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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. Development of CANDU Spent Fuel Sipping System

    International Nuclear Information System (INIS)

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

  3. 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)

  4. Interim storage of CANDU spent fuel and safety performance

    International Nuclear Information System (INIS)

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

  5. 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.

  6. Integrity of spent CANDU fuel during and following dry storage

    International Nuclear Information System (INIS)

    This report examines the issue of CANDU fuel integrity at the back end of the fuel cycle and outlines a program designed to provide assurance that used CANDU fuel will retain its integrity over an extended period. In specific terms, the program is intended to provide assurance that during and following extended dry storage the fuel will remain fit to undergo, without loss of integrity, the handling, packaging and transportation operations that might be necessary until it is placed in disposal containers. The first step in the development of the program was a review of the available technical information on phenomena relevant to fuel integrity. The major conclusions from that review were the following: Under normal storage conditions it is unlikely that the spent fuel will suffer significant degradation during a one-hundred year period and it should be possible to retrieve, repackage and transport the fuel as required, using methods and systems similar to those used today. However, to provide increased confidence regarding the above conclusion, investigations should be conducted in areas where there is higher uncertainty in the prediction of fuel condition and on some degradation processes to which the fuel appears to present higher vulnerability. The proposed program includes, among other tasks, irradiated fuel tests, analytical studies on the most relevant fuel degradation processes and the development of a long-term fuel verification program. (Author)

  7. 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)

  8. 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 %.

  9. 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)

  10. 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)

  11. Thermal Analysis of CANDU Spent Fuel Bay Cooling System

    International Nuclear Information System (INIS)

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

  12. 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.

  13. 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)

  14. Development of CANDU spent fuel verification system using optical fiber scintillator

    International Nuclear Information System (INIS)

    In CANDU, spent fuels discharge 16∼24 bundles from the reactor core at everyday. Those are contained on the tray and the tray is stacked in the spent fuel. Currently, the Agency uses the CANDU Bundle Verifier for Stack (CBVS). It consists of a CZT gamma spectrometric probe which moves vertically along the space in between the columns of trays. Somewhat, spent fuel verification by non-destructive assay has been implemented for safeguards purpose using various radiation detectors such as a gas type detector, a semiconductor detector and so on. However, due to the severe circumstance of spent fuel storage such as high temperature, high radiation intensity and difficult to access area, the applicable radiation detectors and measurement techniques are very limited. An optical fiber scintillator has been known to have a good radiation hardness and physical properties for high temperature and humidity. In order to verify spent fuels stored in difficult to access area, KINAC designed and developed a prototype which was a spent fuel verification equipment using an optical fiber scintillator. The field test was performed at Wolsung NPP (Nuclear Power Plant) pond storage area. And this system will be made an entry for IAEA's verification equipment. For registration, KINAC would be followed the IAEA's QA procedure. At now, KINAC/IAEA developed the user, functional requirement and design specification for System, Hardware and Software separately. After finishing the procedure, it will be used for verification of spent fuel in lieu of CANDU Bundle Verifier for Baskets (CBVS). (author)

  15. CANDU fuel cycle flexibility

    International Nuclear Information System (INIS)

    High neutron economy, on-power refuelling, and a simple bundle design provide a high degree of flexibility that enables CANDU (Canada Deuterium Uranium; registered trademark) reactors to be fuelled with a wide variety of fuel types. Near-term applications include the use of slightly enriched uranium (SEU), and recovered uranium (RU) from reprocessed spent Light Water Reactor (LWR) fuel. Plutonium and other actinides arising from various sources, including spent LWR fuel, can be accommodated, and weapons-origin plutonium could be destroyed by burning in CANDU. In the DUPIC fuel cycle, a dry processing method would convert spent Pressurized Water Reactor (PWR) fuel to CANDU fuel. The thorium cycle remains of strategic interest in CANDU to ensure long-term resource availability, and would be of specific interest to those countries possessing large thorium reserves, but limited uranium resources. (author). 21 refs

