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Sample records for candu spent fuel

  1. CANDU spent fuel dry storage interim technique

    International Nuclear Information System (INIS)

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

  2. Shielding calculations for spent CANDU fuel transport cask

    International Nuclear Information System (INIS)

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

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

    Institute of Scientific and Technical Information of China (English)

    HUO Xiao-Dong; XIE Zhong-Sheng

    2004-01-01

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

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

    International Nuclear Information System (INIS)

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

  5. Development of CANDU Spent Fuel Disposal Concepts for the Improvement of Disposal Efficiency

    International Nuclear Information System (INIS)

    There are two types of spent fuels generated from nuclear power plants, CANDU type and PWR type. PWR spent fuels which include a lot of reusable material can be considered to be recycled. CANDU spent fuels are considered to directly disposed in deep geological formation, since they have little reusable material. In this study, based on the Korean Reference spent fuel disposal System(KRS) which is to dispose both PWR and CANDU spent fuels, the more effective CANDU spent fuel disposal systems have been developed. To do this, the disposal canister has been modified to hold the storage basket which can load 60 spent fuel bundles. From these modified disposal canisters, the disposal systems to meet the thermal requirement for which the temperature of the buffer materials should not be over have been proposed. These new disposals have made it possible to introduce the concept of long term storage and retrievability and that of the two-layered disposal canister emplacement in one disposal hole. These disposal concepts have been compared and analyzed with the KRS CANDU spent fuel disposal system in terms of disposal effectiveness. New CANDU spent fuel disposal concepts obtained in this study seem to improve thermal effectiveness, U-density, disposal area, excavation volume, and closure material volume up to 30 - 40 %.

  6. Photon dose rates estimation for CANDU spent fuel transport and intermediate dry storage

    International Nuclear Information System (INIS)

    The nuclear energy world wide development is accompanied by huge quantities of spent nuclear fuel accumulation. Shielding analyses are an essential component of the nuclear safety, the estimations of radiation doses in order to reduce them under specified limit values being the main task here. According to IAEA data, more than 10 millions packages containing radioactive materials are annually world wide transported. The radioactive material transport safety must be carefully settled. Last decade, both for operating reactors and future reactor projects, a general trend to raise the discharge fuel burnup has been world wide registered. For CANDU type reactors, one of the most attractive solutions seems to be SEU fuel utilization. In the paper there are estimated the CANDU spent fuel photon dose rates at the shipping cask/ storage basket wall for two different fuel projects after a defined cooling period in the NPP pools. The CANDU fuel projects considered were the CANDU standard 37 rod fuel bundle with natural UO2 and SEU fuels. In order to obtain radionuclide inventory and irradiated fuel characteristics, ORIGEN-S code has been used. The spent fuel characteristics are presented, comparatively, for both types of CANDU fuels. By means of the same code the photon source profiles have been calculated. The shielding calculations both for spent fuel transport and intermediate storage have been performed by using Monte Carlo MORSE-SGC code. The ORIGEN-S and MORSE-SGC codes are both included in ORNL's SCALE 4.4a program package. A photon dose rates comparison between the two types of CANDU fuels has been also performed, both for spent fuel transport and intermediate dry storage. (authors)

  7. Estimation of radiation doses characterizing CANDU spent fuel transport and intermediate dry storage

    International Nuclear Information System (INIS)

    Shielding analyses are an essential component of the nuclear safety. The estimations of radiation doses in order to reduce them under specified limitation values is the main task here. In the last decade, a general trend to raise the discharge fuel burnup has been world wide registered for both operating reactors and future reactor projects. For CANDU type reactors, one of the most attractive solutions seems to be SEU fuels utilization. The goal of this paper is to estimate CANDU spent fuel photon dose rates at the shipping cask/storage basket wall and in air, at different distances from the cask/ basket, for two different fuel projects, after a defined cooling period in the NPP pools. Spent fuel inventories and photon source profiles are obtained by means of ORIGEN-S code. The shielding calculations have been performed by using Monte Carlo MORSE-SGC code. A comparison between the two types of CANDU fuels has been also performed. (authors)

  8. Romania Monte Carlo Methods Application to CANDU Spent Fuel Comparative Analysis

    International Nuclear Information System (INIS)

    Romania has a single NPP at Cernavoda with 5 PHWR reactors of CANDU6 type of 705 MW(e) each, with Cernavoda Unit1, operational starting from December 1996, Unit2 under construction while the remaining Unit3-5 is being conserved. The nuclear energy world wide development is accompanied by huge quantities of spent nuclear fuel accumulation. Having in view the possible impact upon population and environment, in all activities associated to nuclear fuel cycle, namely transportation, storage, reprocessing or disposal, the spent fuel characteristics must be well known. The paper aim is to apply Monte Carlo methods to CANDU spent fuel analysis, starting from the discharge moment, followed by spent fuel transport after a defined cooling period and finishing with the intermediate dry storage. As radiation source 3 CANDU fuels have been considered: standard 37 rods fuel bundle with natural UO2 and SEU fuels, and 43 rods fuel bundle with SEU fuel. After a criticality calculation using KENO-VI code, the criticality coefficient and the actinides and fission products concentrations are obtained. By using ORIGEN-S code, the photon source profiles are calculated and the spent fuel characteristics estimation is done. For the shielding calculations MORSE-SGC code has been used. Regarding to the spent fuel transport, the photon dose rates to the shipping cask wall and in air, at different distances from the cask, are estimated. The shielding calculation for the spent fuel intermediate dry storage is done and the photon dose rates at the storage basket wall (active element of the Cernavoda NPP intermediate dry storage) are obtained. A comparison between the 3 types of CANDU fuels is presented. (authors)

  9. Thermal Analysis of CANDU Spent Fuel Bay Cooling System

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

  10. Thermal analysis model for the temperature distribution of the CANDU spent fuel assembly

    International Nuclear Information System (INIS)

    The purpose of this technical is to introduce the methodology and experimental process for the experimental research work with the mock-up test performed to verify and validate the MAXROT code which is a thermal analysis method for Wolsong (CANDU) spent fuel dry storage canister. The experiment was conducted simulating the heat transfer characteristics of combinations of equilateral triangular and square pitch arrays of heater rods, similar to a CANDU spent fuel bundle. After assembly of the heater rod bundle into the containment vessel, the experimental apparatus was operated under the same operating and boundary conditions as an interim dry storage condition at the nuclear power plant site. The reduced data from this experiment has been utilized to verity a model developed to predict the maximum fuel rod surface temperature in a fuel bundle. These test procedures and the experiment can be utilized to establish the fine thermal analysis method applicable to dry storage system for the spent fuel. 12 figs., 5 tabs., 36 refs. (Author) .new

  11. A study on the direct use of spent PWR fuel in CANDU reactors. DUPIC facility engineering

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyun Soo; Lee, Jae Sul; Choi, Jong Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This report summarizes the second year progress of phase II of DUPIC program which aims to verify experimentally the feasibility of direct use of spent PWR fuel in CANDU reactors. The project is to provide the experimental facilities and technologies that are required to perform the DUPIC experiment. As an early part of the project, engineering analysis of those facilities and construction of mock-up facility are described. Another scope of the project is to assess the DUPIC fuel cycle system and facilitate international cooperation. The progresses in this scope of work made during the fiscal year are also summarized in the report. 38 figs, 44 tabs, 8 refs. (Author).

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

    International Nuclear Information System (INIS)

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

  13. Enrichment effects on CANDU-SEU spent fuel Monte Carlo shielding analysis

    International Nuclear Information System (INIS)

    Shielding analyses are an essential component of the nuclear safety, the estimations of radiation doses in order to reduce them under specified limitation values being the main task here. According to IAEA data, more than 10 millions packages containing radioactive materials are annually transported world wide. All the problems arisen from the safe radioactive materials transport assurance must be carefully settled. Last decade, both for operating reactors and future reactor projects, a general trend to raise the discharge fuel burnup has been recorded world wide. For CANDU type reactors, the most attractive solution seems to be SEU and RU fuels utilization. The basic tasks accomplished by the shielding calculations in a nuclear safety analysis consist in dose rates calculation, to prevent any risks both for personnel protection and impact on the environment during the spent fuel manipulation, transport and storage. The paper aims to study the effects induced by fuel enrichment variation on CANDU-SEU spent fuel photon dose rates for a Monte Carlo shielding analysis applied to spent fuel transport after a defined cooling period in the NPP pools. The fuel bundles projects considered here have 43 Zircaloy rods, filled with SEU fuel pellets, the fuel having different enrichment in U-235. All the geometrical and material data related on the cask were considered according to the shipping cask type B model. After a photon source profile calculation by using ORIGEN-S code, in order to perform the shielding calculations, Monte Carlo MORSE-SGC code has been used, both codes being included in the ORNL's SCALE 5 system. The photon dose rates to the shipping cask wall and in air, at different distances from the cask, have been estimated. Finally, a photon dose rates comparison for different fuel enrichments has been performed. (author)

  14. An experimental investigation of the temperature behavior of a CANDU 37-element spent fuel bundle with air backfill

    International Nuclear Information System (INIS)

    As part of the thermal analysis of a CANDU spent fuel dry storage system, a series of experiment has been conducted using a thermal mock-up of a simulated CANDU spent fuel bundle in a dry storage basket. The experimental system was designed to obtain the maximum fuel rod temperature along with the radial and axial temperature distributions within the fuel bundle. The main purpose of these experiments was to characterize the relevant heat transfer mechanisms in a dry, vertically oriented CANDU spent fuel bundle, and to verify the MAXROT code developed for the thermal analysis of a CANDU spent fuel bundle in a dry storage basket. A total of 48 runs were made with 8 different power inputs to the 37-element heater rod bundle ranging from 5 to 40 W, while using 6 different band heaters power inputs from 0 to 250 W to maintain the basket wall at a desired boundary condition temperature at the steady state. The temperature distribution in a heater rod bundle was measured and recorded at the saturated condition for each set of heater rod power and band heaters power. To characterize the heat transfer mechanism involved, the experimental data were corrected analytically for radiation heat transfer and presented as a Nusselt number correlation in terms of the Rayleigh number of the heater rod bundle. The results show that the Nusselt number remains nearly constant and all the experimental dada fall within a conduction regime. The experimental data were compared with the predictions of the MAXROT code to examine the code's accuracy and validity of assumptions used in the code. The MAXROT code explicitly models each representative fuel rod in a CANDU fuel bundle and couples the conductive and radiative heat transfer of the internal gas between rods. Comparisons between the measured and predicted maximum fuel rod temperatures of the simulated CANDU 37-element spent fuel bundle for all 48 tests show that the MAXROT code slightly over-predicts and the agreement is within 2

  15. Current Status and Characterization of CANDU Spent Fuel for Geological Disposal System Design

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Dong Keun; Lee, Seung Woo; Cha, Jeong Hun; Choi, Jong Won; Choi, Heui Joo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Yang [SK Engineering and Construction, Seoul (Korea, Republic of)

    2008-06-15

    Inventories to be disposed of, reference turn up, and source terms for CANDU spent fuel were evaluated for geological disposal system design. The historical and projected inventory by 2040 is expected to be 14,600 MtU under the condition of 30-year lifetime for unit 1 and 40-year lifetime for other units in Wolsong site. As a result of statistical analysis for discharge burnup of the spent fuels generated by 2007, average and stand deviation revealed 6,987 MWD/MtU and 1,167, respectively. From this result, the reference burnup was determined as 8,100 MWD/MtU which covers 84% of spent fuels in total. Source terms such as nuclide concentration for a long-term safety analysis, decay heat, thermo-mechanical analysis, and radiation intensity and spectrum was characterized by using ORIGEN-ARP containing conservativeness in the aspect of decay heat up to several thousand years. The results from this study will be useful for the design of storage and disposal facilities.

  16. Some aspects on security and safety in a potential transport of a CANDU spent nuclear fuel bundle, in Romania

    International Nuclear Information System (INIS)

    Each Member States (MS) is responsible for the security and safety of radioactive material during transport, since radioactive material is most vulnerable during transport. The paper presents some aspects on security and safety related to the potential transport of a CANDU Spent Nuclear Fuel (SNF) bundle from NPP CANDU Cernavoda to INR Pitesti. The possible environmental impact and radiological consequences following a potential event during transportation is analyzed, since the protection of the people and the environment is the essential goal to be achieved. Some testing for the package to be used for transportation will be also given. (author)

  17. Thermal analysis of dry concrete canister storage system for CANDU spent fuel

    International Nuclear Information System (INIS)

    This paper presents the results of a thermal analysis of the concrete canisters for interim dry storage of spent, irradiated Canadian Deuterium Uranium(CANDU) fuel. The canisters are designed to contain 6-year-old fuel safely for periods of 50 years in stainless steel baskets sealed inside a steel-lined concrete shield. In order to assure fuel integrity during the storage, fuel rod temperature shall not exceed the temperature limit. The contents of thermal analysis include the following : 1) Steady state temperature distributions under the conservative ambient temperature and insolation load. 2) Transient temperature distributions under the changes in ambient temperature and insolation load. Accounting for the coupled heat transfer modes of conduction, convection, and radiation, the computer code HEATING5 was used to predict the thermal response of the canister storage system. As HEATING5 does not have the modeling capability to compute radiation heat transfer on a rod-to-rod basis, a separate calculating routine was developed and applied to predict temperature distribution in a fuel bundle. Thermal behavior of the canister is characterized by the large thermal mass of the concrete and radiative heat transfer within the basket. The calculated results for the worst case (steady state with maximum ambient temperature and design insolation load) indicated that the maximum temperature of the 6 year cooled fuel reached to 182.4 .deg. C, slightly above the temperature limit of 180 .deg. C. However,the thermal inertia of the thick concrete wall moderates the internal changes and prevents a rise in fuel temperature in response to ambient changes. The maximum extent of the transient zone was less than 75% of the concrete wall thickness for cyclic insolation changes. When transient nature of ambient temperature and insolation load are considered, the fuel temperature will be a function of the long term ambient temperature as opposed to daily extremes. The worst design

  18. Seismic Soil-Structure Interaction Analysis of a Consolidated Dry Storage Module for CANDU Spent Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Young Gon; Yoon, Jeong Hyoun; Cho, Chun Hyung; Lee, Heung Young [Korea Hydro and Nuclear Power Co., Ltd., Taejon (Korea, Republic of); Choi, Kyu Sup; Jeong, In Su; Kim, Jong Soo [KONES Co., Seoul (Korea, Republic of)

    2005-07-01

    The MACSTOR/KN-400 module has been developed as an effective alternative to the existing stand alone concrete canister for dry storage of CANDU spent fuel. The structure is a concrete monolith of 21.67 m long and 12.66 m wide and has a height equal to 7.518 m including the bottom slab. Inside of the concrete module are built 40 storage cylinders accommodating ten 60- bundle dry storage baskets, which are suspended from the top slab and eventually constrained at 10 cm above the bottom slab with horizontal seismic restraints. The main cooling process of the MACSTOR/KN-400 module is by air convection through air inlets and outlets. The civil design parameters, with respect to meteorological and seismic loads applied to the module are identical to those specified for the Wolsung CANDU 3 and 4 plants except for local geologic characteristics. As per USNRC SRP Section 3.7.2 and current US practices, Soil-Structure Interaction (SSI) effect shall be considered for all structures not supported by a rock or rock-like soil foundation materials. An SSI is a very complicated phenomenon of the structure coupled with the soil medium that is usually semi-infinite in extent and highly nonlinear in its behavior. And the effect of the SSI is noticeable especially for stiff and massive structures resting on relatively soft ground. Thus the SSI effect has to be considered in the seismic design of MACSTOR/KN-400 module resting on soil medium. The scope of the this paper is to carry out a seismic SSI analysis of the MACSTOR/KN-400 module, in order to show how much the SSI gives an effect on the structural responses by comparing with the fixed-base analysis.

  19. A study on the direct use of spent PWR fuel in CANDU reactors -Fuel management and safety analysis-

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyun Soo; Lee, Boh Wook; Choi, Hang Bok; Lee, Yung Wook; Cho, Jae Sun; Huh, Chang Wook [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    The reference DUPIC fuel composition was determined based on the reactor safety, thermal-hydraulics, economics, and refabrication aspects. The center pin of the reference DUPIC fuel bundle is poisoned with natural dysprosium. The worst LOCA analysis has shown that the transient power and heat deposition of the reference DUPIC core are the same as those of natural uranium CANDU core. The intra-code comparison has shown that the accuracy of DUPIC physics code system is comparable to the current CANDU core design code system. The sensitivity studies were performed for the refuelling schemes of DUPIC core and the 2-bundle shift refuelling scheme was selected as the standard refuelling scheme of the DUPIC core. The application of 4-bundle shift refuelling scheme will be studied in parallel as the auto-refuelling method is improved and the reference core parameters of the heterogeneous DUPIC core are defined. The heterogeneity effect was analyzed in a preliminary fashion using 33 fuel types and the random loading strategy. The refuelling simulation has shown that the DUPIC core satisfies the current CANDU 6 operating limits of channel and bundle power regardless of the fuel composition heterogeneity. The 33 fuel types used in the heterogeneity analysis was determined based on the initial enrichment and discharge burnup of the PWR fuel. 90 figs, 62 tabs, 63 refs. (Author).

  20. A study on the radioactive waste management for DUPIC fuel cycle - A study on the direct use of spent PWR fuel in CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyeon Soo; Chun, Kwan Sik; Kim, Jong Ho; Cho, Yeong Hyeon; Paek, Seung Uh; Kang, Hee Seok; Na, Jeong Won; Choi, Jong Won; Lee, Hu Keun; Park, Keun Il; Yang, Seung Yeong [Korea Atomic Energy Res. Inst., Taejon (Korea, Republic of)

    1994-07-01

    The characteristics of the radioactive wastes coming from the DUPIC fuel manufacturing process were analyzed and evaluated. The release amounts of nuclides were estimated, and the characteristics of the oxides of semi-volatile nuclides such as Ru and Cs were analyzed and also these trapping and fixation technologies were assessed. The conceptual methods of experiment including apparatus and procedure were completed to define the vaporization behavior of nuclides. The gross {alpha}-activity and {alpha}-, {gamma}-spectrum of irradiated zircaloy specimens from KORI unit 1 were analyzed and the press to test the compaction of hulls was manufactured. Ingestion hazard index for the disposal of spent DUPIC fuel were calculated using ORIGEN 2 code and compared with the disposal of once-through spent PWR and CANDU fuels, and then the index per kwh for DUPIC cycle was 1/8 {approx} 1/4 lower than that for once-through cycle. (Author).

  1. Fuel for advanced CANDU reactors

    International Nuclear Information System (INIS)

    The CANDU reactor system has proven itself to be a world leader in terms of station availability and low total unit energy cost. In 1985 for example, four of the top ten reactor units in the world were CANDU reactors operating in South Korea and Canada. This excellent operating record requires an equivalent performance record of the low-cost, natural uranium fuel. Future CANDU reactors will be an evolution of the present design. Engineering work is under way to refine the existing CANDU 600 and to incorporate state-of-the-art technology, reducing the capital cost and construction schedule. In addition, a smaller CANDU 300 plant has been designed using proven CANDU 600 technology and components but with an innovative new plant layout that makes it cost competitive with coal fired plants. For the long term, work on advanced fuel cycles and major system improvements is underway ensuring that CANDU plants will stay competitive well into the next century

  2. Demonstration Drop Test and Design Enhancement of the CANDU Spent Fuel Storage Basket in MACSTOR/KN-40

    International Nuclear Information System (INIS)

    A dry interim storage facility named MACSTOR/KN-400 has been constructed at the Wolsung power plant in Korea. The MACSTOR/KN-400 has the separated 7 modules. There are 400 long slender cylinders in one module. In one cylinder, ten baskets where CANDU spent fuels are loaded are stacked and stored. For this MACSTOR/KN-400 facility, analyses and tests for the hypothetical accident conditions that might happen during moving and storing baskets into a cylinder were performed. In a demonstration test, one of test basket models did not satisfy one of the safety-related requirements. So, the revised basket designs were generated by the structural evaluation based on the finite element analyses and specimen tests. Among these revised designs, one design was chosen as the final revised design of the basket. The final revised design is the one that makes the largest reduction of plastic strain in the upper welding region. And it also is the one that makes the smallest design change from the previous basket design and the one easy to adopt the design change. Drop tests and leak tests of the final revised basket were performed and it satisfied all the performance requirements. (author)

  3. Safety Evaluation Report for CANDU Spent Fuel Dry Storage Basket Drop Test

    International Nuclear Information System (INIS)

    This is technical report about the safety demonstration test for the CANDU S/F storage basket. Two demonstration tests were performed such as one case of dropping a basket into the bottom of the cylinder from 7.5 m height and another case of dropping a basket onto the other basket loaded in a cylinder. The existing basket was improved to satisfy the performance requirement. The leak test showed that the enhanced basket had no leak after all the drop tests. The dropped basket was able to be withdrawn through top of the cylinder by a handling tool, which proved the maintenance of retrievability after drop accident

  4. Seismic Structure-Soil-Structure Interaction Analysis of a Consolidated Dry Storage Module for CANDU Spent Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Young Gon; Yoon, Jeong Hyoun; Kim, Sung Hwan; Yang, Ke Hyung; Lee, Heung Young; Cho, Chun Hyung [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of); Choi, Kyu Sup; Jeong, In Su [KONES Corporation, Seoul (Korea, Republic of)

    2005-07-01

    The MACSTOR/KN-400 module has been developed as an effective alternative to the existing stand alone concrete canister for dry storage of CANDU spent fuel. The structure is a concrete monolith of 21.67 m long and 12.66 m wide and has a height equal to 7.518 m including the bottom slab. Inside the concrete module consists of 40 storage cylinders accommodating ten 60-bundle dry storage baskets, which are suspended from the top slab and eventually restrained at 10 cm above the bottom slab with horizontal seismic restraints. The main cooling process of the MACSTOR/KN-400 module shall be by air convection through air inlets and outlets. The civil design parameters, with respect to meteorological and seismic loads applied to the module are identical to those specified for the Wolsung CANDU 3 and 4 plants, except for local site characteristics required for soilstructure interaction (SSI) analysis. It is required for the structural integrity to fulfill the licensing requirements. As per USNRC SRP Section 3.7.2, it shall be reviewed how to consider the phenomenon of coupling of the dynamic response of adjacent structures through the soil, which is referred to as structure-soil-structure interaction (SSSI). The presence of closely spaced multiple structural foundations creates coupling between the foundations of individual structures . Some observations of the actual seismic response of structures have indicated that SSSI effects do exist, but they are generally secondary for the overall structural response motions. SSSI effects, however, may be important for a relatively small structure which is to be close to a relatively large structure, while they may be generally neglected for overall structural response of a large massive structure, such as nuclear power plant. As such the scope of the present paper is to carry out a seismic SSSI analysis in case of the MACSTOR/KN- 400 module, in order to investigate whether or not SSSI effect shall be included in the overall seismic

  5. Second International Conference on CANDU Fuel

    International Nuclear Information System (INIS)

    Thirty-four papers were presented at this conference in sessions dealing with international experience and programs relating to CANDU fuel; fuel manufacture; fuel behaviour; fuel handling, storage and disposal; and advanced CANDU fuel cycles. (L.L.)

