welfare losses brought by taxes. However, when the storage facility is activated, responding to output in the monetary policy rule becomes less desirable.

  9. Manipulator for fuel assemblies in a spent fuel pool, especially for a LMFBR

    International Nuclear Information System (INIS)

    Dalmas, R.

    1988-01-01

    The spent fuel manipulator has - a travelling crane moving longitudinally: - a carriage moving on the travelling crane in a direction perpendicular to its motion so that the carriage is positioned over each assembly, - a telescopic rod carried by the carriage and terminating in a vertically mobile grapple, - a tubular shielded hood on the carriage extending downwards to house the rod, grapple and fuel assembly and maintaining a biologically acceptable level of radiation above the surface of the pool [fr

  10. Materials behavior in interim storage of spent fuel

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.; Bailey, W.J.; Gilbert, E.R.; Inman, S.C.

    1982-01-01

    Interim storage has emerged as the only current spent-fuel management method in the US and is essential in all countries with nuclear reactors. Materials behavior is a key aspect in licensing interim-storage facilities for several decades of spent-fuel storage. This paper reviews materials behavior in wet storage, which is licensed for light-water reactor (LWR) fuel, and dry storage, for which a licensing position for LWR fuel is developing

  11. Analysis of the loss of pool cooling accident in a PWR spent fuel pool with MAAP5

    International Nuclear Information System (INIS)

    Wu, Xiaoli; Li, Wei; Zhang, Yapei; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

    2014-01-01

    Highlights: • A PWR spent fuel pool was modeled by using MAAP5. • Loss of pool cooling severe accident scenarios were studied. • Loss of pool cooling accidents with two mitigation measures were analyzed. - Abstract: The Fukushima Daiichi nuclear accident shows that it is necessary to study potential severe accidents and corresponding mitigation measures for the spent fuel pool (SFP) of a nuclear power plant (NPP). This paper presents the analysis of loss of pool cooling accident scenarios and the discussion of mitigation measures for the SFP at a pressurized water reactor (PWR) NPP with the MAAP5 code. Analysis of uncompensated loss of water due to the loss of pool cooling with different initial pool water levels of 12.2 m (designated as a reference case) and 10.7 m have been performed based on a MAAP5 input model. Scenarios of the accident such as overheating of uncovered fuel assemblies, oxidation of claddings and hydrogen generation, loss of intactness of fuel rod claddings, and release of radioactive fission products were predicted with the assumption that mitigation measures were unavailable. The results covered a broad spectrum of severe accident evaluations in the SFP. Furthermore, as important mitigation measures, the effects of recovering the SFP cooling system and makeup water in SFP on the accident progressions have also been investigated respectively based on the events of pool water boiling and spent fuels uncovery. Based upon the reference case, three cases with the recovery of SFP cooling system and three other cases with makeup water in SFP have been studied. The results showed that, severe accident might happen if SFP cooling system was not restored timely before the spent fuels started to become uncovered; spent fuels could be completely submerged and severe accident might be avoided if SFP makeup water system provided water with a mass flow rate larger than the average evaporation rate defined as the division of pool water mass above the

  12. Compact spent fuel storage at the Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    Antonaccio, Carlos; Conde, Alberto; Flores, Alexis; Masciotra, Humberto; Sala, Guillermo; Zanni, Pablo

    2000-01-01

    The object of this report is to verify the possibility to increase the available storage of irradiated fuel assemblies, placed in the spent fuel pools of the Atucha I nuclear power plant. There is intends the realization of structural modifications in the storage bracket-suspension beam (single and double) for the upper and lower level of the four spent fuel pools. With these modifications that increase the storage capacity 25%, would arrive until the year 2014, it dates dear for the limit of the commercial operation of nuclear power plant. The increase of the capacity in function of the permissible stress for the supports of the bracket-suspension beam. They should be carried out 5000 re-accommodations of irradiated fuel assemblies. The task would demand approximately 3 years. (author)

  13. 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

  14. Conceptual design of reactor TRIGA PUSPATI (RTP) spent fuel storage rack

    International Nuclear Information System (INIS)

    Tonny Lanyau; Mohd Fazli Zakaria; Zaredah Hashim; Ahmad Nabil Ab Rahim; Mohammad Suhaimi Kassim

    2010-01-01

    PUSPATI TRIGA Reactor (RTP) is a pool type research reactor with 1MW thermal power. It has been safely operated since 28 June 1982. During 28 years of safe operation, there are several systems and components of the RTP that have been maintained, repaired, upgraded and replaced in order to maintain its function and safety conditions. RTP has been proposed to be upgraded so that optimum operation of RTP could be achieved as well as fulfill the future needs. Thus, competencies and technical capabilities were needed to design and develop the reactor system. In the meantime, there is system or component need to be maintained such as fuel elements. Since early operation, most of the fuel elements still can be used and none of the fuel elements was replaced or sent for reprocessing and final disposal. Towards the power upgrading, preparation of spent fuel storage is needed for temporary storing of the fuels discharged from the reactor core. The spent fuel storage rack will be located in the spent fuel pool to accommodate the spent fuels before it is send to reprocessing or final disposal. This paper proposes the conceptual design of the spent fuel storage rack. The output of this paper focused on the physical and engineering design of the spent fuel storage. (author)

  15. Storage racks for spent fuels

    International Nuclear Information System (INIS)

    Inoue, Tatsumi.

    1983-01-01

    Purpose: To provide a storage rack which is light in weight, excellent in constructing performance, earthquake proof and allowable for thermal displacement. Constitution: The rack frame is constituted by disposing mounting seat plates at the legs of profiled-steel post members on a rack-mounting plate as a mounting seat by way of mounting liner plates and then weld-joining each of upper, medium and lower beam members and truss members of the profiled-steel posts by the use of connecting plates. Guide plates are laid transversely and joined onto each of the thus formed upper, medium and lower stage frames. Since apertures for mounting guide tubes are opened to each of the guide plates at a predetermined pitch, the storage rack unit is constituted by mounting guide tubes to the apertures. Then, the storage rack is mounted by means of slide keys to the mounting plates where slight amount of sliding or thermal displacement can be absorbed. (Horiuchi, T.)

  16. Subsurface storage of commercial spent nuclear fuel

    International Nuclear Information System (INIS)

    Richards, L.M.; Szulinski, M.J.

    1979-01-01

    The Atlantic Richfield Company has developed the concept of storing spent fuel in dry caissons. Cooling is passive; safety and safeguard features appear promising. The capacity of a caisson to dissipate heat depends on site-specific soil characteristics and on the diameter of the caisson. It is estimated that approx. 2 kW can be dissipated in the length of one fuel element. Fuel elements can be stacked with little effect on temperature. A spacing of approx. 7.5 m (25 ft) between caissons appears rasonable. Business planning indicates a cost of approx. 0.2 mill/kWh for a 15-yr storage period. 12 figures, 4 tables

  17. Spent nuclear fuel storage vessel

    International Nuclear Information System (INIS)

    Watanabe, Yoshio; Kashiwagi, Eisuke; Sekikawa, Tsutomu.

    1997-01-01

    Containing tubes for containing spent nuclear fuels are arranged vertically in a chamber. Heat releasing fins are disposed horizontal to the outer circumference of the containing tubes for rectifying cooling air and promoting cooling of the containing tubes. Louvers and evaporation sides of heat pipes are disposed at a predetermined distance in the chamber. Cooling air flows from an air introduction port to the inside of the chamber and takes heat from the containing tubes incorporated with heat generating spent nuclear fuels, rising its temperature and flows off to an air exhaustion exit. The direction for the rectification plate of the louver is downward from a horizontal position while facing to the air exhaustion port. Since the evaporation sides of the heat pipes are disposed in the inside of the chamber and the condensation side of the heat pipes is disposed to the outside of the chamber, the thermal energy can be recovered from the containing tubes incorporated with spent nuclear fuels and utilized. (I.N.)

  18. Report on the possibilities of long-term storage of irradiated nuclear fuels; Rapport sur les possibilites d'entreposage a long terme de combustibles nucleaires irradies

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    This report aims at giving a legislative aspect to the many rules that govern the activities of the back-end of the fuel cycle in France. These activities concern the unloading of spent nuclear fuels, their reprocessing, storage, recycling and definitive disposal. The following points are reviewed and commented: the management of non-immediately reprocessed fuels (historical reasons of the 'all wastes reprocessing' initial choice, evolution of the economic and political context, the future reprocessing or the definitive disposal of spent fuels in excess); the inevitable long-term storage of part of the spent fuels (quantities and required properties of long-term stored fuels, the eventuality of a definitive disposal of spent fuels); the criteria that long-term storage facilities must fulfill (confinement measures, reversibility, surveillance and control during the whole duration of the storage); storage concept to be retained (increase of storage pools capacity, long-term storage in pools of reprocessing plants, centralized storage in pools, surface dry-storage on power plant sites, reversible underground storage, subsurface storage and storage/disposal in galleries, surface dry-storage facilities); the preliminary studies for the creation of long-term storage facilities (public information, management by a public French organization, clarifying of the conditions of international circulation of spent fuels); problems linked with the presence of foreign spent fuels in France (downstream of the reprocessing cycle, foreign plutonium and wastes re-shipment); conclusions and recommendations. (J.S.)

  19. Report on the possibilities of long-term storage of irradiated nuclear fuels; Rapport sur les possibilites d'entreposage a long terme de combustibles nucleaires irradies

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    This report aims at giving a legislative aspect to the many rules that govern the activities of the back-end of the fuel cycle in France. These activities concern the unloading of spent nuclear fuels, their reprocessing, storage, recycling and definitive disposal. The following points are reviewed and commented: the management of non-immediately reprocessed fuels (historical reasons of the 'all wastes reprocessing' initial choice, evolution of the economic and political context, the future reprocessing or the definitive disposal of spent fuels in excess); the inevitable long-term storage of part of the spent fuels (quantities and required properties of long-term stored fuels, the eventuality of a definitive disposal of spent fuels); the criteria that long-term storage facilities must fulfill (confinement measures, reversibility, surveillance and control during the whole duration of the storage); storage concept to be retained (increase of storage pools capacity, long-term storage in pools of reprocessing plants, centralized storage in pools, surface dry-storage on power plant sites, reversible underground storage, subsurface storage and storage/disposal in galleries, surface dry-storage facilities); the preliminary studies for the creation of long-term storage facilities (public information, management by a public French organization, clarifying of the conditions of international circulation of spent fuels); problems linked with the presence of foreign spent fuels in France (downstream of the reprocessing cycle, foreign plutonium and wastes re-shipment); conclusions and recommendations. (J.S.)

  20. Development of fuel and energy storage technologies

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    Development of fuel cell power plants is intended of high-efficiency power generation using such fuels with less air pollution as natural gas, methanol and coal gas. The closest to commercialization is phosphoric acid fuel cells, and the high in efficiency and rich in fuel diversity is molten carbonate fuel cells. The development is intended to cover a wide scope from solid electrolyte fuel cells to solid polymer electrolyte fuel cells. For new battery power storage systems, development is focused on discrete battery energy storage technologies of fixed type and mobile type (such as electric vehicles). The ceramic gas turbine technology development is purposed for improving thermal efficiency and reducing pollutants. Small-scale gas turbines for cogeneration will also be developed. Development of superconduction power application technologies is intended to serve for efficient and stable power supply by dealing with capacity increase and increase in power distribution distance due to increase in power demand. In the operations to improve the spread and general promotion systems for electric vehicles, load leveling is expected by utilizing and storing nighttime electric power. Descriptions are given also on economical city systems which utilize wide-area energy. 30 figs., 7 tabs.

  1. New developments in dry spent fuel storage

    International Nuclear Information System (INIS)

    Bonnet, C.; Chevalier, Ph.

    2001-01-01

    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)

  2. Regional spent fuel storage facility (RSFSF)

    International Nuclear Information System (INIS)

    Dyck, H.P.

    1999-01-01

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

  3. Storage container for radioactive fuel elements

    International Nuclear Information System (INIS)

    1984-01-01

    The interim storage cask for spent fuel elements or the glass moulds for high-level radioactive waste are made up of heat-resistant, reinforced concrete with chambers and highgrade steel lining. Cooling systems with natural air circulation are connected with the chambers. (HP) [de

  4. High density aseismic spent fuel storage racks

    International Nuclear Information System (INIS)

    Louvat, J.P.

    1985-05-01

    After the reasons of the development of high density aseismic spent fuel racks by FRAMATOME and LEMER, a description is presented, as also the codes, standards and regulations used to design this FRAMATOME storage rack. Tests have been carried out concerning criticality, irradiation of Cadminox, corrosion of the cell, and the seismic behaviour

  5. Studies and research concerning BNFP: converting reprocessing plant's fuel receiving and storage area to an away-from-reactor (AFR) storage facility. Final report

    International Nuclear Information System (INIS)

    Cottrell, J.E.; Shallo, F.A.; Musselwhite, E.L.; Wiedemann, G.F.; Young, M.

    1979-09-01

    Converting a reprocessing plant's fuel receiving and storage station into an Away-From-Reactor storage facility is evaluated in this report. An engineering analysis is developed which includes (1) equipment modifications to the facility including the physical protection system, (2) planning schedules for licensing-related activities, and (3) cost estimates for implementing such a facility conversion. Storage capacities are evaluated using the presently available pools of the existing Barnwell Nuclear Fuel Plant-Fuel Receiving and Storage Station (BNFP-FRSS) as a model

  6. Studies and research concerning BNFP: converting reprocessing plant's fuel receiving and storage area to an away-from-reactor (AFR) storage facility. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Cottrell, Jim E.; Shallo, Frank A.; Musselwhite, E Larry; Wiedemann, George F.; Young, Moylen

    1979-09-01

    Converting a reprocessing plant's fuel receiving and storage station into an Away-From-Reactor storage facility is evaluated in this report. An engineering analysis is developed which includes (1) equipment modifications to the facility including the physical protection system, (2) planning schedules for licensing-related activities, and (3) cost estimates for implementing such a facility conversion. Storage capacities are evaluated using the presently available pools of the existing Barnwell Nuclear Fuel Plant-Fuel Receiving and Storage Station (BNFP-FRSS) as a model.

  7. Calculation of Water Levels in Spent Fuel Pool and Effective Dose Followed by the Worker Geometrically Exposed to Radiation using Gamma-ray Source

    International Nuclear Information System (INIS)

    Lee, Donghee; Park, Kwangheon; Yoon, Hyoungju

    2013-01-01

    If the total effective dose value is lower than the surface dose rate of the water, the worker is able to work in a safe environment. In the case that the level of spent fuel pool is up to 550cm, there exists the limitation for workers to access to the storage pool because the result value is about 8 times higher than surface dose rate. In the case that the level of spent fuel pool is higher than 600cm, however, it can be safe work environment because the result value is lower than surface dose rate. Therefore, in the case of ISO geometry which is the same with practical situation, when considering Gamma-ray emission from spent fuel, effective dose is much higher than surface dose rate when the level of storage pool is lower than the height of fuel, 452.8cm. On the other hand, the level of effective dose decreases rapidly when the level of storage pool is higher than the level of the fuel. This means that it is not the safe environment when the level of fuel below 140cm is lower than surface dose rate. That is why the access of workers should be limited. Whereas, in the case of the level of storage pool above 600cm which is about 140cm higher than the level of the fuel, it is the safe environment for workers because the result value becomes lower than surface dose rate As a result, the level of wet storage of spent fuel should be at least 600cm for workers to work in safe environment because lower dose than surface dose rate makes less radiation exposure

  8. Spent Fuel Storage Operation - Lessons Learned

    International Nuclear Information System (INIS)

    2013-12-01

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

  9. Dry storage of MTR spent fuel from the Argentine radioisotope production reactor RA-3

    International Nuclear Information System (INIS)

    Di Marco, A.; Gillaume, E.J.; Ruggirello, G.; Zaweruchi, A.

    1996-01-01

    The nuclear fuel elements of the RA-3 reactor consist in 19 rectangular fuel plates held in position by two lateral structural plates. The whole assembly is coupled to the lower nozzles that fits in the reactor core grid. The inner plates are 1.5 mm thick, 70.5 mm wide and 655 mm long and the outer plates are 100 mm longer. The fuel plates are formed by a core of an AI-U alloy co-laminated between two plates of Al. Enrichment is 90% 235 U. After being extracted from the reactor, the fuel elements have been let to cool down in the reactor storage pool and finally moved to the storage facility. This facility is a grid of vertical underground channels connected by a piping system. The system is filled with processed and controlled water. At the present the storage capacity of the facility is near to be depleted and some indications of deterioration of the fuel elements has been detected. Due to the present status of the facility and the spent fuel stored there, a decision has been taken to proceed to modify the present underwater storage to dry storage. The project consist in: a) Decontamination and conditioning of the storage channels to prepare them for dry storage. b) Disassembly of the fuel elements in hot cells in order to can only the active fuel plates in an adequate tight canister. c) The remnant structural pieces will be treated as low level waste. (author). 10 figs

  10. Characterization of the neutron sources storage pool of the Neutron Standards Laboratory, using Montecarlo Techniques

    International Nuclear Information System (INIS)

    Campo Blanco, X.

    2015-01-01

    The development of irradiation damage resistant materials is one of the most important open fields in the design of experimental facilities and conceptual nucleoelectric fusion plants. The Neutron Standards Laboratory aims to contribute to this development by allowing the neutron irradiation of materials in its calibration neutron sources storage pool. For this purposes, it is essential to characterize the pool itself in terms of neutron fluence and spectra due to the calibration neutron sources. In this work, the main features of this facility are presented and the characterization of the storage pool is carried out. Finally, an application is shown of the obtained results to the neutron irradiation of material.

  11. Status of Away From Reactor spent fuel storage program

    International Nuclear Information System (INIS)

    King, F.D.

    1979-07-01

    The Away From Reactor (AFR) Spent Fuel Program that the US Department of Energy established in 1977 is intended to preclude the shutting down of commercial nuclear power reactors because of lack of storage space for spent fuel. Legislation now being considered by Congress includes plans to provide storage space for commercial spent fuel beginning in 1983. Utilities are being encouraged to provide as much storage space as possible in their existing storage facilities, but projections indicate that a significant amount of AFR storage will be required. The government is evaluating the use of both existing and new storage facilities to solve this forecasted storage problem for commercial spent fuel

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

    International Nuclear Information System (INIS)

    Shallo, F.A.

    1978-08-01

    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

  13. Radiological protection for spent fuel dry storage at Embalse NPP

    International Nuclear Information System (INIS)

    Carballo, Carlos; Melo, Rodolfo

    2008-01-01

    Embalse NPP dry-stores used fuel elements in concrete silos inside the premises: The fuel elements are kept for at least 6 years in pools located in the controlled area , before being moved into the silos. This paper describes the radiological protection for the different stages of the process, i.e., when the used fuel elements are moved from the pools into the silos, and while they are kept in the concrete silos. The occupational exposure of the personnel operating this system at each stage is showed, as well as the environmental dose rates around the silos, and the dose rate in the shields used during the transfer. These environmental dose rates are assessed with portable instruments and with TLD dosimeters placed around the silos. This paper also describes the periodical routine control performed every two years in the atmosphere inside the silo, the moisture control and the detection of possible aerosols (in some cases, traces of krypton 85 were detected). It is important to point out that the maximum equivalent environmental dose rate H* (10) detected at approximately 20 metres from the silos is overly low: (0.35 micro sievert / hour). Our experience demonstrates that dry storage is totally compatible with the environment and with the ALARA criterion for personnel's doses. (author)

  14. Selected concrete spent fuel storage cask concepts and the DOE/PSN Cooperative Cask Testing Program

    International Nuclear Information System (INIS)

    Creer, J.M.; McKinnon, M.A.; Collantes, C.E.

    1990-01-01

    To date, water pools, metal casks, horizontal concrete modules, and modular vaults have been used to store the major quantity of commercial light water reactor spent nuclear fuel. Recently, vertical concrete dry storage casks have received consideration for storage of spent nuclear fuel. This paper reviews the evolution of the development of selected vertical concrete dry storage casks and outlines a cooperative cask testing (heat transfer and shielding) program involving the US Department of Energy and Pacific Sierra Nuclear Associates. Others participating in the cooperative program are Pacific Northwest Laboratory; EG ampersand G Idaho, Inc.; Wisconsin Electric Power Company; and the Electric Power Research Institute. 28 refs., 14 figs

  15. Bruce used fuel dry storage project evolution from Pickering to Bruce

    International Nuclear Information System (INIS)

    Young, R.E.

    1996-01-01

    Additional fuel storage capacity is required at Bruce Nuclear Generating Station, which otherwise would soon fill up all its pool storage capacity. The recommended option was to use a dry storage container similar to that at Pickering. The changes made to the Pickering type of container included: fuel to be stored in trays; the container's capacity increased to 600 bundles; the container's lid to be changed to a metal one; the single concrete lid to be changed to a double metal lid system; the container not to be transportable; the container would be dry-loaded. 7 figs

  16. Bruce used fuel dry storage project evolution from Pickering to Bruce

    Energy Technology Data Exchange (ETDEWEB)

    Young, R E [Ontario Hydro, Tiverton, ON (Canada). Bruce Nuclear Generating Station-A

    1997-12-31

    Additional fuel storage capacity is required at Bruce Nuclear Generating Station, which otherwise would soon fill up all its pool storage capacity. The recommended option was to use a dry storage container similar to that at Pickering. The changes made to the Pickering type of container included: fuel to be stored in trays; the container`s capacity increased to 600 bundles; the container`s lid to be changed to a metal one; the single concrete lid to be changed to a double metal lid system; the container not to be transportable; the container would be dry-loaded. 7 figs.

  17. Radioactive waste management decommissioning spent fuel storage. V. 3. Waste transport, handling and disposal spent fuel storage

    International Nuclear Information System (INIS)

    1985-01-01

    As part of the book entitled Radioactive waste management decommissioning spent fuel storage, vol. 3 dealts with waste transport, handling and disposal, spent fuel storage. Twelve articles are presented concerning the industrial aspects of nuclear waste management in France [fr

  18. Design considerations, operating and maintenance experience with wet storage of Ontario Hydro's used fuel

    International Nuclear Information System (INIS)

    Frost, C.R.