  16. Disposal costs for advanced CANDU fuel cycles

    International Nuclear Information System (INIS)

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

  17. Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors - II: DUPIC Fuel-Handling Cost

    International Nuclear Information System (INIS)

    The Direct Use of spent Pressurized water reactor fuel In Canada deuterium uranium (CANDU) reactors (DUPIC) fuel-handling technique has been investigated through a conceptual design study to estimate the unit cost that can be used for the DUPIC fuel cycle cost calculation. The conceptual design study has shown that fresh DUPIC fuel can be transferred to the core following the existing spent-fuel discharge route, provided that new fuel-handling equipment, such as the manipulator, opening/sealing tool of shipping casks, new fuel magazine, new fuel ram, dryer, gamma-ray detector, etc., are installed. The reverse path loading option is known to minimize the number of additional pieces of equipment for fuel handling, because it utilizes the existing spent-fuel handling equipment, and the discharge of spent DUPIC fuel can be done through the existing spent-fuel handling system without any modification. However, because the decay heat of spent DUPIC fuel is much higher than that of spent natural uranium fuel, the extra cooling capacity should be supplemented in the spent-fuel storage bay. Based on the conceptual design study, the capital cost for DUPIC fuel handling and extra storage cooling capacity was estimated to be $3 750 000 (as of December 1999) per CANDU plant. The levelized unit cost of DUPIC fuel handling was then obtained by considering the amount of fuel that will be required during the lifetime of a plant, which is 5.13 $/kg heavy metal. Compared with the other unit costs of the fuel cycle components, it is expected that DUPIC fuel handling has only a minor effect on the overall fuel cycle cost

  18. 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

  19. CANDU spent fuel shielding analysis during intermediate dry storage by using Monte Carlo methodology

    International Nuclear Information System (INIS)

    Almost all the countries that operate or construct nuclear power plants have r and d programs for spent nuclear fuel and radioactive waste management. In these programs, optimal solutions for nuclear fuel cycle management are to be identified, geological disposal being one of the main goals here. Romania did not yet adopt a decision for final disposal. Nevertheless, researches and studies are in progress in order to select and characterize the geological formation for spent fuel final disposal. Currently, although there is no comprehensive EU policy in the field of safe spent fuel and radioactive waste management it is desirable to bring and keep the safety of radioactive waste management on a uniform high level among the Member States and the accession countries. The Romanian Cernavoda NPP, of CANDU type, has the following spent fuel management facilities: a spent fuel bay (for spent fuel wet storage) and a spent fuel interim dry storage facility. The dry storage technology is based on MACSTOR system consisting of storage modules located outdoors in the storage site, and equipment operated at the spent fuel storage bay for preparing the spent fuel for dry storage. The spent fuel is transferred from the preparation area to the storage site in a transfer flask. The concrete storage modules have two sealed barriers for storing the spent fuel: a seal welded stainless steel basket containing 60 spent fuel bundles and a seal welded cylinder containing 10 baskets. Twenty storage cylinders are in one storage module for a total capacity of 12,000 bundles per module. In 2003, the first storage module has become operational. The paper has the following contents: Introduction; The paper objectives; Theoretical model Set-Up; Results; Conclusions. In conclusions one notifies that SEU fuel leads to higher burnup degrees associated both with spent fuel and actinides mass reduction for 1 MWh generated electric power (from 7100 MWD/tU for UNAT to 20000 MWD/tU for SEU43

  20. 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

  1. 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).