  6. Using Advanced Fuel Bundles in CANDU Reactors

    International Nuclear Information System (INIS)

    Improving the exit fuel burnup in CANDU reactors was a long-time challenge for both bundle designers and performance analysts. Therefore, the 43-element design together with several fuel compositions was studied, in the aim of assessing new reliable, economic and proliferation-resistant solutions. Recovered Uranium (RU) fuel is intended to be used in CANDU reactors, given the important amount of slightly enriched Uranium (~0.96% w/o U235) that might be provided by the spent LWR fuel recovery plants. Though this fuel has a far too small U235 enrichment to be used in LWR's, it can be still used to fuel CANDU reactors. Plutonium based mixtures are also considered, with both natural and depleted Uranium, either for peacefully using the military grade dispositioned Plutonium or for better using Plutonium from LWR reprocessing plants. The proposed Thorium-LEU mixtures are intended to reduce the Uranium consumption per produced MW. The positive void reactivity is a major concern of any CANDU safety assessment, therefore reducing it was also a task for the present analysis. Using the 43-element bundle with a certain amount of burnable poison (e.g. Dysprosium) dissolved in the 8 innermost elements may lead to significantly reducing the void reactivity. The expected outcomes of these design improvements are: higher exit burnup, smooth/uniform radial bundle power distribution and reduced void reactivity. Since the improved fuel bundles are intended to be loaded in existing CANDU reactors, we found interesting to estimate the local reactivity effects of a mechanical control absorber (MCA) on the surrounding fuel cells. Cell parameters and neutron flux distributions, as well as macroscopic cross-sections were estimated using the transport code DRAGON and a 172-group updated nuclear data library. (author)

  7. A study on the direct use of spent PWR fuel in CANDU -The technology development for nuclear material safeguards-

    International Nuclear Information System (INIS)

    The research contents of the non-destructive assay real time accounting system and the preparation of DIQ. The quantity and neutron emission rate variation of curium and plutonium was investigated with respect to burnup. Also, the neutron detection probability was examined in the neutron slowing down medium and shielding calculation was carried out against the intense gamma rays emitted from the spent fuel. The investigation of the variation of plutonium and curium contents was very helpful to design of nondestructive assay system from the emitted neutrons in the spent fuel. 24 figs, 2 tabs, 20 refs. (Author)

  8. Ninth international conference on CANDU fuel, 'fuelling a clean future'

    International Nuclear Information System (INIS)

    The Canadian Nuclear Society's 9th International Conference on CANDU fuel took place in Belleville, Ontario on September 18-21, 2005. The theme for this year's conference was 'Fuelling a Clean Future' bringing together over 80 delegates ranging from: designers, engineers, manufacturers, researchers, modellers, safety specialists and managers to share the wealth of their knowledge and experience. This international event took place at an important turning point of the CANDU technology when new fuel design is being developed for commercial application, the Advanced CANDU Reactor is being considered for projects and nuclear power is enjoying a renaissance as the source energy for our future. Most of the conference was devoted to the presentation of technical papers in four parallel sessions. The topics of these sessions were: Design and Development; Fuel Safety; Fuel Modelling; Fuel Performance; Fuel Manufacturing; Fuel Management; Thermalhydraulics; and, Spent Fuel Management and Criticalty

  9. Technology development for nuclear material safeguards -A study on the direct use of spent PWR fuel in CANDU-

    International Nuclear Information System (INIS)

    The research contents of the passed one year were the conceptual design of nuclear material measurement points, of near real time accounting system and of unattended monitoring system for detection of nuclear material diversion. The passive neutron detection system was decided as a proper way of detection of plutonium in the spent fuels and the neutrons emitted by each isotopes were investigated. Also, material balance area and major measurement points were selected and related computer code was used for the near real time accounting in DUPIC facility. (Author)

  10. TRIGA spent fuel storage

    International Nuclear Information System (INIS)

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

  11. Economic analysis of alternative options in CANDU fuel cycle

    International Nuclear Information System (INIS)

    In this study, fuel cycle options for CANDU reactor were studied. Three main options in a CANDU fuel cycle involve use of : (1) natural uranium (0.711 weight percent U-235) fuel, (2) slightly enriched uranium (1.2 weight percent U-235) fuel, and (3) recovered uranium (0.83 weight percent U-235) fuel from light water reactor spent fuel. ORIGEN-2 computer code was used to identify composition of the spent fuel for each option , including the standard LWR fuel (3.3 weight percent U-235). Uranium and plutonium credit calculations were performed by using ORIGEN-2 output. WIMSD-5 computer code was used to determine maximum discharge burnup values for each case. Cost estimations were carried out using specially-developed computer programs. Comparison of levelized costs for the fuel cycle options and sensitivity analysis for the cost components are also presented

  12. Proceedings of the fifth international conference on CANDU fuel. V.1,2

    Energy Technology Data Exchange (ETDEWEB)

    Lau, J.H. [ed.

    1997-07-01

    The First International Conference on CANDU Fuel was held in Chalk River in 1986. The CANDU Fuel community has gathered every three years since. The papers presented include topics on international experience, CANFLEX fuel bundles, Fuel design, Fuel modelling, Manufacturing and Quality assurance, Fuel performance and Safety, Fuel cycles and Spent Fuel management. Volume One was published in advance of the conference and Volume Two was printed after the conference.

  13. Proceedings of the fifth international conference on CANDU fuel. V.1,2

    International Nuclear Information System (INIS)

    The First International Conference on CANDU Fuel was held in Chalk River in 1986. The CANDU Fuel community has gathered every three years since. The papers presented include topics on international experience, CANFLEX fuel bundles, Fuel design, Fuel modelling, Manufacturing and Quality assurance, Fuel performance and Safety, Fuel cycles and Spent Fuel management. Volume One was published in advance of the conference and Volume Two was printed after the conference

  14. Implementation of an integrated safeguards approach for transfers of spent fuel to dry storage at multi-unit CANDU generating stations in Canada

    International Nuclear Information System (INIS)

    Full text: The Canadian Nuclear Safety Commission (CNSC) in 2002 undertook an IAEA Support Program task to develop and test an Integrated Safeguards (IS) approach for transfers of spent fuel to dry storage at multi-unit CANDU stations. The IS approach that Canada proposed in 2003 was successfully field tested in Canada by the IAEA in 2004. Since then the IAEA has drawn a positive conclusion for Canada and is preparing to implement IS in Canada based on the State-level Integrated Safeguards Approach (SLA) for Canada that was concluded in November 2005. Implementation of an IS approach for transfers to dry storage, similar to the one proposed by Canada, will be part of the initial phase of the implementation of the SLA. This paper briefly reviews the IS approach for transfers proposed by Canada and explains how the approach may be modified for implementation as part of the SLA to take into account lessons learned from the field test, operator and IAEA concerns, and developments in technology. (author)

  15. Candu advanced fuel cycles: key to energy sustainability

    International Nuclear Information System (INIS)

    A primary rationale for Indonesia to proceed with a nuclear power program is to diversity its energy sources and achieve freedom from future resource constraints. While other considerations, such as economy of power supply, hedging against potential future increases in the price of fossil fuels, fostering the technological development of the Indonesia economy and minimizing greenhouse and other gaseous are important, the strategic resource issue is key. In considering candidate nuclear power technologies upon which to base such a program, a major consideration will be the potential for those technologies to be economically sustained in the face of large future increases in demand for nuclear fuels. the technology or technologies selected should be amenable to evaluation in a rapidly changing technical, economic, resource and environmental policy environment. the world's proven uranium resources which can be economically recovered represent a fairly modest energy resource if utilization is based on the currently commercialized fuel cycles, even with the use of recovered plutonium in mixed oxide fuels. In the long term, fuel cycles relying solely on the use of light water reactors will encounter increasing fuel supply constraints. Because of its outstanding neutron economy and the flexibility of on-power refueling, Candu reactors are the most fuel resource efficient commercial reactors and offer the potential for accommodating an almost unlimited variety of advanced and even more fuel efficient cycles. Most of these cycles utilize nuclear fuel which are too low grade to be used in light water reactors, including many products now considered to be waste, such as spent light water reactor fuel and reprocessing products such as recovered uranium. The fuel-cycle flexibility of the Candu reactor provides a ready path to sustainable energy development in both the short and the long terms. Most of the potential Candu fuel cycle developments can be accommodated in existing

  16. Future fuel cycle development for CANDU reactors

    International Nuclear Information System (INIS)

    The CANDU reactor has proven to be safe and economical and has demonstrated outstanding performance with natural uranium fuel. The use of on-power fuelling, coupled with excellent neutron economy, leads to a very flexible reactor system with can utilize a wide variety of fuels. The spectrum of fuel cycles ranges from natural uranium, through slightly enriched uranium, to plutonium and ultimately thorium fuels which offer many of the advantages of the fast breeder reactor system. CANDU can also burn the recycled uranium and/or the plutonium from fuel discharged from light water reactors. This synergistic relationship could obviate the need to re-enrich the reprocessed uranium and allow a simpler reprocessing scheme. Fule management strategies that will permit future fuel cycles to be used in existing CANDU reactors have been identified. Evolutionary design changes will lead to an even greater flexibility, which will guarantee the continued success of the CANDU system. (author)

  17. Direct reuse of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  18. Thermochemical modelling of advanced CANDU reactor fuel

    Science.gov (United States)

    Corcoran, Emily Catherine

    2009-04-01

    With an aging fleet of nuclear generating facilities, the imperative to limit the use of non-renewal fossil fuels and the inevitable need for additional electricity to power Canada's economy, a renaissance in the use of nuclear technology in Canada is at hand. The experience and knowledge of over 40 years of CANDU research, development and operation in Ontario and elsewhere has been applied to a new generation of CANDU, the Advanced CANDU Reactor (ACR). Improved fuel design allows for an extended burnup, which is a significant improvement, enhancing the safety and the economies of the ACR. The use of a Burnable Neutron Absorber (BNA) material and Low Enriched Uranium (LEU) fuel has created a need to understand better these novel materials and fuel types. This thesis documents a work to advance the scientific and technological knowledge of the ACR fuel design with respect to thermodynamic phase stability and fuel oxidation modelling. For the BNA material, a new (BNA) model is created based on the fundamental first principles of Gibbs energy minimization applied to material phase stability. For LEU fuel, the methodology used for the BNA model is applied to the oxidation of irradiated fuel. The pertinent knowledge base for uranium, oxygen and the major fission products is reviewed, updated and integrated to create a model that is applicable to current and future CANDU fuel designs. As part of this thesis, X-Ray Diffraction (XRD) and Coulombic Titration (CT) experiments are compared to the BNA and LEU models, respectively. From the analysis of the CT results, a number of improvements are proposed to enhance the LEU model and provide confidence in its application to ACR fuel. A number of applications for the potential use of these models are proposed and discussed. Keywords: CANDU Fuel, Gibbs Energy Mimimization, Low Enriched Uranium (LEU) Fuel, Burnable Neutron Absorber (BNA) Material, Coulometric Titration, X-Ray Diffraction

  19. Candu reactors with thorium fuel cycles

    International Nuclear Information System (INIS)

    Over the last decade and a half AECL has established a strong record of delivering CANDU 6 nuclear power plants on time and at budget. Inherently flexible features of the CANDU type reactors, such as on-power fuelling, high neutron economy, fuel channel based heat transport system, simple fuel bundle configuration, two independent shut down systems, a cool moderator and a defence-in-depth based safety philosophy provides an evolutionary path to further improvements in design. The immediate milestone on this path is the Advanced CANDU ReactorTM** (ACRTM**), in the form of the ACR-1000TM**. This effort is being followed by the Super Critical Water Reactor (SCWR) design that will allow water-cooled reactors to attain high efficiencies by increasing the coolant temperature above 5500C. Adaptability of the CANDU design to different fuel cycles is another technology advantage that offers an additional avenue for design evolution. Thorium is one of the potential fuels for future reactors due to relative abundance, neutronics advantage as a fertile material in thermal reactors and proliferation resistance. The Thorium fuel cycle is also of interest to China, India, and Turkey due to local abundance that can ensure sustainable energy independence over the long term. AECL has performed an assessment of both CANDU 6 and ACR-1000 designs to identify systems, components, safety features and operational processes that may need to be modified to replace the NU or SEU fuel cycles with one based on Thorium. The paper reviews some of these requirements and the associated practical design solutions. These modifications can either be incorporated into the design prior to construction or, for currently operational reactors, during a refurbishment outage. In parallel with reactor modifications, various Thorium fuel cycles, either based on mixed bundles (homogeneous) or mixed channels (heterogeneous) have been assessed for technical and economic viability. Potential applications of a

  20. Fuel condition in Canadian CANDU 6 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hu, R.H.; Macici, N [Hydro-Quebec, Montreal, Quebec (Canada); Gibb, R. [New Brunswick Power, Lepreau, NB (Canada); Purdy, P.L.; Manzer, A.M. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Kohn, E. [Ontario Hydro, Toronto, Ontario (Canada)

    1997-07-01

    The cornerstone of the CANDU concept is its natural uranium fuel, and the success of its reactor operation hinges on the fuel condition in the reactor. Neutron economy, on power refuelling, and simple fuel design are among the unique characteristics of CANDU fuel. In Canadian CANDU 6 reactors (Gentilly 2 and Point Lepreau), the 37-element fuel has provided an enviable record of safe, economic and reliable plant operation for 29 reactor years to date. The fuelling cost is among the lowest in the world - a corollary of high neutron economy, simple fuel design, and judicial fuelling scheme. The reliability of fuel is high: only 21 of the 60000 bundles discharged from Gentilly 2 were confirmed defective and the five-year period from March 1992 to February 1997 saw no defect at all at Gentilly-2. Also, thanks to the inherent on-power refuelling capability and an effective defect detection and removal system, the primary coolant loops are kept extremely clean (very low activity level) - benefiting both maintenance and safety. Moreover, the inventories of fission products in the core and in the channel are maintained within the safety analysis envelope, due to on-power fuelling and sophisticated fuel management. In this paper, CANDU 6 fuel performance is reviewed against the feedback from post-irradiation examinations, and the findings from our ongoing R and D program. The results suggest that the fuel behavior m reactor are basically as originally anticipated, despite an evolutionary 3% increase in bundle uranium mass in the 1980's. For operating conditions within the CANDU 6 37-element experience, the average strains are typically 0.09%; and fission gas release, 2.7%. The UO{sub 2} fuel remains stoichiometric after irradiation. In-core measurements of pressure tube fitting are generally low. All these observations are consistent with the excellent fuel performance statistics coming out of the two Canadian CANDU 6 reactors. Additionally, this paper will briefly

  1. Fuel condition in Canadian CANDU 6 reactors

    International Nuclear Information System (INIS)

    The cornerstone of the CANDU concept is its natural uranium fuel, and the success of its reactor operation hinges on the fuel condition in the reactor. Neutron economy, on power refuelling, and simple fuel design are among the unique characteristics of CANDU fuel. In Canadian CANDU 6 reactors (Gentilly 2 and Point Lepreau), the 37-element fuel has provided an enviable record of safe, economic and reliable plant operation for 29 reactor years to date. The fuelling cost is among the lowest in the world - a corollary of high neutron economy, simple fuel design, and judicial fuelling scheme. The reliability of fuel is high: only 21 of the 60000 bundles discharged from Gentilly 2 were confirmed defective and the five-year period from March 1992 to February 1997 saw no defect at all at Gentilly-2. Also, thanks to the inherent on-power refuelling capability and an effective defect detection and removal system, the primary coolant loops are kept extremely clean (very low activity level) - benefiting both maintenance and safety. Moreover, the inventories of fission products in the core and in the channel are maintained within the safety analysis envelope, due to on-power fuelling and sophisticated fuel management. In this paper, CANDU 6 fuel performance is reviewed against the feedback from post-irradiation examinations, and the findings from our ongoing R and D program. The results suggest that the fuel behavior m reactor are basically as originally anticipated, despite an evolutionary 3% increase in bundle uranium mass in the 1980's. For operating conditions within the CANDU 6 37-element experience, the average strains are typically 0.09%; and fission gas release, 2.7%. The UO2 fuel remains stoichiometric after irradiation. In-core measurements of pressure tube fitting are generally low. All these observations are consistent with the excellent fuel performance statistics coming out of the two Canadian CANDU 6 reactors. Additionally, this paper will briefly discuss our

  2. Feasibility study of CANDU-9 fuel handling system

    International Nuclear Information System (INIS)

    CANDU's combination of natural uranium, heavy water and on-power refuelling is unique in its design and remarkable for reliable power production. In order to offer more output, better site utilization, shorter construction time, improved station layout, safety enhancements and better control panel layout, CANDU-9 is now under development with design improvement added to all proven CANDU advantages or applicable technologies. One of its major improvement has been applied to fuel handling system. This system is very similar to that of CANDU-3, and some parts of the system are applied to those of the existing CANDU-6 or CANDU-9. Design concepts and design requirements of fuel handling system for CANDU-9 have been identified to compare with those of the existing CANDU and the design feasibilities have been evaluated. (author). 4 tabs., 13 figs., 9 refs

  3. Thorium-Based Fuels Preliminary Lattice Cell Studies for Candu Reactors

    International Nuclear Information System (INIS)

    The choice of nuclear power as a major contributor to the future global energy needs must take into account acceptable risks of nuclear weapon proliferation, in addition to economic competitiveness, acceptable safety standards, and acceptable waste disposal options. Candu reactors offer a proven technology, safe and reliable reactor technology, with an interesting evolutionary potential for proliferation resistance, their versatility for various fuel cycles creating premises for a better utilization of global fuel resources. Candu reactors impressive degree of fuel cycle flexibility is a consequence of its channel design, excellent neutron economy, on-power refueling, and simple fuel bundle. These features facilitate the introduction and exploitation of various fuel cycles in Candu reactors in an evolutionary fashion. The main reasons for our interest in Thorium-based fuel cycles have been, globally, to extend the energy obtainable from natural Uranium and, locally, to provide a greater degree of energy self-reliance. Applying the once through Thorium (OTT) cycle in existing and advanced Candu reactors might be seen as an evaluative concept for the sustainable development both from the economic and waste management points of view. Two Candu fuel bundles project will be used for the proposed analysis, namely the Candu standard fuel bundle with 37 fuel elements and the CANFLEX fuel bundle with 43 fuel elements. Using the Canadian proposed scheme - loading mixed ThO2-SEU CANFLEX bundles in Candu 6 reactors - simulated at lattice cell level led to promising conclusions on operation at higher fuel burnups, reduction of the fissile content to the end of the cycle, minor actinide content reduction in the spent fuel, reduction of the spent fuel radiotoxicity, presence of radionuclides emitting strong gamma radiation for proliferation resistance benefit. The calculations were performed using the lattice codes WIMS and Dragon (together with the corresponding nuclear data

  4. Advanced CANDU reactors fuel analysis through optimal fuel management at approach to refuelling equilibrium

    International Nuclear Information System (INIS)

    The analysis of alternate CANDU fuels along with natural uranium-based fuel was carried out from the view point of optimal in-core fuel management at approach to refuelling equilibrium. The alternate fuels considered in the present work include thorium containing oxide mixtures (MOX), plutonium-based MOX, and Pressurised Water Reactor (PWR) spent fuel recycled in CANDU reactors (Direct Use of spent PWR fuel in CANDU (DUPIC)); these are compared with the usual natural UO2 fuel. The focus of the study is on the 'Approach to Refuelling Equilibrium' period which immediately follows the initial commissioning of the reactor. The in-core fuel management problem for this period is treated as an optimization problem in which the objective function is the refuelling frequency to be minimized by adjusting the following decision variables: the channel to be refuelled next, the time of the refuelling and the number of fresh fuel bundles to be inserted in the channel. Several constraints are also included in the optimisation problem which is solved using Perturbation Theory. Both the present 37-rod CANDU fuel bundle and the proposed CANFLEX bundle designs are part of this study. The results include the time to reach refuelling equilibrium from initial start-up of the reactor, the average discharge burnup, the average refuelling frequency and the average channel and bundle powers relative to natural UO2. The model was initially tested and the average discharge burnup for natural UO2 came within 2% of the industry accepted 199 MWh/kgHE. For this type of fuel, the optimization exercise predicted the savings of 43 bundles per full power year. In addition to producing average discharge burnups and other parameters for the advanced fuels investigated, the optimisation model also evidenced some problem areas like high power densities for fuels such as the DUPIC. Perturbation Theory has proven itself to be an accurate and valuable optimization tool in predicting the time between

  5. Spent fuel characteristics & disposal considerations

    Energy Technology Data Exchange (ETDEWEB)

    Oversby, V.M.

    1996-06-01

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

  6. A study on the direct use of spent PWR fuel in CANDU reactors -Development of DUPIC fuel on manufacturing and quality control technology-

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Myung Seung; Park, Hyun Soo; Lee, Yung Woo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    Oxidation/reduction process was established after analysis of the effect of process parameter on the sintering behavior using SIMFUEL. Process equipment was studied more detail and some of process equipment items were designed and procured. The chemical analysing method of fission products and fissile content in DUPIC fuel was studied and the behavior and the characteristics of fission products in fuel was also done. Requirement for irradiation in HANARO was analysed to prepare performance evaluation. 100 figs, 48 tabs, 170 refs. (Author).

  7. A study on manufacturing and quality control technology of DUPIC fuel -A study on the direct use of spent PWR fuel in CANDU reactors-

    International Nuclear Information System (INIS)

    As the first year of the experimental research for the manufacturing and irradiation of prototypic DUPIC nuclear fuel, UO2 pellets made from the natural uranium dioxide were used for the study of OREOX (oxidation/reduction of oxide fuel) process. The reference oxidation and reduction processes were established from the evaluation of the characteristics of produced powders and pellets. The manufacturing process and the layout of the manufacturing equipment were established in consideration of the high radioactivity of DUPIC process and the function of hot cells. The properties of materials to be used for DUPIC manufacturing were evaluated, and the punching machine for decladding was also developed. (Author)

  8. Feasible advanced fuel cycle options for CANDU reactors in the Republic of Korea

    International Nuclear Information System (INIS)

    Taking into account the view points on nuclear safety, nuclear waste, non-proliferation and economics from the public, international environment, and utilities, the SEU/RU and DUPIC fuel cycles would be feasible options of advanced fuel cycles for CANDU-PHWRs in the Republic of Korea in the mid- and long-terms, respectively. Comparing with NU fuel, 0.9 % or 1.2 % SEU fuel would increase fuel burnup and hence reduce the spent fuel arisings by a factor of 2 or 3, and also could reduce CANDU fuel cycle costs by 20 to 30%. RU offers similar benefits as 0.9% SEU and is very attractive due to the significantly improved fuel cycle economics, substantially increased burnups, large reduction in fuel requirements as well as in spent fuel arisings. For RU use in a CANDU reactor, re-enrichment is not required. There are 25,000 tes RU produced from reprocessing operations in Europe and Japan, which would theoretically provide sufficient fuel for 500 CANDU 6 reactor-years of operation. According to the physics, thermal-hydraulic and thermal-mechanical assessments of CANFLEX-0.9% RU fuel for a CANDU-6 reactor, the fuel could be introduced into the reactor in a straight-forward fashion. A series of assessments of CANFLEX-DUPIC physics on the compatibility of the fuel design in the existing CANDU 6 reactors has shown that the poisoning of the central element of DUPIC with, for example, natural dysprosium, reduces the void reactivity of the fuel, and that a 2 bundle shift refuelling scheme would be the most appropriate in-core fuel management scheme for a CANDU-6 reactor. The average discharge burnup is ∼15 MWd/kgHE. Although these results have shown promising results for the DUPIC fuel cycle, more in-depth studies are required in the areas of ROP system, large LOCA safety analyses, and so on. The recycling fuel cycles of RU and DUPIC for CANDU are expected to achieve the environmental 3R's (Reduce, Reuse, Recycle) as applied to global energy use in the short- and long

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

    International Nuclear Information System (INIS)

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

  10. Research and development for CANDU fuel channels and fuel

    International Nuclear Information System (INIS)

    The CANDU nuclear reactor is distinctly different from BWR and PWR reactors in that it uses many small pressure tubes rather than one large pressure vessel to contain the fuel and coolant. To exploit the advantages of the natural uranium fuel, the pressure tubes, like other core components, are manufactured from zirconium alloys which have low neutron capture cross sections. Also, because natural uranium fuel only achieves a modest burnup, a simple and inexpensive fuel design has been developed. The present paper reviews the features and the research that have led to the very satisfactory performance of the pressure tubes and the fuel in CANDU reactors. Reference is made to current research and development that may lead to further economies in the design and operation of future power reactors. (author)

  11. Development of nuclear fuel. Development of CANDU advanced fuel bundle

    International Nuclear Information System (INIS)

    In order to develop CANDU advanced fuel, the agreement of the joint research between KAERI and AECL was made on February 19, 1991. AECL conceptual design of CANFLEX bundle for Bruce reactors was analyzed and then the reference design and design drawing of the advanced fuel bundle with natural uranium fuel for CANDU-6 reactor were completed. The CANFLEX fuel cladding was preliminarily investigated. The fabricability of the advanced fuel bundle was investigated. The design and purchase of the machinery tools for the bundle fabrication for hydraulic scoping tests were performed. As a result of CANFLEX tube examination, the tubes were found to be meet the criteria proposed in the technical specification. The dummy bundles for hydraulic scoping tests have been fabricated by using the process and tools, where the process parameters and tools have been newly established. (Author)

  12. Inspection of Candu Nuclear Reactor Fuel Channels

    International Nuclear Information System (INIS)

    The Channel Inspection and Gauging Apparatus of Reactors (CIGAR) is a fully atomated, remotely operated inspection system designed to perform multi-channel, multi-task inspection of CANDU reactor fuel channels. Ultrasonic techniques are used for flaw detection, (with a sensitivity capable of detecting a 0.075 mm deep notch with a signal to noise ratio of 10 dB) and pressure tube wall thickness and diameter measurements. Eddy currrent systems are used to detect the presence of spacers between the coaxial pressure tube and calandria tube, as well as to measure their relative spacing. A servo-accelerometer is used to estimate the sag of the fuel channels. This advanced inspection system was commissioned and declared in service in September 1985. The paper describes the inspection systems themselves and discussed the results achieved to-date. (author)