    1989-01-01

    The characteristics of Ontario Hydro's fuel and at-reactor used fuel storage water pools (or used fuel bays) are described. There are two types of bay, known respectively as primary bays and auxiliary bays, used for at- reactor used fuel storage. Used fuel is discharged remotely from Ontario Hydro's reactors to the primary bays for initial storage and cooling. The auxiliary bays are used to receive and store fuel after its initial cooling in the primary bay, and provide additional storage capacity as needed. With on- power fueling of reactors, each reactor of greater than 500 MW(e) net discharges an average of 10 or more used fuel bundles to bay storage every full power day. The logistics of handling such large quantities of used fuel bundles (corresponding to about 300 te/year of uranium for a 4 unit station) present a challenge to designers and operators. The major considerations in used fuel bay design, including site- specific requirements, reliability and quality assurance, are discussed

  19. Storage of spent nuclear fuel: the problem of spent nuclear fuel in Bulgaria

    International Nuclear Information System (INIS)

    Boyadzhiev, Z.; Vapirev, E.

    1995-01-01

    A review of existing technologies for wet and dry storage of spent nuclear fuel (SNF) and the reprocessing policies is presented. The problem of SNF in Bulgaria is arising from nonobservance of the obligation to return SNF back to the former Soviet Union as agreed in the construction contract. In November 1994 approximately 1800 fuel assemblies have been stored in away-from-reactor (AFR) facility and another 1060 in at-reactor (AR) pools. The national policy is to export SNF out of the country. The AFR facility has a limited capacity and it is designed only for WWER-440 fuel although work is going on to extend it in order to store WWER-1000 SNF. 14 refs

  20. Storage of spent nuclear fuel: the problem of spent nuclear fuel in Bulgaria

    Energy Technology Data Exchange (ETDEWEB)

    Boyadzhiev, Z; Vapirev, E [Kombinat Atomna Energetika, Kozloduj (Bulgaria)

    1996-12-31

    A review of existing technologies for wet and dry storage of spent nuclear fuel (SNF) and the reprocessing policies is presented. The problem of SNF in Bulgaria is arising from nonobservance of the obligation to return SNF back to the former Soviet Union as agreed in the construction contract. In November 1994 approximately 1800 fuel assemblies have been stored in away-from-reactor (AFR) facility and another 1060 in at-reactor (AR) pools. The national policy is to export SNF out of the country. The AFR facility has a limited capacity and it is designed only for WWER-440 fuel although work is going on to extend it in order to store WWER-1000 SNF. 14 refs.

  1. MELCOR Modeling of Air-Cooled PWR Spent Fuel Assemblies in Water empty Fuel Pools

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, L. E.; Lopez, C.

    2013-07-01

    The OECD Spent Fuel Project (SFP) investigated fuel degradation in case of a complete Loss-Of- Coolant-Accident in a PWR spent fuel pool. Analyses of the SFP PWR ignition tests have been conducted with the 1.86.YT.3084.SFP MELCOR version developed by SNL. The main emphasis has been placed on assessing the MELCOR predictive capability to get reasonable estimates of time-to-ignition and fire front propagation under two configurations: hot neighbor (i.e., adiabatic scenario) and cold neighbor (i.e., heat transfer to adjacent fuel assemblies). A detailed description of hypotheses and approximations adopted in the MELCOR model are provided in the paper. MELCOR results accuracy was notably different between both scenarios. The reasons are highlighted in the paper and based on the results understanding a set of remarks concerning scenarios modeling is given.

  2. Features and safety aspects of spent fuel storage facility, Tarapur

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  3. Device with pivoting base for the storage of nuclear fuel

    International Nuclear Information System (INIS)

    Raymond, T.E.

    1978-01-01

    A storage rack for nuclear fuel assemblies comprising lower and upper bearers to support and hold fuel assemblies in their vertical position is described. The feature of this rack is the lower supporting device which comprises a pivoting base on which rests each fuel assembly, thereby enabling the fuel assembly not be subjected to any fatigue during storage [fr

  4. Borated concrete for ZPPR fuel storage

    International Nuclear Information System (INIS)

    Gasidlo, J.M.

    1985-01-01

    Fuel handling at the Zero Power Plutonium Reactor (ZPPR) led to two requirements for storage of ZPPR fuel: a low neutron multiplication and shielded storage to minimize personnel doses. Boron-poisoned concrete was chosen as the storge medium with boron frit as the poisoning agent. The calculated effects of water content and boron concentration led to specifying a concrete with a water content that was higher than ordinary concrete. The finite size of the boron frit particles caused concern about reduced effectiveness due to self-shielding. The self-shielding was evaluated using optical path lengths for spheres and tabulated self-shielding for slabs. The results showed that the finite-sized particles were at least 80% as effective as infinitely dilute absorption. Neutron and gamma dose rates measured in the vault verified that personnel could work in the vault on a regular basis without exceeding personnel dose limits. 4 refs., 3 figs., 7 tabs

  5. Special equipment support the fuel storage

    International Nuclear Information System (INIS)

    Vega, M. E.

    2014-01-01

    In the current juncture one of the keys to any company that works in a market that is as demanding as the nuclear, is its ability to developed new technological products that they can adapt to the different special situations/needs of nuclear Power Plants during their operating life. As an example, below are some of the specialized equipment that ENSA has been developing for more than thirty years that has been doing work in the area of fuel storage. (Author)

  6. Safety assessment for spent fuel storage facilities

    International Nuclear Information System (INIS)

    1994-01-01

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

  7. A study for providing additional storage spaces to ET-RR-1 spent fuel

    International Nuclear Information System (INIS)

    El-Kady, A.; Ashoub, N.; Saleh, H.G.

    1995-01-01

    The ET-RR-1 reactor spent fuel storage pool is a trapezoidal aluminum tank concrete shield and of capacity 10 m 3 . It can hold up to 60 fuel assemblies. The long operation history of the ET-RR-1 reactor resulted in a partially filled spent fuel storage with the remaining spaces not enough to host a complete load from the reactor. This work have been initiated to evaluate possible alternative solutions for providing additional storage spaces to host the available EK-10 fuel elements after irradiation and any foreseen fuel in case of reactor upgrading. Several alternate solutions have been reviewed and decision on the most suitable one is under study. These studies include criticality calculation of some suggested alternatives like reracking the present spent fuel storage pool and double tiering by the addition of a second level storage rack above the existing rack. The two levels may have different factor. Criticality calculation of the double tiering possible accident was also studied. (author)

  8. State environmental review of a proposed utility independent spent fuel storage installation

    International Nuclear Information System (INIS)

    Sabel, G.; Halstead, R.

    1991-01-01

    This paper describes the environmental review process which was applied by the State of Minnesota to a proposed dry cask storage facility. An environmental analysis of the proposed project is summarized, as are alternatives including other dry storage technologies, increased in-pool storage, transhipment, reprocessing, use of higher burnup fuel and conservation. Public comments and concerns included potential cask failures, health impacts, and the possibility of the site becoming a open-quotes permanentclose quotes storage facility. State intervention in the federal license process is also described

  9. Storage device for fuel rods of nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Kempf, B.

    1983-01-01

    The storage device, which can be flexibly matched to the number of fuel rods to be stored and is not tied to a space, has a vertical support post situated on the floor and a stiff upright also situated vertically on the floor, which is used to accommodate at least one fuel rod. The stiff upright is connected directly to the support post by connections which can be undone, or form locking via another vertical stiff upright situation on the floor. (orig./HP) [de

  10. Spent nuclear fuel storage device and spent nuclear fuel storage method using the device

    International Nuclear Information System (INIS)

    Tani, Yutaro

    1998-01-01

    Storage cells attachably/detachably support nuclear fuel containing vessels while keeping the vertical posture of them. A ventilation pipe which forms air channels for ventilating air to the outer circumference of the nuclear fuel containing vessel is disposed at the outer circumference of the nuclear fuel containing vessel contained in the storage cell. A shielding port for keeping the support openings gas tightly is moved, and a communication port thereof can be aligned with the upper portion of the support opening. The lower end of the transporting and containing vessel is placed on the shielding port, and an opening/closing shutter is opened. The gas tightness is kept by the shielding port, the nuclear fuel containing vessel filled with spent nuclear fuels is inserted to the support opening and supported. Then, the support opening is closed by a sealing lid. (I.N.)

  11. Spent fuel pool risk analysis for the Dukovany NPP

    Energy Technology Data Exchange (ETDEWEB)

    Hust' ak, S.; Jaros, M.; Kubicek, J. [UJV Rez, a.s., Husinec-Rez (Czech Republic)

    2013-07-01

    UJV Rez, a.s. maintains a Living Probabilistic Safety Assessment (Living PSA) program for Dukovany Nuclear Power Plant (NPP) in the Czech Republic. This project has been established as a framework for activities related to risk assessment and to support for risk-informed decision making at this plant. The most extensively used PSA application at Dukovany NPP is risk monitoring of instantaneous (point-in-time) risk during plant operation, especially for the purpose of configuration risk management during plant scheduled outages to avoid risk significant configurations. The scope of PSA for Dukovany NPP includes also determination of a risk contribution from spent fuel pool (SFP) operation to provide recommendations for the prevention and mitigation of SFP accidents and to be applicable for configuration risk management. This paper describes the analysis of internal initiating events (IEs) in PSA for Dukovany NPP, which can contribute to the risk from SFP operation. The analysis of those IEs was done more thoroughly in the PSA for Dukovany NPP in order to be used in instantaneous risk monitoring. (orig.)

  12. Mechanical behaviour of PWR fuel rods during intermediate storage

    International Nuclear Information System (INIS)

    Bouffioux, P.; Dalmas, R.; Bernaudat, C.

    2000-01-01

    EDF, which owns the irradiated fuel coming from its NPPs, has initiated studies regarding the mechanical behaviour of a fuel rod and the integrity of its cladding, in the case where the spent fuel is stored for a significant duration. During the phases following in-reactor irradiation (ageing in a water-pool, transport and intermediate storage), many phenomena, which are strongly coupled, may influence the cladding integrity: - residual power and temperature decay; - helium production and release in the free volume of the rod (especially for MOX fuel); - fuel column swelling; - cladding creep-out under the inner gas pressure of the fuel rod; - metallurgical changes due to high temperatures during transportation. In parallel, the quantification of the radiological risk is based on the definition of a cladding integrity criterion. Up to now, this criterion requires that the clad hoop strain due to creep-out does not exceed 1%. A more accurate criterion is being investigated. The study and modelling of all the phenomena mentioned above are included in a R and D programme. This programme also aims at redefining the cladding integrity criterion, which is assumed to be too conservative. The R and D programme will be presented. In order to predict the overall behaviour of the rod during the intermediate storage phases, the AVACYC code has been developed. It includes the models developed in the R and D programme. The input data of the AVACYC code are provided by the results of in-reactor rod behaviour simulations, using the thermal-mechanical CYRANO3 code. Its main results are the evolution vs. time of hoop stresses in the cladding, rod internal pressure and cladding hoop strains. Chained CYRANO-AVACYC calculations have been used to simulate the behaviour of MOX fuel rods irradiated up to 40 GWd/t and stored under air during 100 years, or under water during 50 years. For such fuels, where the residual power remains high, we show that a large part of the cladding strain

  13. 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

  14. CNAAA spent fuel complementary storage building (UFC) construction and licensing: an overview of current status

    International Nuclear Information System (INIS)

    Lima Neto, Bertino do Carmo; Pacifi, Cicero Durval

    2013-01-01

    The reprocessing of nuclear fuel assemblies could be a valuable solution in order to make available additional energy resources and also to decrease the volume of discarded materials. After the burning of nuclear fuel assemblies to produce electrical energy, these components have to be stored in the spent fuel pools of each unit, for at least 10 years, in order to decrease their residual heat. Even after this initial 10 year-period, these spent fuel assemblies still have a great amount of energy, which can be reused. Nowadays, the spent fuel materials can be reprocessed in order to produce electrical energy, or be stored to provide, in the future, an opportunity to decide how these materials will be treated. At the present moment, Brazil does not plan to reprocess these spent fuels assemblies, as performed by some other countries. Thus, Brazil intends to build a spent fuel long term intermediate storage facility to allow the chance to make a decision in the future, taking into account the available technology at that time. Considering the three CNAAA units (Angra 1, 2 and 3 of Central Nuclear Almirante Alvaro Alberto, the Brazilian nuclear power plant, located at Angra dos Reis county, Rio de Janeiro state) have a life time estimated in 60 years, and the intrinsical spent fuel pools storage capacity of these units, a Spent Fuel Complementary Storage Building - UFC has to be foreseen in order to increase the storage capacity of CNAAA. Therefore, the Spent Fuel Complementary Storage Building shall be in operation in 2018, capable to receive the first spent fuel assemblies from Angra 2 and, in the next year, from Angra 1. The same procedure will be applied for the spent fuel assemblies of Angra 3, currently in construction. The Spent Fuel Complementary Storage Building will be constructed and operated by Eletrobras Eletronuclear - the CNAAA owner - and will be located at the same site of the plant. Conceptually, the UFC will be built as a wet storage modality

  15. Comparison of concepts for independent spent fuel storage facilities

    International Nuclear Information System (INIS)

    Held, Ch.; Hintermayer, H.P.

    1978-01-01

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

  16. Spent fuel storage at the Rancho Seco Nuclear Generation Station

    International Nuclear Information System (INIS)

    Miller, K.R.; Field, J.J.

    1995-01-01

    The Sacramento Municipal Utility District (SMUD) has developed a strategy for the storage and transport of spent nuclear fuel and is now in the process of licensing and manufacturing a Transportable Storage System (TSS). Staff has also engaged in impact limiter testing, non-fuel bearing component reinsertion, storage and disposal of GTCC waste, and site specific upgrades in support of spent fuel dry storage

  17. Equipment designs for the spent LWR fuel dry storage demonstration

    International Nuclear Information System (INIS)

    Steffen, R.J.; Kurasch, D.H.; Hardin, R.T.; Schmitten, P.F.

    1980-01-01

    In conjunction with the Spent Fuel Handling and Packaging Program (SFHPP) equipment has been designed, fabricated and successfully utilized to demonstrate the packaging and interim dry storage of spent LWR fuel. Surface and near surface storage configurations containing PWR fuel assemblies are currently on test and generating baseline data. Specific areas of hardware design focused upon include storage cell components and the support related equipment associated with encapsulation, leak testing, lag storage, and emplacement operations

  18. Current status on the spent fuel dry storage management in Taiwan

    International Nuclear Information System (INIS)

    Chen, H.T.; Liu, C.H.

    2006-01-01

    Full text: Full text: One of the high priority issues for the continuous operation of nuclear power plants is how to manage and store spent fuel. In recent years, interim dry storage of spent fuel has become a significant solution in extending the storage capacity at a nuclear reactor site that lacks sufficient spent fuel pool storage capacity as in the world, and also in Taiwan. Although the re-racking project for the spent fuel pools has been undertaken, the Taiwan Power Company (TPC) Chinshan nuclear power plant still will lose its full core reserve by the year 2010. TPC has declared to build an on-site interim dry storage facility, this followed by geological disposal represents the most suitable option at this time. TPC is expected to submit the application for construction permit in 2006; preoperational test and storage should be put into operation by the end of 2008. Interim dry storage is a passive system. Materials used play a crucial role in the safety function of cask. The competent authority of spent fuel management in Taiwan, FCMA/AEC, will carry out a confirmatory evaluation regarding heat dissipation, structural seismic analysis, and radiation shielding to assure available safety function for casks after reviewing safety analysis report submitted by TPC. Third party inspection has been required to enhance quality assurance program and foreign technical consultation will be arranged. Although the security level for such facility will be kept to the same level as an NPP, a comprehensive analysis against a commercial airplane attack on cask should be made and addressed in the supplement of SAR. Licensing hearing is also required before issuing the construction permit. The paper presents the review plan and regulatory requirements for the licensing of an interim dry storage of spent fuel, the licensing procedure, and the development of dry storage cask for spent fuel in Taiwan

  19. Safety of interim storage solutions of used nuclear fuel during extended term

    Energy Technology Data Exchange (ETDEWEB)

    Shelton, C.; Bader, S.; Issard, H.; Arslan, M. [AREVA, 7135 Minstrel Way, Suite 300 Columbia, MD 21045 (United States)

    2013-07-01

    In 2013, the total amount of stored used nuclear fuel (UNF) in the world will reach 225,000 T HM. The UNF inventory in wet storage will take up over 80% of the available total spent fuel pool (SFP) capacity. Interim storage solutions are needed. They give flexibility to the nuclear operators and ensure that nuclear reactors continue to operate. However, we need to keep in mind that they are also an easy way to differ final decision and implementation of a UNF management approach (recycling or final disposal). In term of public perception, they can have a negative impact overtime as it may appear that nuclear industry may have significant issues to resolve. In countries lacking an integrated UNF management approach, the UNF are being discharged from the SFPs to interim storage (mostly to dry storage) at the same rate as UNF is being discharged from reactors, as the SFPs at the reactor sites are becoming full. This is now the case in USA, Taiwan, Switzerland, Spain, South Africa and Germany. For interim storage, AREVA has developed different solutions in order to allow the continued operation of reactors while meeting the current requirements of Safety Authorities: -) Dry storage canisters on pads, -) Dual-purpose casks (dry storage and transportation), -) Vault dry storage, and -) Centralized pool storage.

  20. Critical experiments supporting close proximity water storage of power reactor fuel. Technical progress report

    International Nuclear Information System (INIS)

    Baldwin, M.N.; Hoovler, G.S.; Eng, R.L.; Welfare, F.G.

    1979-07-01

    Close-packed storage of LWR fuel assemblies is needed in order to expand the capacity of existing underwater storage pools. This increased capacity is required to accommodate the large volume of spent fuel produced by prolonged onsite storage. To provide benchmark criticality data in support of this effort, 20 critical assemblies were constructed that simulated a variety of close-packed LWR fuel storage configurations. Criticality calculations using the Monte Carlo KENO-IV code were performed to provide an analytical basis for comparison with the experimental data. Each critical configuration is documented in sufficient detail to permit the use of these data in validating calculational methods according to ANSI Standard N16.9-1975

  1. The dry spent RBMK fuel cask storage site at the Ignalina NPP in Lithuania

    International Nuclear Information System (INIS)

    Penkov, V.V.; Diersch, R.

    1999-01-01

    At present, there are about 15,000 spent RBMK fuel assemblies stored in the water pools near the reactors at the Ignalina Nuclear Power Plant (INPP). Part of them are cut in two bundles and stored in standardized baskets in the pools. Each basket is loaded with 102 bundles. For long-term interim storage of this fuel, it was decided to use dry storage in casks. For this reason, the total activity to be stored is split into individual units (casks). Each cask represents a closed and independent safety system, fulfilling all safety-relevant requirements for both normal operational and hypothetical accidental conditions. The main safety relevant features of the storage cask system are: (1) Inherent safety system; (2) Double barrier system; (3) Passive cooling by natural convection; (4) Safety against accidents. The cask dry storage system is a cost effective and multi-functional system for storage, transport after the operation time and final disposal under consideration of additional protective elements. From an economical point of view, cask storage has a number of advantages. Two cask types have been intended for the INPP storage site: (1) The CASTOR RBMK cask made of ductile cast iron; (2) The CONSTOR RBMK sandwich cask made of an inner and outer steel shell and reinforced heavy concrete. The CASTOR RBMK and the CONSTOR RBMK casks are designed to withstand severe storage site accidents and with help of impact limiters - to fulfil the IAEA test criteria for type B(U)F packages. The INPP spent RBMK fuel storage site is designed as an open air storage for an operational time of 50 years. The casks are arranged on the concrete storage pad. The site is equipped with a crane for cask handling and technological buildings and security systems. The safety analyses for fuel and cask handling and for cask handling and for cask technology at the site have been made and accepted by the Lithuanian Competent Authority. (author)

  2. Fuzzy pool balance: An algorithm to achieve a two dimensional balance in distribute storage systems

    International Nuclear Information System (INIS)

    Wu, Wenjing; Chen, Gang

    2014-01-01

    The limitation of scheduling modules and the gradual addition of disk pools in distributed storage systems often result in imbalances among their disk pools in terms of both disk usage and file count. This can cause various problems to the storage system such as single point of failure, low system throughput and imbalanced resource utilization and system loads. An algorithm named Fuzzy Pool Balance (FPB) is proposed here to solve this problem. The input of FPB is the current file distribution among disk pools and the output is a file migration plan indicating what files are to be migrated to which pools. FPB uses an array to classify the files by their sizes. The file classification array is dynamically calculated with a defined threshold named T max that defines the allowed pool disk usage deviations. File classification is the basis of file migration. FPB also defines the Immigration Pool (IP) and Emigration Pool (EP) according to the pool disk usage and File Quantity Ratio (FQR) that indicates the percentage of each category of files in each disk pool, so files with higher FQR in an EP will be migrated to IP(s) with a lower FQR of this file category. To verify this algorithm, we implemented FPB on an ATLAS Tier2 dCache production system. The results show that FPB can achieve a very good balance in both free space and file counts, and adjusting the threshold value T max and the correction factor to the average FQR can achieve a tradeoff between free space and file count.

  3. Safety of spent fuel elements storage under water at La Hague facility

    International Nuclear Information System (INIS)

    Guezenec, J.Y.

    1990-12-01

    Awaiting for a decision about radioactive waste repository, the spent fuel elements are stored in the storage pools at the La Hague facility. The water in the pools is permanently cooled and purified to maintain the temperature, radioactivity and chemical pollution under preset limits. The first safety problem is concerned with the spent fuel transport casks. Opening of the casks is done under water in a number of facilities. The most recent approach is done by the company To, which established dry manipulation which enables to minimise the risk of possible cask failures as well as external contamination of cooling fins of the casks. Another general safety related problem is related to criticality risk caused by possible cooling failures or by external events like earthquakes. Special probability limit is set up for seismic events to be less than 10 -7 /year. Equally, risk of fuel assembly failures due to possible chocs and possibility of defects in pool isolation are taken into account [fr

  4. How Canada has controlled the spent fuel storage problem

    International Nuclear Information System (INIS)

    Mosey, D.

    1985-01-01

    A report on the irradiated fuel storage workshop held in Toronto in October 1984. In particular Canada's attitude to spent fuel is examined. The basic fuel cycle has been envisaged as running from mining and refining, through interim storage to final geologic disposal, with reprocessing as an option to be considered when it looks economically attractive. (U.K.)