  2. In-situ verification of CANDU spent fuel by the Cherenkov technique

    International Nuclear Information System (INIS)

    Multilayered and densely stacked irradiated CANDU fuel bundles in storage ponds make direct viewing of bundles in order to observe Cherenkov glow practically impossible. Ability to defect the source of Cherenkov glow by visual observation invariably suffers from subjective judgment. In the case of CANDU-type storage geometry, the difficulty in drawing conclusions is even greater for a number of reasons including the near neighbour effect. In this paper, the first results of Cherenkov photographic procedure without the isolation of individual trays are presented in which a new model of the Hungarian underwater telescope in combination with a lightmeter for Cherenkov intensity measurements has been used. It is demonstrated by this technique that photographs of bundles with cooling time of up to 2 a provide a satisfactory record for conclusive attribute verification result for irradiated fuel bundles stacked in multilayers. A distinct glow, with a brightness of higher intensity between the rod of a bundle compared to the surroundings of the bundle, is clearly shown by the pictures. Based on the results of the glow intensity measurements, the use of this photographic method for fuel bundles with longer cooling time of up to 15 a or more would require considerably longer exposure times or more sensitive film. Possible impact on IAEA safeguards of CANDU spent fuel bays by a system, which offers simultaneous item counting and NDA attribute test capabilities in a relatively low intrusive manner, is discussed. The limitations are also considered. (author)

  3. 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)

  4. 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

  5. Drop Test of the Candu Spent Fuel Storage Basket in MACSTOR/KN-400

    International Nuclear Information System (INIS)

    The MACSTOR/KN-400 of Wolsung power plant in Korea is a dry interim storage facilities. There are 400 long slender cylinders in MACSTOR/KN-400. In one cylinder, ten baskets where Candu spent fuels are loaded are stacked and stored. For this MACSTOR/KN-400 facilities, analyses and tests for the hypothetical accident conditions that might happen during moving and storing baskets into a cylinder were performed. The hypothetical accident conditions to be considered are two cases. One is the case of basket dropping onto the bottom plate of a cylinder. The other is the case of basket dropping onto the other basket top plate stored in the cylinder. For the drop analyses, the case of hanging cylinder and the case of cylinder on the unyielding target surface were considered. Based on the dropping analysis, testing condition was determined as the latter case that is for the cylinder on the target surface. In a basket, 60 dummy fuel bundles are loaded which have the same weight of real spent fuel bundles. On the external surface of the basket, 8 strain gauges and 4 accelerometers were attached for the data acquisition. In order to measure the velocity when a basket impacts, three different devices were utilized. And the impact velocity results were compared and cross-checked. After the dropping tests, helium leak tests were conducted to evaluate the leakage rate. (authors)

  6. Drop Test of the Candu Spent Fuel Storage Basket in MACSTOR/KN-400

    Energy Technology Data Exchange (ETDEWEB)

    Choi, W.S.; Jeon, J.Y.; Seo, K.S. [KAERI, 1045 Daedeokdaero, Yuseong, Daejeon, 305-353 (Korea, Republic of); Park, J.E.; Yoo, G.S.; Park, W.G. [Korea Hydro Nuclear Power - KHNP (Korea, Republic of)

    2009-06-15

    The MACSTOR/KN-400 of Wolsung power plant in Korea is a dry interim storage facilities. There are 400 long slender cylinders in MACSTOR/KN-400. In one cylinder, ten baskets where Candu spent fuels are loaded are stacked and stored. For this MACSTOR/KN-400 facilities, analyses and tests for the hypothetical accident conditions that might happen during moving and storing baskets into a cylinder were performed. The hypothetical accident conditions to be considered are two cases. One is the case of basket dropping onto the bottom plate of a cylinder. The other is the case of basket dropping onto the other basket top plate stored in the cylinder. For the drop analyses, the case of hanging cylinder and the case of cylinder on the unyielding target surface were considered. Based on the dropping analysis, testing condition was determined as the latter case that is for the cylinder on the target surface. In a basket, 60 dummy fuel bundles are loaded which have the same weight of real spent fuel bundles. On the external surface of the basket, 8 strain gauges and 4 accelerometers were attached for the data acquisition. In order to measure the velocity when a basket impacts, three different devices were utilized. And the impact velocity results were compared and cross-checked. After the dropping tests, helium leak tests were conducted to evaluate the leakage rate. (authors)