  13. General overview of CANDU advanced fuel cycles program

    International Nuclear Information System (INIS)

    The R and D program for CANDU advanced fuel cycles may be roughly divided into two components which have a near-and long-term focus, respectively. The near-term focus is on the technology to implement improved once-through cycles and mixed oxide (plutonium-uranium oxides) recycle in CANDU and on technologies to separate zirconium isotopes. Included is work on those technologies which would allow a CANDU-LWR strategy to be developed in a growing nuclear power system. For the longer-term, activities are focused on those technologies and fuel cycles which would be appropriate in a period when nuclear fuel demand significantly exceeds mined uranium supplies. Fuel cycles and systems under study are thorium recycle, CANDU fast breeder systems and electro-nuclear fissile breeders. The paper will discuss the rationale underlying these activities, together with a brief description of activities currently under way in each of the fuel cycle technology areas

  14. Proceedings of the international conference on CANDU fuel

    International Nuclear Information System (INIS)

    These proceedings contain full texts of all paper presented at the first International Conference on CANDU Fuel. The Conference was organized and hosted by the Chalk River Branch of the Canadian Nuclear Society and utilized Atomic Energy of Canada Limited's facilities at Chalk River Nuclear Laboratories. Previously, informal Fuel Information Meetings were used in Canada to allow the exchange of information and technology associated with CANDU. The Chalk River conference was the first open international forum devoted solely to CANDU and included representatives of overseas countries with current or potential CANDU programs, as well as Canadian participants. The keynote presentation was given by Dr. J.B. Slater, who noted the correlation between past successes in CANDU fuel cycle technology and the co-operation between researchers, fabricators and reactor owner/operators in all phases of the fuel cycle, and outlined the challenges facing the industry today. In the banquet address, Dr. R.E. Green described the newly restructured AECL Research Company and its mission which blends traditional R and D with commercial initiatives. Since this forum for fuel technology has proven to be valuable, a second International CANDU Fuel Conference is planned for the fall of 1989, again sponsored by the Canadian Nuclear Society

  15. Spent nuclear fuel storage

    International Nuclear Information System (INIS)

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

  16. Basic research and industrialization of CANDU advanced fuel

    International Nuclear Information System (INIS)

    Wolsong Unit 1 as the first heavy water reactor in Korea has been in service for 17 years since 1983. It would be about the time to prepare a plan for the solution of problems due to aging of the reactor. The aging of CANDU reactor could lead especially to the steam generator cruding and pressure tube sagging and creep and then decreases the operation margin to make some problems on reactor operations and safety. The counterplan could be made in two ways. One is to repair or modify reactor itself. The other is to develop new advanced fuel to increase of CANDU operation margin effectively, so as to compensate the reduced operation margin. Therefore, the first objectives in the present R and D is to develop the CANFLEX-NU (CANDU Flexible fuelling-Natural Uranium) fuel as a CANDU advanced fuel. The second objectives is to develop CANDU advanced fuel bundle to utilize advanced fuel cycles such as recovered uranium, slightly enriched uranium, etc. and so to raise adaptability for change in situation of uranium market. Also, it is to develop CANDU advanced fuel technology which improve uranium utilization to cope with a world-wide imbalance between uranium supply and demand, without significant modification of nuclear reactor design and refuelling strategies. As the implementations to achieve the above R and D goal, the work contents and scope of technology development of CANDU advanced fuel using natural uranium (CANFLEX-NU) are the fuel element/bundle designs, the nuclear design and fuel management analysis, the thermalhydraulic analysis, the safety analysis, fuel fabrication technologies, the out-pile thermalhydraulic test and in-pile irradiation tests performed. At the next, the work scopes and contents of feasibility study of CANDU advanced fuel using recycled uranium (CANFLEX-RU) are the fuel element/bundle designs, the reactor physics analysis, the thermalhydraulic analysis, the basic safety analysis of a CANDU-6 reactor with CANFLEX-RU fuel, the fabrication and

  17. Basic research and industrialization of CANDU advanced fuel

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Suk Ho; Park, Joo Hwan; Jun, Ji Su [and others

    2000-04-01

    Wolsong Unit 1 as the first heavy water reactor in Korea has been in service for 17 years since 1983. It would be about the time to prepare a plan for the solution of problems due to aging of the reactor. The aging of CANDU reactor could lead especially to the steam generator cruding and pressure tube sagging and creep and then decreases the operation margin to make some problems on reactor operations and safety. The counterplan could be made in two ways. One is to repair or modify reactor itself. The other is to develop new advanced fuel to increase of CANDU operation margin effectively, so as to compensate the reduced operation margin. Therefore, the first objectives in the present R and D is to develop the CANFLEX-NU (CANDU Flexible fuelling-Natural Uranium) fuel as a CANDU advanced fuel. The second objectives is to develop CANDU advanced fuel bundle to utilize advanced fuel cycles such as recovered uranium, slightly enriched uranium, etc. and so to raise adaptability for change in situation of uranium market. Also, it is to develop CANDU advanced fuel technology which improve uranium utilization to cope with a world-wide imbalance between uranium supply and demand, without significant modification of nuclear reactor design and refuelling strategies. As the implementations to achieve the above R and D goal, the work contents and scope of technology development of CANDU advanced fuel using natural uranium (CANFLEX-NU) are the fuel element/bundle designs, the nuclear design and fuel management analysis, the thermalhydraulic analysis, the safety analysis, fuel fabrication technologies, the out-pile thermalhydraulic test and in-pile irradiation tests performed. At the next, the work scopes and contents of feasibility study of CANDU advanced fuel using recycled uranium (CANFLEX-RU) are the fuel element/bundle designs, the reactor physics analysis, the thermalhydraulic analysis, the basic safety analysis of a CANDU-6 reactor with CANFLEX-RU fuel, the fabrication and

  18. CANDU advanced fuel R and D programs for 1997 - 2006 in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Suk, H.C.; Yang, M.S.; Sim, K-S.; Yoo, K.J. [Korea Atomic Energy Research Inst., Yusong, Taejon (Korea, Republic of)

    1997-07-01

    KAERI has a comprehensive product development program of CANFLEX and DUPIC fuels to introduce them into CANDU reactors in Korea and a clear vision of how the product will evolve over the next 10 years. CANDU reactors are not the majority of nuclear power plants in Korea, but they produce significant electricity to contribute Korea's economic growth as well as to satisfy the need for energy. The key targets of the development program are safety enhancement, reduction of spent fuel volume, and economic improvements, using the inherent characteristics and advantages of CANDU technology The CANFLEX and DUPIC R and D programs are conducted currently under the second stage of Korea's Nuclear Energy R and D Project as a national mid- and long-term program over the next 10 years from 1997 to 2006. The specific activities of the programs have taken account of the domestic and international environment concerning on non-proliferation in the Peninsula of Korea. As the first of the development products in the short-term, the CANFLEX-NU fuel will be completely developed jointly by KAERI/AECL and will be useful for the older CANDU-6 Wolsong unit 1. As the second product, the CANFLEX-0.9 % equivalent SEU fuel is expected to be completely developed within the next decade. It will be used in CANDU-6 reactors in Korea immediately after the development, if the existing RU in the world is price competitive with natural uranium. The DUPIC R and D program, as a long term program, is expected to demonstrate the possibility of use of used PWR fuel in CANDU reactors in Korea during the next 10 years. The pilot scale fabrication facility would be completed around 2010. (author)

  19. Spent fuel assembly hardware

    International Nuclear Information System (INIS)

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

  20. Spent fuel storage rack

    International Nuclear Information System (INIS)

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

  1. Conference proceedings of the 4. international conference on CANDU fuel. V. 1,2

    International Nuclear Information System (INIS)

    These proceedings contain the full texts of all 65 papers presented at the 4th International Conference on CANDU fuel. As such, they represent an update on the state-of-the-art in such important CANDU fuel topics as International Development Programs and Operating Experience with CANDU fuel, Performance Assessments and Fuel Behavior Modeling, Fuel Properties, Licensing and Accident Analyses for CANDU fuel, Design, Testing and Manufacturing, and Advanced Fuel Designs. The large number of papers required the use of parallel sessions for the first time at a CANDU Fuel Conference

  2. CANDU fuel quality and how it is achieved

    International Nuclear Information System (INIS)

    In this three part presentation CANDU fuel quality is reviewed from the point of view of a designer/operator and a fabricator. In Part 'A' fuel performance and quality considerations are discussed from the point of view of a designer-operator. In Parts 'B' and 'C' fuel quality is reviewed from the point of view of a fabricator. The presentation was divided in this way to convey the 'team effort' attitude which exists in the Canadian program; the team effort which is an essential part of the CANDU story. (auth)

  3. Fuel management simulation for CANFLEX-RU in CANDU 6

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Joon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    Fuel management simulations have been performed for CANFLEX-09% RU fuel in the CANDU 6 reactor. In this study, the bi-directional 4-bundle shift fuelling scheme was assumed. The lattice cell and time-average calculation were carried out. The refuelling simulation calculations were performed for 600 full power days. Time-averaged results show good axial power profile with the CANFLEX-RU fuel. During the simulation period, the maximum channel and bundle power were maintained below the licensing limit of CANDU 6 reactor. 7 refs., 4 figs. (Author)

  4. Spent fuel reprocessing options

    International Nuclear Information System (INIS)

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

  5. Spent Fuel in Chile

    International Nuclear Information System (INIS)

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

  6. CANDU type fuel behavior evaluation - a probabilistic approach

    International Nuclear Information System (INIS)

    In order to realistically assess the behavior of the fuel elements during in-reactor operation, probabilistic methods have recently been introduced in the analysis of fuel performance. The present paper summarizes the achievements in this field at the Institute for Nuclear Research (INR), pointing out some advantages of the utilized method in the evaluation of CANDU type fuel behavior in steady state conditions. The Response Surface Method (RSM) has been selected for the investigation of the effects of the variability in fuel element computer code inputs on the code outputs (fuel element performance parameters). A new developed version of the probabilistic code APMESRA based on RSM is briefly presented. The examples of application include the analysis of the results of an in-reactor fuel element experiment and the investigation of the calculated performance parameter distribution for a new CANDU type extended burnup fuel element design. (author)

  7. Spent Fuel Management of NPPs in Argentina

    International Nuclear Information System (INIS)

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

  8. Spent fuel workshop'2002

    Energy Technology Data Exchange (ETDEWEB)

    Poinssot, Ch

    2002-07-01

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

  9. Modelling nuclear fuel vibrations in horizontal CANDU reactors

    International Nuclear Information System (INIS)

    Flow-induced fuel vibrations in the pressure tubes of CANDU reactors are of vital interest to designers because fretting damage may result. Computer simulation is being used to study how bundles vibrate and to identify bundle design features which will reduce vibration and hence fretting. (author)

  10. Spent fuel management in Argentina

    International Nuclear Information System (INIS)

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

  11. Temperature effect of DUPIC fuel in CANDU reactor

    International Nuclear Information System (INIS)

    The fuel temperature coefficient (FTC) of DUPIC fuel was calculated by WIMS-AECL with ENDF/B-V cross-section library. Compared to natural uranium CANDU fuel, the FTC of DUPIC fuel is less negative when fresh and is positive after 10,000 MWD/T of irradiation. The effect of FTC on the DUPIC core performance was analyzed using the pace-time kinetics module in RFSP for the refueling transient which occurs daily during normal operation of CANDU reactors. In this study, the motion of zoen controller units (ZCU) was modeled externally to describe the reactivity control during the refueling transient. Refueling operation was modeled as a linear function of time by changing the fuel burnup incrementally and the average fuel temperature was calculated based on the bundle power during the transient. The analysis showed that the core-wide FTC is negative and local positive FTC of the DUPIC fuel can be accommodated in the CANDU reactor because the FTC is very small, the refueling operation occurs slowly, and the channel-front-peaked axial power profile weakens the contribution of the positive FTC. (author). 11 refs., 31 tabs., 10 figs

  12. Development of defueling device for CANDU fuel channel (modeling)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. Y.; Yu, K. H.; Yang, J. S.; Lee, H. S.; Chang, K. J.; Kim, Y. J. [CNEC Technical Office, Taejon (Korea, Republic of); Lee, S. K. [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    2004-07-01

    Commercial CANDU reactors use D{sub 2}O for moderator and heat transfer material and also have Fueling Machines(F/M) and related system equipment in order to assist on-power refueling operation. A Defuelling Device(DFD) is developed for the proper defuelling of all fuels in all fuel channels during shutdown condition of plant. This device is considered more efficient in defuelling compared to the existing Fuel Grapple System for its use of existing D{sub 2}O flow in the fuel channel. In this study, computational fluid dynamic software is used for optimize and evaluation of the design for its applicability.

  13. New developments in dry spent fuel storage

    International Nuclear Information System (INIS)

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

  14. Used CANDU fuel waste consumed and eliminated: environmentally responsible, economically sound, energetically enormous

    International Nuclear Information System (INIS)

    The 43,800 tonnes of currently stored CANDU nuclear fuel waste can all be consumed in fast-neutron reactors (FNRs) to reduce its long-term radioactive burden 100,000 times while extracting about 130 times more nuclear energy than the prodigious amounts that have already been gained from the fuel in CANDU reactors. The cost of processing CANDU fuel for use in FNRs plus the cost of recycling the FNR fuel is about 2.5 times less on a per kWh energy basis than the currently projected cost of disposal of 3.6 million used CANDU fuel bundles in a deep geological repository. (author)

  15. Third international conference on CANDU fuel

    International Nuclear Information System (INIS)

    These proceedings contain full texts of all 49 papers from the ten sessions and the banquet address. The sessions were on the following subjects: International experience and programs; Fuel behaviour and operating experience; Fuel modelling; Fuel design; Advanced fuel and fuel cycle technology; AECL's concept for the disposal of nuclear fuel waste. The individual papers have been abstracted separately

  16. Spent-fuel-storage alternatives

    International Nuclear Information System (INIS)

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

  17. Spent Fuel Management in Bulgaria

    International Nuclear Information System (INIS)

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

  18. Development of the advanced CANDU technology -Development of CANDU advanced fuel fabrication technology-

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Chang Bum; Park, Choon Hoh; Park, Chul Joo; Kwon, Woo Joo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This project is carrying out jointly with AECL to develop CANFLEX fuel which can enhance reactor safety, fuel economy and can be used with various fuel cycles (natural U, slightly enriched U, other advanced fuel). The final goal of this research is to load the CANFLEX fuel in commercial CANDU reactor for demonstration irradiation. The annual portion of research activities performed during this year are followings ; The detail design of CANFLEX-NU fuel was determined. Based on this design, various fabrication drawings and process specifications were revised. The seventeen CANFLEX-NU fuel bundles for reactivity test in ZED-2 and out-pile test, two CANFLEX-SEU fuel bundles for demo-irradiation in NRU were fabricated. Advanced tack welding machine was designed and sequence control software of automatic assembly welder was developed. The basic researches related to fabrication processes, such as weld evaluation by ECT, effect of additives in UO{sub 2}, thermal stabilities of Zr based metallic glasses, were curried out. 51 figs, 22 tabs, 42 refs. (Author).

  19. CANDU 6 fuel channel stress analysis using ANSYS fatigue module

    International Nuclear Information System (INIS)

    Design reliability can be confirmed by the stress analysis, and its results become the basis of the structural integrity for components. The report presents the development of CANDU 6 fuel channel stress analysis methodology and procedure per ASME Code using the ANSYS fatigue module. Stress analysis was performed in accordance with the procedure developed on the basis of ASME Code Section III NB-3200. FORTRAN programs and ANSYS macros used in data processing were developed to systematized the analysis. Stresses were separately analyzed for mechanical and thermal load respectively, and then combined in the post-processing stage for the various conditions. Maximum stress intensity range was then calculated at selected nodes by using the ANSYS fatigue module for the sum of mechanical and thermal stress values. As a results, structural integrity of CANDU 6 fuel channel was proved in this report and analysis reliability for CANDU reactor was shown to be enhanced by the establishment of analysis procedure bases upon ASME Code. (Author) 11 refs., 11 figs., 6 tabs

  20. Core analysis during transition from 37-element fuel to CANFLEX-NU fuel in CANDU 6

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Joon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    An 1200-day time-dependent fuel-management for the transition from 37-element fuel to CANFLEX-NU fuel in a CANDU 6 reactor has been simulated to show the compatibility of the CANFLEX-NU fuel with the reactor operation. The simulation calculations were carried out with the RFSP code, provided by cell averaged fuel properties obtained from the POWDERPUFS-V code. The refueling scheme for both fuels was an eight bundle shift at a time. The simulation results show that the maximum channel and bundle powers were maintained below the license limit of the CANDU 6. This indicates that the CANFLEX-NU fuel bundle is compatible with the CANDU 6 reactor operation during the transition period. 3 refs., 2 figs., 1 tab. (Author)

  1. Public health risks associated with the CANDU nuclear fuel cycle

    International Nuclear Information System (INIS)

    This report analyzes in a preliminary way the risks to the public posed by the CANDU nuclear fuel cycle. Part 1 considers radiological risks, while part 2 (published as INFO-0141-2) evaluates non-radiological risks. The report concludes that, for radiological risks, maximum individual risks to members of the public are less than 10-5 per year for postulated accidents, are less than 1 percent of regulatory limits for normal operation and that collective doses are small, less than 3 person-sieverts. It is also concluded that radiological risks are much smaller than the non-radiological risks posed by activities of the nuclear fuel cycle

  2. Cost analysis and economic comparison for alternative fuel cycles in the heavy water cooled canadian reactor (CANDU)

    International Nuclear Information System (INIS)

    Three main options in a CANDU fuel cycle involve use of: (1) natural uranium (0.711 weight percent U-235) fuel, (2) slightly enriched uranium (1.2 weight percent U-235) fuel, and (3) recovered uranium (0.83 weight percent U-235) fuel from light water reactor spent fuel. ORIGEN-2 computer code was used to identify composition of the spent fuel for each option, including the standard LWR fuel (3.3 weight percent U-235). Uranium and plutonium credit calculations were performed using ORIGEN-2 output. WIMSD-5 computer code was used to determine maximum discharge burnup values for each case. For the 3 cycles selected (natural uranium, slightly enriched uranium, recovered uranium), levelized fuel cycle cost calculations are performed over the reactor lifetime of 40 years, using unit process costs obtained from literature. Components of the fuel cycle costs are U purchase, conversion, enrichment, fabrication, SF storage, SF disposal, and reprocessing where applicable. Cost parameters whose effects on the fuel cycle cost are to be investigated are escalation ratio, discount rate and SF storage time. Cost estimations were carried out using specially developed computer programs. Share of each cost component on the total cost was determined and sensitivity analysis was performed in order to show how a change in a main cost component affects the fuel cycle cost. The main objective of this study has been to find out the most economical option for CANDU fuel cycle by changing unit prices and cost parameters

  3. Fuel Temperature Characteristics for Fuel Channels using Burnable Poison in the CANDU reactor

    International Nuclear Information System (INIS)

    Although the CANFLEX RU fuel bundle loaded 11.0 wt% Er2O3 are originally designed focused on the safety characteristics, the fuel temperature characteristics is revealed to be not deteriorated but rather is slightly enhanced by the decreased fuel temperature in the outer ring compared with that of standard 37 fuel bundle. Recently, for an equilibrium CANDU core, the power coefficient was reported to be slightly positive when newly developed Industry Standard Tool set reactor physics codes were used. Therefore, it is required to find a new way to effectively decrease the positive power coefficient of CANDU reactor without seriously compromising the economy. In order to make the power coefficient of the CANDU reactor negative at the operating power, Roh et al. have evaluated the various burnable poison (BP) materials and its loading scheme in terms of the fuel performance and reactor safety characteristics. It was shown that reactor safety characteristics can be greatly improved by the use of the BP in the CANDU reactor. In a view of safety, the fuel temperature coefficient (FTC) is an important safety parameter and it is dependent on the fuel temperature. For an accurate evaluation of the safety-related physics parameters including FTC, the fuel temperature distribution and its correlation with the coolant temperature should be accurately identified. Therefore, we have evaluated the fuel temperature distribution of a CANFLEX fuel bundle loaded with a burnable poison and compared the standard 37 element fuel bundle and CANFELX-NU fuel bundle

  4. Management of Spent Fuel in Germany

    International Nuclear Information System (INIS)

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

  5. Disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

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

  6. Disposal of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    1979-12-01

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

  7. Safe, permanent disposal of used CANDU fuel

    International Nuclear Information System (INIS)

    AECL's assessment of nuclear fuel waste disposal deep in plutonic rock of the Canadian Precambrian Shield is now well advanced. A comprehensive understanding has evolved of the chemical and physical processes controlling the containment of radionuclides in used fuel. The following conclusions have been reached: containers with outer shells of titanium and copper can be expected to isolate used fuel from contact with groundwater for at least 500 years, the period during which the hazard is greatest; uranium oxide fuel can be expected to dissolve at a rate less than 10-8 per day, resulting in uranium concentrations less that 1 μg/L, which is consistent with observations of uranium oxide deposits in the earth's crust; movement of dissolved radionuclides away from the containers can be delayed for thousands of years by placing a compacted bentonite-clay layer between the container and the rock mass; and, the granite plutons of interest consist of relatively large rock volumes of low permeability separated by relatively thin fracture zones, and the low permeability volumes are sufficiently large to accommodate a vault design that will ensure radionuclides do not reach the surface in unacceptable concentrations

  8. Assessment of spent fuel cooling

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-02-01

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

  9. 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)

  10. Procurement and supply of CANDU fuels

    International Nuclear Information System (INIS)

    In 1955 a decision was made to proceed with construction of a Nuclear Power Demonstration Station (NPD) near Rolfton, Ontario. This project, headed by Atomic Energy of Canada with major involvement of private industry, was the genesis for the development of nuclear electric generation in Canada. This paper reviews one aspect of the Canadian program: the evolution of fuel procurement and supply, which in itself has been a remarkable Canadian achievement. (author)

  11. A model for fission product distribution in CANDU fuel

    International Nuclear Information System (INIS)

    This paper describes a model to estimate the distribution of active fission products among the UO2 grains, grain-boundaries, and the free void spaces in CANDU fuel elements during normal operation. This distribution is required for the calculation of the potential release of activity from failed fuel sheaths during a loss-of-coolant accident. The activity residing in the free spaces (''free'' inventory) is available for release upon sheath rupture, whereas relatively high fuel temperatures and/or thermal shock are required to release the activity in the grain boundaries or grains. A preliminary comparison of the model with the data from in-reactor sweep-gas experiments performed in Canada yields generally good agreement, with overprediction rather than under prediction of radiologically important isotopes, such as I131. The model also appears to generally agree with the ''free'' inventory release calculated using ANS-5.4. (author)

  12. Development of the CANDU high-burnup fuel design/analysis technology

    International Nuclear Information System (INIS)

    This report contains all the information related to the development of the CANDU advanced fuel, so-called CANFLEX-NU, which is composed of 43 elements with natural uranium fuel. Also, it contains the compatibility study of CANFLEX-RU which is considered as a CANDU high burnup fuel. This report describes the mechanical design, thermalhydraulic and safety evaluations of CANFLEX fuel bundle. (author). 38 refs., 24 tabs., 74 figs

  13. Joint studies on large CANDU

    International Nuclear Information System (INIS)

    from economic, safety and strategic viewpoints. A large number of research and development programs are now in place at AECL and KAERI that will permit substantial improvements to be realized in the next generation of CANDU okabts, Furthermore, opportunities exist for engineered improvements based on the research and development in advancing the generic CANDU Technology. Final Large CANDU joint study report with technical deliverables will be issued 1994 October. Phase 2 R and D program of the joint studies will be determined this year and implemented in next year. CANDU neutron economy permits versatility in choices of fuel cycles. This allows a utility to choose fuel cycle options for lower fuelling cost, better security of supply, and ultimately for much lower spent-fuel volume, than with PWR's alone. To meet Korea's strategic requirements, CANDU should be an integral part of the electricity supply mix.

  14. Quality control of CANDU6 fuel element in fabrication process

    International Nuclear Information System (INIS)

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

  15. Iraq spent fuel removal program

    International Nuclear Information System (INIS)

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

  16. Intermodal transportation of spent fuel

    International Nuclear Information System (INIS)

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

  17. Spent fuel management in Argentina

    International Nuclear Information System (INIS)

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

  18. Active Interrogation for Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-05

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

  19. Transportation of spent MTR fuels

    Energy Technology Data Exchange (ETDEWEB)

    Raisonnier, D.