  5. Storage rack for nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Wachter, W.J.

    1988-01-01

    A storage rack for nuclear fuel assemblies is described comprising storage tubes, each having a polygon cross-section. The tubes being nested with cell walls of one tube aligned with and confronting cell walls of other tubes. Each cell wall having an array of embossed buttons so arranged that buttons of one cell wall engage buttons of a confronting cell wall, and the engage buttons are welded together to secure the tubes. At least one layer of neutron-poison material comprises a flexible, resilient pad interprosed between the aligned cell walls; whereby a major portion of the total outer surface area of each confronting cell wall is engaged with the layer of neutron-poison material

  6. Dry storage of irradiated nuclear fuel

    International Nuclear Information System (INIS)

    Tolmie, R.D.

    1983-01-01

    In transferring radioactive material between the preparation and clean chambers of a dry storage complex, irradiated nuclear fuel is posted from the preparation chamber to a sealable canister supported in a closable bucket in the clean chamber, or a contaminated sealed canister is posted from a closed bucket in the clean chamber into the preparation chamber by using a facility comprising two coaxial tubes constituting a closable orifice between the two chambers, the tubes providing sealing means for the bucket, and masking means for the bucket and canister closures together with means for withdrawing the closures into the preparation chamber. (author)

  7. International safeguards for spent fuel storage

    International Nuclear Information System (INIS)

    Kratzer, M.; Wonder, E.; Immerman, W.; Crane, F.

    1981-08-01

    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

  8. Spent fuel management: reprocessing or storage

    International Nuclear Information System (INIS)

    Lima Soares, M.L. de; Oliveira Lopes, M.J. de

    1986-01-01

    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 plants for LWR fuel are being planned by Spain, Pakistan and Argentina. (Author) [pt

  9. Spent fuel management: reprocessing or storage

    International Nuclear Information System (INIS)

    Lima Soares, M.L. de; Oliveira Lopes, M.J. de.

    1986-01-01

    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) [pt

  10. Criticality safety analysis of spent fuel storage for NPP Mochovce using MCNP5

    International Nuclear Information System (INIS)

    Farkas, G.; Hascik, J.; Lueley, J.; Vrban, B.; Petriska, M.; Slugen, V.; Urban, P.

    2011-01-01

    The paper presents results of nuclear criticality safety analysis of spent fuel storage for the first and second unit of NPP Mochovce. The spent fuel storage pool (compact and reserve grid) was modeled using the Monte Carlo code MCNP5. Conservative approach was applied and calculation of k eff values was performed for normal and various postulated emergency conditions in order to evaluate the final maximal k eff values. The requirement of current safety regulations to ensure 5% subcriticality was met except one especially conservative case. (Authors)

  11. The NRC activities concerning Boraflex use in spent-fuel storage racks

    International Nuclear Information System (INIS)

    Kopp, L.I.

    1996-01-01

    The U.S. Nuclear Regulatory Commission (NRC) has issued several generic communications to the nuclear industry identifying two issues with respect to using Boraflex in spent-fuel storage racks. The first issue related to gamma-radiation-induced shrinkage of Boraflex and the potential to develop tears or gaps in the material. This phenomenon is typically accounted for in criticality analyses of spent-fuel storage racks. The second issue concerned long-term Boraflex performance throughout the intended service life of the racks as a result of both gamma irradiation and exposure to the wet pool environment

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

    International Nuclear Information System (INIS)

    1988-01-01

    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

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  14. Redesign of the spent fuel storage racks at the Trojan Nuclear Plant

    International Nuclear Information System (INIS)

    Stump, K.

    1987-01-01

    The spent fuel pool (SFP) at the Trojan Nuclear Plant located near Prescott, Oregon, was originally designed to hold 1.33 cores worth of spent fuel assemblies. Due to the delay in the site selection and preparation process for the spent fuel repository, the SFP storage capacity was increased in 1978 from 260 assemblies to 651 assemblies and in 1983 was increased again from 651 to 1408 assemblies to allow Trojan to continue operations through the year 2003 with a full core reserve in the SFP. Now it appears unlikely that a high level waste repository will be in operation before 2010. This indicates that a further capacity increase in the SFP is required to allow commercial operation until 2010, at which time the repository should be open to receive spent fuel. To accomplish this, an increase of seven times the original SFP capacity of 260 assemblies is needed. This paper presents a spent fuel assembly rack design that enables the required capacity increase in the SFP to be met. By the use of a boron carbide - silicon polymer inside a titanium/vanadium honeycomb as a neutron absorber between the fuel assemblies and by increasing the metal to water ratio of the spent fuel pool to harden the neutron energy spectrum the capacity of the SFP is increased to 1880 assemblies for an increase of 7.23 times the original spent fuel pool capacity. The multiplication factor for the pool with every fuel assembly slot filled in the new rack system is 0.62; well below the NRC regulatory limit of keff < 0.95. The capacity increase with allow the commercial operation of the Trojan Nuclear Plant through 2010 with a full core reserve in the spent fuel pool

  15. Fission product concentration evolution in sodium pool following a fuel subassembly failure in an LMFBR

    International Nuclear Information System (INIS)

    Natesan, K.; Velusamy, K.; Selvaraj, P.; Kasinathan, N.; Chellapandi, P.; Chetal, S.; Bhoje, S.

    2003-01-01

    During a fuel element failure in a liquid metal cooled fast breeder reactor, the fission products originating from the failed pins mix into the sodium pool. Delayed Neutron Detectors (DND) are provided in the sodium pool to detect such failures by way of detection of delayed neutrons emitted by the fission products. The transient evolution of fission product concentration is governed by the sodium flow distribution in the pool. Transient hydraulic analysis has been carried out using the CFD code PHOENICS to estimate fission product concentration evolution in hot pool. k- ε turbulence model and zero laminar diffusivity for the fission product concentration have been considered in the analysis. Times at which the failures of various fuel subassemblies (SA) are detected by the DND are obtained. It has been found that in order to effectively detect the failure of every fuel SA, a minimum of 8 DND in hot pool are essential

  16. Fukushima - calculation of the reactor core inventory and storage pools Dai-ichi 1 to Dai-ichi 4, an estimation of a source term

    International Nuclear Information System (INIS)

    Krpelanova, M.; Carny, P.

    2011-01-01

    Inventory of the reactor core and spent fuel storage pool of the reactors at Dai-ichi 1 to Dai-ichi 4 was determined to need a realistic estimate of the source (released into the atmosphere environment) and modelling of radiological impact of the events in Fukushima NPP. Calculations of inventories were carried out by the methodology that is used in systems to support emergency response and crisis management anymore. Calculations were made based on a model that respects knowledge of real fuels and fuel cycles for individual reactors Dai-ichi. Necessary input data for training the model and calculate inventories are obtained from the IAEA PRIS database.

  17. Cost and implications of a middle-term program for storage of spent fuel in a nuclear power station (BWR)

    International Nuclear Information System (INIS)

    Mochon, J.L.; Quintana, R.

    1978-01-01

    The experience gained with the Cofrentes Nuclear Power Station Project is presented. Originally the station had two spent fuel storage pools, in the fuel building, plus a little pool inside the containment, and all were to be fitted with extensive aluminium storage racks with a total capacity for 1+-1/3 cores. Due to the present world situation with regard to the ''back-end''of the fuel cycle, it was decided to enlarge the pools size and to change the design of the racks, to obtain a final storage capacity of 5+-1/4 cores, so covering over 18 years of operation. The changes introduced in the project, as well as its costs, and the possibilities of election still open are examined in the paper. (author)

  18. Cost and implications of a middle-term program for storage of spent fuel in a nuclear power station (BWR)

    International Nuclear Information System (INIS)

    Mochon, J.L.; Quintana, R.

    1978-01-01

    The paper is based on the experience gained with the Cofrentes Nuclear Power Station Project. Originally, the station had two spent fuel storage pools, in the fuel building, plus a little pool inside the containment, and all were to be fitted with extensive aluminum storage racks with a total capacity for 1+1/3 cores. Due to the present world situation with regard to the 'back-end' of the fuel cycle, it was decided to enlarge the pools' size and to change the design of the racks, to obtain a final storage capacity of 5+1/4 cores, so covering over 18 years of operation. The changes introduced in the project, as well as its costs, and the possibilities of election still open are examined in the paper

  19. Behavior of spent nuclear fuel and storage system components in dry interim storage.

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, A.B. Jr.; Gilbert, E.R.; Guenther, R.J.

    1982-08-01

    Irradiated nuclear fuel has been handled under dry conditions since the early days of nuclear reactor operation, and use of dry storage facilities for extended management of irradiated fuel began in 1964. Irradiated fuel is currently being stored dry in four types of facilities: dry wells, vaults, silos, and metal casks. Essentially all types of irradiated nuclear fuel are currently stored under dry conditions. Gas-cooled reactor (GCR) and liquid metal fast breeder reactor (LMFBR) fuels are stored in vaults and dry wells. Certain types of fuel are being stored in licensed dry storage facilities: Magnox fuel in vaults in the United Kingdom and organic-cooled reactor (OCR) fuel in silos in Canada. Dry storage demonstrations are under way for Zircaloy-clad fuel from boiling water reactors BWR's, pressurized heavy-water reactors (PHWRs), and pressurized water reactors (PWRs) in all four types of dry storage facilities. The demonstrations and related hot cell and laboratory tests are directed toward expanding the data base and establishing a licensing basis for dry storage of water reactor fuel. This report reviews the scope of dry interim storage technology, the performance of fuel and facility materials, the status of programs in several countries to license dry storage of water reactor fuel, and the characteristics of water reactor fuel that relate to dry storage conditions.

  20. Behavior of spent nuclear fuel and storage-system components in dry interim storage

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.; Gilbert, E.R.; Guenther, R.J.

    1982-08-01

    Irradiated nuclear fuel has been handled under dry conditions since the early days of nuclear reactor operation, and use of dry storage facilities for extended management of irradiated fuel began in 1964. Irradiated fuel is currently being stored dry in four types of facilities: dry wells, vaults, silos, and metal casks. Essentially all types of irradiated nuclear fuel are currently stored under dry conditions. Gas-cooled reactor (GCR) and liquid metal fast breeder reactor (LMFBR) fuels are stored in vaults and dry wells. Certain types of fuel are being stored in licensed dry storage facilities: Magnox fuel in vaults in the United Kingdom and organic-cooled reactor (OCR) fuel in silos in Canada. Dry storage demonstrations are under way for Zircaloy-clad fuel from boiling water reactors BWR's, pressurized heavy-water reactors (PHWRs), and pressurized water reactors (PWRs) in all four types of dry storage facilities. The demonstrations and related hot cell and laboratory tests are directed toward expanding the data base and establishing a licensing basis for dry storage of water reactor fuel. This report reviews the scope of dry interim storage technology, the performance of fuel and facility materials, the status of programs in several countries to license dry storage of water reactor fuel, and the characteristics of water reactor fuel that relate to dry storage conditions

  1. Comparison of wet and dry storage of spent nuclear fuels

    International Nuclear Information System (INIS)

    Soederman, E.

    1998-06-01

    Technologies for interim storage of spent nuclear fuels are reviewed. Pros and cons of wet and dry storage are discussed. No conclusions about preferences for one or the other technologies can be made

  2. Options for the interim storage of spent fuel

    International Nuclear Information System (INIS)

    Kromar, M.; Kurincic, B.

    1995-01-01

    Different concepts for the interim storage of spent fuel arising from operation of a NPP are discussed. We considered at reactor as well as away from reactor storage options. Included are enhancements of existing storage capabilities and construction of a new wet or dry storage facility. (author)

  3. Determination of maximum water temperature within the spent fuel pool of Angra Nuclear Power Plant - Unit 3

    Energy Technology Data Exchange (ETDEWEB)

    Werner, F.L., E-mail: fernanda.werner@poli.ufrj.br [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Departamento de Engenharia Nuclear; Alves, A.S.M., E-mail: asergi@eletronuclear.gov.br [Eletrobras Termonuclear (Eletronuclear), Rio de Janeiro, RJ (Brazil); Frutuoso e Melo, P.F., E-mail: frutuoso@nuclear.ufrj.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    In this paper, a mathematical model for the determination of the maximum water temperature within the spent fuel pool of Angra Nuclear Power Plant – Unit 3 was developed. The model was obtained from the boundary layer analysis and the application of Navier-Stokes equation to a vertical flat plate immersed in a water flow under free convection regime. Both types of pressure loss coefficients through the flow channel were considers in the modeling, the form coefficient for fuel assemblies (FAs) and the loss due to rod friction. The resulting equations enabled the determination of a mixed water temperature below the storage racks (High Density Storage Racks) as well as the estimation of a temperature gradient through the racks. The model was applied to the authorized operation of the plant (power operation, plant outage and upset condition) and faulted conditions (loss of coolant accidents and external events). The results obtained are in agreement with Brazilian and international standards. (author)

  4. Determination of maximum water temperature within the spent fuel pool of Angra Nuclear Power Plant - Unit 3

    International Nuclear Information System (INIS)

    Werner, F.L.; Frutuoso e Melo, P.F.

    2017-01-01

    In this paper, a mathematical model for the determination of the maximum water temperature within the spent fuel pool of Angra Nuclear Power Plant – Unit 3 was developed. The model was obtained from the boundary layer analysis and the application of Navier-Stokes equation to a vertical flat plate immersed in a water flow under free convection regime. Both types of pressure loss coefficients through the flow channel were considers in the modeling, the form coefficient for fuel assemblies (FAs) and the loss due to rod friction. The resulting equations enabled the determination of a mixed water temperature below the storage racks (High Density Storage Racks) as well as the estimation of a temperature gradient through the racks. The model was applied to the authorized operation of the plant (power operation, plant outage and upset condition) and faulted conditions (loss of coolant accidents and external events). The results obtained are in agreement with Brazilian and international standards. (author)

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  6. Transportation and storage of foreign spent power reactor fuel

    International Nuclear Information System (INIS)

    1979-01-01

    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

  7. Thermal analyses for the spend fuel pool of Taiwan BWR plants during the loss of cooling accident

    Energy Technology Data Exchange (ETDEWEB)

    Chen, B-Y.; Yeh, C-L.; Wei, W-C.; Chen, Y-S., E-mail: onepicemine@iner.gov.tw, E-mail: clinyeh@iner.gov.tw, E-mail: hn150456@iner.gov.tw, E-mail: yschen@iner.gov.tw [Inst. of Nuclear Energy Research, Longtan Township, Taoyuan County, Taiwan (China)

    2014-07-01

    After the Fukushima nuclear accident, the safety of the spent fuel pool has become an important concern. In this study, thermal analysis of the spent fuel pool under a loss of cooling accident is performed. The BWR spent fuel pools in Taiwan are investigated, including the Chinshan, Kuosheng, and Lungmen plants. The transient pool temperature and level behaviors are calculated based on lumped energy balance. After the pool level drops below the top of the fuel, the peak cladding temperature is predicted by the Computational Fluid Dynamics (CFD) analysis. The influence to the cladding temperature of the uniform and checkboard fuel loading patterns is also investigated. (author)

  8. Pyramid mountain diesel fuel storage site remediation

    Energy Technology Data Exchange (ETDEWEB)

    Brolmsa, M.; Sandau, C. [Jacques Whitford Environment Ltd., Burnaby, BC (Canada)

    2005-07-01

    Remediation activities during the decommissioning of a microwave tower facility where a tram line was used to transfer diesel fuel from the base of a mountain to its summit were described. As the site was leased from Parks Canada, federal guidelines were used to assess levels of contamination. Underground storage tanks (USTs) used for diesel storage had been replaced with aboveground storage tanks (AST) in 1994. Remediation was also complicated by the remote location and altitude of the site, as well as by extreme weather conditions. Hand auguring and test pitting were used at both the summit and base to allow characterization and preliminary delineation of impacted soils. A heavy lift helicopter was used to place demolition and excavation equipment on the summit. An excavator was used to remove hydrocarbon impacted soils. Following the remedial excavation for the summit diesel AST, residual soil impacts in excess of the applicable remediation guidelines were present at the bottom of the tank nest and under a floor slab. An environmental liner was installed, and a quantitative screening level risk assessment demonstrated the low level of risk for the area, as well as for waste oil impacted soils on the slope below the summit. Contaminants of potential concern were barium, zinc, naphthalene, and petroleum hydrocarbon fractions F1-F4. It was concluded that there are now no unacceptable ecological or human risks from residual impacts at the site. 1 tab., 19 figs.

  9. The Canadian long-term experimental used fuel storage program

    International Nuclear Information System (INIS)

    Wasywich, K.M.; Taylor, P.

    1993-01-01

    The Canadian experimental fuel storage program consists of four components: (1) storage of used CANDU (CANadian Deuterium Uranium, registered trademark of AECL) fuel under water, with periodic examination; (2) storage of used CANDU fuel in dry air at seasonally varying temperatures, and in both dry and moisture-saturated air at 150 C, also with periodic examination; (3) underlying research on the oxidation of unused and used UO 2 in dry and moist air at temperatures up to 300 C; and (4) modeling of UO 2 oxidation in dry air. The primary objective of the fuel-storage experiments is to investigate the stability of used CANDU fuel during long-term storage. Burnup of the fuel in these experiments ranges from ∼43 to 582 MW h/kg U, while the outer-element linear power ratings range from 22 to 79 kW/m. The storage behavior of intact and intentionally defected fuel, and fuel that defected in-reactor, is being investigated in the above experiments. Since differences in UO 2 oxidation behavior were observed between dry-air, moisture-saturated air and wet storage of intentionally defected used CANDU fuel, underlying research was initiated on oxidation of unused and used fuel to develop a better understanding of the different mechanisms. Modeling of UO 2 oxidation based on the results of the dry-storage experiments is also under way

  10. Seismic and structural analysis of high density/consolidated spent fuel storage racks

    International Nuclear Information System (INIS)

    Shah, S.J.; Biddle, J.R.; Bennett, S.M.; Schechter, C.B.; Harstead, G.A.; Kopecky, B.

    1995-01-01

    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. 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, ad the type of the seismic event. This paper presents several of the mathematical models usually used. The models include features to allow sliding and tipping of the racks and to represent the hydrodynamic coupling which can occur between fuel assemblies and rack cells, between adjacent racks, and between the racks and the reinforced concrete walls. 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 analyzed for the fuel loading conditions of consolidated, full, empty, and half-loaded with fuel assemblies

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  12. A COMPARISON OF CHALLENGES ASSOCIATED WITH SLUDGE REMOVAL & TREATMENT & DISPOSAL AT SEVERAL SPENT FUEL STORAGE LOCATIONS

    Energy Technology Data Exchange (ETDEWEB)

    PERES, M.W.

    2007-01-09

    Challenges associated with the materials that remain in spent fuel storage pools are emerging as countries deal with issues related to storing and cleaning up nuclear fuel left over from weapons production. The K Basins at the Department of Energy's site at Hanford in southeastern Washington State are an example. Years of corrosion products and piles of discarded debris are intermingled in the bottom of these two pools that stored more 2,100 metric tons (2,300 tons) of spent fuel. Difficult, costly projects are underway to remove radioactive material from the K Basins. Similar challenges exist at other locations around the globe. This paper compares the challenges of handling and treating radioactive sludge at several locations storing spent nuclear fuel.

  13. Evaluation of burnup credit for fuel storage analysis -- Experience in Spain

    International Nuclear Information System (INIS)

    Conde, J.M.; Recio, M.

    1995-01-01

    Several Spanish light water reactor commercial nuclear power plants are close to maximum spent-fuel pool storage capacity. The utilities are working on the implementation of state-of-the-art methods to increase the storage capacity, including both changes in the pool design (recracking) and the implementation of new analysis approaches with reduced conservation (burnup credit). Burnup credit criticality safety analyses have been approved for two pressurized water reactor plants (four units) and one boiling water reactor (BWR); an other BWR storage analysis is being developed at this moment. The elimination of the ''fresh fuel assumption'' increases the complexity of the criticality analysis to be performed, sometimes putting into question the capability of the analytic tools to properly describe this new situation and increasing the scope of the scenarios to be analyzed. From a regulatory perspective, the reactivity reduction associated with burnup of the fuel can be given credit only if the exposure of each fuel bundle can be known with enough accuracy. Subcriticality of spent-fuel storage depends mainly on the initial fuel enrichment, storage geometry, fuel exposure history, and cooling time. The last two aspects introduced new uncertainties in the criticality analysis that should be quantified in an adequate way. In addition, each and every fuel bundle has its own specific exposure history, so that strong assumptions and simplified calculational schemes have to be developed to undertake the analysis. The Consejo de Seguridad Nuclear (CSN), Spanish regulatory authority on the matter of nuclear safety and radiation protection, plays an active role in the development of analysis methods to support burnup credit, making proposals that may be beneficial in terms of risk and cost while keeping the widest safety margins possible

  14. Storage of Spent Nuclear Fuel. Specific Safety Guide

    International Nuclear Information System (INIS)

    2012-01-01

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

  15. The analysis of the RA reactor irradiated fuel cooling in the spent fuel pool; Analiza hladjenja ozracenog goriva u bazenu za odlaganje reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Vrhovac, M; Afgan, N; Spasojevic, D; Jovic, V [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1985-07-01

    According to the RA reactor exploitation plan the great quantity of the irradiated spent fuel will be disposed in the reactor spent fuel pool after each reactor campaign which will including the present spent fuel inventory increase the residual power level in the pool and will soon cause the pool capacity shortage. To enable the analysis of the irradiated fuel cooling the pool and characteristic spent fuel canister temperature distribution at the residual power maximum was done. The results obtained under the various spent fuel cooling conditions in the pit indicate the normal spent fuel thermal load even in the most inconvenient cooling conditions. (author)

  16. Water chemistry management of the spent-fuel pool in Thailand

    International Nuclear Information System (INIS)

    Suparit, Nitaya; Sukharn, Sumalee; Busamongkol, Arporn; Laoharojanaphand, Sirinart

    1999-01-01

    Water chemistry of the OAEP spent-fuel pool has been closely monitored without any pre-treatment for its conductivity, pH, temperature, chloride ion, sulfate ion, nitrate ion, phosphate ion, silver ion, and copper ion as well as its gamma activity of Cs-137. Conductivity, pH and temperature were measured using a portable pH and conductivity meter with built in temperature probe. Chloride ion was monitored by an automatic micro-titrator with silver nitrate as titrant and platinum indicator electrode. Nitrate, sulfate and phosphate were analysed by ion-exchange chromatographic method using an anion separator column and salicylate buffer as eluant. Gamma activity of Cs-137 was measured using a Canberra gamma spectrometer with HpGe detector. Silver and copper were analysed by ICP-AES technique within 6 hours after collection. During the study period from March 1996-September 1998, the conductivity was between l.25-4.80 μ/cm, pH in the range of 5-8.1, and temperature from 26.4-29.6 degree celsius. Chloride ion was found between 0.l-0.8 ppm. Silver, copper, nitrate, sulfate and phosphate ions were undetectable. Overall chemical composition of the water shows that the water is kept in standard condition recommended for safety storage. However, the presence of gamma activity of Cs-137 (average value of 138 Bq/l) indicates a slight leak of the spent fuel. (author)

  17. Design and operational experience of the NUHOMS-24P spent fuel storage system

    International Nuclear Information System (INIS)

    McConaghy, W.J.; Lehnert, R.A.; Rasmussen, R.W.