  7. Candu fuel and fuel cycles

    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 emissions 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. 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 fuels 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

  8. Development of a spent fuel bay operator support system for PHWR-CANDU nuclear power plant

    International Nuclear Information System (INIS)

    The safety advantages of the CANDU 600 NPPs could be further enhanced by the supplementation of the Canadian experience on plant operation with Computerised Operator Support System (COSSs) adjusted to the actual plant configuration, the idiosyncrasies of the given equipment, the errors during engineering and construction phases and the psychological features of the operation personnel. One of the most relevant systems with an important decision component that involves relatively complex procedures - thereby related human errors - is Cernavoda NPP's Spent Fuel Bay Cooling and Purification System. Strong motivations to consider a flexible COSS, both for operation and for intervention purpose, are given by the global activity stored in the systems bays and the location of the bays outside the containment. A Spent Fuel Bay Operator Support System (SFOSS) is at a research-grade at the Institute of Atomic Physics and at the Center of Technology and Engineering for Nuclear Projects in compliance with the general principles of the expert systems and under a IAEA Co-ordinate Research Program. The paper illustrates a generic description of the system and also of the SFOSS structure. (Author) 10 Refs

  9. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Vieru, G., E-mail: gheorghe.vieru@nuclear.ro [Inst. for Nuclear Research, Pitesti (Romania)

    2010-07-01

    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)

  11. 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

  12. 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.

  13. CANDU: The fuel conserving reactor

    International Nuclear Information System (INIS)

    Because of their high neutron economy and unique design features, CANDU heavy water moderated reactors are the only established commercial reactors able to use directly low fissile content fuels such as natural uranium or uranium recovered from spent light water reactor fuel (RU). These features also help them to achieve the highest fuel utilization of all commercially available reactors, whether the fuel is based on natural uranium or mixed oxides of plutonium, uranium or thorium. As nuclear capacity growth increases demands on the world's finite uranium resources, AECL envisages near term use in CANDU reactors of a fuel incorporating RU and fuels containing thorium, with either plutonium or low enriched uranium (LEU) as the fissile 'driver' fuel. In the long term, AECL proposes the use of future 'Generation X' CANDU reactors with enhanced neutron economy to achieve a near-Self-Sufficient Equilibrium Thorium (SSET) fuel cycle. This CANDU SSET would have a conversion ratio of unity and be able to produce power indefinitely, with the need for little additional fissile material once equilibrium is reached (the amount of 233U needed in the fresh fuel is the same as is present in the discharged fuel, including processing losses.) This would also enable a CANDU-Fast Breeder Reactor (FBR) synergism that would allow each fuel-generating, though expensive, FBR to supply the initial fissile requirements of several less-expensive, CANDU SSET reactors operating on the thorium cycle. The closer the approach to an SSET that CANDUs can achieve, the higher the ratio of CANDUs to breeders in an economically optimized reactor fleet. CANDU reactors thereby become natural partners of both light water-cooled thermal reactors and fast breeder reactors, in both cases making optimum use of their spent fuel components and enhancing the overall sustainability of nuclear power. (authors)

  14. Synergistic CANDU-LWR fuel cycles

    International Nuclear Information System (INIS)