    1997-08-01

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

  20. HFIR spent fuel management alternatives

    International Nuclear Information System (INIS)

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

  1. Development of spent fuel remote handling technology

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-01

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

  2. Development of spent fuel remote handling technology

    International Nuclear Information System (INIS)

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

  3. Intercomparison of safety evaluations for CANDU or LWR burned fuel disposed in a salt massive

    International Nuclear Information System (INIS)

    A safety analysis of a generic vault, located into a salt massive, was carried out for for spent fuel from CANDU or LWR NPPs. Three scenarios were considered for the evolution of the system: - sub-erosion, as normal evolution, - a combined process of water intrusion through an anhydrous vein and from brine bags (remaining undetected in the vicinity of the repository areas), - human intrusion. As key parameter in evaluation of long-term repository's safety the biosphere exposure (dose) was chosen. For the first scenario considered the maximum dose, due mainly to U-234, was found below the German standard value of 3 x 10-4 Sv/y. The effects of sub-erosion rate and salt concentration in ground water on the maximum dose were calculated and found rather serious. Although, having in view the rather excessive conservative assumptions adopted (the barrier effect of the geosphere was neglected) more conclusive results should be based upon a more realistic approach of the issue. In the case of human intrusion scenario the maximum exposure would be 8 x 10-5 Sv/y for CANDU fuel and 1.3 x 10-4 Sv/y for LWR fuel, as due mainly to Np-237. The analyses of local sensitivity carried out to investigate the influence of input parameters upon the release in geosphere took into consideration four parameters: a. the solubility limits; b. the diffusion coefficients; c. reference convergence rate, Kr; d. the maximum brine pressure pmax. Due to the low probability of the human intrusion scenario the effects appear to be acceptable. For the case of combined scenario the maximum doses,were found to be 9 x 10-6 Sv/y and 1 x 10-7 Sv/y for CANDU and LWR, respectively, mainly due to I-129 and Ra-225, and to I-129 and Cs-135, respectively. The effects of brine bags upon the temperature in the repository and the radiological consequences are presented for the two types of spent fuels

  4. Intermodal transfer of spent fuel

    International Nuclear Information System (INIS)

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

  5. CANDU fuel research and development in Korea: current status and future prospects

    International Nuclear Information System (INIS)

    The current status and future prospect of CANDU fuel R and D in Korea is always subjected to the consideration of domestic and international environments concerning nuclear safety, nuclear waste, nonproliferation and economy in favor of the arguments from public acceptance, international environments, and utilities. Considering that, at the end of 2000, the procurement of additional CANDU units at the Shin Wolsong site was decided not to proceed, the current and future CANDU fuel R and D would be oriented to the safety and economy of fuel and reactor operations rather than the national strategy of nuclear fuel cycle and reactor programs in Korea. Therefore, the current CANDU advanced fuel R and D programs such as CANFLEX-NU fuel industrialization, CANFLEX-0.9% SEU/RU fuel R and D, and DUPIC fuel cycle development in a laboratory-scale will be continued for the time being as it was. But the R and D of CANDU innovative fuels such as CANFLEX-1.2% ∼ 1.5 % SEU fuel, thorium oxide fuel and DUPIC fuel would have some difficulties to continue in the mid- and long-term if they would not have the justifications in the points of the nonproliferation, economic and safety views of fuel, fuel cycle and reactor. (author)

  6. Current status and future prospect of Candu fuel research and development in Korea

    International Nuclear Information System (INIS)

    The current status and future prospect of CANDU fuel R and D in Korea is always subjected to the consideration of domestic and international environments concerning nuclear safety, nuclear waste, non-proliferation and economy in favor of the arguments from public acceptance, international environments, and utilities. Considering that, at the end of 2000, the procurement of additional CANDU units at the Shin Wolsong site was decided not to proceed, the current and future CANDU fuel R and D would be oriented to the safety and economy of fuel and reactor operations rather than the national strategy of nuclear fuel cycle and reactor programs in Korea. Therefore, the current CANDU advanced fuel R and D programs such as CANFLEX-NU fuel industrialization, CANFLEX-0.9% SEU/RU fuel R and D, and DUPIC fuel cycle development in a laboratory-scale will be continued for the time being as it was. But the R and D of CANDU innovative fuels such as CANFLEX- 1.2% ∼ 1.5 % SEU fuel, thorium oxide fuel and DUPIC fuel would have some difficulties to continue in the mid-and long-term if they would not have the justifications in the points of the non-proliferation, economic and safety views of fuel, fuel cycle and reactor. (author)

  7. Spent fuel management in Argentina

    International Nuclear Information System (INIS)

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

  8. Subchannel analysis of CANDU 37-element fuel bundles

    International Nuclear Information System (INIS)

    The subchannel analysis codes COBRA-IV and ASSERT-4 have been used to predict the mass and enthalpy imbalance within a CANDU 37-element fuel channel under various system conditions. The objective of this study was to assess the various capabilities of the ASSERT code and highlight areas where further validation or development may be needed. The investigation indicated that the ASSERT code has all the basic models required to accurately predict the flow and enthalpy imbalance for complex rod bundles. The study also showed that the code modelling of void drift and diffusion requires refinement to some coefficients and that further validation is needed at high flow rate and high void fraction conditions, where ASSERT and COBRA are shown to predict significantly different trends. The results of a recent refinement of ASSERT modelling are also discussed

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

    International Nuclear Information System (INIS)

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

  10. Spent fuel storage

    International Nuclear Information System (INIS)

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

  11. GNS spent fuel cask experience

    International Nuclear Information System (INIS)

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

  12. GNS spent fuel cask experience

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-05-01

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

  13. Creep and shrinkage analysis for concrete spent fuel dry storage module

    International Nuclear Information System (INIS)

    CANDU reactors are designed in Canada and are built and operated worldwide to produce electricity economically with no emission of green house gases. This paper presents creep and shrinkage analysis for a concrete spent fuel dry storage module of a CANDU nuclear power plant. Creep and shrinkage analysis was performed using a method outlined in American Concrete Institute (ACI) code, and then the creep and shrinkage strains were analyzed in a finite element model to obtain the structural behavior of the concrete module. This demonstrated that the creep and shrinkage analysis for concrete spent fuel dry storage is reasonable. AECL's spent fuel dry storage module is adequate to resist the time-dependent effects due to creep and shrinkage of concrete. (author)

  14. Creep and shrinkage analysis for concrete spent fuel dry storage module

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, D. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)], E-mail: zhangd@aecl.ca

    2009-07-01

    CANDU reactors are designed in Canada and are built and operated worldwide to produce electricity economically with no emission of green house gases. This paper presents creep and shrinkage analysis for a concrete spent fuel dry storage module of a CANDU nuclear power plant. Creep and shrinkage analysis was performed using a method outlined in American Concrete Institute (ACI) code, and then the creep and shrinkage strains were analyzed in a finite element model to obtain the structural behavior of the concrete module. This demonstrated that the creep and shrinkage analysis for concrete spent fuel dry storage is reasonable. AECL's spent fuel dry storage module is adequate to resist the time-dependent effects due to creep and shrinkage of concrete. (author)

  15. Thermalhydraulic analysis of spent fuel baskets

    International Nuclear Information System (INIS)

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

  16. Dry storage of irradiated CANDU fuel at Pickering NGS

    International Nuclear Information System (INIS)

    Ontario Hydro generates about 86 million MW-h/year from its 20 nuclear CANDU reactors. The combination of a large generating capacity and relatively low fuel burn-up means that Ontario Hydro must manage very large volume of its used fuel. Irradiated fuel bays at Pickering NGS will be full by mid 1995. Additional storage capacity will be required by this date for the station to continue operation. Several long term storage options to supplement existing on-site facilities were studied. The dry storage system, based on the modular storage container, as an option, was found to be economical and operationally simple. The dry storage facility at Pickering NGS is planned to provide additional on-site storage capacity for fuel generated from 1994 to the end of the station's operating life (year 2025). This paper describes the design and operation of the Dry Storage System at Ontario Hydro's Pickering Nuclear Power Generating Station. The facility is planned as a two phased project. Phase I will provide a storage space for 700 dry storage containers (268,800 fuel bundles) or about 12 years of station's operation. Phase II will have the capacity for additional 800 containers (307,200 fuel bundles) and will provide storage until station decommissioning in 2025. Seventy containers will be required annually to meet storage requirements of the station's operation at 100% capacity factor. Staff of six persons will be required to operate the facility. Normal operation includes activities such as receiving and commissioning new containers, loading them with 4 modules of used fuel in the bays, draining and drying the cavity, decontaminating the container surface and lid welding. A helium leak test is performed before the container is placed in the storage. (author) 6 refs., 6 tabs., 7 figs

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

    CERN Document Server

    Jonkmans, G; Jewett, C; Thompson, M

    2012-01-01

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

  18. The Study of Nuclear Fuel Cycle Options Based On PWR and CANDU Reactors

    International Nuclear Information System (INIS)

    The study of nuclear fuel cycle options based on PWR and CANDU type reactors have been carried out. There are 5 cycle options based on PWR and CANDU reactors, i.e.: PWR-OT, PWR-OT, PWR-MOX, CANDU-OT, DUPIC, and PWR-CANDU-OT options. While parameters which assessed in this study are fuel requirement, generating waste and plutonium from each cycle options. From the study found that the amount of fuel in the DUPIC option needs relatively small compared the other options. From the view of total radioactive waste generated from the cycles, PWR-MOX generate the smallest amount of waste, but produce twice of high level waste than DUPIC option. For total plutonium generated from the cycle, PWR-MOX option generates smallest quantity, but for fissile plutonium, DUPIC options produce the smallest one. It means that the DUPIC option has some benefits in plutonium consumption aspects. (author)

  19. Fuel bundle geometry and composition influence on coolant void reactivity reduction in ACR and CANDU reactors

    International Nuclear Information System (INIS)

    It is very well known that the CANDU reactor has positive Coolant Void Reactivity (CVR), which is most important criticisms about CANDU. The most recent innovations based on using a thin absorbent Hafnium shell in the central bundle element were successfully been applied to the Advanced CANDU Reactor (ACR) project. The paper's objective is to analyze elementary lattice cell effects in applying such methods to reduce the CVR. Three basic fuel designs in their corresponding geometries were chosen to be compared: the ACR-1000TM, the RU-43 (developed in INR Pitesti) and the standard CANDU fuel. The bundle geometry influence on void effect was also evaluated. The WIMS calculations proved the Hafnium absorber suitability (in the latest 'shell design') to achieve the negative CVR target with great accuracy for the ACR-1000 fuel bundle design than for the other two projects. (authors)

  20. Overview of the CANDU fuel handling system for advanced fuel cycles

    International Nuclear Information System (INIS)

    Because of its neutron economies and on-power re-fuelling capabilities the CANDU system is ideally suited for implementing advanced fuel cycles because it can be adapted to burn these alternative fuels without major changes to the reactor. The fuel handling system is adaptable to implement advanced fuel cycles with some minor changes. Each individual advanced fuel cycle imposes some new set of special requirements on the fuel handling system that is different from the requirements usually encountered in handling the traditional natural uranium fuel. These changes are minor from an overall plant point of view but will require some interesting design and operating changes to the fuel handling system. Some preliminary conceptual design has been done on the fuel handling system in support of these fuel cycles. Some fuel handling details were studies in depth for some of the advanced fuel cycles. This paper provides an overview of the concepts and design challenges. (author)

  1. Dry spent nuclear fuel transfer

    International Nuclear Information System (INIS)

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

  2. Thermalhydraulic characteristics for fuel channels using burnable poison in the CANDU reactor

    International Nuclear Information System (INIS)

    The power coefficient is one of the most important physics parameters governing nuclear reactor safety and operational stability, and its sign and magnitude have a significant effect on the safety and control characteristics of the power reactor. Recently, for an equilibrium CANDU core, the power coefficient was reported to be slightly positive when newly developed Industry Standard Tool set reactor physics codes were used. Therefore, it is required to find a new way to effectively decrease the positive power coefficient of CANDU reactor without seriously compromising the economy. In order to make the power coefficient of the CANDU reactor negative at the operating power, Roh et al. have evaluated the various burnable poison (BP) materials and its loading scheme in terms of the fuel performance and reactor safety characteristics. It was shown that reactor safety characteristics can be greatly improved by the use of the BP in the CANDU reactor. However, the previous study has mainly focused on the safety characteristics by evaluating the power coefficient for the fuel channel using BP in the CANDU reactor. Together with the safety characteristics, the economic performance is also important in order to apply the newly designed fuel channel to the power plant. In this study, the economic performance has been evaluated by analyzing the thermal hydraulic characteristics for the fuel channel using BP in the CANDU reactor

  3. Spent fuel storage racks

    International Nuclear Information System (INIS)

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

  4. Return of spent TRIGA fuel

    International Nuclear Information System (INIS)

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

  5. CANDU RU fuel manufacturing basic technology development and advanced fuel verification tests

    International Nuclear Information System (INIS)

    A PHWR advanced fuel named the CANFLEX fuel has been developed through a KAERI/AECL joint Program. The KAERI made fuel bundle was tested at the KAERI Hot Test Loop for the performance verification of the bundle design. The major test activities were the fuel bundle cross-flow test, the endurance fretting/vibration test, the freon CHF test, and the fuel bundle heat-up test. KAERI also has developing a more advanced PHWR fuel, the CANFLEX-RU fuel, using recovered uranium to extend fuel burn-up in the CANDU reactors. For the purpose of proving safety of the RU handling techniques and appraising feasibility of the CANFLEX-RU fuel fabrication in near future, a physical, chemical and radiological characterization of the RU powder and pellets was performed. (author). 54 refs., 46 tabs., 62 figs

  6. CANDU RU fuel manufacturing basic technology development and advanced fuel verification tests

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Chang Hwan; Chang, S.K.; Hong, S.D. [and others

    1999-04-01

    A PHWR advanced fuel named the CANFLEX fuel has been developed through a KAERI/AECL joint Program. The KAERI made fuel bundle was tested at the KAERI Hot Test Loop for the performance verification of the bundle design. The major test activities were the fuel bundle cross-flow test, the endurance fretting/vibration test, the freon CHF test, and the fuel bundle heat-up test. KAERI also has developing a more advanced PHWR fuel, the CANFLEX-RU fuel, using recovered uranium to extend fuel burn-up in the CANDU reactors. For the purpose of proving safety of the RU handling techniques and appraising feasibility of the CANFLEX-RU fuel fabrication in near future, a physical, chemical and radiological characterization of the RU powder and pellets was performed. (author). 54 refs., 46 tabs., 62 figs.

  7. Interactive hypermedia training manual for spent-fuel bundle counters

    International Nuclear Information System (INIS)

    Spent-fuel bundle counters, developed by the Canadian Safeguards Support Program for the International Atomic Energy Agency, provide a secure and independent means of counting the number of irradiated fuel bundles discharged into the fuel storage bays at CANDU nuclear power stations. Paper manuals have been traditionally used to familiarize IAEA inspectors with the operation, maintenance and extensive reporting capabilities of the bundle counters. To further assist inspectors, an interactive training manual has been developed on an Apple Macintosh computer using hypermedia software. The manual uses interactive animation and sound, in conjunction with the traditional text and graphics, to simulate the underlying operation and logic of the bundle counters. This paper presents the key features of the interactive manual and highlights the advantages of this new technology for training

  8. Interim storage facility for spent fuel

    International Nuclear Information System (INIS)

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

  9. Spent fuel corrosion and dissolution

    International Nuclear Information System (INIS)

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

  10. Status of Spent Fuel Management in Korea

    International Nuclear Information System (INIS)

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

  11. Developments in spent fuel storage

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-04-01

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

  12. Spent fuel canister docking station

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-01-15

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

  13. Probability of spent fuel transportation accidents

    Energy Technology Data Exchange (ETDEWEB)

    McClure, J. D.

    1981-07-01

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

  14. An analytical assessment of the longitudinal ridging of CANDU type fuel element

    International Nuclear Information System (INIS)

    There are 380 fuel channels in a CANDU-6 reactor, and twelve fuel bundles are loaded into each fuel channel. High-pressure, heavy water coolant passes through the fuel bundle string to remove heat generated from the fuel. Fuel sheath collapses down around the uranium dioxide pellet due to the coolant pressure when the fuel is loaded into the reactor. Longitudinal ridges may form in CANDU fuel element sheaths as a result of sheath collapse onto the pellets. A static analysis, finite-element (FE) model was developed to simulate the longitudinal ridging of the fuel element with use of the structural analysis computer code ABAQUS. Collapse pressures were calculated for the fifty-one cases for which test results of WCL in 1973 and 1975 are available. Calculation results under-predicted the critical collapse pressure but it showed significant relationship against test results

  15. Proceedings of the 1. international conference on CANDU fuel handling systems

    International Nuclear Information System (INIS)

    Besides information on fuel loading and handling systems for CANDU and PHWR reactors, the 25 papers in these proceedings also include some on dry storage, modification to fuel strings at Bruce A, and on the SLAR (spacer location and repositioning) system for finding and moving garter springs. The individual papers have been abstracted separately

  16. Simulation of CANDU Fuel Behaviour into In-Reactor LOCA Tests

    International Nuclear Information System (INIS)

    The purpose of this work is to simulate the behaviour of an instrumented, unirradiated, zircaloy sheathed UO2 fuel element assembly of CANDU type, subjected to a coolant depressurization transient in the X-2 pressurized water loop of the NRX reactor at the Chalk River Nuclear Laboratories in 1983. The high-temperature transient conditions are such as those associated with the onset of a loss of coolant accident (LOCA). The data and the information related to the experiment are those included in the OECD/NEA-IFPE Database (IFPE/CANDU-FIO-131 NEA-1783/01). As tool for this simulation is used the TRANSURANUS fuel performance code, developed at ITU, Germany, along with the corresponding fabrication and in-reactor operating conditions specific of the CANDU PHWR fuel. The results, analyzed versus the experimental ones, are encouraging and perfectible. (author)

  17. Destructive Examination of Experimental Candu Fuel Elements Irradiated in TRIGA-SSR Reactor

    International Nuclear Information System (INIS)

    The object of this work is the behaviour of CANDU fuel elements under power cycling conditions. The tests were run in the 14 MW(th) TRIGA-SSR (Steady State Reactor) reactor from Institute for Nuclear Research (INR) Pitesti. zircaloy-4 is the material used for CANDU fuel sheath. The importance of studying its behaviour results from the fact that the mechanical properties of the CANDU fuel sheath suffer modifications during normal and abnormal operation. In the nuclear reactor the fuel elements endure dimensional and structural changes as well as cladding oxidation, hydriding and corrosion. These changes can lead to defects and even to the loss of integrity of the cladding. This paper presents the results of examinations performed in the Post- irradiation Examination Laboratory (PIEL) from INR Pitesti, on samples from a fuel element irradiated in TRIGA-SSR reactor: (i) Dimensional and macrostructural characterization; (ii) Microstructural characterization by metallographic analyses; (iii) Determination of mechanical properties; (iv) Fracture surface analysis by scanning electron microscopy (SEM). The obtained data could be used to evaluate the security, reliability and nuclear fuel performance, and for CANDU fuel improvement. (author)

  18. Post Irradiation Examination of Experomental CANDU Fuel Elements Irradiated in TRIGA-SSR Reactor

    International Nuclear Information System (INIS)

    The object of this work is the behaviour of CANDU fuel elements under power cycling conditions. The tests were run in the 14 MW (th) TRIGA-SSR (Steady State Reactor) reactor from Institute for Nuclear Research (INR) Pitesti. Zircaloy-4 is the material used for CANDU fuel sheath. The importance of studying its behaviour results from the fact that the mechanical properties of the CANDU fuel sheath suffer modifications during normal and abnormal operation. In the nuclear reactor the fuel elements endure dimensional and structural changes as well as cladding oxidation, hydriding and corrosion. These changes can lead to defects and even to the loss of integrity of the cladding. This paper presents the results of examinations performed in the Post Irradiation Examination Laboratory (PIEL) from INR Pitesti, on samples from a fuel element irradiated in TRIGA-SSR reactor: (i) Dimensional and macrostructural characterization; (ii) Gamma scanning and tomography; (iii) Measurement of pressure, volume and isotopic composition of fission gas; (iv) Microstructural characterization by metallographic analyses; (v) Determination of mechanical properties; amd (vi) Fracture surface analysis by scanning electron microscopy (SEM). The obtained data could be used to evaluate the security, reliability and nuclear fuel performance, and for CANDU fuel improvement. (author)

  19. Load following tests on CANDU-type fuel elements in TRIGA research reactor of INR Pitesti

    International Nuclear Information System (INIS)

    Two load following (LF) tests on CANDU-type fuel elements were performed in the TRIGA Research Reactor of INR Pitesti, where the tests were designed to represent fuel in a CANDU reactor operating in a load following regime. In the first LF test the designated '78R' fuel element successfully experienced 367 power cycles, mostly between 23 and 56 kW/m average linear power. In the second LF test, developed under INR-AECL co-operation, the fuel element designated as 'ME01' withstood 200 power cycles from 27 to 54 kW/m average linear power, as well as additional ramps due to reactor trips and restarts during the test period. This experimental program is ongoing at INR Pitesti. Both LF tests were simulated with finite element computer codes in order to evaluate Stress Corrosion Fatigue (SCF) of the cladding arising from expansion and contraction of the pellets. New LF tests are planned to be performed in order to establish the limits and capabilities for CANDU fuel in LF conditions. This paper presents the results of the LF tests performed in the INR TRIGA Research Reactor compared with the analytical assessment for SCF conditions and their relation to CANDU fuel performance in LF conditions. (author)

  20. Development of the advanced PHWR technology -Design and analysis of CANDU advanced fuel-

    Energy Technology Data Exchange (ETDEWEB)

    Suk, Hoh Chun; Shim, Kee Sub; Byun, Taek Sang; Park, Kwang Suk; Kang, Heui Yung; Kim, Bong Kee; Jung, Chang Joon; Lee, Yung Wook; Bae, Chang Joon; Kwon, Oh Sun; Oh, Duk Joo; Im, Hong Sik; Ohn, Myung Ryong; Lee, Kang Moon; Park, Joo Hwan; Lee, Eui Joon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This is the `94 annual report of the CANDU advanced fuel design and analysis project, and describes CANFLEX fuel design and mechanical integrity analysis, reactor physics analysis and safety analysis of the CANDU-6 with the CANFLEX-NU. The following is the R and D scope of this fiscal year : (1) Detail design of CANFLEX-NU and detail analysis on the fuel integrity, reactor physics and safety. (a) Detail design and mechanical integrity analysis of the bundle (b) CANDU-6 refueling simulation, and analysis on the Xe transients and adjuster system capability (c) Licensing strategy establishment and safety analysis for the CANFLEX-NU demonstration demonstration irradiation in a commercial CANDU-6. (2) Production and revision of CANFLEX-NU fuel design documents (a) Production and approval of CANFLEX-NU reference drawing, and revisions of fuel design manual and technical specifications (b) Production of draft physics design manual. (3) Basic research on CANFLEX-SEU fuel. 55 figs, 21 tabs, 45 refs. (Author).

  1. Spent fuel. Dissolution and oxidation

    International Nuclear Information System (INIS)

    Data from studies of the low temperature air oxidation of spent fuel were retrieved in order to provide a basis for comparison between the mechanism of oxidation in air and corrosion in water. U3O7 is formed by diffusion of oxygen into the UO2 lattice. A diffusion coefficient of oxygen in the fuel matric was calculated for 25 degree C to be in the range of 10-23 to 10-25 m2/s. The initial rates of U release from spent fuel and from UO2 appear to be similar. The lowest rates (at 25 degree c >10-4 g/(m2d)) were observed under reducing conditions. Under oxidizing conditions the rates depend mainly of the nature and concentraion of the oxidant and/or on corbonate. In contact with air, typical initial rates at room temperature were in the range between 0.001 and 0.1 g/(m2d). A study of apparent U solubility under oxidizing conditions was performed and it was suggested that the controlling factor is the redox potential at the UO2 surface rather than the Eh of the bulk solution. Electrochemical arguments were used to predict that at saturation, the surface potential will eventually reach a value given by the boundaries at either the U3O7/U3O8 or the U3O7/schoepite stability field, and a comparison with spent fuel leach data showed that the solution concentration of uranium is close to the calculated U solubility at the U3O7/U3O8 boundary. The difference in the cumulative Sr and U release was calculated from data from Studsvik laboratory. The results reveal that the rate of Sr release decreases with the square root of time under U-saturated conditions. This time dependence may be rationalized either by grain boundary diffusion or by diffusion into the fuel matrix. Hence, there seems to be a possibility of an agreement between the Sr release data, structural information and data for oxygen diffusion in UO2. (G.B.)