    1991-01-01

    The NUHOMS spent fuel storage system provides a safe and economical method for the dry storage of spent fuel assemblies either at an independent spent fuel storage installation (ISFSI) at reactor or at a centralized storage facility away from reactor. The system consists of three major safety-related components: a dry shielded canister (DSC) which provides a high integrity containment boundary and a controlled storage environment for the fuel; a reinforced concrete horizontal storage module (HSM) which houses the stored DSCs and provides radiation shielding, protection against natural phenomena and an efficient means for decay heat removal; and a transfer cask which provides for the safe shielded transfer of DSCs from a plant spent fuel pool to a HSM. The NUHOMS system is designed and licensed to the requirements of 10 CFR 72 and ANS/ANSI 57.9 for ISFSIs. The NUHOMS concept was developed in early 1980s, and in 1987, a larger version of the NUHOMS system, 24P, was developed. The operational features of NUHOMS and the loading experience at Oconee are reported. (K.I.)

  18. Thermal analyses for the rack design with spent fuel pool during the loss of cooling accident

    Energy Technology Data Exchange (ETDEWEB)

    Yeh, C-L.; Chen, Y-S.; Chen, B-Y., E-mail: clinyeh@iner.gov.tw, E-mail: yschen@iner.gov.tw, E-mail: onepicemine@iner.gov.tw [Inst. of Nuclear Energy Research, Taoyuan County, Taiwan (China); Tseng, Y-S., E-mail: ystseng@mx.nthu.edu.tw [National Tsing Hua Univ., Engineering and System Science, Hsinchu, Taiwan (China); Wei, W-C., E-mail: hn150456@iner.gov.tw [Inst. of Nuclear Energy Research, Taoyuan County, Taiwan (China)

    2014-07-01

    Alternative fuel arrangements separating the latest fuels discharge from the reactor core are proposed, such as the 1x4 configuration in which the hot assembly is surrounded by 4 assemblies with much lower decay heat. For the rack design in the BWR spent fuel pool design, the lateral flow is eliminated by solid walls. In this study, cooling enhancement of splitting fuel rack is investigated using Computational Fluid Dynamics (CFD). The fuels in the pool are modeled by porous medium. Separating the fuel rack by a distance of 10 cm can lower the peak cladding temperature and the natural convection between the fuels and then earns more response time for the site people to implement necessary mitigation actions. (author)

  19. Analysis of dose rates received around the storage pool for irradiated control rods in a BWR nuclear power plant

    International Nuclear Information System (INIS)

    Rodenas, J.; Abarca, A.; Gallardo, S.

    2011-01-01

    BWR control rods are activated by neutron reactions in the reactor. The dose produced by this activity can affect workers in the area surrounding the storage pool, where activated rods are stored. Monte Carlo (MC) models for neutron activation and dose assessment around the storage pool have been developed and validated. In this work, the MC models are applied to verify the expected reduction of dose when the irradiated control rod is hanged in an inverted position into the pool.

  20. Safe transport of spent fuels after long-term storage

    International Nuclear Information System (INIS)

    Aritomi, M.; Takeda, T.; Ozaki, S.

    2004-01-01

    Considering the scarcity of energy resources in Japan, a nuclear energy policy pertaining to the spent fuel storage has been adopted. The nuclear energy policy sets the rules that spent fuels generated from LWRs shall be reprocessed and that plutonium and unburnt uranium shall be recovered and reused. For this purpose, a reprocessing plant, which has a reprocessing capability of 800 ton/yr, is under construction at Rokkasho Village. However, it is anticipated that the start of its operation will be delayed. In addition, the amount of spent fuels generated from nuclear power plants exceeds its reprocessing capability. Therefore, the establishment of storage technology for spent fuels becomes an urgent problem in Japan in order to continue smoothly the LWR operations. In this paper, the background of nuclear power generation in Japan is introduced at first. Next, the policy of spent fuel storage in Japan and circumstances surrounding the spent fuels in Japan are mentioned. Furthermore, the major subjects for discussions to settle and improve 'Standard for Safety Design and Inspection of Metal Casks for Spent Fuel Interim Storage Facility' in Atomic Energy Society of Japan are discussed, such as the integrity of fuel cladding, basket, shielding material and metal gasket for the long term storage for achieving safe transport of spent fuels after the storage. Finally, solutions to the unsolved subject in establishing the spent fuel interim storage technologies ase introduced accordingly

  1. CFD Simulation of Heat and Fluid Flow for Spent Fuel in a Dry Storage

    International Nuclear Information System (INIS)

    In, Wangkee; Kwack, Youngkyun; Kook, Donghak; Koo, Yanghyun

    2014-01-01

    A dry storage system is used for the interim storage of spent fuel prior to permanent depository and/or recycling. The spent fuel is initially stored in a water pool for more than 5 years at least after dispatch from the reactor core and is transported to dry storage. The dry cask contains a multiple number of spent fuel assemblies, which are cooled down in the spent fuel pool. The dry cask is usually filled up with helium gas for increasing the heat transfer to the environment outside the cask. The dry storage system has been used for more than a decade in United States of America (USA) and the European Union (EU). Korea is also developing a dry storage system since its spent fuel pool is anticipated to be full within 10 years. The spent fuel will be stored in a dry cask for more than 40 years. The integrity and safety of spent fuel are important for long-term dry storage. The long-term storage will experience the degradation of spent fuel such as the embrittlement of fuel cladding, thermal creep and hydride reorientation. High burn-up fuel may expedite the material degradation. It is known that the cladding temperature has a strong influence on the material degradation. Hence, it is necessary to accurately predict the local distribution of the cladding temperature using the Computational Fluid Dynamics (CFD) approach. The objective of this study is to apply the CFD method for predicting the three-dimensional distribution of fuel temperature in a dry cask. This CFD study simulated the dry cask for containing the 21 fuel assemblies under development in Korea. This paper presents the fluid velocity and temperature distribution as well as the fuel temperature. A two-step CFD approach was applied to simulate the heat and fluid flow in a dry storage of 21 spent fuel assemblies. The first CFD analysis predicted the helium flow and temperature in a dry cask by a assuming porous body of the spent fuel. The second CFD analysis was to simulate a spent fuel assembly in the

  2. Corrosion surveillance program of aluminum spent fuel elements in wet storage sites

    International Nuclear Information System (INIS)

    Linardi, E; Haddad, R

    2012-01-01

    Due to different degradation issues observed in aluminum-clad spent fuel during long term storage in water, the IAEA implemented in 1996 a Coordinated Research Project (CRP) and a Regional Project for Latin America, on Corrosion of Research Reactor Aluminum Clad Spent Fuel in Water. Argentine has been among the participant countries of these projects, carrying out spent fuel corrosion surveillance activities in its storage facilities. As a result of the research a large database on corrosion of aluminum-clad fuel has been generated. It was determined that the main types of corrosion affecting the spent fuel are pitting and galvanic corrosion due to contact with stainless steel. It was concluded that the quality of the water is the critical factor to control in a spent fuel storage facility. Another phase of the program is being conducted currently, which began in 2011 with the immersion of test racks in the RA1 reactor pool, and in the Research Reactor Spent Fuel Storage Facility (FACIRI), located in Ezeiza Atomic Center. This paper presents the results of the chemical analysis of the water performed so far, and its relationship with the examination of the coupons extracted from the sites (author)

  3. Realistic evaluation of new fuel storage criticality

    International Nuclear Information System (INIS)

    Hamasaki, M.; Itahara, K.; Shimada, S.; Etsu, M.; Kuroda, M.; Watanabe, H.; Kuragasaki, M.; Kawaguchi, K.

    1987-01-01

    In the criticality safety design of dry fuel storage facility, the optimum moderation condition which appears at rather low water density (0.05 ∼ 0.1 gH 2 O/cc) has been a barrier to incorporate economical design. In this study we have shown that this optimum moderation does not occur, even in fire-fighting with water supply system, by criticality analyses and experiments. Evaluated critical amount of water in the space has found itself far from the level estimated actually attainable through the experiment with fire-fighting system. In another experiment we have taken the holograms of 0.003 ∼ 0.35 g/cc water which was realized in the very narrow space by the nozzle manufactured for this special purpose. Those holograms demonstrate that the droplets in the low density water are more closely packed than we anticipate they are. (author)

  4. Current activities on improving storage conditions of the research reactor RA spent fuel - Part II

    International Nuclear Information System (INIS)

    Matausek, M.V.; Kopecni, M.; Vukadin, Z.; Plecas, I.; Pavlovic, R.; Sotic, O.; Marinkovic, N.

    1998-01-01

    To minimize further corrosion and preserve integrity of aluminum barrels and the stainless steel channel-type containers that were found to contain leaking spent fuel, actions to improve conditions in the existing spent fuel storage pool at the RA research reactor were initiated. Technology was elaborated and equipment was produced and applied for removal of sludge and other debris from the bottom of the pool, filtration of the pool water, sludge conditioning in cement matrix and disposal at the low and medium waste repository at VINCA site. More sophisticated operations are to be performed together with foreign experts. Safety measures and precautions were determined. Subcriticality was proved under normal and/or possible abnormal conditions. (author)

  5. Storage experience in Hungary with fuel from research reactors

    International Nuclear Information System (INIS)

    Gado, J.; Hargitai, T.

    1996-01-01

    In Hungary several critical assemblies, a training reactor and a research reactor have been in operation. The fuel used in the research and training reactors are of Soviet origin. Though spent fuel storage experience is fairly good, medium and long term storage solutions are needed. (author)

  6. Management and storage of nuclear fuel from Belgian research reactors

    International Nuclear Information System (INIS)

    Gubel, P.

    1996-01-01

    Experiences and problems with the storage of irradiated fuel at research reactors in Belgium are described. In particular, interim storage problems exist for spent fuel elements at the BR2 and the shut down BR3 reactors in Mol. (author). 1 ref

  7. Spent fuel receipt and storage at the Morris Operation

    International Nuclear Information System (INIS)

    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

  8. Corrosion assessment of dry fuel storage containers

    Energy Technology Data Exchange (ETDEWEB)

    Graves, C.E.

    1994-09-01

    The structural stability as a function of expected corrosion degradation of 75 dry fuel storage containers located in the 200 Area Low-Level Waste Burial Grounds was evaluated. These containers include 22 concrete burial containers, 13 55-gal (208-l) drums, and 40 Experimental Breeder Reactor II (EBR-II) transport/storage casks. All containers are buried beneath at least 48 in. of soil and a heavy plastic tarp with the exception of 35 of the EBR-II casks which are exposed to atmosphere. A literature review revealed that little general corrosion is expected and pitting corrosion of the carbon steel used as the exterior shell for all containers (with the exception of the concrete containers) will occur at a maximum rate of 3.5 mil/yr. Penetration from pitting of the exterior shell of the 208-l drums and EBR-II casks is calculated to occur after 18 and 71 years of burial, respectively. The internal construction beneath the shell would be expected to preclude containment breach, however, for the drums and casks. The estimates for structural failure of the external shells, large-scale shell deterioration due to corrosion, are considerably longer, 39 and 150 years respectively for the drums and casks. The concrete burial containers are expected to withstand a service life of 50 years.

  9. Three-dimensional seismic analysis for spent fuel storage rack

    International Nuclear Information System (INIS)

    Lee, Gyu Mahn; Kim, Kang Soo; Park, Keun Bae; Park, Jong Kyun

    1998-01-01

    Time history analysis is usually performed to characterize the nonlinear seismic behavior of a spent fuel storage rack (SFSR). In the past, the seismic analyses of the SFSR were performed with two-dimensional planar models, which could not account for torsional response and simultaneous multi-directional seismic input. In this study, three-dimensional seismic analysis methodology is developed for the single SFSR using the ANSY code. The 3-D model can be used to determine the nonlinear behavior of the rack, i.e., sliding, uplifting, and impact evaluation between the fuel assembly and rack, and rack and the pool wall. This paper also reviews the 3-D modeling of the SFSR and the adequacy of the ANSYS for the seismic analysis. As a result of the adequacy study, the method of ANSYS transient analysis with acceleration time history is suitable for the seismic analysis of highly nonlinear structure such as an SFSR but it isn't appropriate to use displacement time history of seismic input. (author)

  10. Spent nuclear fuel assembly storage vessel

    International Nuclear Information System (INIS)

    Yagishita, Takuya

    1998-01-01

    The vessel of the present invention promotes an effect of removing after heat of spent nuclear fuel assemblies so as not to give force to the storage vessel caused by expansion of heat removing partitioning plates. Namely, the vessel of the present invention comprises a cylinder body having closed upper and lower portions and a plurality of heat removing partitioning cylinders disposed each at a predetermined interval in the circumferential direction of the above-mentioned cylinder body. The heat removing partitioning cylinders comprises (1) first heat removing partitioning plates extended in the radial direction of the cylinder body and opposed at a predetermined gap in the circumferential direction of the cylinder body, and having the base ends on the side of the inner wall of the cylinder body being secured to the inner wall of the cylinder body and (2) a second heat removing plate for connecting the top ends of both opposed heat removing partitioning plates on the central side of the cylinder body with each other. Spent nuclear fuel assemblies are contained in a plurality of closed spaces surrounded by the first heat removing partitioning plates and the second heat removing partitioning plate. With such constitution, since after heat is partially transferred from the heat removing partitioning plates to the cylindrical body directly by heat conduction, the heat removing effect can be promoted compared with the prior art. (I.S.)

  11. Storage rack for spent nuclear fuels

    International Nuclear Information System (INIS)

    Kiyama, Yoichi.

    1996-01-01

    A storage rack comprises a number of rack cells for containing spent nuclear fuels and two upper and lower rack support plates. Small through holes are formed to lateral walls of the rack cell each at a position slightly above the position of the upper rack support plate. Finger members each having a protrusion which fits the small through hole is secured at the upper surface of the upper rack support plate. The finger member is a metal leaf-spring erected at the periphery of a rack insertion hole of the rack support plate. Gaps for allowing thermal expansion of the rack cell are formed each between the edge of the rack cell insertion hole of the rack support plate and the rack cell, and between the lower edge of the small through hole on a side wall of the rack cell and the lower portion of the protrusion of the finger member. If the rack cell is inserted to a bottom, the protrusion of the finger member fits the small through hole on the side of the rack cell. With such a constitution, the rack cell is prevented from withdrawing in conjunction with removal of fuels. (I.N.)

  12. Dry storage of irradiated nuclear fuels and vitrified wastes

    International Nuclear Information System (INIS)

    Deacon, D.

    1982-01-01

    A review is given of the work of GEC Energy Systems Ltd. over the years in the dry storage of irradiated fuel. The dry-storage module (designated as Cell 4) for irradiated magnox fuel recently constructed at Wylfa nuclear power station is described. Development work on the long-term dry storage of irradiated oxide fuels is reported. Four different methods of storage are compared. These are the pond, vault, cask and caisson stores. It is concluded that there are important advantages with the passive air-cooled ESL dry stove. (U.K.)

  13. Failed (leaking) spent fuel management and storage in the Paks NPP

    International Nuclear Information System (INIS)

    Burjan, T.

    2011-01-01

    At the cycle 22, unit 4, Paks NPP the fissile contents raised irregularly in the water of the primary circuit. At the end of the cycle sipping tests were performed for the entire core to find out the leaking fuel assembly primarily responsible for this phenomenon. The identified leaking assembly temporarily was placed in the Spent Fuel Relaxing Pool. For measuring environmental impact of leaking assemblies an investigation program was developed and implemented. The assessment covered the following: effects of the leaking fuel on the water of relaxing pool and on the gaseous emissions in case open storage; in case when the leaking cassette is in a special hermetical storage case, how much gas is collected in the locked case and what is its composition; how to change the measured sipping test signal depending on relaxing time of leaking fuel cassettes. Based on the evaluation of the investigation program results the NPP modified the operational instructions for the treatment and storage of failed fuel assemblies. (author)

  14. Bidding strategy for pumped-storage plant in pool-based electricity market

    International Nuclear Information System (INIS)

    Kanakasabapathy, P.; Shanti Swarup, K.

    2010-01-01

    This paper develops optimal bidding strategies for a pumped-storage plant in a pool-based electricity market. In the competitive regime, when compared to simple hydroelectric generator, profit of the pumped-storage plant is maximized by operating it as a generator when market clearing price is high and as a pump when the price is low. Based on forecasted hourly market clearing price, a multistage looping algorithm to maximize the profit of a pumped-storage plant is developed, considering both the spinning and non-spinning reserve bids and meeting the technical operating constraints of the plant. The proposed model is adaptive for the nonlinear three-dimensional relationship between the power produced, the energy stored, and the head of the associated reservoir. Different operating cycles for a realistic pumped-storage plant are considered and simulation results are reported and compared. (author)

  15. Calculation of the external dose rate in the spent fuel pool for the case to use compact racks

    International Nuclear Information System (INIS)

    Passos, E.M. dos; Alves, A.S.M.

    1988-01-01

    The possible introduction of compact racks in the spent fuel pool of the Angra 1 Nuclear Power Plant largely inreases its storage capacity, but originates an increase of the gamma radiation sources. The precise evaluation of the effects of the adoption of this option on the external gamma dose rates and also on the thickness of the concrete shielding requires the utilization of sofisticated computer codes (QAD, ANISN), which allow the calculation of the gamma dose rates through thick shielding walls. This paper describes the utilized methodology for the calculation of the modified pool shieldings, showing the obtained results for the Angra 1 NPP case. The gamma dose rate was calculated with the point Kernel model, first analytically, and later through utilization of the tridimensional multigroup QAD computer code. (author) [pt

  16. At-reactor storage of spent fuel for life-of-plant

    International Nuclear Information System (INIS)

    Fuierer, A.A.

    1990-01-01

    The management of commercial spent fuel is a fairly broad topic beginning with the discharge from a reactor, its storage on-site, its transport from the reactor site to a U.S. Department of Energy (DOE) facility, and its ultimate disposal in a geologic repository. This paper discusses spent-fuel management in the at-reactor phase. There are two basic methods for at-reactor storage of spent fuel. The first is wet storage in a pool, and the second is dry storage external to the plant in some form of cask or vault. Spent-fuel consolidation will impact the utility and the DOE waste system. Some of these impacts have a positive effect and some have a negative effect, and each will vary somewhat for each utility. Spent-fuel consolidation and life-of-plant storage will be an increased burden to utilities but will likely result in significant cost savings to the overall waste management system and by proper integration can result in significant institutional benefits

  17. Verification of criticality safety in on-site spent fuel storage systems

    International Nuclear Information System (INIS)

    Rasmussen, R.W.

    1989-01-01

    On February 15, 1984, Duke Power Company received approval for a two-region, burnup credit, spent fuel storage rack design at both Units 1 and 2 of the McGuire Nuclear Station. Duke also hopes to obtain approval by January of 1990 for a dry spent fuel storage system at the Oconee Nuclear Station, which will incorporate the use of burnup credit in the criticality analysis governing the design of the individual storage units. While experiences in burnup verification for criticality safety for their dry storage system at Oconee are in the future, the methods proposed for burnup verification will be similar to those currently used at the McGuire Nuclear Station in the two-region storage racks installed in both pools. In conclusion, the primary benefit of the McGuire rerack effort has obviously been the amount of storage expansion it provided. A total increase of about 2,000 storage cells was realized, 1,000 of which were the result of pursuing the two-region rather than the conventional poison rack design. Less impacting, but equally as important, however, has been the experience gained during the planning, installation, and operation of these storage racks. This experience should prove useful for future rerack efforts likely to occur at Duke's Catawba Nuclear Station as well as for the current dry storage effort underway for the Oconee Nuclear Station

  18. Characterization of the storage pool of the Neutron Standards Laboratory of CIEMAT, using Monte Carlo techniques

    Energy Technology Data Exchange (ETDEWEB)

    Campo B, X.; Mendez V, R.; Embid S, M. [Centro de Investigaciones Energeticas, Medioambientales y Tecnologicas, Av. Complutense 40, 28040 Madrid (Spain); Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98060 Zacatecas (Mexico); Sanz G, J., E-mail: xandra.campo@ciemat.es [Universidad Nacional de Educacion a Distancia, Escuela Tecnica Superior de Ingenieros Industriales, C. Juan del Rosal 12, 28040 Madrid (Spain)

    2014-08-15

    Neutron Standards Laboratory of CIEMAT in Spain is a brand new irradiation facility, with {sup 241}Am-Be (185 GBq) and {sup 252}Cf (5 GBq) calibrated neutron sources which are stored in a water pool with a concrete cover. From this storage place an automated system is able to take the selected source and place it in the irradiation position, 4 m over the ground level and in the geometrical center of the Irradiation Room with 9 m (length) x 7.5 m (width) x 8 m (height). For calibration or irradiation purposes, detectors or materials can be placed on a bench but it is possible to use the pool (1.0 m x 1.5 m and more than 1.0 m depth) for long time irradiations in thermal neutron fields. For this reason it is essential to characterize the pool itself in terms of neutron spectrum. In this document, the main features of this facility are presented and the characterization of the storage pool in terms of neutron fluence rate and neutron spectrum has been carried out using simulations with MCNPX-2.7.e code. The MCNPX-2.7.e model has been validated using experimental measurements outside the pool (Bert hold LB6411). Inside the pool, the fluence rate decreases and the spectra is thermalized with the distance to the {sup 252}Cf source. This source predominates and the effect of the {sup 241}Am-Be source in these magnitudes is not shown until positions closer than 20 cm from it. (author)

  19. Characterization of the storage pool of the Neutron Standards Laboratory of CIEMAT, using Monte Carlo techniques

    International Nuclear Information System (INIS)

    Campo B, X.; Mendez V, R.; Embid S, M.; Vega C, H. R.; Sanz G, J.