  2. Transportation accident scenarios for commercial spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Wilmot, E L

    1981-02-01

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

  3. Spent fuel shipping cask accident evaluation

    International Nuclear Information System (INIS)

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

  4. Overview on spent fuel management strategies

    International Nuclear Information System (INIS)

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

  5. Behavior of spent fuel under unsaturated conditions

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-05-15

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

  7. Automated refueling simulations of a CANDU for the exploitation of thorium fuels

    Science.gov (United States)

    Holmes, Bradford

    CANDU nuclear reactors are in a unique circumstance where they are able to utilize and exploit a number of different fuel options to provide power as a utility. Thorium, a fertile isotope found naturally, is one option that should be explored. Thorium is more abundant than uranium, which is the typical fuel in the reactor and the availability of thorium makes nuclear energy desirable to more countries. This document contains the culmination of a project that explores, tests, and analyzes the feasibility of using thorium in a CANDU reactor. The project first develops a set of twodimensional lattice and three dimensional control rod simulations using the DRAGON Version 4 nuclear physics codes. This step is repeated for many concentrations of thorium. The data generated in these steps is then used to determine a functional enrichment of thorium. This is done via a procedural elimination and optimization of certain key parameters including but not limited to average exit burnup and reactivity evolution. For the purposes of this project, an enrichment of 1 % thorium was found viable. Full core calculations were done using the DONJON 4 code. CANFUEL, a program which simulates the refueling operations of a CANDU reactor for this fuel type was developed and ran for a simulation period of one hundred days. The program and the fuel selection met all selected requirements for the entirety of the simulation period. CANFUEL requires optimization for fuel selection before it can be used extensively. The fuel selection was further scrutinized when a reactivity insertion event was simulated. The adjuster rod 11 withdrawal from the core was analyzed and compared to classical CANDU results in order to ensure no significant deviations or unwanted evolutions were encountered. For this case, the simulation results were deemed acceptable with no significant deviations from the classical CANDU case.

  8. Spent nuclear fuel reprocessing modeling

    International Nuclear Information System (INIS)

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

  9. Spent fuel management newsletter. No. 2

    International Nuclear Information System (INIS)

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

  10. A modal method for transient thermal analysis of CANDU fuel channel

    International Nuclear Information System (INIS)

    The classical modal expansion technique has been applied to predict transient fuel and coolant temperatures under on-power conditions in a CANDU fuel channel. The temperature profile across the fuel pellet is assumed to be parabolic and fuel and coolant temperatures are expanded with Fourier series. The coefficient derivatives are written in state space form and solved by the Runge-Kutta method of fifth order. To validate the present model, the calculated fuel temperatures for several sample cases were compared with HOTSPOT-II, which employs a more rigorous finite-difference model. The agreement was found to be reasonable for the operational transients simulated. The advantage of the modal method is the fast computation speed for application to the real-time system such as the CANDU simulator which is being currently developed at the Institute for Advanced Engineering (IAE). (author)

  11. Remote technology applications in spent fuel management

    International Nuclear Information System (INIS)

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

  12. Designing and calculating the pressure loses for different geometries of CANDU type fuel clusters

    International Nuclear Information System (INIS)

    It is well known that circulation of the coolant through the pressure tube of a CANDU type reactor must ensure, through its flow rate values, the optimal conditions of heat transfer from the fuel clusters towards the heavy water. The flow rate through fuel channels differs from one another (up to 24 kg/s) depending on the fuel element sheath temperature, the latter depending in turn one the channels/clusters positions in the calandria vessel. In these conditions, one of the main problem of design in the CANDU type reactor plants is related to the hydraulic resistance represented by the fuel clusters loading the pressure tube or, in other words, the problem of pressure losses (pressure drops) over the length of the fuel cluster column. More precisely, this hydraulic resistance should not exceed a given value imposed by the performance calculations for the pumps used. A sustained activity of analysing comparatively the different geometry types of the fuel clusters was developed at INR Pitesti, a special attention being paid to their behavior as hydraulic resistances. The paper presents a set of computation programs devoted on one hand to the design of fuel clusters of different types and to an estimating computation of the pressure losses resulting from loading these clusters into a specific fuel channel of the CANDU type reactor, on the other hand. During the presentation of the work, different computing codes will be run for demonstration

  13. Spent Nuclear Fuel (SNF) Project Execution Plan

    Energy Technology Data Exchange (ETDEWEB)

    LEROY, P.G.

    2000-11-03

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

  14. Spent Nuclear Fuel (SNF) Project Execution Plan

    International Nuclear Information System (INIS)

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

  15. Swelling of spent fuel storage tubes

    International Nuclear Information System (INIS)

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

  16. ASSERT-PV 3.2: Advanced subchannel thermalhydraulics code for CANDU fuel bundles

    International Nuclear Information System (INIS)

    Highlights: • Introduction to a new version of the Canadian subchannel code, ASSERT-PV 3.2. • Enhanced models for flow-distribution, CHF and post-dryout heat transfer prediction. • Model changes focused on unique features of horizontal CANDU bundles. • Detailed description of model changes for all major thermalhydraulics models. • Discussion on rationale and limitation of the model changes. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The most recent release version, ASSERT-PV 3.2 has enhanced phenomenon models for improved predictions of flow distribution, dryout power and CHF location, and post-dryout (PDO) sheath temperature in horizontal CANDU fuel bundles. The focus of the improvements is mainly on modeling considerations for the unique features of CANDU bundles such as horizontal flows, small pitch to diameter ratios, high mass fluxes, and mixed and irregular subchannel geometries, compared to PWR/BWR fuel assemblies. This paper provides a general introduction to ASSERT-PV 3.2, and describes the model changes or additions in the new version to improve predictions of flow distribution, dryout power and CHF location, and PDO sheath temperatures in CANDU fuel bundles

  17. ASSERT-PV 3.2: Advanced subchannel thermalhydraulics code for CANDU fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Rao, Y.F., E-mail: raoy@aecl.ca; Cheng, Z., E-mail: chengz@aecl.ca; Waddington, G.M., E-mail: waddingg@aecl.ca; Nava-Dominguez, A., E-mail: navadoma@aecl.ca

    2014-08-15

    Highlights: • Introduction to a new version of the Canadian subchannel code, ASSERT-PV 3.2. • Enhanced models for flow-distribution, CHF and post-dryout heat transfer prediction. • Model changes focused on unique features of horizontal CANDU bundles. • Detailed description of model changes for all major thermalhydraulics models. • Discussion on rationale and limitation of the model changes. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The most recent release version, ASSERT-PV 3.2 has enhanced phenomenon models for improved predictions of flow distribution, dryout power and CHF location, and post-dryout (PDO) sheath temperature in horizontal CANDU fuel bundles. The focus of the improvements is mainly on modeling considerations for the unique features of CANDU bundles such as horizontal flows, small pitch to diameter ratios, high mass fluxes, and mixed and irregular subchannel geometries, compared to PWR/BWR fuel assemblies. This paper provides a general introduction to ASSERT-PV 3.2, and describes the model changes or additions in the new version to improve predictions of flow distribution, dryout power and CHF location, and PDO sheath temperatures in CANDU fuel bundles.

  18. Spent Fuel Management in the Slovak Republic

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-07-01

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

  20. Long term wet spent nuclear fuel storage

    International Nuclear Information System (INIS)

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

  1. Spent nuclear fuel disposal liability insurance

    International Nuclear Information System (INIS)

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

  2. Validation of the COBRA code for dry out power calculation in CANDU type advanced fuels

    International Nuclear Information System (INIS)

    Stern Laboratories perform a full scale CHF testing of the CANFLEX bundle under AECL request. This experiment is modeled with the COBRA IV HW code to verify it's capacity for the dry out power calculation . Good results were obtained: errors below 10 % with respect to all data measured and 1 % for standard operating conditions in CANDU reactors range . This calculations were repeated for the CNEA advanced fuel CARA obtaining the same performance as the CANFLEX fuel. (author)

  3. 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)

  4. Nuclear spent fuel management. Experience and options

    International Nuclear Information System (INIS)

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

  5. Spent Fuel Management Newsletter. No. 1

    International Nuclear Information System (INIS)

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

  6. Performance evaluation of two CANDU fuel elements tested in the TRIGA reactor

    International Nuclear Information System (INIS)

    Nuclear Research Institute at Pitesti has a set of facilities, which allow the testing, manipulation and examination of nuclear fuel and structure materials irradiated in CANDU reactors from Cernavoda NPP. These facilities consist of TRIGA materials testing reactor and Post-Irradiation Examination Laboratory (LEPI). The purpose of this work is to describe the post-irradiation examination, of two experimental CANDU fuel elements (EC1 and EC2). The fuel elements were mounted into a pattern port, one in extension of the other in a measuring test for the central temperature evolution. The results of post-irradiation examination are obtained from: Visual inspection and photography of the outer appearance of sheath; Profilometry (diameter, bending, ovalization) and length measuring; Determination of axial and radial distribution of the fission products activity by gamma scanning; Measurement of pressure, volume and isotopic composition of fission gas; Microstructural characterization by metallographic and ceramographic analyzes; Isotopic composition and burn-up determination. The post-irradiation examination results are used, on one hand, to confirm the security, reliability and performance of the irradiated fuel, and on the other hand, for further development of CANDU fuel. (authors)

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

    International Nuclear Information System (INIS)

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

  8. International experience in conditioning spent fuel elements

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  10. Advances in HTGR spent fuel treatment technology

    International Nuclear Information System (INIS)

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

  11. Spent fuel disposal impact on plant decommissioning

    International Nuclear Information System (INIS)

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

  12. Spent Fuel Background Report Volume I

    International Nuclear Information System (INIS)

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

  13. Enhanced candu 6 reactor: status

    International Nuclear Information System (INIS)

    operating plants. The EC6 will utilize modern computers and control systems housed in an Advanced Control Room which, along with automated testing, will make the plants easier to operate, with minimal operator intervention. Improvements to the fire protection system and enhanced security features will further protect the assets. The CANDU 6s inherent ability to utilize different fuel cycles and the Advanced MACSTOR module to store spent dry fuel further add to the attractiveness of the EC6 product. This paper describes the various enhancements that are being made to the EC6 and explains these features in more detail

  14. LWR spent fuel management in Germany

    International Nuclear Information System (INIS)

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

  15. Safeguards issues in spent fuel consolidation facilities

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-01-01

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

  16. Aging effect on the fuel behaviors for CANDU fuel safety analysis

    International Nuclear Information System (INIS)

    Because of the aging of heat transport system components, the reactor thermalhydraulic conditions can vary, which may affect the safety response. In a recent safety analysis for the refurbished Wolsong 1 NPP, various aging effects were incorporated into the hydraulic models of the components in the primary heat transport system (PHTS) for conservatism. The aging data of the thermal-hydraulic components for an 11 EFPY of Wolsong 1 were derived based on the site operation data and were modified to the appropriate input data for the thermal-hydraulic code for a safety analysis of a postulated accident. This paper deals with the aging effect of the PHTS of the CANDU reactor on the fuel performance during normal operation and transient period following a postulated accident such as a feeder stagnation break. (author)

  17. Development of spent fuel storage process equipment

    International Nuclear Information System (INIS)

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

  18. Operating Experience in Spent Fuel Storage Casks

    International Nuclear Information System (INIS)

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

  19. Timely topics on spent fuel storage

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-10-01

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

  20. Timely topics on spent fuel storage

    International Nuclear Information System (INIS)

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

  1. Study of CANDU thorium-based fuel cycles by deterministic and Monte Carlo methods

    International Nuclear Information System (INIS)

    In the framework of the Generation IV forum, there is a renewal of interest in self-sustainable thorium fuel cycles applied to various concepts such as Molten Salt Reactors [1, 2] or High Temperature Reactors [3, 4]. Precise evaluations of the U-233 production potential relying on existing reactors such as PWRs [5] or CANDUs [6] are hence necessary. As a consequence of its design (online refueling and D2O moderator in a thermal spectrum), the CANDU reactor has moreover an excellent neutron economy and consequently a high fissile conversion ratio [7]. For these reasons, we try here, with a shorter term view, to re-evaluate the economic competitiveness of once-through thorium-based fuel cycles in CANDU [8]. Two simulation tools are used: the deterministic Canadian cell code DRAGON [9] and MURE [10], a C++ tool for reactor evolution calculations based on the Monte Carlo code MCNP [11]. (authors)

  2. CANDU fuel attribution through the analysis of delayed neutron temporal behaviour

    International Nuclear Information System (INIS)

    Delayed Neutron Counting (DNC) is an established technique in the Canadian nuclear industry as it is used for the detection of defective fuel in several CANDU reactors and the assay of uranium in geological samples. This paper describes the possible expansion of DNC to the discipline of nuclear forensics analysis. The temporal behaviour of experimentally measured delayed neutron spectra were used to determine the relative contributions of 233U and 235U to the overall fissile content present in mixtures with average absolute errors of ±4 %. The characterization of fissile content in current and proposed CANDU fuels (natural UO2, thoria and mixed oxide (MOX) based) by DNC analysis is evaluated through Monte Carlo simulations. (author)

  3. Antineutrino monitoring of spent nuclear fuel

    OpenAIRE

    Brdar, Vedran; Huber, Patrick; Kopp, Joachim

    2016-01-01

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

  4. Investigation of Spent Nuclear Fuel Pool Coolability

    OpenAIRE

    Nimander, Fredrik

    2011-01-01

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

  5. Foreign experience on effects of extended dry storage on the integrity of spent nuclear fuel

    International Nuclear Information System (INIS)

    This report summarizes the results of a survey of foreign experience in dry storage of spent fuel from nuclear power reactors that was carried out for the US Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM). The report reviews the mechanisms for degradation of spent fuel cladding and fuel materials in dry storage, identifies the status and plans of world-wide experience and applications, and documents the available information on the expected long-term integrity of the dry-stored spent fuel from actual foreign experience. Countries covered in this survey are: Argentina, Canada, Federal Republic of Germany (before reunification with the former East Germany), former German Democratic Republic (former East Germany), France, India, Italy, Japan, South Korea, Spain, Switzerland, United Kingdom, and the former USSR (most of these former Republics are now in the Commonwealth of Independent States [CIS]). Industrial dry storage of Magnox fuels started in 1972 in the United Kingdom; Canada began industrial dry storage of CANDU fuels in 1980. The technology for safe storage is generally considered to be developed for time periods of 30 to 100 years for LWR fuel in inert gas and for some fuels in oxidizing gases at low temperatures. Because it will probably be decades before countries will have a repository for spent fuels and high-level wastes, the plans for expanded use of dry storage have increased significantly in recent years and are expected to continue to increase in the near future

  6. Foreign experience on effects of extended dry storage on the integrity of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, K.J.; Mitchell, S.J.

    1992-04-01

    This report summarizes the results of a survey of foreign experience in dry storage of spent fuel from nuclear power reactors that was carried out for the US Department of Energy`s (DOE) Office of Civilian Radioactive Waste Management (OCRWM). The report reviews the mechanisms for degradation of spent fuel cladding and fuel materials in dry storage, identifies the status and plans of world-wide experience and applications, and documents the available information on the expected long-term integrity of the dry-stored spent fuel from actual foreign experience. Countries covered in this survey are: Argentina, Canada, Federal Republic of Germany (before reunification with the former East Germany), former German Democratic Republic (former East Germany), France, India, Italy, Japan, South Korea, Spain, Switzerland, United Kingdom, and the former USSR (most of these former Republics are now in the Commonwealth of Independent States [CIS]). Industrial dry storage of Magnox fuels started in 1972 in the United Kingdom; Canada began industrial dry storage of CANDU fuels in 1980. The technology for safe storage is generally considered to be developed for time periods of 30 to 100 years for LWR fuel in inert gas and for some fuels in oxidizing gases at low temperatures. Because it will probably be decades before countries will have a repository for spent fuels and high-level wastes, the plans for expanded use of dry storage have increased significantly in recent years and are expected to continue to increase in the near future.

  7. Foreign experience on effects of extended dry storage on the integrity of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, K.J.; Mitchell, S.J.

    1992-04-01

    This report summarizes the results of a survey of foreign experience in dry storage of spent fuel from nuclear power reactors that was carried out for the US Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM). The report reviews the mechanisms for degradation of spent fuel cladding and fuel materials in dry storage, identifies the status and plans of world-wide experience and applications, and documents the available information on the expected long-term integrity of the dry-stored spent fuel from actual foreign experience. Countries covered in this survey are: Argentina, Canada, Federal Republic of Germany (before reunification with the former East Germany), former German Democratic Republic (former East Germany), France, India, Italy, Japan, South Korea, Spain, Switzerland, United Kingdom, and the former USSR (most of these former Republics are now in the Commonwealth of Independent States (CIS)). Industrial dry storage of Magnox fuels started in 1972 in the United Kingdom; Canada began industrial dry storage of CANDU fuels in 1980. The technology for safe storage is generally considered to be developed for time periods of 30 to 100 years for LWR fuel in inert gas and for some fuels in oxidizing gases at low temperatures. Because it will probably be decades before countries will have a repository for spent fuels and high-level wastes, the plans for expanded use of dry storage have increased significantly in recent years and are expected to continue to increase in the near future.

  8. A study of coolant thermal mixing within CANDU fuel bundles using ASSERT-PV

    International Nuclear Information System (INIS)

    This paper presents the results of a study of the thermal mixing of single-phase coolant in 28-element CANDU fuel bundles. The approach taken in the present work is to identify the physical mechanisms contributing to coolant mixing, and to systematically assess the importance of each mechanism. Coupled effects were also considered by flow simulation with mixing mechanisms modelled simultaneously. For the limited range of operating conditions considered and when all mixing mechanisms were modelled simultaneously, the flow was found to be very close to fully mixed. A preliminary model of coolant mixing, suitable for use in the fuel and fuel channel code FACTAR, is also presented. (author)

  9. Feasibility study of modeling a CANDU fuel element using a multiphysics object-oriented simulation environment

    International Nuclear Information System (INIS)

    The first phase of the feasibility study of using a Multiphysics Object-Oriented Simulation Environment (MOOSE) for modeling a CANDU fuel element is presented. A two-dimensional model of a fuel pellet sheath was created to examine the contact algorithm within MOOSE. The results obtained show the expected behaviour of contact pressure and penetration in 2D. Preliminary results for a 3D model of a quarter fuel pellet and sheath are provided but at present contain anomalies currently being investigated. The next steps in the feasibility study are outlined. (author)

  10. Feasibility study of modeling a CANDU fuel element using a multiphysics object-oriented simulation environment

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, K., E-mail: Kyle.Gamble@rmc.ca [Royal Military College of Ontario, Kingston, Ontario (Canada); Williams, A. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Chan, P.K. [Royal Military College of Ontario, Kingston, Ontario (Canada)

    2013-07-01

    The first phase of the feasibility study of using a Multiphysics Object-Oriented Simulation Environment (MOOSE) for modeling a CANDU fuel element is presented. A two-dimensional model of a fuel pellet sheath was created to examine the contact algorithm within MOOSE. The results obtained show the expected behaviour of contact pressure and penetration in 2D. Preliminary results for a 3D model of a quarter fuel pellet and sheath are provided but at present contain anomalies currently being investigated. The next steps in the feasibility study are outlined. (author)

  11. Spent fuel transport in Romania by road: An approach considering safety, risk and radiological consequences

    International Nuclear Information System (INIS)

    The transport of high-level radioactive wastes, involving Type B packages, is a part of the safety of the Romanian waste management programme and the overall aim of this activity is to promote the safe transport of radioactive materials in Romania. The paper presents a safety case analysis of the transport of a single spent fuel CANDU bundle, using a Romanian built Type B package, from the CANDU type nuclear power plant Cernavoda to the INR Pitesti, in order to be examined within INR's hot-cells facilities. The safety assessment includes the following main aspects: (1) evaluation and analysis of available data on road traffic accidents; (2) estimation of the expected frequency for severe road accident scenarios resulting in potential radionuclide release; and (3) evaluation of the expected radiological consequences and accident risks of transport operations. (author)

  12. OECD NEA Benchmark Database of Spent Nuclear Fuel Isotopic Compositions for World Reactor Designs

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, Ian C [ORNL; Sly, Nicholas C [ORNL; Michel-Sendis, Franco [OECD Nuclear Energy Agency

    2014-01-01

    Experimental data on the isotopic concentrations in irradiated nuclear fuel represent one of the primary methods for validating computational methods and nuclear data used for reactor and spent fuel depletion simulations that support nuclear fuel cycle safety and safeguards programs. Measurement data have previously not been available to users in a centralized or searchable format, and the majority of accessible information has been, for the most part, limited to light-water-reactor designs. This paper describes a recent initiative to compile spent fuel benchmark data for additional reactor designs used throughout the world that can be used to validate computer model simulations that support nuclear energy and nuclear safeguards missions. Experimental benchmark data have been expanded to include VVER-440, VVER-1000, RBMK, graphite moderated MAGNOX, gas cooled AGR, and several heavy-water moderated CANDU reactor designs. Additional experimental data for pressurized light water and boiling water reactor fuels has also been compiled for modern assembly designs and more extensive isotopic measurements. These data are being compiled and uploaded to a recently revised structured and searchable database, SFCOMPO, to provide the nuclear analysis community with a centrally-accessible resource of spent fuel compositions that can be used to benchmark computer codes, models, and nuclear data. The current version of SFCOMPO contains data for eight reactor designs, 20 fuel assembly designs, more than 550 spent fuel samples, and measured isotopic data for about 80 nuclides.

  13. Silicon carbide TRIPLEX materials for CANDU fuel cladding and pressure tubes

    International Nuclear Information System (INIS)

    Ceramic Tubular Products has developed a superior silicon carbide (SiC) material TRIPLEX, which can be used for both fuel cladding and other zirconium alloy materials in light water reactor (LWR) and heavy water reactor (CANDU) systems. The fuel cladding can replace Zircaloy cladding and other zirconium based alloy materials in the reactor systems. It has the potential to provide higher fuel performance levels in currently operating natural UO2 (NEU) fuel design and in advanced fuel designs (UO2(SEU), MOX thoria) at higher burnups and power levels. In all the cases for fuel designs TRIPLEX has increased resistance to severe accident conditions. The interaction of SiC with steam and water does not produce an exothermic reaction to produce hydrogen as occurs with zirconium based alloys. In addition the absence of creep down eliminates clad ballooning during high temperature accidents which occurs with Zircaloy blocking water channels required to cool the fuel. (author)

  14. 78 FR 3853 - Retrievability, Cladding Integrity and Safe Handling of Spent Fuel at an Independent Spent Fuel...

    Science.gov (United States)

    2013-01-17

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

  15. Study on the use of slightly enriched uranium fuel cycle in an existing CANDU 6 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yeom, Choong Sub; Kim, Hyun Dae [Institute for Advanced Engineering, Seoul (Korea, Republic of)

    1997-12-31

    To test the viability of CANFLEX-SEU bundles in an existing CANDU 6 reactor, core follow-up simulation has been carried out using the reactor fueling simulation program of the CANDU 6, RFSP computer code, and a lattice physics code, WIMS-AECL. During the core follow-up, bundle and channel powers and zone levels have been checked against their operating limits at each simulation. It is observed from the simulation results that an equilibrium core loaded with 0.9 w/o CANFLEX-SEU bundles could be refueled and maintained for 550 FPD without any significant violations in the channel and bundle power limits and the permissible operating range of the liquid zone controllers. 8 refs., 2 figs., 1 tab. (Author)

  16. Neutron intensity of fast reactor spent fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  17. Integrated spent nuclear fuel database system

    International Nuclear Information System (INIS)

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

  18. Spent fuel transportation and storage experience

    International Nuclear Information System (INIS)

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

  19. Thermal-hydraulics performance optimization of Candu fuel using Assert subchannel code

    International Nuclear Information System (INIS)

    An optimization of fuel bundle geometry using the subchannel code ASSERT is performed in support of Candu fuel design to enhance the thermohydraulics performance. The new bundle design is based on a reference CANFLEX bundle with changes to the centre and inner-ring element diameters and pitch-circle diameters (PCDs) of various element rings. Different methods of varying the PCDs for reaching the optimized geometry are considered in an attempt to minimize the optimization effort. The optimized geometry in the present analysis is the one that maximizes the dryout power and that has simultaneous CHF (critical heat flux) initiation involving more than one subchannel rings. (authors)

  20. Power pulse tests on CANDU type fuel elements in TRIGA reactor of INR Pitesti

    International Nuclear Information System (INIS)

    Pulse irradiation tests on short fuel elements have been carried out in TRIGA Annular Core Pulse Reactor (TRIGA ACPR) of INR Pitesti to investigate aspects related to the thermal and mechanical behavior of CANDU type fuel elements under short duration and large amplitude power pulse conditions. Short test fuel elements were instrumented with thermocouples for cladding surface temperature measurements and pressure sensor for element internal pressure measurement. Transient histories of reactor power, cooling water pressure, fuel element internal pressure and cladding temperature were recorded during tests. The fuel elements were subjected to total energy deposition from 70 to 280cal g-1 UO2. Rapid fuel pellet expansion due to a power excursion caused radial and longitudinal deformation of the cladding. Cladding failure mechanism and the failure threshold have been established. This paper presents some recent results obtained from these power pulse tests performed in TRIGA ACPR of INR Pitesti. (author)

  1. Behavior of CANDU fuel under power pulse conditions at the TRIGA reactor of INR Pitesti

    International Nuclear Information System (INIS)

    Pulse irradiation tests on short fuel elements have been carried out in TRIGA Annular Core Pulse Reactor (TRIGA ACPR) of INR Pitesti to investigate aspects related to the thermal and mechanical behavior of CANDU type fuel elements under short duration and large amplitude power pulse conditions. Short test fuel elements were instrumented with thermocouples for cladding surface temperature measurements and pressure sensors for element internal pressure measurement. Transient histories of reactor power, cooling water pressure, fuel element internal pressure and cladding temperature were recorded during tests. The fuel elements were subjected to total energy deposition from 70 to 280 cal g-1 UO2. Rapid fuel pellet expansion due to a power excursion caused radial and longitudinal deformation of the cladding. Cladding failure mechanism and the failure threshold have been established. This paper presents some recent results obtained from these power pulse tests performed in TRIGA ACPR of INR Pitesti. (orig.)