    2014-08-01

    Neutron Standards Laboratory of CIEMAT in Spain is a brand new irradiation facility, with 241 Am-Be (185 GBq) and 252 Cf (5 GBq) calibrated neutron sources which are stored in a water pool with a concrete cover. From this storage place an automated system is able to take the selected source and place it in the irradiation position, 4 m over the ground level and in the geometrical center of the Irradiation Room with 9 m (length) x 7.5 m (width) x 8 m (height). For calibration or irradiation purposes, detectors or materials can be placed on a bench but it is possible to use the pool (1.0 m x 1.5 m and more than 1.0 m depth) for long time irradiations in thermal neutron fields. For this reason it is essential to characterize the pool itself in terms of neutron spectrum. In this document, the main features of this facility are presented and the characterization of the storage pool in terms of neutron fluence rate and neutron spectrum has been carried out using simulations with MCNPX-2.7.e code. The MCNPX-2.7.e model has been validated using experimental measurements outside the pool (Bert hold LB6411). Inside the pool, the fluence rate decreases and the spectra is thermalized with the distance to the 252 Cf source. This source predominates and the effect of the 241 Am-Be source in these magnitudes is not shown until positions closer than 20 cm from it. (author)

  20. Analysis of the impact of retrievable spent fuel storage

    International Nuclear Information System (INIS)

    Merrill, E.T.; White, M.K.; Fleischman, R.M.

    1978-03-01

    The impact of retrievably storing spent fuel is measurable in terms of the contribution the stored spent fuel makes to implementing the fuel management option selected. For the case of a decision to recycle LWR fuel in LWRs, a useful indicator of impact is the ratio of energy production with varying degrees of spent fuel retrievability to that achievable with total spent fuel retrievability. For a decision made in the year 2000, this ratio varies from 0.81 (10 yr storage in reactor basins) to 0.97 (retrievable storage for 25 years after fuel discharge). An earlier decision to recycle in LWRs results in both of these ratios being nearer to 1.0. If a decision is reached to implement a breeder reactor economy, the chosen comparison is the installed breeder capacity achievable with varying degrees of spent fuel retrievability. If a decision to build breeder reactors is reached in the year 2000, the maximum possible installed breeder capacity in 2040 varies from 490 GWe (10 yr storage in reactor basins) to 660 GWe (all fuel retrievably stored). If all fuel is retrievably stored 25 years, 635 GWe of breeder capacity is achievable by 2040. For an earlier decision date, such as 1985, the maximum possible installed breeder capacity in 2040 ranges from 740 GWe (no retrievable storage) to 800 GWe (all fuel retrievably stored). As long as a decision to reprocess is reached before 2000, most of the potential benefit of retrievable storage may be realized by implementing retrievable storage after such a decision is made. Neither providing retrievable spent fuel storage prior to a decision to reprocess, nor designing such storage for more than 25 years of retrievability appear to offer significant incremental benefit

  1. Long-term storage of spent nuclear fuel

    International Nuclear Information System (INIS)

    Kempe, T.F.; Martin, A.; Thorne, M.C.

    1980-06-01

    This report presents the results of a study on the storage of spent nuclear fuel, with particular reference to the options which would be available for long-term storage. Two reference programmes of nuclear power generation in the UK are defined and these are used as a basis for the projection of arisings of spent fuel and the storage capacity which might be needed. The characteristics of spent fuel which are relevant to long-term storage include the dimensions, materials and physical construction of the elements, their radioactive inventory and the associated decay heating as a function of time after removal from the reactor. Information on the behaviour of spent fuel in storage ponds is reviewed with particular reference to the corrosion of the cladding. The review indicates that, for long-term storage, both Magnox and AGR fuel would need to be packaged because of the high rate of cladding corrosion and the resulting radiological problems. The position on PWR fuel is less certain. Experience of dry storage is less extensive but it appears that the rate of corrosion of cladding is much lower than in water. Unit costs are discussed. Consideration is given to the radiological impact of fuel storage. (author)

  2. Nuclear criticality assessment of LEU and HEU fuel element storage

    International Nuclear Information System (INIS)

    Pond, R.B.; Matos, J.E.

    1984-01-01

    Criticality aspects of storing LEU (20%) and HEU (93%) fuel elements have been evaluated as a function of 235 U loading, element geometry, and fuel type. Silicide, oxide, and aluminide fuel types have been evaluated ranging in 235 U loading from 180 to 620 g per element and from 16 to 23 plates per element. Storage geometry considerations have been evaluated for fuel element separations ranging from closely packed formations to spacings of several centimeters between elements. Data are presented in a form in which interpolations may be made to estimate the eigenvalue of any fuel element storage configuration that is within the range of the data. (author)

  3. Contingency strategy for insufficient full core off load capability in spent fuel pool for Chinshan nuclear power station

    International Nuclear Information System (INIS)

    Huang, Pinghue

    2012-01-01

    The spent fuel pool (SFP) at Taiwan Power Company's (TUC's) Chinshan plant lost the full core off load (FCO) capability in 2010, even with the second SFP repacking project to expand the capacity as reported in 12PBNC. The TEPC had originally planned to move some spent fuel assemblies from SFP to dry storage facility, however, the dry storage project had seriously fell behind. Thus, it is required to address insufficient FCO capability, and the following contingency measures have been employed: The first step was to explore whether there was a specific regulatory requirement for FCO capability, and none were identified. Also, the industrial experiences were explored. The refueling strategy is changed from FCO to in-core shuffling. A feasibility evaluation performed indicates the Technical Specifications require: alternate method of decay heat removal, and verification of shutdown margin for each in vessel fuel movement. Specific methods have been successfully established. A safety evaluation for operation without FCO capability was performed, and no safety concerns were identified. The risk for operation without FCO capability was assessed. The previous operational experiences were identified. Moreover, such works are not expected in subsequent cycles. The new fuel vault is used to store new fuel assemblies. The criticality analysis has been performed and some new approaches are proposed to enhance the storage flexibility as reported in 17PBNC. An inter-unit transfer cask has been designed to transfer spent fuel from the SFP of one unit to the other. The FCO capability can be effectively extended for three more years with this consideration. The TPC discussed the contingency strategy with the ROCAEC in May 2006, and the ROCAEC's concurrence was attained. With the proposed strategy, Chinshan units have been operating smoothly

  4. Contingency strategy for insufficient full core off load capability in spent fuel pool for Chinshan nuclear power station

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Pinghue [Taiwan Power Company, Taipei (China)

    2012-03-15

    The spent fuel pool (SFP) at Taiwan Power Company's (TUC's) Chinshan plant lost the full core off load (FCO) capability in 2010, even with the second SFP repacking project to expand the capacity as reported in 12PBNC. The TEPC had originally planned to move some spent fuel assemblies from SFP to dry storage facility, however, the dry storage project had seriously fell behind. Thus, it is required to address insufficient FCO capability, and the following contingency measures have been employed: The first step was to explore whether there was a specific regulatory requirement for FCO capability, and none were identified. Also, the industrial experiences were explored. The refueling strategy is changed from FCO to in-core shuffling. A feasibility evaluation performed indicates the Technical Specifications require: alternate method of decay heat removal, and verification of shutdown margin for each in vessel fuel movement. Specific methods have been successfully established. A safety evaluation for operation without FCO capability was performed, and no safety concerns were identified. The risk for operation without FCO capability was assessed. The previous operational experiences were identified. Moreover, such works are not expected in subsequent cycles. The new fuel vault is used to store new fuel assemblies. The criticality analysis has been performed and some new approaches are proposed to enhance the storage flexibility as reported in 17PBNC. An inter-unit transfer cask has been designed to transfer spent fuel from the SFP of one unit to the other. The FCO capability can be effectively extended for three more years with this consideration. The TPC discussed the contingency strategy with the ROCAEC in May 2006, and the ROCAEC's concurrence was attained. With the proposed strategy, Chinshan units have been operating smoothly.

  5. Fuel handling and storage systems in nuclear power plants

    International Nuclear Information System (INIS)

    1984-01-01

    The scope of this Guide includes the design of handling and storage facilities for fuel assemblies from the receipt of fuel into the nuclear power plant until the fuel departs from that plant. The unirradiated fuel considered in this Guide is assumed not to exhibit any significant level of radiation so that it can be handled without shielding or cooling. This Guide also gives limited consideration to the handling and storage of certain core components. While the general design and safety principles are discussed in Section 2 of this Guide, more specific design requirements for the handling and storage of fuel are given in detailed sections which follow the general design and safety principles. Further useful information is to be found in the IAEA Technical Reports Series No. 189 ''Storage, Handling and Movement of Fuel and Related Components at Nuclear Power Plants'' and No. 198 ''Guide to the Safe Handling of Radioactive Wastes at Nuclear Power Plants''. However, the scope of the Guide does not include consideration of the following: (1) The various reactor physics questions associated with fuel and absorber loading and unloading into the core; (2) The design aspects of preparation of the reactor for fuel loading (such as the removal of the pressure vessel head for a light water reactor) and restoration after loading; (3) The design of shipping casks; (4) Fuel storage of a long-term nature exceeding the design lifetime of the nuclear power plant; (5) Unirradiated fuel containing plutonium

  6. Conceptual aspects of the safety evaluation of a project of complementary spent nuclear fuel dry storage unit

    Energy Technology Data Exchange (ETDEWEB)

    Freitas, Rafaela da S. A.; Fontes, Gladson S., E-mail: rafaaelaandrade@hotmail.com, E-mail: gsfontes@hotmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil); Saldanha, Pedro L. C., E-mail: saldanha@cnen.gov.br [Comissão Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    Based on the number of cycles and the amount of new fuel elements exchanged in the reactor cores at each cycle, the forecast for the exhaustion of the spent nuclear fuel pools of the Brazil plants has provision until 2021. As are still in the studies the availability of a long-term storage facility for spent fuel, the short-term solution will be the construction of the Complementary Storage Spent Nuclear Fuel Unit, it will build inside the site in Angra Plants. The dry cask is a method of storage in which the fuel elements of high-level radioactive waste are stored, such as spent nuclear fuel, which already cooled in the fuel pool for at least one year and up to ten years. The purpose of the present paper is to discuss a conceptual study of the safety analysis of a project of licensing of a Dry Storage Unit (DSU) with the objective of verifying the application of national and international criteria, requirements and standards. The safety analysis will make on the principles adopted by the US Nuclear USNRC and the standards adopted at CNEN for dry storage. The concept of installation, seismic, geological and other analysis will be approached for approval of the site to be installed at DSU, the approved permit for the construction and finally the external and internal events that may occur being incidents and / or accidents and which are The necessary mitigations if something occurs within a period of time. (author)

  7. Conceptual aspects of the safety evaluation of a project of complementary spent nuclear fuel dry storage unit

    International Nuclear Information System (INIS)

    Freitas, Rafaela da S. A.; Fontes, Gladson S.; Saldanha, Pedro L. C.

    2017-01-01

    Based on the number of cycles and the amount of new fuel elements exchanged in the reactor cores at each cycle, the forecast for the exhaustion of the spent nuclear fuel pools of the Brazil plants has provision until 2021. As are still in the studies the availability of a long-term storage facility for spent fuel, the short-term solution will be the construction of the Complementary Storage Spent Nuclear Fuel Unit, it will build inside the site in Angra Plants. The dry cask is a method of storage in which the fuel elements of high-level radioactive waste are stored, such as spent nuclear fuel, which already cooled in the fuel pool for at least one year and up to ten years. The purpose of the present paper is to discuss a conceptual study of the safety analysis of a project of licensing of a Dry Storage Unit (DSU) with the objective of verifying the application of national and international criteria, requirements and standards. The safety analysis will make on the principles adopted by the US Nuclear USNRC and the standards adopted at CNEN for dry storage. The concept of installation, seismic, geological and other analysis will be approached for approval of the site to be installed at DSU, the approved permit for the construction and finally the external and internal events that may occur being incidents and / or accidents and which are The necessary mitigations if something occurs within a period of time. (author)

  8. The Impact of Microbially Influenced Corrosion on Spent Nuclear Fuel and Storage Life

    International Nuclear Information System (INIS)

    Wolfram, J. H.; Mizia, R. E.; Jex, R.; Nelson, L.; Garcia, K. M.

    1996-01-01

    A study was performed to evaluate if microbial activity could be considered a threat to spent nuclear fuel integrity. The existing data regarding the impact of microbial influenced corrosion (MIC) on spent nuclear fuel storage does not allow a clear assessment to be made. In order to identify what further data are needed, a literature survey on MIC was accomplished with emphasis on materials used in nuclear fuel fabrication, e.g., A1, 304 SS, and zirconium. In addition, a survey was done at Savannah River, Oak Ridge, Hanford, and the INEL on the condition of their wet storage facilities. The topics discussed were the SNF path forward, the types of fuel, ramifications of damaged fuel, involvement of microbial processes, dry storage scenarios, ability to identify microbial activity, definitions of water quality, and the use of biocides. Information was also obtained at international meetings in the area of biological mediated problems in spent fuel and high level wastes. Topics dis cussed included receiving foreign reactor research fuels into existing pools, synergism between different microbes and other forms of corrosion, and cross contamination

  9. The Impact of Microbially Influenced Corrosion on Spent Nuclear Fuel and Storage Life

    Energy Technology Data Exchange (ETDEWEB)

    J. H. Wolfram; R. E. Mizia; R. Jex; L. Nelson; K. M. Garcia

    1996-10-01

    A study was performed to evaluate if microbial activity could be considered a threat to spent nuclear fuel integrity. The existing data regarding the impact of microbial influenced corrosion (MIC) on spent nuclear fuel storage does not allow a clear assessment to be made. In order to identify what further data are needed, a literature survey on MIC was accomplished with emphasis on materials used in nuclear fuel fabrication, e.g., A1, 304 SS, and zirconium. In addition, a survey was done at Savannah River, Oak Ridge, Hanford, and the INEL on the condition of their wet storage facilities. The topics discussed were the SNF path forward, the types of fuel, ramifications of damaged fuel, involvement of microbial processes, dry storage scenarios, ability to identify microbial activity, definitions of water quality, and the use of biocides. Information was also obtained at international meetings in the area of biological mediated problems in spent fuel and high level wastes. Topics dis cussed included receiving foreign reactor research fuels into existing pools, synergism between different microbes and other forms of corrosion, and cross contamination.

  10. Scheme of higher-density storage of spent nuclear fuel in Chernobyl NPP interim storage facility no. 1

    International Nuclear Information System (INIS)

    Britan, P.M.

    2008-01-01

    On 29. March 2000 the Cabinet of Ministers of Ukraine issued a decree prescribing that the last operating unit of Chernobyl NPP be shut down before its design lifetime expiry. In accordance with the Contract concluded on 14 June 1999 between the National Energy-generating Company 'Energoatom' and the Consortium of Framatome, Campenon Bernard-SGE and Bouygues, in order to store the spent ChNPP fuel a new interim dry storage facility (ISF-2) for spent ChNPP fuel would be built. Currently the spent nuclear fuel (spent fuel assemblies - SFAs) is stored in reactor cooling pools and in the reactors on Units 1, 2, 3, as well as in the wet Interim Storage Facility (ISF-1). Taking into account the expected delay with the commissioning of ISF-2, and in connection with the scheduled activities to build the New Safe Confinement (including the taking-down of the existing ventilation stack of ChNPP Units 3 and 4) and the expiry of the design operation life of Units 1 and 2, it is expedient to remove the nuclear fuel from Units 1, 2 and 3. This is essential to improve nuclear safety and ensure that the schedule of construction of the New Safe Confinement is met. The design capacity of ISF-1 (17 800 SFAs) is insufficient to store all SFAs (21 284) currently on ChNPP. A technically feasible option that has been applied on other RBMK plants is denser storage of spent nuclear fuel in the cooling ponds of the existing ISF-1. The purpose of the proposed modifications is to introduce changes to the ISF-1 design supported by necessary justifications required by the Ukrainian codes with the objective of enabling the storage of additional SFAs in the existing storage space (cooling pools). The need for the modification is caused by the requirement to remove nuclear fuel from the ChNPP units as soon as possible, before the work begins to decommission these units, as well as to create safe conditions for the construction of the New Safe Confinement over the existing Shelter Unit. (author)

  11. Structural analyses of the fuel receiving station pool at the Nuclear Fuel Service reprocessing plant, West Valley, New York

    International Nuclear Information System (INIS)

    Dong, R.G.; Ma, S.M.

    1978-01-01

    The FRS is a pool structure and enclosing building constructed in 1966 for storing spent nuclear fuel. The enclosing building was not analyzed. The pool structure's responses to operating loads, seismic excitation, and an accidentally dropped cask were determined. Locations in the FRS pool were identified where structural strength would be exceeded in the event of an earthquake of 0.2 g maximum ground acceleration or an accident in which a cask dropped from the maximum height of the crane hook used to maneuver it. 25 figures, 4 tables

  12. Survey of wet and dry spent fuel storage

    International Nuclear Information System (INIS)

    1999-07-01

    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

  13. American proposals for long range storage of irradiated fuel

    International Nuclear Information System (INIS)

    Sugier, Annie

    1978-01-01

    The American politics of irradiated fuel management is reviewed, the short-range storage of huge amounts of wastes being the fundamental problem. Two steps are considered: the ''At the Reactor'' storage, ensured by the electricity companies, and the ''Away From Reactor'' storage on the DOE's responsibility. A technical and economical study has been carried out in order to estimate the cost of the AFR provisory storage and a project of taxation has been established on this basis [fr

  14. American proposals for long range storage of irradiated fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sugier, A [CEA, 75 - Paris (France). Dept. des Programmes

    1978-12-01

    The American politics of irradiated fuel management is reviewed, the short-range storage of huge amounts of wastes being the fundamental problem. Two steps are considered: the ''At the Reactor'' storage, ensured by the electricity companies, and the ''Away From Reactor'' storage on the DOE's responsibility. A technical and economical study has been carried out in order to estimate the cost of the AFR provisory storage and a project of taxation has been established on this basis.

  15. Fuel consolidation and compaction and storage of NFBC

    International Nuclear Information System (INIS)

    Fuierer, T.

    1992-01-01

    Rochester Gas and Electric Corporation (RG ampersand E) has been involved in two separate fuel consolidation demonstration programs. One of those programs resulted in identifying some problems that may be resolved in consolidation hardware compaction and storage in order for consolidation to be attractive. In conjunction with the Electric Power Research Institute (EPRI), a study was recently performed on hardware compaction and storage. Consolidation is probably not a commercial alternative at this point in time because there are still several problems that must be resolved. There are some potential advantages of fuel consolidation. Consolidation has attractive economics and can minimize the institutional impacts of expanding spent fuel storage by internalizing spent fuel storage operations. The licensing effort is fairly simple. Consolidation may be less likely to have public intervention since the storage expansion will occur inside the plant. Consolidation can be subcontracted and the equipment is temporary. It can be used in conjunction with other storage expansion technologies such as dry storage. Fewer dry storage casks would be needed to store consolidated fuel than would be necessary for intact spent fuel

  16. Current state and perspectives of spent fuel storage in Russia

    International Nuclear Information System (INIS)

    Kurnosov, V.A.; Tichonov, N.S.; Makarchuk, T.F.

    1999-01-01

    Twenty-nine power units at nine nuclear power plants, having a total installed capacity of 22 GW(e), are now in operation in the Russian Federation. They produce approximately 12% of the generated electricity in the country. The annual spent fuel arising is approximately 790 tU. The concept of the closed fuel cycle was adopted as the basis for nuclear power development in the Russian Federation, but until now this concept is only implemented for the fuel cycles of WWER-440 and BN-600 reactors. The WWER-1000 spent fuel is planned to be reprocessed at the reprocessing plant RT-2 which is under construction near Krasnoyarsk. The RBMK-1000 spent fuel is not reprocessed. It is meant to be stored in intermediate storage facilities at the NPP sites. The status of the spent fuel (SF) stored in the storage facilities is given in the paper. The principal characteristics of the fuel cycles of the Russian NPPs in the period up to 2015 is also given in the report. The key variant of the current spent fuel management at RBMK-1000 NPPs is storage in at-reactor and in away-from-reactor wet storage facilities at the power plant site with a capacity of 2,000 W. The storage capacity at the operating RBMKs (including the increase due to denser fuel assembly arrangement) will provide SF reception from the NPPs only up to 2005. For RBMK spent fuel, intermediate dry storage is foreseen at power plant sites in metallic concrete casks and thereafter transportation to the central storage facility at the RT-2 plant for long-term storage. The SF will be reprocessing after completion of the reprocessing plant at RT-2. In the Programme of Nuclear Power Development in the Russian Federation for the period 1998 to 2005 and for the period until 2010 year, provisions are made for the construction of a central dry storage facility before 2010. The facility will have a design capacity of 30,000 tU for WWER-1000 and RBMK-1000 spent fuel and is part of the reprocessing plant RT-2. The paper considers

  17. Maintaining Continuity of Knowledge of Spent Fuel Pools: Tool Survey

    Energy Technology Data Exchange (ETDEWEB)

    Benz, Jacob M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Smartt, Heidi A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Tanner, Jennifer E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); MacDougall, Matthew R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-08-30

    This report examines supplemental tools that can be used in addition to optical surveillance cameras to maintain CoK in low-to-no light conditions, and increase the efficiency and effectiveness of spent fuel CoK, including item counting and ID verification, in challenging conditions.

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

    International Nuclear Information System (INIS)

    Hodgson, Z.; Hambley, D.I.; Gregg, R.; Ross, D.N.

    2013-01-01

    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. Safety assessment of OPG's used fuel for dry storage

    International Nuclear Information System (INIS)

    Roman, H.; Khan, A.