  2. Investigations on flow induced vibration of simulated CANDU fuel bundles in a pipe

    International Nuclear Information System (INIS)

    In this paper, vibration of a two-bundle string consisting of simulated CANDU fuel bundles subjected to turbulent liquid flow is investigated through numerical simulations and experiments. Large eddy simulation is used to solve the three-dimensional turbulent flow surrounding the fuel bundles for determining fluid excitations. The CFD model includes pipe flow, flow through the inlet fuel bundle along with its two endplates, half of the second bundle and its upstream endplate. The fluid excitation obtained from the fluid model is subsequently fed into a fuel bundle vibration code written in FORTRAN. Fluid structure interaction terms for the fuel elements are approximated using the slender body theory. Simulation results are compared to measurements conducted on the simulated fuel bundles in a testing hydraulic loop. (author)

  3. Corrosion of spent Advanced Test Reactor fuel

    International Nuclear Information System (INIS)

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

  4. Subchannel Analysis for enhancing the fuel performance in CANDU reactor

    International Nuclear Information System (INIS)

    The effect of the fuel rod geometry in a fuel bundle using the subchannel code ASSERT has been evaluated to design the fuel bundle having the advanced fuel performance. Based on the configuration of standard 37-element fuel bundle, the element diameter of fuel rods in each ring has been changed while that of fuel rods in other rings is kept as the original size. The dryout power of each element in a fuel bundle has been obtained for the modified fuel bundle and compared with that of a standard fuel bundle. From the calculated mixture enthalpy and void fraction of each subchannel, it was found that the modification of element diameter largely affects to the thermal characteristics of the subchannel on the upper region of a modified element by the buoyancy drift effect. The optimized geometry in a fuel bundle has been suggested from the consideration of the change of void reactivity as well as the dryout power of a bundle. The dependency of the transverse interchange model on the present results has been checked by examining the dryout power of a bundle for the different mixing coefficient and buoyancy drift model

  5. Spent fuel management of Jose Cabrera NPP

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-07-01

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

  6. Reference Spent Fuel and Its Source Terms for a Design of Deep Geological Disposal System

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Dong Keun

    2005-12-15

    In this study, current status and future trend of domestic spent fuels were analyzed to propose reference spent nuclear fuel. And then, source terms needed for design of a deep geological disposal system were calculated using ORIGEN-ARP. The reference spent fuels selected based on assembly physical dimension, inventory projection, trend of initial enrichment of 235U, discharge burnup are as follows; The 17x17 Korean Optimized Fuel Assembly with initial enrichment of 4.0 wt.% 235U and discharge burnup of 45 GWD/MTU was adopted as a low-burnup representative fuel. For the high-burnup representative fuel, 16x16 Korean Standard Fuel Assembly with initial enrichment of 4.5 wt.% 235U and discharge burnup of 55 GWD/MTU was chosen. CANDU fuel with initial enrichment of 0.711 wt.% 235U and discharge burnup of 7.5 GWD/MTU was also considered. For these reference fuels, decay heat, radiation intensity and spectrum, nuclide concentration, and individual nuclide radioactivity were calculated using ORIGEN-ARP for a disposal system design. It is expected that the source terms estimated in this study will be applied to the disposal system development in the future.

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

    International Nuclear Information System (INIS)

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

  8. CFD analysis of the 37-element fuel channel for CANDU6 reactor

    International Nuclear Information System (INIS)

    We analyzed the thermal-hydraulic behavior of coolant flow along fuel bundles with appendages of end support plate, spacer pad, and bearing pad, which are the CANDU6 characteristic design. The computer code used is a commercial CFD code, CFX-12. The present CFD analysis model calculates the conjugate heat transfer between the fuel and coolant. Using the same volumetric heat source as the O6 channel, the CFD predictions of the axial temperature distributions of the fuel element are compared with those by the CATHENA (one-dimensional safety analysis code for CANDU6 reactor). It is shown that CFX-12 predictions are in good agreement with those by the CATHENA code for the single liquid convection region (especially before the axial position of the first half of the channel length). However, the CFD analysis at the second half of the fuel channel, where the two-phase flow is expected to occur, over-predicts the fuel temperature, since the wall boiling model is not considered in the present CFD model. (author)

  9. Hanford spent fuel inventory baseline

    Energy Technology Data Exchange (ETDEWEB)

    Bergsman, K.H.

    1994-07-15

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

  10. Development of spent fuel remote handling technology

    International Nuclear Information System (INIS)

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

  11. Spent fuel management: Current status and prospects

    International Nuclear Information System (INIS)

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

  12. Spent Nuclear Fuel Alternative Technology Decision Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Shedrow, C.B.

    1999-11-29

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

  13. Concrete Spent Fuel Cask Criticality Calculation

    International Nuclear Information System (INIS)

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

  14. Spent Fuel Behaviour During Interim Storage

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  16. Classification of spent nuclear fuel (SNF)

    International Nuclear Information System (INIS)

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

  17. Understanding CANDU fuel bowing in dryout: an industry approach

    International Nuclear Information System (INIS)

    Fuel element bow induced by dryout could potentially perturb the coolant flow distribution and heat transfer from the fuel element to the coolant. Some accident scenarios leading to dryout of the fuel element are: loss of power regulation pump trip, pump seizure, small and large break loss of coolant accidents. In these accidents, it is desirable to show with confidence that the fuel remains sufficiently cooled to maintain its geometry, even if it is in dryout. This can be demonstrated if fuel elements are separated from each other and from the pressure tube, with a sufficient (and stable) gap. Therefore, the prediction of the amount of bow, and its effect on heat transfer conditions is required for the assessments. The utilities have joined force in launching an experimental investigation at Stern Laboratories to characterize the bowing phenomena. This program will investigate the amount of deflection, transient and permanent, that results from accident conditions which cause a dry patch on one side of the sheath. This is expected to bound the consequences of fuel bowing due to dryout. Since the accident transients begin at full power and high coolant pressure (about 10 MPa) they generate sharp thermal gradients (dry patch) and it is necessary to develop a simulation with representative dry fuel sheath conditions initiated from a normal full power and coolant state. The amount of bow is driven by thermal gradients in both the fuel pellets and the sheath, therefore, the thermal gradients should be representative. This program is structured in a series of tests progressing from simple representation to complex simulation. It is divided into 3 experimental phases: Phase 1 - Thermalhydraulic simulation of fuel element bow by a heated tube; Phase 2 - Thermal and mechanical bow with a simulator which accounts for pellet / fuel sheath interaction with internal pellet temperature distributions; and Phase 3 - Fuel element bow with a simulator using Zircaloy-4 fuel sheath

  18. Plutonium Consumption Program, CANDU Reactor Project final report

    Energy Technology Data Exchange (ETDEWEB)

    1994-07-31

    DOE is investigating methods for long term dispositioning of weapons grade plutonium. One such method would be to utilize the plutonium in Mixed OXide (MOX) fuel assemblies in existing CANDU reactors. CANDU (Canadian Deuterium Uranium) reactors are designed, licensed, built, and supported by Atomic Energy of Canada Limited (AECL), and currently use natural uranium oxide as fuel. The MOX spent fuel assemblies removed from the reactor would be similar to the spent fuel currently produced using natural uranium fuel, thus rendering the plutonium as unattractive as that in the stockpiles of commercial spent fuel. This report presents the results of a study sponsored by the DOE for dispositioning the plutonium using CANDU technology. Ontario Hydro`s Bruce A was used as reference. The fuel design study defined the optimum parameters to disposition 50 tons of Pu in 25 years (or 100 tons). Two alternate fuel designs were studied. Safeguards, security, environment, safety, health, economics, etc. were considered. Options for complete destruction of the Pu were also studied briefly; CANDU has a superior ability for this. Alternative deployment options were explored and the potential impact on Pu dispositioning in the former Soviet Union was studied. An integrated system can be ready to begin Pu consumption in 4 years, with no changes required to the reactors other than for safe, secure storage of new fuel.

  19. Plutonium Consumption Program, CANDU Reactor Project final report

    International Nuclear Information System (INIS)

    DOE is investigating methods for long term dispositioning of weapons grade plutonium. One such method would be to utilize the plutonium in Mixed OXide (MOX) fuel assemblies in existing CANDU reactors. CANDU (Canadian Deuterium Uranium) reactors are designed, licensed, built, and supported by Atomic Energy of Canada Limited (AECL), and currently use natural uranium oxide as fuel. The MOX spent fuel assemblies removed from the reactor would be similar to the spent fuel currently produced using natural uranium fuel, thus rendering the plutonium as unattractive as that in the stockpiles of commercial spent fuel. This report presents the results of a study sponsored by the DOE for dispositioning the plutonium using CANDU technology. Ontario Hydro's Bruce A was used as reference. The fuel design study defined the optimum parameters to disposition 50 tons of Pu in 25 years (or 100 tons). Two alternate fuel designs were studied. Safeguards, security, environment, safety, health, economics, etc. were considered. Options for complete destruction of the Pu were also studied briefly; CANDU has a superior ability for this. Alternative deployment options were explored and the potential impact on Pu dispositioning in the former Soviet Union was studied. An integrated system can be ready to begin Pu consumption in 4 years, with no changes required to the reactors other than for safe, secure storage of new fuel

  20. Radioactivity of spent TRIGA fuel

    International Nuclear Information System (INIS)

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

  1. Radioactivity of spent TRIGA fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-04-29

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

  2. Radioactivity of spent TRIGA fuel

    Science.gov (United States)

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

    2015-04-01

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

  3. Spent fuel surveillance and monitoring methods

    International Nuclear Information System (INIS)

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

  4. Data base for a CANDU-PHW operating on a once-through, natural uranium fuel cycle

    International Nuclear Information System (INIS)

    This report, prepared for INFCE, describes a standard 600 MW(e) CANDU-PHW reactor operating on a once-through natural uranium fuel cycle. Subsequently, data are given for an extrapolated 1000 MW(e) design (the nominal capacity adopted for the INFCE study) operating on the same fuel cycle. (author)

  5. TRIGA spent-fuel storage criticality analysis

    International Nuclear Information System (INIS)

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

  6. ITER SAFETY TASK NID-10A:CANDU occupational exposure experience: ORE for ITER fuel cycle and cooling systems

    International Nuclear Information System (INIS)

    This report contains information on TRITIUM Occupational Exposure (Internal Dose) from typical CANDU Nuclear Generating Stations. In addition to dose, airborne tritium levels are provided, as these strongly influence operational exposure. The exposure dose data presented in this report cover a period of five years of operation and maintenance experience from four CANDU Reactors and are considered representative of other CANDU reactors. The data are broken down according to occupational function ( Operators, Maintenance and Support Service etc.). The referenced systems are mainly centered on CANDU Hear Transport System, Moderator System, Tritium Removal Facility and Heavy Water (D20) Upgrading System. These systems contain the bulk part of tritium contamination in the CANDU Reactor. Because of certain similarities between ITER and CANDU systems, this data can be used as the most relevant TRITIUM OCCUPATIONAL DOSE information for ITER COOLING and FUEL CYCLE systems dose assessment purpose, if similar design and operation principles as described in the report are adopted. (author). 16 refs., 8 tabs., 13 figs

  7. Assessment of Welding System Modification of The Candu and PWR Fuel Element Types end Plug

    International Nuclear Information System (INIS)

    To anticipate future possibility of a nuclear fuel element industry in Indonesia, research on other types of nuclear fuel element beside Cirene type has to be done. It can be accomplished, one of them, by modifying the already available equipment. Based on the sheath material, the sheath dimension and the welding process parameters such as welding current and welding cycles, the available Magnetic Force Welding can be used for welding end plug of Candu nuclear fuel element by modifying some of its components (tube clamp, plug clamp, etc). The available Pellet drying and element filling furnace with its supporting system with includes helium gas filling, welding chamber, argon gas supply, vacuum system, sheath clamp and sheath driving system can be used for welding end plug with sheath of PWR nuclear fuel element by adding og Tungsten inert Gas (TIG) welding machine in the welding chamber and modifying a few components (seal clamp, sheath clamp)

  8. Spent nuclear fuel project integrated schedule plan

    Energy Technology Data Exchange (ETDEWEB)

    Squires, K.G.

    1995-03-06

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

  9. The CANDU irradiated fuel safeguards sealing system at the threshold of implementation

    International Nuclear Information System (INIS)

    The development of a safeguards containment and surveillance system for the irradiated fuel discharged from CANDU nuclear generating stations has inspired the development of three different sealing technologies. Each seal type utilizes a random seal identity of different design. The AECL Random Coil (ARC) Seal combines the identity and integrity elements in the ultrasonic signature of a wire coil. Two variants of an optical seal have been developed which features identity elements of crystalline zirconium and aluminum. The sealed cap-seal uses a conventional IAEA 'Type X Seal' (wire seal). The essential features and relative merits of each seal design are described

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

    International Nuclear Information System (INIS)

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

  11. Self-interrogation of spent fuel

    International Nuclear Information System (INIS)

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

  12. Spent Nuclear Fuel Storage Program user's guide

    International Nuclear Information System (INIS)

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

  13. Operation of spent fuel storage facilities

    International Nuclear Information System (INIS)

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

  14. Regional spent fuel storage facility (RSFSF)

    International Nuclear Information System (INIS)

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

  15. Foreign encapsulation concepts of spent fuel

    International Nuclear Information System (INIS)

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

  16. The cascad spent fuel dry storage facility

    International Nuclear Information System (INIS)

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

  17. Reactor BN-350 spent fuel handling

    International Nuclear Information System (INIS)

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

  18. Numerical Estimation of the Spent Fuel Ratio

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-01-01

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

  19. ASSERT/NUCIRC commissioning for CANDU 6 fuel channel CCP analysis

    International Nuclear Information System (INIS)

    CANDU PHWR fuel channel pressure tubes will expand or creep under long-term (aging process) influence of temperature, pressure, and neutron flux. This diametral pressure tube creep will influence the critical channel power (CCP), or conditions that lead to dryout. In order to provide safety analysis models to quantify the effect of diametral pressure tube creep on CCP, a COG (AECL/NBP/HQ) project is underway to commission the ASSERT and NUCIRC codes to establish reliable production tools for the assessment of CANDU6 CCP in nominal (uncrept) and crept pressure tube fuel channels. This paper gives an overview of the background and objectives of the project along with a brief introduction into the subchannel analysis code ASSERT and the 1-D thermalhydraulics code NUCIRC. This project is a multistage endeavour, for which the first stage results are presented. A detailed cross-comparison of the 1-D (NUCIRC) and subchannel (ASSERT) models of pressure drop (ΔP) and critical heat flux (CHF) has been undertaken and has led to several enhancements and refinements to the respective models. These results are presented in addition to results of ASSERT commissioning against NUCIRC for a matrix of ΔP and dryout cases in a nominal pressure tube, which are based upon Gentilly 2 and Point Lepreau site area. Additionally, the initial results of an assessment, using ASSERT, of the effects of creep on ΔP are presented. In concluding, the status and future directions for ASSERT/NUCIRC CANDU 6 CCP analysis project are summarized. (author). 2 refs., 12 figs

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

    International Nuclear Information System (INIS)

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

  1. The economics of advanced fuel cycles in CANDU (PHW) reactors

    International Nuclear Information System (INIS)

    The economic assessments of advanced fuel cycles performed within Ontario Hydro are collated and summarized. The results of the analyses are presented in a manner designed to provide a broad perspective of the economic issues regarding the advanced cycles. The enriched uranium fuel cycle is shown to be close to competitive at today's uranium prices, and its relative position vis-a-vis the natural uranium cycle will improve as uranium prices continue to rise. In the longer term, the plutonium-topped thorium cycle is identified as being the most economically desirable. It is suggested that this cycle may not be commercially attractive until the second or third decade of the next century. (auth)

  2. Evolution of procurement and supply conditions for CANDU fuels

    International Nuclear Information System (INIS)

    In 1955 a decision was made to proceed with construction of a Nuclear Power Demonstration Station (NPD) near Rolphton, Ontario. This project, headed by Atomic Energy of Canada with major involvement of private industry, was the genesis for the development of nuclear electric generation in Canada. This paper reviews one aspect of the Canadian program: the evolution of fuel procurement and supply, which in itself has been a remarkable Canadian achievement. (author)

  3. Spent fuel assembly source term parameters

    International Nuclear Information System (INIS)

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

  4. Burnup credit implementation in spent fuel management

    International Nuclear Information System (INIS)

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

  5. Historical overview of domestic spent fuel shipments

    International Nuclear Information System (INIS)

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

  6. Progress in developing an on-line fuel-failure monitoring tool for CANDU reactors

    International Nuclear Information System (INIS)

    This paper describes the continued development of an on-line defected fuel diagnostic tool for CANDU reactors. One of the key capabilities of this tool is the ability to estimate the power and number of defects in the core based on the Gaseous Fission Product Monitoring System (GFP), and grab sample data. To perform this analysis, a clear understanding of the empirical diffusion coefficient D' [s-1] is required. This paper examines two existing models for D' and presents a new model based on 133Xe release data from commercial reactor experience. The new model is successfully applied to commercial data to demonstrate a novel technique for extracting defected fuel element power from GFP data during a reactor power change. The on-line defected fuel diagnostic tool is in a developmental stage, and this paper reports the latest enhancements. (author)

  7. The E.D.F spent fuel

    International Nuclear Information System (INIS)

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

  8. Overview of Spent Fuel Management In China

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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)

  10. Stressmeter placement at spent fuel test in climax granite

    International Nuclear Information System (INIS)

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

  11. Report on interim storage of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    1993-04-01

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

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

    International Nuclear Information System (INIS)

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

  13. 77 FR 75065 - Rescinding Spent Fuel Pool Exclusion Regulations

    Science.gov (United States)

    2012-12-19

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

  14. Development of spent fuel remote handling technology

    International Nuclear Information System (INIS)

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

  15. Development of spent fuel remote handling technology

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-02-15

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

  16. LONGER: a computer program for longitudinal ridging and axial collapse assessment of CANDU fuel

    International Nuclear Information System (INIS)

    CANDU® fuel element sheath is designed to be thin and flexible for the benefit of enhanced heat transfer from the pellet to the coolant through the sheath. The flexibility of the sheath may allow the formation of longitudinal ridges on the sheath or collapse of the sheath into an axial gap under certain conditions. For both cases of deformations, the sheath may experience significant strains, and may result in sheath failure. To ensure the sheath mechanical integrity, the fuel element design needs to be assessed to preclude the conditions for longitudinal ridging and sheath collapse into the axial gap. The AECL developed LONGER computer program is used in fuel design analysis for such purpose. The LONGER code contains a number of models derived based on measurements (empirical models) and based on analytical equations, to predict the following parameters related to the deformations of CANDU nuclear fuel element sheaths. For longitudinal ridging: The critical diametral clearance for sheath longitudinal ridging, and The critical pressure for longitudinal ridging of the sheath. For axial collapse: The critical pressure for instantaneous sheath collapse into an axial gap. For circumferential collapse: The critical pressure for elastic collapse of the sheath, and The effective circumferential collapse pressure of the sheath by taking into account the axial and radial loads and the ovality of the sheath. The LONGER code has been qualified in accordance with the CSA standard N286.7-99 compliant AECL Software Quality Assurance (SQA) program. This paper describes the features and capabilities of the LONGER code that are used in CANDU fuel design analysis. (author)

  17. LONGER: a computer program for longitudinal ridging and axial collapse assessment of CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Paul, U.K.; Xu, Z.; Xu, S.; Wang, X.; Chakraborty, K. [Atomic Energy of Canada Limited, Mississauga (Canada)

    2010-07-01

    CANDU® fuel element sheath is designed to be thin and flexible for the benefit of enhanced heat transfer from the pellet to the coolant through the sheath. The flexibility of the sheath may allow the formation of longitudinal ridges on the sheath or collapse of the sheath into an axial gap under certain conditions. For both cases of deformations, the sheath may experience significant strains, and may result in sheath failure. To ensure the sheath mechanical integrity, the fuel element design needs to be assessed to preclude the conditions for longitudinal ridging and sheath collapse into the axial gap. The AECL developed LONGER computer program is used in fuel design analysis for such purpose. The LONGER code contains a number of models derived based on measurements (empirical models) and based on analytical equations, to predict the following parameters related to the deformations of CANDU nuclear fuel element sheaths. For longitudinal ridging: The critical diametral clearance for sheath longitudinal ridging, and The critical pressure for longitudinal ridging of the sheath. For axial collapse: The critical pressure for instantaneous sheath collapse into an axial gap. For circumferential collapse: The critical pressure for elastic collapse of the sheath, and The effective circumferential collapse pressure of the sheath by taking into account the axial and radial loads and the ovality of the sheath. The LONGER code has been qualified in accordance with the CSA standard N286.7-99 compliant AECL Software Quality Assurance (SQA) program. This paper describes the features and capabilities of the LONGER code that are used in CANDU fuel design analysis. (author)

  18. Semiautomatic spent-fuel-handling machine

    International Nuclear Information System (INIS)

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

  19. Benchmark calculations of a radiation heat transfer for a CANDU fuel channel analysis using the CFD code

    International Nuclear Information System (INIS)

    To justify the use of a commercial Computational Fluid Dynamics (CFD) code for a CANDU fuel channel analysis, especially for the radiation heat transfer dominant conditions, the CFX-10 code is tested against three benchmark problems which were used for the validation of a radiation heat transfer in the CANDU analysis code, a CATHENA. These three benchmark problems are representative of the CANDU fuel channel configurations from a simple geometry to a whole fuel channel geometry. For the solutions of the benchmark problems, the temperature or the net radiation heat flux boundary conditions are prescribed for each radiating surface to determine the radiation heat transfer rate or the surface temperature, respectively by using the network method. The Discrete Transfer Model (DTM) is used for the CFX-10 radiation model and its calculation results are compared with the solutions of the benchmark problems. The CFX-10 results for the three benchmark problems are in close agreement with those solutions, so it is concluded that the CFX-10 with a DTM radiation model can be applied to the CANDU fuel channel analysis where a surface radiation heat transfer is a dominant mode of the heat transfer. (author)

  20. International safeguards for spent fuel storage

    International Nuclear Information System (INIS)

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

  1. Cost analysis methodology of spent fuel storage

    International Nuclear Information System (INIS)

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

  2. Design of spent fuel storage facilities

    International Nuclear Information System (INIS)

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

  3. Radiological assessment of 36Cl in the disposal of used CANDU fuel

    International Nuclear Information System (INIS)

    An assessment of the potential radiological impact of 36Cl in the disposal of used CANDU fuel has been performed. The assessment was based on new data on chlorine impurity levels in used fuel. Data bases for the vault, geosphere, and biosphere models used in the EIS postclosure assessment case study (Goodwin et al. 1994) were modified to include the necessary 36Cl data. The resulting safety analysis shows that estimated radiological risks from 36Cl are forty times lower than from 129I at 104 a; this, incorporation of 36Cl into the models does not change the overall conclusions of the study of Goodwin et al. (1994a). For human intrusion scenarios, an analysis using the methodology of Wuschke (1992) showed that the maximum risk is unaffected by the inclusion of 36Cl. (author). 51 refs., 5 tabs., 15 figs

  4. The Conflux Fuel bundle: An Economic and Pragmatic Route to the use of Advanced Fuel Cycles in CANDU Reactors

    International Nuclear Information System (INIS)

    The CANFLEX1 bundle is being developed jointly by AECL and KAERI as a vehicle for introducing the use of enrichment and advanced fuel cycles in CANDU2 reactors. The bundle design uses smaller diameter fuel elements in the outer ring of a 43-element bundle to reduce the maximum element ratings in a CANDU fuel bundle by 20% compared to the 37-element bundle currently in use. This facilitates burnups of greater than 21,000 MW d/TAU to optimize the economic benefit available from the use of enrichment and advanced fuel cycles. A combination of this lower fuel rating, plus development work underway at Aecl to enhance the thermalhydraulic characteristics of the bundle (including both CHF3 and bundle. This provides extra flexibility in the fuel management procedures required for fuel bundles with higher fissile contents. The different bundle geometry requires flow tests to demonstrate acceptable vibration and fretting behavior of the Conflux bundle. A program to undertake the necessary range of flow tests has started at KAERI, involving the fabrication of the required bundles, and setting up for the actual tests. A program to study the fuel management requirements for slightly enriched (0.9 wt % 235 in total U) Conflux fuel has been undertaken by both Aecl and KAERI staff, and further work has started for higher enrichments. Irradiation testing of the Conflux bundle started in the NUR reactor in 1989, and a second irradiation test is due to start shortly. This paper describes the program, and reviews the status of key parts of the program

  5. Management and disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

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

  6. Wet spent fuel interim storage facility

    International Nuclear Information System (INIS)

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

  7. Safety assessment for spent fuel storage facilities

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-06-01

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

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

    International Nuclear Information System (INIS)

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

  10. Remote technologies for handling spent fuel

    International Nuclear Information System (INIS)

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

  11. Considerations for the transportation of spent fuel

    International Nuclear Information System (INIS)

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

  12. Spent Fuel Working Group Report. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    O`Toole, T.

    1993-11-01

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

  13. Spent Fuel Dry Storage Cask Thermal Test

    International Nuclear Information System (INIS)

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

  14. Optimization of the self-sufficient thorium fuel cycle for CANDU power reactors

    Directory of Open Access Journals (Sweden)

    Bergelson Boris R.

    2008-01-01

    Full Text Available The results of optimization calculations for CANDU reactors operating in the thorium cycle are presented in this paper. Calculations were performed to validate the feasibility of operating a heavy-water thermal neutron power reactor in a self-sufficient thorium cycle. Two modes of operation were considered in the paper: the mode of preliminary accumulation of 233U in the reactor itself and the mode of operation in a self-sufficient cycle. For the mode of accumulation of 233U, it was assumed that enriched uranium or plutonium was used as additional fissile material to provide neutrons for 233U production. In the self-sufficient mode of operation, the mass and isotopic composition of heavy nuclei unloaded from the reactor should provide (after the removal of fission products the value of the multiplication factor of the cell in the following cycle K>1. Additionally, the task was to determine the geometry and composition of the cell for an acceptable burn up of 233U. The results obtained demonstrate that the realization of a self-sufficient thorium mode for a CANDU reactor is possible without using new technologies. The main features of the reactor ensuring a self-sufficient mode of operation are a good neutron balance and moving of fuel through the active core.

  15. 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.

  16. Status and trends in spent fuel reprocessing

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  18. TRIGA spent fuel bundles safe storage

    International Nuclear Information System (INIS)

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

  19. Radionuclide release from research reactor spent fuel

    International Nuclear Information System (INIS)

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

  20. Selection of instruments used for vibration measurement of fuel bundles in a pressure tube under CANDU reactor operating conditions

    International Nuclear Information System (INIS)

    Vibration characteristics of CANDU fuel bundle and fuel elements is a key parameter considered in the design of a fuel bundle. Out-reactor frequency and temperature sweep tests, under reactor operating conditions, are performed to verify vibration characteristics of CANDU fuel bundles. Several options have been considered in the selection of vibration instrumentation to perform out-reactor frequency and temperature sweep tests. This paper compares the benefits and disadvantages of various vibration instruments and summarizes the rationale behind the selection of instruments used for vibration measurements over a range of temperature and pressure pulsation frequencies. The conclusions are presented from the bench tests performed, which confirm the use of the selected instruments. (author)

  1. Release of 14C from the gap and grain-boundary regions of used CANDU fuels to aqueous solutions

    International Nuclear Information System (INIS)

    This study was undertaken as part of the Canadian Nuclear Fuel Waste Management Program (CNFWMP), to measure 14C inventories of used CANDU fuel. Other objectives were to measure the fraction of the total 14C inventory that would be instantly released to solution from used CANDU fuels upon sheath failure and to determine if the assumptions made in safety assessment calculations of used fuel waste disposal regarding instant release of 14C were correct. Results showed that the measured 14C inventories were a factor of 11.5 ± 3.9 lower than the estimated 14C inventory values used in safety assessment calculations. Measured instant release values for 14C ranged from 0.06 to 5.04% (of total 14C inventories) with an average of 2.7 ± 1.6%, indicating that instant release fractions for 14C used in safety assessment calculations (1.2--25%) were overestimated

  2. Spent Nuclear Fuel Vibration Integrity Study

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-01-01

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

  3. Nondestructive verification and assay systems for spent fuels

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  5. Advanced waste forms from spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-31

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

  6. Assessment of the performance of used CANDU fuel under disposal conditions

    International Nuclear Information System (INIS)

    Results of the work conducted since 1991 on determining gap and/or grain-boundary inventories for several important radionuclides such as 137Cs , 129I, 14C, 90Sr , 99Tc and 36Cl in used CANDU fuel and investigation of effect of parameters such as fuel power and burnup on their release rates is summarized in the report. Since the great majority of radionuclides are contained within the grains of the fuel pellets, the long-term release rate is governed by the dissolution rate of the uranium oxide matrix. Although UO2 is highly insoluble, the solubility of uranium increases by many orders of magnitude under oxidizing conditions. The rate of UO2 dissolution, and thus release of fission products from the fuel, is most sensitive to vault redox conditions, radiation field, groundwater composition and temperature and these factors have been investigated if justifiable assurances are to be given that radionuclide releases from a waste vault will be very limited. The redox conditions within a waste vault will evolve with time from initially oxidizing to eventually non-oxidizing as oxygen, trapped within the vault on sealing, is consumed and radiation fields, which can produce oxidants by the radiolysis of water, decay. Effective containment of the fuel should prevent its contact with groundwater until this redox evolution is complete. (author)

  7. Operation of CANDU power reactor in thorium self-sufficient fuel cycle

    Indian Academy of Sciences (India)

    B R Bergelson; A S Gerasimov; G V Tikhomirov

    2007-02-01

    This paper presents the results of calculations for CANDU reactor operation in thorium fuel cycle. Calculations are performed to estimate the feasibility of operation of heavy-water thermal neutron power reactor in self-sufficient thorium cycle. Parameters of active core and scheme of fuel reloading were considered to be the same as for standard operation in uranium cycle. Two modes of operations are discussed in the paper: mode of preliminary accumulation of 233U and mode of operation in self-sufficient cycle. For the mode of accumulation of 233U it was assumed for calculations that plutonium can be used as additional fissile material to provide neutrons for 233U production. Plutonium was placed in fuel channels, while 232Th was located in target channels. Maximum content of 233U in target channels was estimated to be ∼ 13 kg/t of ThO2. This was achieved by irradiation for six years. The start of the reactor operation in the self-sufficient mode requires 233U content to be not less than 12 kg/t. For the mode of operation in self-sufficient cycle, it was assumed that all channels were loaded with identical fuel assemblies containing ThO2 and certain amount of 233U. It is shown that nonuniform distribution of 233U in fuel assembly is preferable.

  8. The mode of operation of CANDU power reactor in thorium self-sufficient fuel cycle

    Directory of Open Access Journals (Sweden)

    Bergelson Boris R.

    2008-01-01

    Full Text Available This paper presents the results of calculations for CANDU reactor operation in the thorium fuel cycle. The calculations were performed to estimate feasibility of operation of a heavy-water thermal neutron power reactor in the self-sufficient thorium cycle. The parameters of the active core and the scheme of fuel reloading were considered to be the same as for the standard operation in the uranium cycle. Two modes of operation are discussed in the paper: the mode of preliminary accumulation of 233U and the mode of operation in the self-sufficient cycle. For calculations for the mode of accumulation of 233U, it was assumed that plutonium was used as the additional fissile material to provide neutrons for 233U production. Plutonium was placed in fuel channels, while 232Th was located in target channels. The maximum content of 233U in the target channels was about 13 kg/t of ThO2. This was achieved by six year irradiation. The start of reactor operation in the self-sufficient mode requires content of 233U not less than 12 kg/t. For the mode of operation in the self-sufficient cycle, it was assumed that all the channels were loaded with the identical fuel assemblies containing ThO2 and a certain amount of 233U. It was shown that the non-uniform distribution of 233U in a fuel assembly is preferable.

  9. Depleting a CANDU-6 fuel assembly using detailed burnup data and reactionwise energy release

    International Nuclear Information System (INIS)

    Temporal behavior of reactor fuel assembly due to neutron exposure is an integral part of lattice analysis. It is important to estimate the production of actinides and fission products as a function of burnup so as to decide the quality of fuel for further energy production. It is also important from the point of view of post irradiation behavior of fuel. The information on heat production during and after irradiation helps in determining the amount of time a fuel assembly needs to be cooled before taking it up for storage or reprocessing. In the present study we have considered the CANDU-6 fuel assembly as reference. Lattice analysis has been performed using development version of code DRAGON. A total of 192 nuclides have been selected as part of the analysis, of which 19 are actinides, 151 are fission products and the rest are structural elements. The fission products have been treated explicitly. There is no pseudo fission product. Using DRAGR module, a multigroup microscopic cross section library in DRAGLIB format has been generated. An important aspect of this library is the explicit treatment of most neutron induced reactions. We have for the first time attempted to perform power normalization due to energy from various neutron induced reactions including (n, γ), (n, f), (n, 2n), (n, 3n), (n, 4n), (n, α), (n, p), (n, 2α), (n, np), (n, d), (n, t). Energy due to decay has also been considered explicitly. Even though the decay energy contributes very little relative to the neutron induced reactions, the information will be very useful for post irradiation behavior of fuel. It was observed that the maximum contributing reactions for the power normalization are (n, f), (n, γ) and (n, 2n). We have assessed the contribution of fission products and actinides towards power normalization as a function of burnup. We have also studied the pinwise contribution towards power normalization in each ring of CANDU-6 fuel. We have attempted to compare the effect of

  10. Exporting technology for CANDU fuel manufacturing to the People's Republic of China - a stimulating experience for the Romanian nuclear fuel plant

    International Nuclear Information System (INIS)

    Adopting CANDU type reactors to produce nuclear-generated electricity, Romania has also developed his capability to produce nuclear fuel. Since 1995, FCN Pitesti is the unique nuclear fuel supplier for Cernavoda CANDU Power Station. Fuel plant upgrading and qualification was achieved in co-operation with AECL and Zircatec Precision Industries Inc. The fuel bundles manufactured at FCN Pitesti proved to be of excellent quality, operating with a very low defect rate, all defected fuel being reported in the first period of the reactor operation. It is a fact now that FCN has the capability to solve a wide variety of aspects one of the most significant being the development of new equipment and the increase of the capacity in order to cover the future nuclear fuel needs. On this basis FCN was invited to contribute with his potential to a supplying contract with China National Nuclear Corporation - 202 Plant, for CANDU nuclear fuel technology. Following an offer including several categories of equipment and technology, the option was for beryllium coaters and coating technology and training for end cap manufacturing. The arrangements consider Romanian company as a sub-supplier, this option ensuring the consistence with the largest part of the supply for CANDU fuel technology, offered by Zircatec. Two pieces of beryllium coaters have been produced and tested in Romania and the operating demonstration was made in the presence of Zircatec staff and Chinese delegates. The Chinese delegated were trained for complete operating modes and their ability to handle the equipment was certified accordingly. They also have been trained in the end cap technology and related quality inspection. The paper includes a short presentation of the equipment and associated work to fit the specified needs. The involvement of the Romanian fuel plant in this contract could be considered as an extension of the previous co-operation with the Canadian partners on CANDU nuclear fuel and finally

  11. Path forward of spent fuel management in the world

    International Nuclear Information System (INIS)

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

  12. Calibration of spent fuel measurement assembly

    International Nuclear Information System (INIS)

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

  13. Systems impacts of spent fuel disassembly alternatives

    International Nuclear Information System (INIS)

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

  14. Systems impacts of spent fuel disassembly alternatives

    Energy Technology Data Exchange (ETDEWEB)

    1984-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-02-01

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

  17. Optimizing in-bay fuel inspection capability to meet the needs of today's CANDU fleet

    International Nuclear Information System (INIS)

    With the recent return to service of many CANDU units, aging of all others, increasingly competitive energy market and aging hot cell infrastructure - there exists now a greater need for timely, cost-effective and reliable collection of irradiated fuel performance information from fuel bay inspections. The recent development of simple in-bay tools, used in combination with standardized technical specifications, inspection databases and assessment techniques, allows utilities to characterize the condition of irradiated fuel and any debris lodged in the bundle in a more timely fashion and more economically than ever. Use of these tools and 'advanced' techniques permits timely engineering review and disposition of emerging issues to support reliable operation of the CANDU fleet. (author)

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

    International Nuclear Information System (INIS)

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

  19. Investigation of the grain-boundary chemistry in used CANDU fuel by x-ray photoelectron spectroscopy (XPS)

    International Nuclear Information System (INIS)

    The grain-boundary chemistry of used CANDU fuel is being systematically investigated by X-ray photoelectron spectroscopy (XPS) using a McPherson ESCA-36 instrument that has been adapted for routine studies of highly radioactive materials. Initial stages of fuel corrosion under various storage and disposal conditions can be identified from chemical-shift effects for uranium. For example, pervasive but highly selective grain-boundary oxidation has been revealed in CANDU fuels exposed to moist air at 150 deg. C for extended periods, suggesting aggressive radiolytic processes operating in a thin film of adsorbed water. Pronounced segregation of a number of fission products to cracks and grain boundaries in used CANDU fuels has been explicitly demonstrated by XPS as well. Model calculations and composition depth profiles are indicative of near monolayer films. Some correlations between fuel power history and fission-product distributions have been established and possible evidence of migration during moist-air exposure has been obtained. The key advantages and limitations of XPS in this context are discussed and illustrated with selected results. (author). 23 refs, 8 figs, 1 tab

  20. Overview of the United States spent nuclear fuel program

    International Nuclear Information System (INIS)

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

  1. 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.

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

    International Nuclear Information System (INIS)

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

  3. Potential Interface Issues in Spent Fuel Management

    International Nuclear Information System (INIS)

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

  4. Spent fuel management: reprocessing or storage

    International Nuclear Information System (INIS)

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

  5. Development of Spent Fuel Examination Technology

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-04-15

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

  6. Antineutrino monitoring of spent nuclear fuel

    CERN Document Server

    Brdar, Vedran; Kopp, Joachim

    2016-01-01

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

  7. Buckling analysis of spent fuel basket

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  9. Transportation and storage of foreign spent power reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    1979-09-30

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

  10. Transportation and storage of foreign spent power reactor fuel

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-07-01

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

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

    International Nuclear Information System (INIS)

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

  13. Technical bases for interim storage of spent nuclear fuel

    International Nuclear Information System (INIS)

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

  14. Management of spent fuel in the Slovak Republic

    International Nuclear Information System (INIS)

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

  15. Micro-focus x-ray inspection of the bearing pad welded by laser for CANDU fuel element

    International Nuclear Information System (INIS)

    To attach the bearing pads on the surface of CANDU fuel element, laser welding technique has been reviewed to replace brazing technology which is complicate process and makes use of the toxic beryllium. In this study, to evaluate the soundness of the weld of the bearing pad of CANDU fuel element, a precise X-ray inspection system was developed using a micro-focus X-ray generator with an image intensifier and a real time camera system. The weld of the bearing pad welded by Nd:YAG laser has been inspected by the developed inspection system. Image processing technique has been applied to reduce random noise and to enhance the contrast of the X-ray image. A few defects on the weld of the bearing pads have been detected by the X-ray inspection process

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-03-01

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

  17. Survey of wet and dry spent fuel storage

    International Nuclear Information System (INIS)

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

  18. Nuclear Spent Fuel Management in Spain

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

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

    International Nuclear Information System (INIS)

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

  2. 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

  3. Explicit core-follow simulations for a CANDU 6 reactor fuelled with recovered-uranium CANFLEX bundles

    International Nuclear Information System (INIS)

    Recovered uranium (RU) is a by-product of many light-water reactor (LWR) fuel recycling programs. After fission products and plutonium (Pu) have been removed from spent LWR fuel, RU is left. A fissile content in the RU of 0.9 to 1.0% makes it impossible for reuse in an LWR without re-enrichment, but CANDU reactors have a sufficiently high neutron economy to use RU as fuel. Explicit core-follow simulations were run to analyse the viability of RU as a fuel for existing CANDU 6 cores. The core follow was performed with RFSP, using WIMS-AECL lattice properties. During the core follow, channel powers and bundle powers were tracked to determine the operating envelope for RU in a CANFLEX bundle. The results show that RU fits the operating criteria of a generic CANDU 6 core and is a viable fuel option in CANDU reactors. (author)

  4. Storage of Spent Nuclear Fuel. Specific Safety Guide

    International Nuclear Information System (INIS)

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

  5. Investigation of coolant thermal mixing within 28-element CANDU fuel bundles using the ASSERT-PV thermal hydraulics code

    International Nuclear Information System (INIS)

    This paper presents the results of a study of the thermal mixing of single-phase coolant in 28-element CANDU fuel bundles under steady-state conditions. The study, which is based on simulations performed using the ASSERT-PV thermal hydraulic code, consists of two main parts. In the first part the various physical mechanisms that contribute to coolant mixing are identified and their impact is isolated via ASSERT-PV simulations. The second part is concerned with development of a preliminary model suitable for use in the fuel and fuel channel code FACTAR to predict the thermal mixing that occurs between flow annuli. (author)

  6. Investigation of the CANLUB/sheath interface in CANDU fuel at extended burnup by XPS and SEM/WDX

    Energy Technology Data Exchange (ETDEWEB)

    Hocking, W.H. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Behnke, R.; Duclos, A.M.; Gerwing, A.F. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada); Chan, P.K. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)

    1997-07-01

    A systematic investigation of the fuel-sheath interface in CANDU fuel as a function of extended burnup has been undertaken by XPS and SEM/WDX analysis. Adherent deposits of UO{sub 2} and fission products, including Cs, Ba, Rb, I, Te, Cd and possibly Ru, have been routinely identified on CANLUB coated and bare Zircaloy surfaces. Some trends in the distribution and chemistry of key fission products have begun to emerge. Several potential mechanisms for degradation of the CANLUB graphite layer at high burnup have been practically excluded. New evidence of carbon relocation within the fuel element and limited reaction with excess oxygen has also been obtained. (author)

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

    International Nuclear Information System (INIS)

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

  8. Analytical methodology and facility description spent fuel policy

    International Nuclear Information System (INIS)

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

  9. Analytical methodology and facility description spent fuel policy

    Energy Technology Data Exchange (ETDEWEB)

    1978-08-01

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

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

    International Nuclear Information System (INIS)

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

  11. Romanian-Canadian joint program for qualification of FCN as a CANDU fuel supplier

    International Nuclear Information System (INIS)

    RENEL (Romania Power Authority), the co-ordinator of Romanian Nuclear Program, have decided to improve, starting 1990 the existing capability to produce CANDU nuclear fuel at FCN Pitesti. The objective of the program was defined with AAC (AECL - ANSALDO Consortium) for the qualification of FCN fuel plant according to Canadian Z299.2 standard. The Qualification Program was performed under AAC Work Order C-003. The co-ordination was assumed by AECL, as overall Design Authority. ZPI (Zircatec Precision Industries Inc., Canada), were designated to supply technical assistance, equipments and know how where necessary. After a preliminary verification of the FCN fuel plant, including the processes and system investigation, performed under AECL and ZPI assistance, the Qualification Program was defined in all details. The upgrading of documentation on all aspects required by Z299.2 was performed. Few processes needed to be reconsidered and equipment was delivered by ZPI or other suppliers. This includes mainly welding equipments and special inspection equipments. Health Physics was practically fully reconsidered. New equipment and practice were adapted to provide adequate control on health conditions. Every manufacturing and inspection process was checked to determine their performance during a Qualification Run based on acceptance criteria which have been established in the Qualification Plan. Manufacturing Demonstration Run was an important step to prove that all plant functions have been accomplished during the fabrication of 200 fuel bundles. These bundles have been fully accepted and 66 of them have been loaded in the first charge of Unit 1 Cemavoda NPS. The surveillance and audit actions made by AECL and ZPI during this period confirmed the FCN capability to operate an adequate system meeting the to required quality assurance standard. The very open attitude of AECL, Zircatec and FCN staff have stimulated the progress of the project and a successful achievement of the

  12. Monitoring instrumentation spent fuel management program. Final report

    International Nuclear Information System (INIS)

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

  13. Radwaste management and spent fuel management in JAVYS

    International Nuclear Information System (INIS)

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

  14. An approach to meeting the spent fuel standard

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-05-01

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

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

    International Nuclear Information System (INIS)

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

  16. 77 FR 28406 - Spent Fuel Transportation Risk Assessment

    Science.gov (United States)

    2012-05-14

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

  17. Spent-fuel photon and neutron source spectra

    International Nuclear Information System (INIS)

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

  18. 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

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

    International Nuclear Information System (INIS)

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

  20. Spent nuclear fuel project quality assurance program plan

    Energy Technology Data Exchange (ETDEWEB)

    Lacey, R.E.

    1997-05-09

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1988-12-01

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

  2. CANDU 9 design

    International Nuclear Information System (INIS)

    AECL has made significant design improvements in the latest CANDU nuclear power plant (NPP) - the CANDU 9. The CANDU 9 operates with the energy efficient heavy water moderated reactor and natural uranium fuel and utilizes proven technology. The CANDU 9 NPP design is similar to the world leading CANDU 6 but is based upon the single unit adaptation of the 900 MWe class reactors currently operating in Canada as in integrated four-unit configurations. The evolution of the CANDU family of heavy water reactors (HAIR) is based on a continuous product improvement approach. Proven equipment and systems from operating stations are standardized and used in new products. As a result of the flexibility of the technology, evolution of the current design will ensure that any new requirements can be met, and there is no need to change the basic concept. This paper will provide an overview for some of the key features of the CANDU 9 NPP such as nuclear systems and equipment, advanced control and computer systems, safety design and protection features, and plant layout. The safety enhancements and operability improvements implemented in this design are described and some of the advantages that can be expected by the operating utility are highlighted. (author)

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

    International Nuclear Information System (INIS)

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

  4. The TVO concept for direct disposal of spent fuel

    International Nuclear Information System (INIS)

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

  5. Assessment of the storage concept for conditioned spent fuel

    International Nuclear Information System (INIS)

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

  6. Spent fuel management: Current status and prospects 1993

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  8. Spent fuel management: Current status and prospects 1990

    International Nuclear Information System (INIS)

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

  9. MTR radiological database for SRS spent nuclear fuel facilities

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  11. Spent fuel data base: commercial light water reactors

    International Nuclear Information System (INIS)

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

  12. Status of spent fuels in Japanese research reactors

    International Nuclear Information System (INIS)

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

  13. Volatile fission product distributions in LWR spent fuel rods

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-12-01

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

  15. Packaging of spent fuel elements into special containers

    International Nuclear Information System (INIS)

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

  16. Overview of methods to increase dryout power in CANDU fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Groeneveld, D.C., E-mail: degroeneveld@gmail.com [Chalk River Laboratories, AECL, Chalk River (Canada); University of Ottawa, Department of Mechanical Engineering, Ottawa (Canada); Leung, L.K.H. [Chalk River Laboratories, AECL, Chalk River (Canada); Park, J.H. [Korean Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-06-15

    Highlights: • Small changes in bundle geometry can have noticeable effects on the bundle CHF. • Rod spacing devices can results in increases of over 200% in CHF. • CHF enhancement decays exponentially downstream from spacers. • CHF-enhancing bundle appendages also increase the post-CHF heat transfer. - Abstract: In CANDU reactors some degradation in the CCP (critical channel power, or power corresponding to the first occurrence of CHF in any fuel channel) will occur with time because of ageing effects such as pressure-tube diametral creep, increase in reactor inlet-header temperature, increased hydraulic resistance of feeders. To compensate for the ageing effects, various options for recovering the loss in CCP are described in this paper. They include: (i) increasing the bundle heated perimeter, (ii) optimizing the bundle configuration, (iii) optimizing core flow and flux distribution, (iv) reducing the bundle hydraulic resistance, (v) use of CHF-enhancing bundle appendages, (vi) more precise experimentation, and (vii) redefining CHF. The increase in CHF power has been quantified based on experiments on full-scale bundles and subchannel code predictions. The application of several of these CHF enhancement principles has been used in the development of the 43-rod CANFLEX bundle.