    2005-01-01

    'Full text:' Ontario Power Generation (OPG) operates the Pickering Waste Management Facility (PWMF) and Western Waste Management Facility (WWMF) where OPG has been storing 10-year or older used fuel in the Dry Storage Containers (DSCs) since 1996 and 2003 respectively. The construction licence for the Darlington Used Fuel Dry Storage Facility (DUFDSF) was obtained in August 2004. Safety assessment of the used fuel for dry storage is required to support each request for regulatory approval to construct and operate a dry storage facility. The objective of the safety assessment is to assess the used fuel performance under normal operation and postulated credible accident scenarios. A reference used fuel bundle is defined based on the operating history and data on fuel discharged from the reactors of the specific nuclear generating station. The characteristics of the reference used fuel bundle are used to calculate the nuclide inventory, source term and decay heat used for the assessment. When assessing malfunctions and accidents, postulated external and internal events are considered. Consideration is also given to the design basis accidents of the specific nuclear generating station that could affect the used fuel under dry storage. For those events deemed credible (i.e. probability > 10 -7 ), a bounding fuel failure consequence is predicted. Given the chemical characteristics of the radionuclides in used fuel, the design of the CANDU fuel and the conditions inside the DSC, in the event that a used fuel bundle should become damaged during used fuel dry storage operations, the only significant radionuclides species that are volatile are krypton-85 and tritium. Release of these radionuclides is considered in calculating public and worker doses. (author)

  20. Concept study for interim storage of research reactor fuel elements in transport and storage casks. Transport and storage licensing procedure for the CASTOR MTR 2 cask. Final report

    International Nuclear Information System (INIS)

    Weiss, M.

    2001-01-01

    As a result of the project, a concept was to be developed for managing spent fuel elements from research reactors on the basis of the interim storage technology existing in Germany, in order to make the transition to direct disposal possible in the long term. This final report describes the studies for the spent fuel management concept as well as the development of a transport and storage cask for spent fuel elements from research reactors. The concept analyses were based on data of the fuel to be disposed of, as well as the handling conditions for casks at the German research reactors. Due to the quite different conditions for handling of casks at the individual reactors, it was necessary to examine different cask concepts as well as special solutions for loading the casks outside of the spent fuel pools. As a result of these analyses, a concept was elaborated on the basis of a newly developed transport and storage cask as well as a mobile fuel transfer system for the reactor stations, at which a direct loading of the cask is not possible, as the optimal variant. The cask necessary for this concept with the designation CASTOR trademark MTR 2 follows in ist design the tried and tested principles of the CASTOR trademark casks for transport and interim storage of spent LWR fuel. With the CASTOR trademark MTR 2, it is possible to transport and to place into long term interim storage various fuel element types, which have been and are currently used in German research reactors. The technical development of the cask has been completed, the documents for the transport license as type B(U)F package design and for obtaining the storage license at the interim storage facility of Ahaus have been prepared, submitted to the licensing authorities and to a large degree already evaluated positively. The transport license of the CASTOR trademark MTR 2 has been issued for the shipment of VKTA-contents and FRM II compact fuel elements. (orig.)

  1. Cna 1 spent fuel element interim dry storage system thermal analysis

    International Nuclear Information System (INIS)

    Hilal, R. E; Garcia, J. C; Delmastro, D. F

    2006-01-01

    At the moment, the Atucha I Nuclear Power Plant (Cnea-I) located in the city of Lima, has enough room to store its spent fuel (Sf) in their two pools spent fuel until about 2015.In case of life extension a spend fuel element interim dry storage system is needed.Nucleolectrica Argentina S.A. (N A-S A) and the Comision Nacional de Energia Atomica (Cnea), have proposed different interim dry storage systems.These systems have to be evaluated in order to choose one of them.The present work's objective is the thermal analysis of one dry storage alternative for the Sf element of Cna 1.In this work a simple model was developed and used to perform the thermal calculations corresponding to the system proposed by Cnea.This system considers the store of sealed containers with 37 spent fuels in concrete modules.Each one of the containers is filled in the pool houses and transported to the module in a transference cask with lead walls.Fulfill the maximum cladding temperature requirement ( [es

  2. Existing and near future practices of spent fuel storage in Slovak Republic

    International Nuclear Information System (INIS)

    Mizov, J.

    1999-01-01

    In this paper existing and near future practices of spent fuel storage in Slovak Republic are discussed: (1) Reactor operation and spent fuel production; (2) Past policy in spent fuel storage; (3) Away-from-reactor (AFR) storage facility at Bohunice NPP site; (4) Present policy in spent fuel storage; (5) Final disposal of spent fuel

  3. Status and operational experience report of spent fuel storage facility in Kozloduy NPP for the period 1990 - 1994

    Energy Technology Data Exchange (ETDEWEB)

    Kalimanov, M [Kombinat Atomna Energetika, Kozloduj (Bulgaria)

    1994-12-31

    Spent Fuel Storage Facility (SFSF) of Kozloduy NPP is designed for a long-term storage of 4920 spent fuel assemblies which are generated by all units for ten year operational period. The assemblies are stored in SFSF after 3 year storage in the reactor cooling pool. The SFSF operational safety is ensured by a number of strictly followed regulations related to: arrangement of the assemblies and conditions at which they are stored; transportation of the assemblies to the facility; residual heat removal; quality of the water used in the storage pool; water temperature and level control. Two independent groups of experts have carried out investigations to study the building safety. Their reports have been considered and accepted by the council of the Ministry of Environment which was the final step in licensing the SFSF.

  4. Sensitivity studies of some important parameters in the criticality of nuclear fuel pool

    International Nuclear Information System (INIS)

    Pina, C.

    1981-01-01

    The criticality study of spent fuels storage aiming to verify this security. A small description of the optimal moderation phenomena as well as the effect of stainless steel plates as neutron absorbed material, arre presented [pt

  5. Comparative economics for DUCRETE spent fuel storage cask handling, transportation, and capital requirements

    International Nuclear Information System (INIS)

    Powell, F.P.

    1995-04-01

    This report summarizes economic differences between a DUCRETE spent nuclear fuel storage cask and a conventional concrete storage cask in the areas of handling, transportation, and capital requirements. The DUCRETE cask is under evaluation as a new technology that could substantially reduce the overall costs of spent fuel and depleted U disposal. DUCRETE incorporates depleted U in a Portland cement mixture and functions as the cask's primary radiation barrier. The cask system design includes insertion of the US DOE Multi-Purpose Canister inside the DUCRETE cask. The economic comparison is from the time a cask is loaded in a spent fuel pool until it is placed in the repository and includes the utility and overall US system perspectives

  6. Compacted spent-fuel storage--designs and problems

    International Nuclear Information System (INIS)

    Rubinstein, H.J.; Gilcrest, J.D.; Kendall, W.R.

    1979-01-01

    Typical rack designs, licensing, contracting methods, installation and operational problems are described. Due to the lack of reprocessing and independent fuel storage facilities, new plants built in the United States will continue to install high-density spent-fuel storage racks. As to the rack designs, the most significant feature is the introduction of freestanding rack designs. The trends in spent-fuel storage appear to be toward the use of high-density racks, either with or without absorber, for all plants in the design, construction, or operation phase; the use of freestanding rack designs; and the separation of engineering and fabrication during procurement

  7. Verification of heat removal capability of a concrete cask system for spent fuel storage

    International Nuclear Information System (INIS)

    Sakai, Mikio; Fujiwara, Hiroaki; Sakaya, Tadatugu

    2001-01-01

    The reprocessing works comprising of a center of nuclear fuel cycle in Japan is now under construction at Rokkasho-mura in Aomori prefecture, which is to be operated in 2005. However, as reprocessing capacity of the works is under total forming amount of spent nuclear fuels, it has been essential to construct a new facility intermediately to store them at a period before reprocessing them because of prediction to reach limit of pool storage in nuclear power stations. There are some intermediate storage methods, which are water pool method for wet storage, and bolt method, metal cask method, silo method and concrete cask method for dry storage. Among many methods, the dry storage is focussed at a standpoint of its operability and economy, the concrete cask method which has a lot of using results in U.S.A. has been focussed as a method expectable in its cost reduction effect among it. The Ishikawajima-Harima Heavy Industries Co., Ltd. produced, in trial, a concrete cask with real size to confirm productivity when advancing design work on concrete cask. By using the trial product, a heat removal test mainly focussing temperature of concrete in the cask was carried out to confirm heat conductive performances of the cask. And, analysis of heat conductivity was also carried out to verify validity of its analysis model. (G.K.)

  8. The status of spent fuel storage in the UK

    International Nuclear Information System (INIS)

    Dunn, M.J.; Topliss, I.R.

    1999-01-01

    Nuclear generating capacity in the UK is static with no units currently under construction. There are three main nuclear fuel types used in the UK for Magnox reactors, AGRs and PWRs. All Magnox fuel will ultimately be reprocessed following a short period of interim storage. AGR fuel will either be reprocessed or long term stored in ponds. PWR fuel will be stored underwater at the reactor site for the foreseeable future, with no decision as yet made to its ultimate management route. (author)

  9. International long-term interim storage for spent fuel. An independent storage service investor model

    International Nuclear Information System (INIS)

    Leister, P.

    1999-01-01

    Thinking globally the obvious world-wide demands for large storage capacities for spent fuel within the next decades and the newly arising demands for long-term interim storage of spent fuel urges to respond by international interim storage facilities of high capacity. Low cost storage can be achieved only by arranging the storage facility underground in a suitable host rock formation and by selecting the geographical are by an international competition under those countries, who are willing to offer their land. The investor and operator of an international storage facility selected and realised by a competition on the free market as well as the country where the storage is built are both bound by two different kinds of contacts. The main contract is between the offering country/region and the independent operator. The independent operator has in addition a series of contracts with various utilities, which are interested to have their spent fuel stored for a longer period

  10. Performance evaluation of METAMIC neutron absorber in spent fuel storage rack

    Directory of Open Access Journals (Sweden)

    Kiyoung Kim

    2018-06-01

    Full Text Available High-density spent fuel (SF storage racks have been installed to increase SF pool capacity. In these SF racks, neutron absorber materials were placed between fuel assemblies allowing the storage of fuel assemblies in close proximity to one another. The purpose of the neutron absorber materials is to preclude neutronic coupling between adjacent fuel assemblies and to maintain the fuel in a subcritical storage condition. METAMIC neutron absorber has been used in high-density storage racks. But, neutron absorber materials can be subject to severe conditions including long-term exposure to gamma radiation and neutron radiation. Recently, some of them have experienced degradation, such as white spots on the surface. Under these conditions, the material must continue to serve its intended function of absorbing neutrons. For the first time in Korea, this article uses a neutron attenuation test to examine the performance of METAMIC surveillance coupons. Also, scanning electron microscope analysis was carried out to verify the white spots that were detected on the surface of METAMIC. In the neutron attenuation test, there was no significant sign of boron loss in most of the METAMIC coupons, but the coupon with white spots had relatively less B-10 content than the others. In the scanning electron microscope analysis, corrosion material was detected in all METAMIC coupons. Especially, it was confirmed that the coupon with white spots contains much more corrosion material than the others. Keywords: Blister, Criticality, METAMIC, Neutron Absorber, Neutron Attenuation Test, Scanning Electron Microscope

  11. Guide for subdivision of spent fuel pool. Project UNESA MAAP5-SFP

    International Nuclear Information System (INIS)

    Martinez Barrios, M.; Garcia Gonzalez, M.; Perez Martin, F. J.

    2013-01-01

    The main goal of the UNESA MAAP5-SFP project is to analyze the capabilities of MAAP5 code and, particularly, the Spent Fuel Pool (SFP) module in order to tackle its modeling and facilitate the development of specific SFP models of Spanish NPPs. Within the project, Empresarios Agrupados (EEAA) is the responsible for the development of the Guide for the subdivision of the Spent Fuel Pool (SFP). This Guide includes a theoretical description of the model that is used by the code and a sequence of practical cases with the aim to evaluate the influence of specific parameters

  12. Post-accident cooling capacity analysis of the AP1000 passive spent fuel pool cooling system

    International Nuclear Information System (INIS)

    Su Xia

    2013-01-01

    The passive design is used in AP1000 spent fuel pool cooling system. The decay heat of the spent fuel is removed by heating-boiling method, and makeup water is provided passively and continuously to ensure the safety of the spent fuel. Based on the analysis of the post-accident cooling capacity of the spent fuel cooling system, it is found that post-accident first 72-hour cooling under normal refueling condition and emergency full-core offload condition can be maintained by passive makeup from safety water source; 56 hours have to be waited under full core refueling condition to ensure the safety of the core and the spent fuel pool. Long-term cooling could be conducted through reserved safety interface. Makeup measure is available after accident and limited operation is needed. Makeup under control could maintain the spent fuel at sub-critical condition. Compared with traditional spent fuel pool cooling system design, the AP1000 design respond more effectively to LOCA accidents. (authors)

  13. Results on Technical and Consultants Service Meetings on Lessons Learned from Operating Experience in Wet and Dry Spent Fuel Storage

    International Nuclear Information System (INIS)

    White, B.; Zou, X.

    2015-01-01

    Spent fuel storage has been and will continue to be a vital portion of the nuclear fuel cycle, regardless of whether a member state has an open or closed nuclear fuel cycle. After removal from the reactor core, spent fuel cools in the spent fuel pool, prior to placement in dry storage or offsite transport for disposal or reprocessing. Additionally, the inventory of spent fuel at many reactors worldwide has or will reach the storage capacity of the spent fuel pool; some facilities are alleviating their need for additional storage capacity by utilizing dry cask storage. While there are numerous differences between wet and dry storage; when done properly both are safe and secure. The nuclear community shares lessons learned worldwide to gain knowledge from one another’s good practices as well as events. Sharing these experiences should minimize the number of incidents worldwide and increase public confidence in the nuclear industry. Over the past 60 years, there have been numerous experiences storing spent fuel, in both wet and dry mediums, that when shared effectively would improve operations and minimize events. These lessons learned will also serve to inform countries, who are new entrants into the nuclear power community, on designs and operations to avoid and include as best practices. The International Atomic Energy Agency convened a technical and several consultants’ meetings to gather these experiences and produce a technical document (TECDOC) to share spent fuel storage lessons learned among member states. This paper will discuss the status of the TECDOC and briefly discuss some lessons learned contained therein. (author)

  14. International auspices for the storage of spent nuclear fuel as a nonproliferation measure

    International Nuclear Information System (INIS)

    O'Brien, J.N.

    1981-01-01

    The maintenance of spent nuclear fuel from power reactors will pose problems regardless of how or when the debate over reprocessing is resolved. At present, many reactor sites contain significant buildups of spent fuel stored in holding pools, and no measure short of shutting down reactors with no remaining storage capacity will alleviate the need for away-from-reactor storage. Although the federal government has committed itself to dealing with the spent fuel problem, no solution has been reached, largely because of a debate over differing projections of storage capacity requirements. Proliferation of weapons grade nuclear material in many nations presents another pressing issue. If nations with small nuclear programs are forced to deal with their own spent fuel accumulations, they will either have to reprocess it indigenously or contract to have it reprocessed in a foreign reprocessing plant. In either case, these nations may eventually possess sufficient resources to assemble a nuclear weapon. The problem of spent fuel management demands real global solutions, and further delay in solving the problem of spent nuclear fuel accumulation, both nationally and globally, can benefit only a small class of elected officials in the short term and may inflict substantial costs on the American public, and possibly the world

  15. Fuel-assembly behavior under dynamic impact loads due to dry-storage cask mishandling

    International Nuclear Information System (INIS)

    1991-07-01

    Continued operation of nuclear power plants is contingent on the ability to provide adequate storage of spent fuel. Until recently, utilities have been able to maintain interim in-pool spent fuel storage. However, many facilities have reached their capacity and are now faced with shipping their spent fuel in dry casks to alternate storage facilities. The objective of this report is to provide estimates of the structural integrity of irradiated LWR fuel rods subjected to impact loads resulting from postulated cask handling accidents. This is accomplished in five stages: (1) Material properties for irradiated fuel are compiled for use in the structural analyses. (2) Results from parametric analyses of representative assembly designs are used to determine the most limiting case for end and side drop postulated handling accidents. (3) Detailed structural analysis results are presented for these critical designs. The detailed analyses include the coupling of assembly interaction with the cask and cask internals. (4) Criteria for both ultimate stress and brittle fracture failure modes of fuel rod cladding are established. (5) Safe cask handling drop height limits are computed based on items 2 through 4 above. 44 figs., 18 tabs

  16. Improved method to demonstrate the structural integrity of high density fuel storage racks

    International Nuclear Information System (INIS)

    Hinderks, M.; Ungoreit, H.; Kremer, G.

    2001-01-01

    Reracking of existing fuel pools to the maximum extent is desirable from an economical point of view. This goal can be achieved by minimizing the gaps between the spent fuel storage racks. Since the rack design is aimed at enabling consolidated fuel rod storage, additional requirements arise with respect to the design and the structural analysis. The loads resulting from seismic events are decisive for the structural analysis and require a specially detailed and in-depth analysis for high seismic loads. The verification of structural integrity and functionality is performed in two phases. In the first phase the motional behavior of single racks, rows of racks and, where required, of all racks in the pool is simulated by excitation with displacement time histories under consideration of the fluid-structure interaction (FSI). The displacements from these simulations are evaluated, while the loads are utilized as input data for the structural analysis of the racks and the pool floor. The structural analyses for the racks comprise substantially stress analyses for base material and welds as well as stability analyses for the support channels and the rack outside walls. The analyses are performed in accordance with the specified codes and standards

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

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  18. K Basins fuel encapsulation and storage hazard categorization

    International Nuclear Information System (INIS)

    Porten, D.R.

    1994-12-01

    This document establishes the initial hazard categorization for K-Basin fuel encapsulation and storage in the 100 K Area of the Hanford site. The Hazard Categorization for K-Basins addresses the potential for release of radioactive and non-radioactive hazardous material located in the K-Basins and their supporting facilities. The Hazard Categorization covers the hazards associated with normal K-Basin fuel storage and handling operations, fuel encapsulation, sludge encapsulation, and canister clean-up and disposal. The criteria categorizes a facility based on total curies per radionuclide located in the facility. Tables 5-3 and 5-4 display the results in section 5.0. In accordance with DOE-STD-1027 and the analysis provided in section 5.0, the K East Basin fuel encapsulation and storage activity and the K West Basin storage are classified as a open-quotes Category 2close quotes Facility

  19. Eddy current testing of PWR fuel pencils in the pool of the Osiris reactor

    International Nuclear Information System (INIS)

    Faure, M.; Marchand, L.

    1983-12-01

    A nondestructive testing bench is described. It is devoted to examination of high residual power fuel pencils without stress on the cladding nor interference with cooling. Guiding by fluid bearings decrease the background noise. Scanning speed is limited only by safety criteria and data acquisition configuration. Simultaneous control of various parameters is possible. Associated to an irradiation loop, loaded and unloaded in a reactor swinning pool, this bench can follow fuel pencil degradation after each irradiation cycle [fr

  20. Analysis of the LBLOCAs in the HANARO pool for the 3-pin fuel test loop

    International Nuclear Information System (INIS)

    Park, S. K.; Chi, D. Y.; Sim, B. S.; Park, K. N.; Ahn, S. H.; Lee, J. M.; Lee, C. Y.; Kim, Y. J.

    2004-12-01

    The Fuel Test Loop(FTL) has been developed to meet the increasing demand on fuel irradiation and burn up test required the development of new fuels in Korea. It is designed to provide the test conditions of high pressure and temperature like the commercial PWR and CANDU power plants. And also the FTL have the cooling capability to sufficiently remove the thermal power of the in-pile test section for normal operation, Anticipated Operational Occurrences(AOOs), and Design Basis Accidents(DBAs). This report deals with the Large Break Loss of Coolant Accidents (LBLOCAs) in HANARO pool for the 3-pin fuel test loop. The MARS code has been used for the prediction of the emergency core cooling capability of the FTL and the peak cladding temperature of the test fuels for the LBLOCAs. The location of the pipe break is assumed at the hill taps connecting the cold and hot legs in HANARO pool to the inlet and outlet nozzles of the In-Pile test Section (IPS). Double ended guillotine break is assumed for the large break loss of coolant accidents. The discharge coefficients of 0.1, 0.33, 0.67, 1.0 are investigated for the LBLOCAs. The test fuels for PWR and CANDU test modes are not heated up for the LBLOCAs caused by the double ended guillotine break in the HANARO pool. The reason is that the sufficient emergency cooling water to cool down the test fuels is supplied continuously to the in-pile test section. Therefore the PCTs for the LBLOCAs in the HANARO pool meet the design criterion of commercial PWR fuel that maximum PCT is lower than 1204 .deg. C

  1. Apparatus for the storage of transport- and storage-containers containing radioactive fuel elements

    International Nuclear Information System (INIS)

    Vox, A.

    1983-01-01

    The invention concerns an apparatus for the storage of transport and storage containers containing radioactive fuel elements. For each transport or storage container there is a separate silo-type container of steel, concrete, prestressed concrete or suchlike breakproof and fireproof material, to be placed in the open, that can be opened for removal and placing of the transport or storage container respectively. (orig.) [de

  2. Scientific basis for storage criteria for interim dry storage of aluminum-clad fuels

    International Nuclear Information System (INIS)

    Sindelar, R.L.; Peacock, H.B. Jr.; Lam, P.S.; Iyer, N.C.; Louthan, M.R. Jr.; Murphy, J.R.

    1996-01-01

    An engineered system for dry storage of aluminum-clad foreign and domestic research reactor spent fuel owned by the US Department of Energy is being considered to store the fuel up to a nominal period of 40 years prior to ultimate disposition. Scientifically-based criteria for environmental limits to drying and storing the fuels for this system are being developed to avoid excessive degradation in sealed and non-sealed (open to air) dry storage systems. These limits are based on consideration of degradation modes that can cause loss of net section of the cladding, embrittlement of the cladding, distortion of the fuel, or release of fuel and fission products from the fuel/clad system. Potential degradation mechanisms include corrosion mechanisms from exposure to air and/or sources of humidity, hydrogen blistering of the aluminum cladding, distortion of the fuel due to creep, and interdiffusion of the fuel and fission products with the cladding. The aluminum-clad research reactor fuels are predominantly highly-enriched aluminum uranium alloy fuel which is clad with aluminum alloys similar to 1100, 5052, and 6061 aluminum. In the absence of corrodant species, degradation due to creep and diffusion mechanisms limit the maximum fuel storage temperature to 200 C. The results of laboratory scale corrosion tests indicate that this fuel could be stored under air up to 200 C at low relative humidity levels (< 20%) to limit corrosion of the cladding and fuel (exposed to the storage environment through assumed pre-existing pits in the cladding). Excessive degradation of fuels with uranium metal up to 200 C can be avoided if the fuel is sufficiently dried and contained in a sealed system; open storage can be achieved if the temperature is controlled to avoid excessive corrosion even in dry air

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-10-01

    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.

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

    International Nuclear Information System (INIS)

    2003-01-01

    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

  6. Development of INSPCT-S for inspection of spent fuel pool

    Energy Technology Data Exchange (ETDEWEB)

    Walters, W.; Haghighat, A. [Nuclear Engineering Program, Mechanical Engineering Dept., Virginia Tech., Blacksburg, VA 24061 (United States); Sitaraman, S.; Ham, Y. [Lawrence Livermore National Laboratory, 7000 East Ave., Livermore, CA 94550 (United States)

    2011-07-01

    In this paper, we discuss an accurate and fast software tool (INSPCT-S, Inspection of Nuclear Spent fuel-Pool Calculation Tool, version Spreadsheet) developed for calculation of the response of fission chambers placed in a spent fuel pool, such as Atucha-I. INSPCT-S is developed for identification of suspicious regions of the pool that may have missing or substitute assemblies. INSPCT-S uses a hybrid algorithm based on the adjoint function methodology. The neutron source is comprised of spontaneous fission, ({alpha}, n) interactions, and subcritical multiplication. The former is evaluated using the ORIGEN-ARP code, and the latter is obtained with the fission matrix (FM) formulation. The FM coefficients are determined using the MCNP Monte Carlo code, and the importance function is determined using the PENTRAN 3-D parallel Sn code. Three databases for the neutron source, FM elements, and adjoint flux are prepared as functions of different parameters including burnup, cooling time, enrichment, and pool lattice size. INSPCT-S uses the aforementioned databases and systems of equations to calculate detector responses, which are subsequently compared with normalized experimental data. If this comparison is not satisfied, INSPCT-S utilizes color coding to identify the suspicious regions of a spent fuel pool. (authors)

  7. Degradation of EBR-II driver fuel during wet storage

    International Nuclear Information System (INIS)

    Pahl, R. G.

    2000-01-01

    Characterization data are reported for sodium bonded EBR-II reactor fuel which had been stored underwater in containers since the 1981--1982 timeframe. Ten stainless steel storage containers, which had leaked water during storage due to improper sealing, were retrieved from the ICPP-603 storage basin at the Idaho National Engineering and Environmental Laboratory (INEEL) in Idaho. In the container chosen for detailed destructive analysis, the stainless steel cladding on the uranium alloy fuel had ruptured and fuel oxide sludge filled the bottom of the container. Headspace gas sampling determined that greater than 99% hydrogen was present. Cesium 137, which had leached out of the fuel during the aqueous corrosion process, dominated the radionuclide source term of the water. The metallic sodium from the fuel element bond had reacted with the water, forming a concentrated caustic solution of NaOH

  8. Behaviour of spent fuel assemblies during extended storage

    International Nuclear Information System (INIS)

    1987-04-01

    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

  9. Burnup credit calculations for criticality safety justification for RBMK-1000 spent fuel of transport and storage systems

    Directory of Open Access Journals (Sweden)

    V. V. Galchenko

    2010-12-01

    Full Text Available In present paper the burnup credit calculations for TK-8 transport container and SVJP-1 spent fuel storage fa-cility of pool type with RBMK-1000 spent fuel during 100-years of cooling time were performed for criticality safety analysis purpose using MCNP and SCALE codes. Only actinides were taken into account for these critical systems. Two approaches were analyzed with isotopes distribution calculations along fuel assembly height and without it. The results show that subcriticality margin is increased considerably using burnup credit and isotopes distribution along fuel assembly height made this value more reasonable.

  10. Spent fuel storage process equipment development

    International Nuclear Information System (INIS)

    Park, Hyun Soo; Lee, Jae Sol; Yoo, Jae Hyung

    1990-02-01

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

  11. Validation of the EIR LWR calculation methods for criticality assessment of storage pools

    International Nuclear Information System (INIS)

    Grimm, P.; Paratte, J.M.

    1986-11-01

    The EIR code system for the calculation of light water reactors is presented and the methods used are briefly described. The application of the system to various types of critical experiments and benchmark problems proves its good accuracy, even for heterogeneous configurations containing strong neutron absorbers such as Boral. Since the multiplication factor k eff is normally somewhat overpredicted and the spread of the results is small, this code system is validated for the calculation of storage pools, taking into account a safety margins of 1.5% on k eff . (author)

  12. Fission product release in conditions of a spent fuel pool severe accident

    International Nuclear Information System (INIS)

    Ohai, Dumitru

    2007-01-01

    Full text: Depending on the residence time, fuel burnup, and fuel rack configuration, there may be sufficient decay heat for the fuel clad to heat up, swell, and burst in case of a loss of pool water. Initiating event categories can be: loss of offsite power from events initiated by severe weather, internal fire, loss of pool cooling, loss of coolant inventory, seismic event, aircraft impact, tornado, missile attack. The breach in the clad releases the radioactive gases present in the gap between the fuel and clad, what is called 'gap release'. If the fuel continues to heat up, the zirconium clad will reach the point of rapid oxidation in air. This reaction of zirconium and air, or zirconium and steam is exothermic. The energy released from the reaction, combined with the fuel's decay energy, can cause the reaction to become self-sustaining and ignite the zirconium. The increase in heat from the oxidation reaction can also raise the temperature in adjacent fuel assemblies and propagate the oxidation reaction. Simultaneously, the sintered UO 2 pellets resulting from pins destroying are oxidized. Due to the self-disintegration of pellets by oxidation, fission gases and low volatile fission products are released. The release rate, the chemical nature and the amount of fission products depend on powder granulation distribution and environmental conditions. The zirconium burning and pellets self-disintegration will result in a significant release of spent fuel fission products that will be dispersed from the reactor site. (author)

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

    International Nuclear Information System (INIS)

    Bennett, Peter; Larsen, Erlend

    2014-01-01

    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 (UO 2 ). 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

  14. Assessment of S(α, β) libraries for criticality safety evaluations of wet storage pools by refined trend analyses

    International Nuclear Information System (INIS)

    Kolbe, E.; Vasiliev, A.; Ferroukhi, H.

    2009-01-01

    In a recent criticality safety evaluation (CSE) of a commercial wet storage pool applying MCNPX-2.5.0 in combination with the ENDF/B-VII.0 and JEFF-3.1 continuous energy cross section libraries, the maximum permissible initial fuel-enrichment limit for water reflected configurations was found to be dependant upon the applied neutron cross section library. More detailed investigations indicated that the difference is mainly caused by different sub-libraries for thermal neutron scattering based on parameterizations of the S(α, β) scattering matrix. Hence an analysis of trends was done with respect to the low energy neutron flux in order to assess the S(α, β) data sets. First, when performing the trend analysis based on the full set of 149 benchmarks that were employed for the validation, significant trends could not be found. But by analyzing a selected subset of benchmarks clear trends with respect to the low energy neutron flux could be detected. The results presented in this paper demonstrate the sensitivity of specific configurations to the parameterizations of the S(α, β) scattering matrix and thus may help to improve CSE of wet storage pools. Finally, in addition to the low energy neutron flux, we also refined the trend analyses with respect to other key (spectrum-related) parameters by performing them with various selected subsets of the full suite of 149 benchmarks. The corresponding outcome using MCNPX 2.5.0 in combination with the ENDF/B-VII.0, ENDF/B-VI.8, JEFF-3.1, JEF-2.2, and JENDL-3.3 neutron cross section libraries are presented and discussed. (authors)

  15. A software tool integrated risk assessment of spent fuel transpotation and storage

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Mi Rae; Almomani, Belal; Ham, Jae Hyun; Kang, Hyun Gook [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Christian, Robby [Dept. of Mechanical, Aerospace, and Nuclear Engineering, Rensselaer Polytechnic Institute, Troy (Korea, Republic of); Kim, Bo Gyung [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Lee, Sang Hoon [Dept. of Mechanical and Automotive Engineering, Keimyung University, Daegu (Korea, Republic of)

    2017-06-15

    When temporary spent fuel storage pools at nuclear power plants reach their capacity limit, the spent fuel must be moved to an alternative storage facility. However, radioactive materials must be handled and stored carefully to avoid severe consequences to the environment. In this study, the risks of three potential accident scenarios (i.e., maritime transportation, an aircraft crashing into an interim storage facility, and on-site transportation) associated with the spent fuel transportation process were analyzed using a probabilistic approach. For each scenario, the probabilities and the consequences were calculated separately to assess the risks: the probabilities were calculated using existing data and statistical models, and the consequences were calculated using computation models. Risk assessment software was developed to conveniently integrate the three scenarios. The risks were analyzed using the developed software according to the shipment route, building characteristics, and spent fuel handling environment. As a result of the risk analysis with varying accident conditions, transportation and storage strategies with relatively low risk were developed for regulators and licensees. The focus of this study was the risk assessment methodology; however, the applied model and input data have some uncertainties. Further research to reduce these uncertainties will improve the accuracy of this mode.

  16. A software tool integrated risk assessment of spent fuel transpotation and storage

    International Nuclear Information System (INIS)

    Yun, Mi Rae; Almomani, Belal; Ham, Jae Hyun; Kang, Hyun Gook; Christian, Robby; Kim, Bo Gyung; Lee, Sang Hoon

    2017-01-01

    When temporary spent fuel storage pools at nuclear power plants reach their capacity limit, the spent fuel must be moved to an alternative storage facility. However, radioactive materials must be handled and stored carefully to avoid severe consequences to the environment. In this study, the risks of three potential accident scenarios (i.e., maritime transportation, an aircraft crashing into an interim storage facility, and on-site transportation) associated with the spent fuel transportation process were analyzed using a probabilistic approach. For each scenario, the probabilities and the consequences were calculated separately to assess the risks: the probabilities were calculated using existing data and statistical models, and the consequences were calculated using computation models. Risk assessment software was developed to conveniently integrate the three scenarios. The risks were analyzed using the developed software according to the shipment route, building characteristics, and spent fuel handling environment. As a result of the risk analysis with varying accident conditions, transportation and storage strategies with relatively low risk were developed for regulators and licensees. The focus of this study was the risk assessment methodology; however, the applied model and input data have some uncertainties. Further research to reduce these uncertainties will improve the accuracy of this mode

  17. Proposal of a dry storage installation in Angra NPP for spent nuclear fuel

    International Nuclear Information System (INIS)

    Romanato, Luiz S.; Rzyski, Barbara M.

    2009-01-01

    When nuclear fuel is removed from a power reactor core after the depletion of efficiency in generating energy is called Spent Nuclear Fuel (SNF). After its withdrawal from the reactor core, SNF is temporarily stored in pools usually at the same site of the reactor. During this time, short-living radioactive elements and generated heat undergo decay until levels that allow removing the SNF from the pool and sending it for reprocessing or a temporary storage whether any of its final destinations has not yet been defined. It can be loaded in casks and disposed during years in a dry storage installations until be sent to a reprocessing plant or deep repositories. Before any decision, reprocessing or disposal, the SNF needs to be safely and efficiently isolated in one of many types of storages that exist around the world. Worldwide, the amount of SNF increases annually and in the next years this amount will be higher as a consequence of new Nuclear Power Plants (NPP) construction. In Brazil, that is about to construct the Angra 3 nuclear power reactor, a project about the final destination of the SNF is not yet ready. This paper presents a proposal for a dry storage installation in the Angra NPP site since it can be an initial solution for the Brazilian's SNF, until a final decision is taken. (author)

  18. Dry Storage of Research Reactor Spent Nuclear Fuel - 13321

    Energy Technology Data Exchange (ETDEWEB)

    Adams, T.M.; Dunsmuir, M.D.; Leduc, D.R.; Severynse, T.F.; Sindelar, R.L. [Savannah River National Laboratory (United States); Moore, E.N. [Moore Nuclear Energy, LLC (United States)

    2013-07-01

    Spent fuel from domestic and foreign research reactors is received and stored at the Savannah River Site's L Area Material Storage (L Basin) Facility. This DOE-owned fuel consists primarily of highly enriched uranium in metal, oxide or silicide form with aluminum cladding. Upon receipt, the fuel is unloaded and transferred to basin storage awaiting final disposition. Disposition alternatives include processing via the site's H Canyon facility for uranium recovery, or packaging and shipment of the spent fuel to a waste repository. A program has been developed to provide a phased approach for dry storage of the L Basin fuel. The initial phase of the dry storage program will demonstrate loading, drying, and storage of fuel in twelve instrumented canisters to assess fuel performance. After closure, the loaded canisters are transferred to pad-mounted concrete overpacks, similar to those used for dry storage of commercial fuel. Unlike commercial spent fuel, however, the DOE fuel has high enrichment, very low to high burnup, and low decay heat. The aluminum cladding presents unique challenges due to the presence of an oxide layer that forms on the cladding surface, and corrosion degradation resulting from prolonged wet storage. The removal of free and bound water is essential to the prevention of fuel corrosion and radiolytic generation of hydrogen. The demonstration will validate models predicting pressure, temperature, gas generation, and corrosion performance, provide an engineering scale demonstration of fuel handling, drying, leak testing, and canister backfill operations, and establish 'road-ready' storage of fuel that is suitable for offsite repository shipment or retrievable for onsite processing. Implementation of the Phase I demonstration can be completed within three years. Phases II and III, leading to the de-inventory of L Basin, would require an additional 750 canisters and 6-12 years to complete. Transfer of the fuel from basin storage

  19. Integrity of spent CANDU fuel during and following dry storage

    International Nuclear Information System (INIS)

    Villagran, J.E.

    2004-01-01

    This report examines the issue of CANDU fuel integrity at the back end of the fuel cycle and outlines a program designed to provide assurance that used CANDU fuel will retain its integrity over an extended period. In specific terms, the program is intended to provide assurance that during and following extended dry storage the fuel will remain fit to undergo, without loss of integrity, the handling, packaging and transportation operations that might be necessary until it is placed in disposal containers. The first step in the development of the program was a review of the available technical information on phenomena relevant to fuel integrity. The major conclusions from that review were the following: Under normal storage conditions it is unlikely that the spent fuel will suffer significant degradation during a one-hundred year period and it should be possible to retrieve, repackage and transport the fuel as required, using methods and systems similar to those used today. However, to provide increased confidence regarding the above conclusion, investigations should be conducted in areas where there is higher uncertainty in the prediction of fuel condition and on some degradation processes to which the fuel appears to present higher vulnerability. The proposed program includes, among other tasks, irradiated fuel tests, analytical studies on the most relevant fuel degradation processes and the development of a long-term fuel verification program. (Author)

  20. Spent Fuel Transfer to Dry Storage Using Unattended Monitoring System

    International Nuclear Information System (INIS)

    Park, Jae Hwan; Park, Soo Jin

    2009-01-01

    There are 4 CANDU reactors at Wolsung site together with a spent fuel dry storage associated with unit 1. These CANDU reactors, classified as On-Load Reactor (OLR) for Safeguards application, change 16- 24 fuel bundles with fresh fuel in everyday. Especially, the spent fuel bundles are transferred from spent fuel bays to dry storage throughout a year because of the insufficient capacity of spent fuel pond. Safeguards inspectors verify the spent fuel transfer to meet safeguards purposes according to the safeguards criteria by means of inspector's presence during the transfer campaign. For the verification, 60-80 person-days of inspection (PDIs) are needed during approximately 3 months for each unit. In order to reduce the inspection effort and operators' burden, an Unattended Monitoring System (UMS) was designed and developed by the IAEA for the verification of spent fuel bundles transfers from wet storage to dry storage. Based on the enhanced cooperation of CANDU reactors between the ROK and the IAEA, the IAEA installed the UMS at Wolsung unit 2 in January 2005 at first. After some field trials during the transfer campaign, this system is being replaced the traditional human inspection since September 1, 2006 combined with a Short Notice Inspection (SNI) and a near-real time Mailbox Declaration

  1. Assembly for transport and storage of radioactive nuclear fuel elements

    International Nuclear Information System (INIS)

    Myers, G.

    1978-01-01

    The invention concerns the self-control of coolant deficiencies on the transport of spent fuel elements from nuclear reactors. It guarantees that drying out of the fuel elements is prevented in case of a change of volume of the fluid contained in storage tanks and accumulators and serving as coolant and shielding medium. (TK) [de

  2. Experimental study on heat pipe heat removal capacity for passive cooling of spent fuel pool

    International Nuclear Information System (INIS)

    Xiong, Zhenqin; Wang, Minglu; Gu, Hanyang; Ye, Cheng

    2015-01-01

    Highlights: • A passively cooling SFP heat pipe with an 8.2 m high evaporator was tested. • Heat removed by the heat pipe is in the range of 3.1–16.8 kW. • The heat transfer coefficient of the evaporator is 214–414 W/m 2 /K. • The heat pipe performance is sensitive to the hot water temperature. - Abstract: A loop-type heat pipe system uses natural flow with no electrically driven components. Therefore, such a system was proposed to passively cool spent fuel pools during accidents to improve nuclear power station safety especially for station blackouts such as those in Fukushima. The heat pipe used for a spent fuel pool is large due to the spent fuel pool size. An experimental heat pipe test loop was developed to estimate its heat removal capacity from the spent fuel pool during an accident. The 7.6 m high evaporator is heated by hot water flowing vertically down in an assistant tube with a 207-mm inner diameter. R134a was used as the potential heat pipe working fluid. The liquid R134a level was 3.6 m. The tests were performed for water velocities from 0.7 to 2.1 × 10 −2 m/s with water temperatures from 50 to 90 °C and air velocities from 0.5 m/s to 2.5 m/s. The results indicate significant heat is removed by the heat pipe under conditions that may occur in the spent fuel pool

  3. International management and storage of plutonium and spent fuel

    International Nuclear Information System (INIS)

    1978-09-01

    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

  4. Loss of cooling accident simulation of nuclear power station spent-fuel pool

    Energy Technology Data Exchange (ETDEWEB)

    Lee, M.; Liang, K-S., E-mail: mlee@ess.nthu.edu.tw, E-mail: ksliang_1@hotmail.com [National Tsing Hua Univ., Hsinchu, Taiwan (China); Lin, K-Y., E-mail: syrup760914@gmail.com [Taiwan Power Company, Taiwan (China)

    2014-07-01

    The core melt down accident of Fukushima Nuclear Power Station on March 11th, 2011 alerted nuclear industry that the long term loss of cooling of spent fuel pool may need some attention. The target plant analyzed is the Chinshan Nuclear Power Station of Taiwan Power Company. The 3-Dimensional RELAP5 input deck of the spent fuel pool of the station is built. The results indicate that spent fuel of Chinshan Nuclear Power Station is uncovered at 6.75 days after an accident of loss cooling takes place and cladding temperature rises above 2,200{sup o}F around 8 days. The time is about 13 hours earlier than the results predicted using simple energy balance method. The results also show that the impact of Counter Current Flow Limitation (CCFL) and radiation heat transfer model is marginal. (author)

  5. Assessment of fuel damage of pool type research reactor in the case of fuel plates blockage

    Energy Technology Data Exchange (ETDEWEB)

    Jalil, Jafari; Samad, Khakshournia [AEOI, Karegar Ave. School of R and D of Nuclear Reactors and Accelerators, Teheran (Iran, Islamic Republic of); D' Auria, F. [Pisa Univ., DIMNP (Italy)

    2007-07-01

    Tehran Research Reactor (TRR) is a pool type 5 MW research reactor. It is assumed that external objects or debris that may fall down to reactor core cause obstruction of coolant flow through one of the fuel assemblies. Thermal hydraulic analysis of this event, using the RELAP5 system code has been studied. The reported transient is related to the partial and total obstruction of a single Fuel Element (FE) cooling channel of 27 FE equilibrium core of TRR. Such event constitutes a severe accident for this type of reactor since it may lead to local dryout and eventually to loss of the FE integrity. Two scenarios are analysed to emphasize the severity of the accident. The first one is a partial blockage of an average FE considering four different obstruction levels: 25%, 50%, 75% and 97% of nominal flow area. The second one is an extreme scenario consisting of total blockage of the same FE. This study constitutes the first step of a larger work which consists of performing a 3-dimensional simulation using the Best Estimate coupled code technique. However, as a first approach the instantaneous reactor power is derived through the point kinetic calculation included in the RELAP5 code. Main results obtained from the RELAP5 calculations are as following. First, in the case of flow blockage under 97% of the nominal flow area of an average FE, only an increase of the coolant and clad temperatures is observed without any consequences for the integrity of the FE. The mass flow rate remains sufficient to cool the clad safely. Secondly, in the case of total obstruction of the nominal flow area, it is seen that transient turns out to be a severe accident due to the dryout conditions are reached shortly and melting of the cladding occurs. Thirdly, the use of the point kinetic approach leads to conservative results. A best estimate simulation of such kind of transients requires the use of 3-dimensional kinetic calculations, which could be done using the current Coupled Codes

  6. Analysis of the SBLOCAs in HANARO pool for the 3-pin fuel test loop

    International Nuclear Information System (INIS)

    Park, S. K.; Chi, D. Y.; Sim, B. S.; Park, K. N.; Ahn, S. H.; Lee, J. M.; Lee, C. Y.; Kim, Y. J.

    2004-09-01

    Fuel Test Loop(FTL) has been developed to meet the increasing demand on fuel irradiation and burn up test required the development of new fuels in Korea. It is designed to provide the test conditions of high pressure and temperature like the commercial PWR and CANDU power plants. And also the FTL have the cooling capability to sufficiently remove the thermal power of the in-pile test section for normal operation, Anticipated Operational Occurrences(AOOs), and Design Basis Accidents(DBAs). This report deals with the Small Break Loss Of Coolant Accidents (SBLOCAs) in HANARO pool for the 3-pin fuel test loop. The MARS code has been used for the prediction of the emergency core cooling capability of the FTL and the peak cladding temperature of the test fuels for the SBLOCAs. The location of the pipe break is assumed at the hill taps connecting the cold and hot legs in HANARO pool to the inlet and outlet nozzles of the In-Pile test Section (IPS). The break size is also assumed less than 20% of the cross section area of the pipe. The test fuels are heated up when the cold leg break occur. However, they are not heated up when the hot leg break occur. The maximum Peak Cladding Temperatures (PCT) are predicted to be about 906.9 .deg. C for the cold leg break accident in PWR fuel test mode and 971.9 .deg. C in CANDU fuel test mode respectively. The critical break size is about the 6% of the cross section area of the pipe for PWR fuel test mode and the 8% for CANDU fuel test mode. The PCTs meet the design criterion of commercial PWR fuel that the maximum PCT is lower than 1204 .deg. C

  7. Behavior of spent fuel and cask components after extended periods of dry storage

    International Nuclear Information System (INIS)

    Kenneally, R.; Kessler, J.