  17. An elasto-plastic model for mechanical contact between the pellets and sheath in CANDU nuclear fuel elements

    International Nuclear Information System (INIS)

    During high-temperature transients, increased mechanical contact can occur between the fuel stack and the sheath in the axial and/or radial direction. As well, there is an axial linear power gradient and an axial gradient in mechanical properties of the sheath specific to a CANDU-type fuel element. This requires a code with the capability to treat multiple axial segments. This paper describes a contact model that allows the elasto-plastic mechanical contact in radial/axial direction for multiple axial segments. (14 figs., 5 refs.)

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

    International Nuclear Information System (INIS)

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

  19. 75 FR 9452 - Solicitation of Topics for Discussion at a Spent Fuel Storage and Transportation Licensing...

    Science.gov (United States)

    2010-03-02

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

  20. Safety and Licensing of Spent Fuel Storage Facilities

    International Nuclear Information System (INIS)

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

  1. The U.S.-Russian joint studies on using power reactors to disposition surplus weapons plutonium as spent fuel

    International Nuclear Information System (INIS)

    In 1996, the US and the Russian Federation completed an initial joint study of the candidate options for the disposition of surplus weapons plutonium in both countries. The options included long term storage, immobilization of the plutonium in glass or ceramic for geologic disposal, and the conversion of weapons plutonium to spent fuel in power reactors. For the latter option, the US is only considering the use of existing light water reactors (LWRs) with no new reactor construction for plutonium disposition, or the use of Canadian deuterium uranium (CANDU) heavy water reactors. While Russia advocates building new reactors, the cost is high, and the continuing joint study of the Russian options is considering only the use of existing VVER-1000 LWRs in Russia and possibly Ukraine, the existing BN-60O fast neutron reactor at the Beloyarsk Nuclear Power Plant in Russia, or the use of the Canadian CANDU reactors. Six of the seven existing VVER-1000 reactors in Russia and the eleven VVER-1000 reactors in Ukraine are all of recent vintage and can be converted to use partial MOX cores. These existing VVER-1000 reactors are capable of converting almost 300 kg of surplus weapons plutonium to spent fuel each year with minimum nuclear power plant modifications. Higher core loads may be achievable in future years

  2. The U.S.-Russian joint studies on using power reactors to disposition surplus weapon plutonium as spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Chebeskov, A.; Kalashnikov, A. [State Scientific Center, Obninsk (Russian Federation). Inst. of Physics and Power Engineering; Bevard, B.; Moses, D. [Oak Ridge National Lab., TN (United States); Pavlovichev, A. [State Scientific Center, Moscow (Russian Federation). Kurchatov Inst.

    1997-09-01

    In 1996, the US and the Russian Federation completed an initial joint study of the candidate options for the disposition of surplus weapons plutonium in both countries. The options included long term storage, immobilization of the plutonium in glass or ceramic for geologic disposal, and the conversion of weapons plutonium to spent fuel in power reactors. For the latter option, the US is only considering the use of existing light water reactors (LWRs) with no new reactor construction for plutonium disposition, or the use of Canadian deuterium uranium (CANDU) heavy water reactors. While Russia advocates building new reactors, the cost is high, and the continuing joint study of the Russian options is considering only the use of existing VVER-1000 LWRs in Russia and possibly Ukraine, the existing BN-60O fast neutron reactor at the Beloyarsk Nuclear Power Plant in Russia, or the use of the Canadian CANDU reactors. Six of the seven existing VVER-1000 reactors in Russia and the eleven VVER-1000 reactors in Ukraine are all of recent vintage and can be converted to use partial MOX cores. These existing VVER-1000 reactors are capable of converting almost 300 kg of surplus weapons plutonium to spent fuel each year with minimum nuclear power plant modifications. Higher core loads may be achievable in future years.

  3. Regeneration of ammonia borane spent fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-02-01

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

  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. Thermal Cooling Limits of Sabotaged Spent Fuel Pools

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1978-05-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-15

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

  11. Incentives for improvement of CANDU

    International Nuclear Information System (INIS)

    CANDU is a relatively young technology which has demonstrated many achievements as an electrical power generation system. These achievements include an unsurpassed safety record, high annual and lifetime capacity factors, low electricity cost and a broad range of other performance strengths which together indicate that the CANDU technology is fundamentally sound. Known capabilities not yet fully exploited, such as advanced fuel cycle options, indicate that CANDU technology will continue to pay strong dividends on research, development and design investment. This provides a strong incentive for the improvement of CANDU on a continuing basis

  12. Status of dry storage of spent nuclear fuel

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  14. Optimization of thorium-uranium content in a 54-element fuel bundle for use in a CANDU-SCWR

    International Nuclear Information System (INIS)

    A new 54-element fuel bundle design has been proposed for use in a pressure-tube supercritical water-cooled reactor, a pre-conceptual evolution of existing CANDU reactors. Pursuant to the goals of the Generation IV International Forum regarding advancement in nuclear fuel cycles, optimization of the thorium and uranium content in each ring of fuel elements has been studied with the objectives of maximizing the achievable fuel utilization (burnup) and total thorium content within the bundle, while simultaneously minimizing the linear element ratings and coolant void reactivity. The bundle was modeled within a reactor lattice cell using WIMS-AECL, and the uranium and thorium content in each ring of fuel elements was optimized using a weighted merit function of the aforementioned criteria and a metaheuristic search algorithm. (author)

  15. Behaviour of spent fuel assemblies during extended storage

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  17. Transporting spent reactor fuel: allegations and responses

    International Nuclear Information System (INIS)

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

  18. Development of Advanced Spent Fuel Management Process

    International Nuclear Information System (INIS)

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

  19. Development of Advanced Spent Fuel Management Process

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-06-15

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

  20. Existing Condition Analysis of Dry Spent Fuel Storage Technology

    Institute of Scientific and Technical Information of China (English)

    LI Ning; XU Lan; HAO Jian-sheng

    2016-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  4. LWR spent fuel storage technology: Advances and experience

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  6. Development of information management system on LWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-10-01

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

  7. Analysis of Fuel Temperature Reactivity Coefficients According to Burn-up and Pu-239 Production in CANDU Reactor

    International Nuclear Information System (INIS)

    The resonances for some kinds of nuclides such as U-238 and Pu-239 are not easy to be accurately processed. In addition, the Pu-239 productions from burnup are also significant in CANDU, where the natural uranium is used as a fuel. In this study, the FTCs were analyzed from the viewpoints of the resonance self-shielding methodology and Pu-239 build-up. The lattice burnup calculations were performed using the TRITON module in the SCALE6 code system, and the BONAMI module was executed to obtain self-shielded cross sections using the Bondarenko approach. Two libraries, ENDF/B-VI.8 and ENDF/B-VII.0, were used to compare the Pu-239 effect on FTC, since the ENDF/B-VII has updated the Pu-239 cross section data. The FTCs of the CANDU reactor were newly analyzed using the TRITON module in the SCALE6 code system, and the BONAMI module was executed to apply the Bondarenko approach for self-shielded cross sections. When compared with some reactor physics codes resulting in slightly positive FTC in the specific region, the FTCs evaluated in this study showed a clear negativity over the entire fuel temperature range on fresh/equilibrium fuel. In addition, the FTCs at 960.15 K were slightly negative during the entire burnup. The effects on FTCs from the library difference between ENDF/B-VI.8 and ENDF/B-VII.0 are recognized to not be large; however, they appear more positive when more Pu-239 productions with burnup are considered. This feasibility study needs an additional benchmark evaluation for FTC calculations, but it can be used as a reference for a new FTC analysis in CANDU reactors

  8. Scenarios for the transmutation of actinides in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hyland, Bronwyn, E-mail: hylandb@aecl.ca [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada); Gihm, Brian, E-mail: gihmb@aecl.ca [Atomic Energy of Canada Limited, 2251 Speakman Drive, Mississauga, Ontario, L5K 1B2 (Canada)

    2011-12-15

    With world stockpiles of used nuclear fuel increasing, the need to address the long-term utilization of this resource is being studied. Many of the transuranic (TRU) actinides in nuclear spent fuel produce decay heat for long durations, resulting in significant nuclear waste management challenges. These actinides can be transmuted to shorter-lived isotopes to reduce the decay heat period or consumed as fuel in a CANDU(R) reactor. Many of the design features of the CANDU reactor make it uniquely adaptable to actinide transmutation. The small, simple fuel bundle simplifies the fabrication and handling of active fuels. Online refuelling allows precise management of core reactivity and separate insertion of the actinides and fuel bundles into the core. The high neutron economy of the CANDU reactor results in high TRU destruction to fissile-loading ratio. This paper provides a summary of actinide transmutation schemes that have been studied in CANDU reactors at AECL, including the works performed in the past. The schemes studied include homogeneous scenarios in which actinides are uniformly distributed in all fuel bundles in the reactor, as well as heterogeneous scenarios in which dedicated channels in the reactor are loaded with actinide targets and the rest of the reactor is loaded with fuel. The transmutation schemes that are presented reflect several different partitioning schemes. Separation of americium, often with curium, from the other actinides enables targeted destruction of americium, which is a main contributor to the decay heat 100-1000 years after discharge from the reactor. Another scheme is group-extracted transuranic elements, in which all of the transuranic elements, plutonium (Pu), neptunium (Np), americium (Am), and curium (Cm) are extracted together and then transmuted. This paper also addresses ways of utilizing the recycled uranium, another stream from the separation of spent nuclear fuel, in order to drive the transmutation of other actinides.

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

    Gwynne, Peter

    2016-07-01

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

  11. Microbiology of spent nuclear fuel storage basins.

    Science.gov (United States)

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

    1998-12-01

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

  12. Microbiology of spent nuclear fuel storage basins.

    Science.gov (United States)

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

    1998-12-01

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

  13. An integrated CANDU system

    International Nuclear Information System (INIS)

    Twenty years of experience have shown that the early choices of heavy water as moderator and natural uranium as fuel imposed a discipline on CANDU design that has led to outstanding performance. The integrated structure of the industry in Canada, incorporating development, design, supply, manufacturing, and operation functions, has reinforced this performance and has provided a basis on which to continue development in the future. These same fundamental characteristics of the CANDU program open up propsects for further improvements in economy and resource utilization through increased reactor size and the development of the thorium fuel cycle

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Dana, W.P.

    1995-12-01

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

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

    International Nuclear Information System (INIS)

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

  17. Public Opinion Surveys in Spent Nuclear Fuel Management

    OpenAIRE

    Vasilieva, E.

    2002-01-01

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

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

    OpenAIRE

    Sihm Kvenangen, Karen

    2007-01-01

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

  19. Economic and Innovative Spent Fuel Pool Level Instrumentation

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-07-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1988-03-01

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

  3. The clearance potential index and hazard factors of CANDU fuel bundle and a comparison of experimental-calculated inventories

    International Nuclear Information System (INIS)

    In the field of radioactive waste management, the radiotoxicity can be characterized by two different approaches: 1) IAEA, 2004 RS-G-1.7 clearance concept and 2) US, 10CFR20 radioactivity concentration guides in terms of ingestion / inhalation hazard expressed in m3 of water/air. A comparison between the two existing safety concepts was made in the paper. The modeled case was a CANDU natural uranium, 37 elements fuel bundle with a reference burnup of 685 GJ/kgU (7928.24 MWd/tU). The radiotoxicity of the light nuclide inventories, actinide, and fission-products was calculated in the paper. The calculation was made using the ORIGEN-S from ORIGEN4.4a in conjunction with the activation-burnup library and an updated decay data library with clearance levels data in ORIGEN format produced by WIMS-AECL/SCALENEA-1 code system. Both the radioactivity concentration expressed in Curie and Becquerel, and the clearance index and ingestion / inhalation hazard were calculated for the radionuclides contained in 1 kg of irradiated fuel element at shutdown and for 1, 50, 1500 years cooling time. This study required a complex activity that consisted of various phases such us: the acquisition, setting up, validation and application of procedures, codes and libraries. For the validation phase of the study, the objective was to compare the measured inventories of selected actinide and fission products radionuclides in an element from a Pickering CANDU reactor with inventories predicted using a recent version of the ORIGEN-ARP from SCALE 5 coupled with the time dependent cross sections library, CANDU 28.lib, produced by the sequence SAS2H of SCALE 4.4a. In this way, the procedures, codes and libraries for the characterization of radioactive material in terms of radioactive inventories, clearance, and biological hazard factors are being qualified and validated, in support for the safety management of the radioactive wastes

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-02-15

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

  6. Spent nuclear fuel discharges from US reactors 1993

    Energy Technology Data Exchange (ETDEWEB)

    1995-02-01

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

  7. 77 FR 60482 - Yankee Atomic Electric Company; Yankee Rowe Independent Spent Fuel Storage Installation, Staff...

    Science.gov (United States)

    2012-10-03

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

  8. 10 CFR 171.15 - Annual fees: Reactor licenses and independent spent fuel storage licenses.

    Science.gov (United States)

    2010-01-01

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

  9. 10 CFR 72.230 - Procedures for spent fuel storage cask submittals.

    Science.gov (United States)

    2010-01-01

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

  10. 75 FR 33853 - Maine Yankee Atomic Power Company; Independent Spent Fuel Storage Installation; Issuance of...

    Science.gov (United States)

    2010-06-15

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

  11. 77 FR 33005 - Connecticut Yankee Atomic Power Company; Haddam Neck Independent Spent Fuel Storage Installation...

    Science.gov (United States)

    2012-06-04

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

  12. International safeguards concerns of Spent Fuel Disposal Program

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  16. Studies on spent nuclear fuel evolution during storage

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    JOHNSON, D.M.

    2000-01-27

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

  19. Spent fuel receipt and storage at the Morris Operation

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

    Bucholz, James A.

    1983-01-01

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

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

    International Nuclear Information System (INIS)

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

  2. Spent fuel receipt and storage at the Morris Operation

    Energy Technology Data Exchange (ETDEWEB)

    Astrom, K A; Eger, K J

    1978-06-01

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

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

    OpenAIRE

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

    2016-01-01

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

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

    OpenAIRE

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

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-01-01

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

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

    International Nuclear Information System (INIS)

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

  7. Dry Refabrication Technology Development of Spent Nuclear Fuel

    International Nuclear Information System (INIS)

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

  8. Handling of final storage of unreprocessed spent nuclear fuel

    International Nuclear Information System (INIS)

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

  9. Disposal of spent nuclear fuel in geological repositories

    International Nuclear Information System (INIS)

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

  10. Neutron resonance transmission analysis of reactor spent fuel assemblies

    International Nuclear Information System (INIS)

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

  11. A general computing code devoted to the analysis of bending vibrations specific to the CANDU type fuel channel

    International Nuclear Information System (INIS)

    It is known that circulation of the coolant through the pressure tube of a CANDU type reactor initiates and maintains bending vibrations in: individual fuel elements, fuel cluster, cluster column and in the pressure tube. The driving forces are either aleatory, due to turbulent flow, or harmonical due to the pressure pulsations from the circulation pumps. The vibrations induced by laminar flow in case of excessive intensities may induce both a acceleration of the fretting wear phenomena in the fuel elements and pressure tubes and a premature aging of the latter. In these conditions an important problem in the cluster design is that of obtaining, based on knowledge of laminar flow frequency structure, the eigenfrequencies for the four categories of oscillatory systems mentioned above and thus to avoid by construction the resonance phenomenon or at least to diminish its impairing effects. An activity of comparative analysis in different fuel cluster types is underway at INR Pitesti, a special attention being of course directed toward their vibrational behavior. The paper presents a general computational code devoted to characterization of bending vibration for: individual fuel elements, fuel element cluster, pressure tube loaded or not with fuel clusters and filled or not with coolant; fuel channel. During the presentation of the work the computing code will be run for demonstration

  12. CANDU market prospects

    International Nuclear Information System (INIS)

    This 1994 survey of prospective markets for CANDU reactors discusses prospects in Turkey, Thailand, the Philippines, Korea, Indonesia, China and Egypt, and other opportunities, such as in fuel cycles and nuclear safety. It was concluded that foreign partners would be needed to help with financing

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-12-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  18. Fission product release assessment for end fitting failure in Candu reactor loaded with CANFLEX-NU fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Dirk Joo; Jeong, Chang Joon; Lee, Kang Moon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    Fission product release (FPR) assessment for End Fitting Failure (EFF) in CANDU reactor loaded with CANFLEX-natural uranium (NU) fuel bundles has been performed. The predicted results are compared with those for the reactor loaded with standard 37-element bundles. The total channel I-131 release at the end of transient for EFF accident is calculated to be 380.8 TBq and 602.9 TBq for the CANFLEX bundle and standard bundle channel cases, respectively. They are 4.9% and 7.9% of total inventory, respectively. The lower total releases of the CANFLEX bundle O6 channel are attributed to the lower initial fuel temperatures caused by the lower linear element power of the CANFLEX bundle compared with the standard bundle. 4 refs., 1 fig., 4 tabs. (Author)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

    Veliscek-Carolan, Jessica

    2016-11-15

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

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

    Veliscek-Carolan, Jessica

    2016-11-15

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

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

    International Nuclear Information System (INIS)

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

  5. Present status and prospect of spent fuel transportation

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-12-01

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

  7. 77 FR 37937 - License Renewal Application for Prairie Island Nuclear Generating Plant Independent Spent Fuel...

    Science.gov (United States)

    2012-06-25

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

  8. K Basin spent nuclear fuel characterization

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  11. Mission Need Statement: Idaho Spent Fuel Facility Project

    Energy Technology Data Exchange (ETDEWEB)

    Barbara Beller

    2007-09-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  16. Information handbook on independent spent fuel storage installations

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-01

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

  17. Alteration of spent fuel matrix under unsaturated water conditions

    International Nuclear Information System (INIS)

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

  18. Spent fuel and high-level radioactive waste storage

    International Nuclear Information System (INIS)

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

  19. Application of spent fuel treatment technology to plutonium immobilization

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Ferrada, J.J.

    2004-10-04

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  4. GT-MHR spent fuel storage disposal without processing

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  6. Shielding Performance Measurements of Spent Fuel Transportation Container

    Directory of Open Access Journals (Sweden)

    SUN Hong-chao

    2015-11-01

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

  7. International management and storage of plutonium and spent fuel

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-07-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1985-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Sihm Kvenangen, Karen

    2007-06-15

    In the future, the demand for energy is predicted to grow and more countries plan to utilize nuclear energy as their source of electric energy. This gives rise to many important issues connected to nuclear energy, such as finding methods that can verify that the spent nuclear fuel has been handled safely and used in ordinary power producing cycles as stated by the operators. Gamma ray spectroscopy is one method used for identification and verification of spent nuclear fuel. In the specific gamma ray spectroscopy method called gamma scanning the gamma radiation from the fission products Cs-137, Cs-134 and Eu-154 are measured in a spent fuel assembly. From the results, conclusions can be drawn about the fuels characteristics. This degree project examines the possibilities of using alternative measuring approaches when using the gamma scanning method. The focus is on examining how to increase the quality of the measured data. How to decrease the measuring time as compared with the present measuring strategy, has also been investigated. The main part of the study comprises computer simulations of gamma scanning measurements. The simulations have been validated with actual measurements on spent nuclear fuel at the central interim storage, Clab. The results show that concerning the quality of the measuring data the conventional strategy is preferable, but with other starting positions and with a more optimized equipment. When focusing on the time aspect, the helical measuring strategy can be an option, but this needs further investigation.

  13. Mechanistic modeling of bearing pad to pressure tube contact under localized high temperature conditions in a CANDU fuel channel

    International Nuclear Information System (INIS)

    During a postulated critical break LOCA (loss of coolant accident) in a CANDU reactor the coolant flow rates in the fuel channels of the flow pass of the reactor core downstream of the pipe break can rapidly reduce to very low values and remain very low for a period of tens of seconds following the break. Under the sustained low flow conditions, the fuel sheath (cladding) temperature in the affected channels rapidly increases and the coolant in the channels becomes significantly voided. The pressure tubes in the affected pass heat up under a combination of convection heat transfer from the low flows of superheated seam and thermal radiation heat transfer from the hot fuel. Additionally, hot spots may develop on the inner surface of pressure tubes at locations where the fuel bearing pads are in direct contact with the pressure tube. Localized thermal creep strain deformation at the hot spots is a potential pressure tube failure mechanism which could challenge fuel channel integrity. This paper evaluates the local thermal-mechanical deformation of a pressure tube in a CANDU reactor under critical break LOCA conditions tube using a coupled thermal-mechanical finite element COMSOL multi-physics model and investigates the conditions resulting in fuel channel failure due to localized contact between bearing pad and pressure. The mechanistic models are validated against data obtained from COG funded experiments performed at WRL (Whiteshell Research Laboratory). Multiphysics calculations are performed in which the heat transfer, thermal-mechanical and creep strain equations are solved, simultaneously. Heat conduction from bearing pads to the inner surface of the pressure tube is modeled with appropriate convective and radiation heat transfer boundary conditions. Thermal creep strain deformation of the Zr-2.5%Nb pressure tube is modeled using correlations derived from separate uniaxial tests that are reported in the literature. Contact conductance models based on

  14. 10 CFR 72.236 - Specific requirements for spent fuel storage cask approval and fabrication.

    Science.gov (United States)

    2010-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  17. Loviisa NPP. Extension of spent fuel storage capacity

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  19. Test plan for spent fuel cladding containment credit tests

    International Nuclear Information System (INIS)

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

  20. The development of a remote gauging and inspection capability for fuel channels in Candu reactors

    International Nuclear Information System (INIS)

    Equipment under development for the inspection and gauging of pressure tubes in CANDU (Canadian Deuterium Uranium) type reactors is described. A brief overview of the mechanical scanning system is presented followed by a detailed description of the measurement and data processing systems for the gauging of diameter and wall thickness, volumetric inspection of the tube wall and gauging of the annular gap between the pressure tube and the calandria tube. Experience of testing ultrasonic transducers in very high (106 Roentgens/hour)(R/h) radiation fields is reviewed. (author)

  1. Synthesis on the spent fuel long term evolution

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-06-01

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

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

    International Nuclear Information System (INIS)

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

  4. Synthesis on the spent fuel long term evolution

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  9. Characterization of activated metals in spent fuel hardware

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  11. Modelling of radiation field around spent fuel container

    NARCIS (Netherlands)

    Kryuchkov, EF; Opalovsky, VA; Tikhomirov, GV

    2005-01-01

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

  12. Chemical reprocessing of spent nuclear fuels

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    2012-06-01

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

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

    International Nuclear Information System (INIS)

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

  18. The small (or large) modular CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Meneley, D.; Harvel, G. [Univ. of Ontario Inst. of Tech., Oshawa, Ontario (Canada)

    2013-07-01

    This presentation outlines the design for small (or large) modular CANDU. The origins of this work go back many years to a comment by John Foster, then President of AECL CANDU. Foster noted that the CANDU reactor, with its many small fuel channels, was like a wood campfire. To make a bigger fire, just throw on some more logs (channels). If you want a smaller fire, just use fewer logs. The design process is greatly simplified.

  19. Characteristics of DUPIC fuel fabrication technology

    International Nuclear Information System (INIS)

    The concept of DUPIC (Direct use of spent PWR fuel in CANDU reactor) fuel is recently being taken a growing interest as an innovative fuel in terms of proliferation resistant fuel cycle for reusing spent PWR fuel in CANDU reactor without wet reprocessing process. The fabrication processes are composed of decladding of spent PWR fuel rods, OREOX processing to produce the powder feedstock from the spent PWR fuel meat. Once the resinterable powder is prepared, the fuel fabrication processes such as pelletizing, rod and bundle manufacturing processes are similar to the conventional powder/pellet routes for the fuel fabrication except that all the processes should be performed in hot cell by remote manner. The characteristics of the DUPIC fuel fabrication processes and the status of the equipment development at KAERI is described. (author)

  20. Thai Strategic Plan for Spent Fuel Management

    International Nuclear Information System (INIS)

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