    2001-01-01

    The U.S. Nuclear Regulatory Commission (NRC) promulgated 10 CFR Part 72, Title 10, for the independent storage of spent nuclear fuel and high-level radioactive waste outside reactor spent fuel pools. Part 72 currently limits the license term for an independent spent fuel storage installation to 20 years from the date of issuance. Licenses may be renewed by the Commission at or before the expiration of the license term. Applications for renewal of a license should be filed at least two years prior to the expiration of the existing license. In preparation for possible license renewal, the NRC Office of Nuclear Material and Safeguards, Spent Fuel Project Office, is developing the technical basis for renewals of licenses and Certificates of Compliance for dry storage systems for spent nuclear fuel and high-level radioactive waste at independent spent fuel storage installation sites. An analysis of past performance of selected components of these systems is required as part of that technical basis. In the years 1980 through the early 1990, the Department of Energy (DOE) procured four prototype dry storage casks for testing at the Idaho National Engineering and Environmental Laboratory (INEEL): Castor-V/21, MC-10, TN-24P, and VSC-17. The primary purpose of the testing was to benchmark thermal and radiological codes and to determine the thermal and radiological characteristics of the casks. A series of examinations in 1999 and early 2000 to investigate the integrity of the Castor V/21 cask were undertaken. There is no evidence of significant degradation of the Castor V/21 cask systems important to safety from the time of initial loading of the cask in 1985 up to the time of testing in 1999. (author)

  8. Safety issues of dry fuel storage at RSWF

    International Nuclear Information System (INIS)

    Clarksean, R.L.; Zahn, T.P.

    1995-01-01

    Safety issues associated with the dry storage of EBR-II spent fuel are presented and discussed. The containers for the fuel have been designed to prevent a leak of fission gases to the environment. The storage system has four barriers for the fission gases. These barriers are the fuel cladding, an inner container, an outer container, and the liner at the RSWF. Analysis has shown that the probability of a leak to the environment is much less than 10 -6 per year, indicating that such an event is not considered credible. A drop accident, excessive thermal loads, criticality, and possible failure modes of the containers are also addressed

  9. Capabilities for processing shipping casks at spent fuel storage facilities

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  10. Generic environmental impact statement on handling and storage of spent light water power reactor fuel. Appendices

    International Nuclear Information System (INIS)

    1978-03-01

    Detailed appendices are included with the following titles: light water reactor fuel cycle, present practice, model 1000MW(e) coal-fired power plant, increasing fuel storage capacity, spent fuel transshipment, spent fuel generation and storage data (1976-2000), characteristics of nuclear fuel, and ''away-from-reactor'' storage concept

  11. Security measurements and radiological protection in the source panoramic irradiators and storage in pool

    International Nuclear Information System (INIS)

    Del Valle O, C.

    1996-01-01

    The aim of this paper is to investigate and to study the safety and protecting measurements that must be taken into account in the design and the use of panoramic source irradiators with wet storage or pool, concerning to category IV. The generic characteristics in plants of kind, as well as their description, are mentioned in this paper. The devices, that comply the security and control systems based on their redundancy, diversity and independence, are examined. Likewise, it describes the design requirements of the overcast, of the irradiators, of the source frame, of the transporting system of product, of the procedure access, of the security system of the irradiator shelf control, of the irradiation room, of the irradiation storage pool, of the ventilation system, for the protection in case of fire of fire, for electric energy failures, for the warning symbols and signs. It contains scope about the organization and responsibilities that must be taken into account in plants of this type. A detailed plan has been made for its operation and maintenance, enclosing instructions and registers for this reason. The statement of emergency events and their respective answers, the analysis of cases and reasons that causes accidents and its implementation and regular inspection procedures for the improvement of the plant are also studied. (author). 2 refs

  12. Advanced surveillance technologies for used fuel long-term storage and transportation - 59032

    International Nuclear Information System (INIS)

    Tsai, Hanchung; Liu, Yung Y.; Nutt, Mark; Shuler, James

    2012-01-01

    Utilities worldwide are using dry-cask storage systems to handle the ever-increasing number of discharged fuel assemblies from nuclear power plants. In the United States and possibly elsewhere, this trend will continue until an acceptable disposal path is established. The recent Fukushima nuclear power plant accident, specifically the events with the storage pools, may accelerate the drive to relocate more of the used fuel assemblies from pools into dry casks. Many of the newer cask systems incorporate dual-purpose (storage and transport) or multiple-purpose (storage, transport, and disposal) canister technologies. With the prospect looming for very long term storage - possibly over multiple decades - and deferred transport, condition- and performance-based aging management of cask structures and components is now a necessity that requires immediate attention. From the standpoint of consequences, one of the greatest concerns is the rupture of a substantial number of fuel rods that would affect fuel retrievability. Used fuel cladding may become susceptible to rupture due to radial-hydride-induced embrittlement caused by water-side corrosion during the reactor operation and subsequent drying/transfer process, through early stage of storage in a dry cask, especially for high burnup fuels. Radio frequency identification (RFID) is an automated data capture and remote-sensing technology ideally suited for monitoring sensitive assets on a long-term, continuous basis. One such system, called ARG-US, has been developed by Argonne National Laboratory for the U.S. Department of Energy's Packaging Certification Program for tracking and monitoring drums containing sensitive nuclear and radioactive materials. The ARG-US RFID system is versatile and can be readily adapted for dry-cask monitoring applications. The current built-in sensor suite consists of seal, temperature, humidity, shock, and radiation sensors. With the universal asynchronous receiver/transmitter interface in

  13. Storage, transportation and disposal system for used nuclear fuel assemblies

    Science.gov (United States)

    Scaglione, John M.; Wagner, John C.

    2017-01-10

    An integrated storage, transportation and disposal system for used fuel assemblies is provided. The system includes a plurality of sealed canisters and a cask sized to receive the sealed canisters in side by side relationship. The plurality of sealed canisters include an internal basket structure to receive a plurality of used fuel assemblies. The internal basket structure includes a plurality of radiation-absorbing panels and a plurality of hemispherical ribs generally perpendicular to the canister sidewall. The sealed canisters are received within the cask for storage and transportation and are removed from the cask for disposal at a designated repository. The system of the present invention allows the handling of sealed canisters separately or collectively, while allowing storage and transportation of high burnup fuel and damaged fuel to the designated repository.

  14. Spent LWR fuel encapsulation and dry storage demonstration

    International Nuclear Information System (INIS)

    Bahorich, R.J.; Durrill, D.C.; Cross, T.E.; Unterzuber, R.

    1980-01-01

    In 1977 the Spent Fuel Handling and Packaging Program (SFHPP) was initiated by the Department of Energy to develop and test the capability to satisfactorily encapsulate typical spent fuel assemblies from commercial light-water nuclear power plants and to establish the suitability of one or more surface and near surface concepts for the interim dry storage of the encapsulated spent fuel assemblies. The E-MAD Facility at the Nevada Test Site, which is operated for the Department of Energy by the Advanced Energy Systems Division (AESD) of the Westinghouse Electric Corporation, was chosen as the location for this demonstration because of its extensive existing capabilities for handling highly radioactive components and because of the desirable site characteristics for the proposed storage concepts. This paper describes the remote operations related to the process steps of handling, encapsulating and subsequent dry storage of spent fuel in support of the Demonstration Program

  15. Modular vault dry storage system for interim storage of irradiated fuel

    International Nuclear Information System (INIS)

    Cundill, B.R.; Ealing, C.J.; Agarwal, B.K.

    1988-01-01

    The Foster Wheeler Energy Application (FWEA) Modular Vault Dry Store (MVDS) is a dry storage concept for the storage of all types of irradiated reactor fuel. For applications in the US, FWEA submitted an MVDS Topical Report to the US NRC during 1986. Following NRC approval of the MVDS Topical Report concept for unconsolidated LWR fuel, US utilities have available a new, compact, economic and flexible system for the storage of irradiated fuel at the reactor site for time periods of at least 20 years (the period of the first license). The MVDS concept jointly developed by FWEA and GEC in the U.K., has other applications for large central away from reactor storage facilities such as a Monitorable Retrievable Storage (MRS) installation. This paper describes the licensed MVDS design, aspects of performance are discussed and capital costs compared with alternative concepts. Alternative configurations of MVDS are outlined

  16. Liquid waste processing from TRIGA spent fuel storage pits

    International Nuclear Information System (INIS)

    Buchtela, Karl

    1988-01-01

    At the Atominstitute of the Austrian Universities and also at other facilities running TRIGA reactors, storage pits for spent fuel elements are installed. During the last revision procedure, the reactor group of the Atominstitute decided to refill the storage pits and to get rid of any contaminated storage pit water. The liquid radioactive waste had been pumped to polyethylene vessels for intermediate storage before decontamination and release. The activity concentration of the storage pit water at the Aominstitute after a storage period of several years was about 40 kBq/l, the total amount of liquid in the storage pits was about 0.25 m 3 . It was attempted to find a simple and inexpensive method to remove especially the radioactive Cesium from the waste solution. Different methods for decontamination like distillation, precipitation and ion exchange are discussed

  17. Spent fuel storage and transportation - ANSTO experience

    International Nuclear Information System (INIS)

    Irwin, Tony

    2002-01-01

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

  18. Study of a brazilian cask and its installation for PWR spent nuclear fuel dry storage

    International Nuclear Information System (INIS)

    Romanato, Luiz Sergio

    2009-01-01

    Spent nuclear fuel (SNF) is removed from the nuclear reactor after the depletion on efficiency in generating energy. After the withdrawal from the reactor core, the SNF is temporarily stored in pools at the same site of the reactor. At this time, the generated heat and the short and medium lived radioactive elements decay to levels that allow removing SNF from the pool and sending it to temporary dry storage. In that phase, the fuel needs to be safely and efficiently stored, and then, it can be retrieved in a future, or can be disposed as radioactive waste. The amount of spent fuel increases annually and, in the next years, will still increase more, because of the construction of new nuclear plants. Today, the number of new facilities back up to levels of the 1970's, since it is greater than the amount of decommissioning in old installations. As no final decision on the back-end of the nuclear fuel cycle is foreseen in the near future in Brazil, either to recover the SNF or to consider it as radioactive waste, this material has to be isolated in some type of storage model existing around the world. In the present study it is shown that dry SNF storage is the best option. A national cask model for SNF as well these casks storage installation are proposed. It is a multidisciplinary study in which the engineering conceptual task was developed and may be applied to national SNF removed from the Brazilian power reactors, to be safely stored for a long time until the Brazilian authorities will decide about the site for final disposal. (author)

  19. Neutron physical aspects of the storage of BWR fuel elements

    International Nuclear Information System (INIS)

    Woloch, F.; Sdouz, G.; Suda, M.

    1980-01-01

    For the storage of BWR fuel elements in a high density fuel rack using boronated steel absorbers and in a fuel rack with a larger pitch without absorber, criticality calculations are performed. The cooling water density is varied for the storage without absorbers. For the selected pitches of 16.5 cm for the high density fuel rack and 25 cm for the fuel rack without absorber respectively the ksub(infinitely) values of 0.933 and 0.748 are obtained. The dependence of the results on different calculational methods and on the influence of the variation of three important design parameters, i.e. of the concentration of boron, of the thickness of the boronated steel and of the watergap is investigated for the high density fuel rack. The average isothermal temperature coefficient is obtained for the high density fuel rack as -4.5 x 10 -40 sup(0)C -1 and as approx. 2.0 x 10 -40 sup(0)C -1 for the fuel rack without absorbers. For both ways of storage the aspects of safety of the results are discussed thoroughly. (orig.) 891 RW/orig. 892 CKA [de

  20. Storage pool water condition of multipurpose compact 60Co irradiator of IPEN-CNEN/SP

    International Nuclear Information System (INIS)

    Kodama, Yasko; Rela, Paulo R.

    2008-01-01

    A new type compact gamma irradiator was developed and installed at IPEN-CNEN/SP, it is classified as category IV irradiator according to International Atomic Energy Agency-IAEA. The originality of its design remains on the rotating concrete door, which integrates the shielding system with the product handling system, permitting the input and output of the products in a continuous way, without the necessity to lower the sources and open the irradiator chamber to change the batch. The licensed capacity of this irradiator is 37 P Bq (1 MCi) and the water pool liner, manufactured in stainless steel, has 7.0 m deep and 2.7 m diameter. A sealed radiation source may be defined as a quantity of radioactive material sealed within non-radioactive material, which the confining barrier is intended to prevent leakage or escape of the radioactive material under normal handling conditions and also under foreseeable mishaps. The use of cobalt-60 radiation sources in a wet storage environment requires care in the control of that environment to avoid the potential for deterioration of the stainless steel. In the literature, the tests performed demonstrated that stainless steel had excellent performance under controlled environment conditions. The inspection includes assessing the condition of storage pool water and measuring the pool water conductivity. The demineralized water normally used in storage pools is usually an effective and secure medium in which to store stainless steel. However, there are certain constituents and contaminants in the water which can affect this behaviour. For instance, the pH has a small but definite effect on the tendency for pitting corrosion. Nevertheless, there is also evidence that both strongly alkaline and strongly acid environments can encourage pitting corrosion. In general, an around neutral pH is recommended. The conductivity is used as a measure of the extent to which ionic species are present in the pool water. International recommendations

  1. Advantages on dry interim storage for spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Romanato, L.S. [Centro Tecnologico da Marinha em Sao Paulo, Av. Professor Lineu Prestes 2468, 05508-900 Sao Paulo (Brazil); Rzyski, B.M. [IPEN/ CNEN-SP, 05508-000 Sao Paulo (Brazil)]. e-mail: romanato@ctmsp.mar.mil.br

    2006-07-01

    When the nuclear fuel lose its ability to efficiently create energy it is removed from the core reactor and moved to a storage unit waiting for a final destination. Generally, the spent nuclear fuel (SNF) remains inside concrete basins with water within the reactors facility for the radioactive activity decay. Water cools the generated heat and shields radioactivity emissions. After some period of time in water basins the SNF can be sent to a definitive deposition in a geological repository and handled as radioactive waste or to reprocessing installations, or still wait for a future solution. Meanwhile, SNF remains stored for a period of time in dry or wet installations, depending on the method adopted by the nuclear power plant or other plans of the country. In many SNF wet storage sites the capacity can be fulfilled very quickly. If so, additional area or other alternative storage system should be given. There are many options to provide capacity increase in the wet storage area, but dry storages are worldwide preferred since it reduces corrosion concerns. In the wet storage the temperature and water purity should be constantly controlled whereas in the dry storage the SNF stands protected in specially designed canisters. Dry interim storages are practical and approved in many countries especially that have the 'wait and see' philosophy (wait to see new technologies development). This paper shows the advantages of dry interim storages sites in comparison with the wet ones and the nowadays problems as terrorism. (Author)

  2. Advantages on dry interim storage for spent nuclear fuel

    International Nuclear Information System (INIS)

    Romanato, L.S.; Rzyski, B.M.

    2006-01-01

    When the nuclear fuel lose its ability to efficiently create energy it is removed from the core reactor and moved to a storage unit waiting for a final destination. Generally, the spent nuclear fuel (SNF) remains inside concrete basins with water within the reactors facility for the radioactive activity decay. Water cools the generated heat and shields radioactivity emissions. After some period of time in water basins the SNF can be sent to a definitive deposition in a geological repository and handled as radioactive waste or to reprocessing installations, or still wait for a future solution. Meanwhile, SNF remains stored for a period of time in dry or wet installations, depending on the method adopted by the nuclear power plant or other plans of the country. In many SNF wet storage sites the capacity can be fulfilled very quickly. If so, additional area or other alternative storage system should be given. There are many options to provide capacity increase in the wet storage area, but dry storages are worldwide preferred since it reduces corrosion concerns. In the wet storage the temperature and water purity should be constantly controlled whereas in the dry storage the SNF stands protected in specially designed canisters. Dry interim storages are practical and approved in many countries especially that have the 'wait and see' philosophy (wait to see new technologies development). This paper shows the advantages of dry interim storages sites in comparison with the wet ones and the nowadays problems as terrorism. (Author)

  3. Storage rack for fuel cell receiving shrouds

    International Nuclear Information System (INIS)

    Mollon, L.

    1978-01-01

    Disclosed is a rack for receiving a multiplicity of vertical tubular shrouds or tubes for storing spent nuclear fuel cells. The rack comprises a plurality of horizontally reticulated frames interconnected by tension rods and spacing tubes surrounding the rods

  4. Effect of long-term storage of LWR spent fuel on Pu-thermal fuel cycle

    International Nuclear Information System (INIS)

    Kurosawa, Masayoshi; Naito, Yoshitaka; Suyama, Kenya; Itahara, Kuniyuki; Suzuki, Katsuo; Hamada, Koji

    1998-01-01

    According to the Long-term Program for Research, Development and Utilization of Nuclear Energy (June, 1994) in Japan, the Rokkasho Reprocessing Plant will be operated shortly after the year 2000, and the planning of the construction of the second commercial plant will be decided around 2010. Also, it is described that spent fuel storage has a positive meaning as an energy resource for the future utilization of Pu. Considering the balance between the increase of spent fuels and the domestic reprocessing capacity in Japan, it can be expected that the long-term storage of UO 2 spent fuels will be required. Then, we studied the effect of long-term storage of spent fuels on Pu-thermal fuel cycle. The burnup calculation were performed on the typical Japanese PWR fuel, and the burnup and criticality calculations were carried out on the Pu-thermal cores with MOX fuel. Based on the results, we evaluate the influence of extending the spent fuel storage term on the criticality safety, shielding design of the reprocessing plant and the core life time of the MOX core, etc. As the result of this work on long-term storage of LWR spent fuels, it becomes clear that there are few demerits regarding the lifetime of a MOX reactor core, and that there are many merits regarding the safety aspects of the fuel cycle facilities. Furthermore, long-term storage is meaningful as energy storage for effective utilization of Pu to be improved by technological innovation in future, and it will allow for sufficient time for the important policymaking of nuclear fuel cycle establishment in Japan. (author)

  5. Criticality safety assessment on the RSG-GAS spent fuel storage for anticipating the next core conversion program

    International Nuclear Information System (INIS)

    Sembiring, Tagor Malem; Kuntoro, Iman; Zuhair; Liem, Peng Hong

    2003-01-01

    Criticality assessment on the spent fuel storage racks of the RSG-GAS multipurpose reactor has been conducted to support the undergoing core conversion program, in which higher uranium fuel densities of silicide (up to 4.8 gU.cm -3 ) and molybdenum (up to 8.3 gU.cm -3 ) fuel elements are adopted to enhance the reactor performance, core cycle length and reactor utilization. In the assessment, the k eff of the rack as a function of fuel density is calculated for fresh fuel elements which is a very conservative approach recommended by IAEA. Besides fuel densities, effects of water densities due to pool water temperature variation, and the fuel elements' orientation on the k eff are analyzed as well. The criticality calculations are all carried out by using MNCP4B2 Monte Carlo code with ENDF/B-VI library. For the library sensitivity, JENDL-3.3 library is also used and compared. The calculation results show the most reactive condition is for the case when the spent fuel racks are filled with fresh U-6Mo fuel element with meat density of 8.30 gU.cm -3 . For all fuel types, density and operating condition, the calculated k eff with 3 times standard deviations are confirmed less than the allowable value of 0.95. It can be concluded that the existing spent fuel storage racks can be safely used for storing the planned high density uranium fuels. (author)

  6. Development of the vacuum drying process for the PWR spent nuclear fuel dry storage

    Energy Technology Data Exchange (ETDEWEB)

    Baeg, Chagn Yeal; Cho, Chun Hyung [Korea Radioactive Waste Agency, Daejeon (Korea, Republic of)

    2016-12-15

    This paper describes the development of a dry operation process for PWR spent nuclear fuel, which is currently stored in the domestic NPP's storage pool, using a dual purpose metal cask. Domestic NNPs have had experience with wet type transportation of PWR spent nuclear fuel between neighboring NPPs since the early 1990s, but no experience with dry type operation. For this reason, we developed a specific operation process and also confirmed the safety of the major cask components and its spent nuclear fuel during the dual purpose metal cask operation process. We also describe the short term operation process that was established to be completed within 21 hours and propose the allowable working time for each step (15 hours for wet process, 3 hours for drain process and 3 hours for vacuum drying process)

  7. Large scale experiments simulating hydrogen distribution in a spent fuel pool building during a hypothetical fuel uncovery accident scenario

    Energy Technology Data Exchange (ETDEWEB)

    Mignot, Guillaume; Paranjape, Sidharth; Paladino, Domenico; Jaeckel, Bernd; Rydl, Adolf [Paul Scherrer Institute, Villigen (Switzerland)

    2016-08-15

    Following the Fukushima accident and its extended station blackout, attention was brought to the importance of the spent fuel pools' (SFPs) behavior in case of a prolonged loss of the cooling system. Since then, many analytical works have been performed to estimate the timing of hypothetical fuel uncovery for various SFP types. Experimentally, however, little was done to investigate issues related to the formation of a flammable gas mixture, distribution, and stratification in the SFP building itself and to some extent assess the capability for the code to correctly predict it. This paper presents the main outcomes of the Experiments on Spent Fuel Pool (ESFP) project carried out under the auspices of Swissnuclear (Framework 2012–2013) in the PANDA facility at the Paul Scherrer Institut in Switzerland. It consists of an experimental investigation focused on hydrogen concentration build-up into a SFP building during a predefined scaled scenario for different venting positions. Tests follow a two-phase scenario. Initially steam is released to mimic the boiling of the pool followed by a helium/steam mixture release to simulate the deterioration of the oxidizing spent fuel. Results shows that while the SFP building would mainly be inerted by the presence of a high concentration of steam, the volume located below the level of the pool in adjacent rooms would maintain a high air content. The interface of the two-gas mixture presents the highest risk of flammability. Additionally, it was observed that the gas mixture could become stagnant leading locally to high hydrogen concentration while steam condenses. Overall, the experiments provide relevant information for the potentially hazardous gas distribution formed in the SFP building and hints on accident m