    CANDU is the most neutron-efficient reactor available commercially, allowing utilization of a range of fuel cycles. The flexibility of on-line refuelling allows fuel management to accommodate these different fuels. A synergism with light-water reactors (LWR) is possible through the use in CANDU of uranium and/or plutonium recovered from spent LWR fuel. In the TANDEM fuel cycle, the unseparated uranium and plutonium (1.5% fissile) would give a burnup in CANDU of about 25 MW.d/kg HE, producing four times more energy than that available from simply recycling the plutonium in an LWR. In another potential fuel cycle, uranium recovered from spent LWR fuel during conventional reprocessing is also recycled in CANDU, without re-enrichment. An average recovered uranium (RU) enrichment of 0.9% in U-235 results in a CANDU burnup of at least 13 MW.d/kg U, allowing twice as much energy to be extracted, compared with that from an LWR. The fuelling cost for RU in CANDU are about 35% lower than for natural uranium. Additionally, direct use of spent LWR fuel in CANDU is theoretically possible, but requires practical demonstration. AECL and KAERI are developing the CANFLEX (CANDU Flexible Fuelling) advanced fuel bundle as the optimal carrier for all extended burnup fuel cycles envisaged for CANDU

  15. CANDU fuel performance

    International Nuclear Information System (INIS)

    The paper presents a review of CANDU fuel performance including a 28-element bundle for Pickering reactors, a 37-element bundle for the Bruce and Darlington reactors, and a 37-element bundle for the CANDU-6 reactors. Special emphasis is given to the analysis of fuel defect formation and propagation and definition of fuel element operating thresholds for normal operation and accident conditions. (author)

  16. CANDU advanced fuel cycles

    International Nuclear Information System (INIS)

    This report is based on informal lectures and presentations made on CANDU Advanced Fuel Cycles over the past year or so, and discusses the future role of CANDU in the changing environment for the Canadian and international nuclear power industry. The changing perspectives of the past decade lead to the conclusion that a significant future market for a CANDU advanced thermal reactor will exist for many decades. Such a reactor could operate in a stand-alone strategy or integrate with a mixed CANDU-LWR or CANDU-FBR strategy. The consistent design focus of CANDU on enhanced efficiency of resource utilization combined with a simple technology to achieve economic targets, will provide sufficient flexibility to maintain CANDU as a viable power producer for both the medium- and long-term future

  17. Thermal behavior of the CANDU type spent fuel dry-storage concrete canister

    International Nuclear Information System (INIS)

    This paper describes a simple model developed for calculation of the temperature distribution and thermal behavior analysis of the spent fuel dry-storage concrete canister. The model takes into account the relevant heat transfer processes and the cylindrical geometry of the concrete canister. (author)

  18. 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).

  19. CANDU fuel-cycle vision

    International Nuclear Information System (INIS)

    The fuel-cycle path chosen by a particular country will depend on a range of local and global factors. The CANDU reactor provides the fuel-cycle flexibility to enable any country to optimize its fuel-cycle strategy to suit its own needs. AECL has developed the CANFLEX fuel bundle as the near-term carrier of advanced fuel cycles. A demonstration irradiation of 24 CANFLEX bundles in the Point Lepreau power station, and a full-scale critical heat flux (CHF) test in water are planned in 1998, before commercial implementation of CANFLEX fuelling. CANFLEX fuel provides a reduction in peak linear element ratings, and a significant enhancement in thermalhydraulic performance. Whereas natural uranium fuel provides many advantages, the use of slightly enriched uranium (SEU) in CANDU reactors offers even lower fuel-cycle costs and other benefits, such as uprating capability through flattening the channel power distribution across the core. Recycled uranium (RU) from reprocessing spent PWR fuel is a subset of SEU that has significant economic promise. AECL views the use of SEU/RU in the CANFLEX bundle as the first logical step from natural uranium. High neutron economy enables the use of low-fissile fuel in CANDU reactors, which opens up a spectrum of unique fuel-cycle opportunities that exploit the synergism between CANDU reactors and LWRs. At one end of this spectrum is the use of materials from conventional reprocessing: CANDU reactors can utilize the RU directly without re-enrichment, the plutonium as conventional Mixed-oxide (MOX) fuel, and the actinide waste mixed with plutonium in an inert-matrix carrier. At the other end of the spectrum is the DUPIC cycle, employing only thermal-mechanical processes to convert spent LWR fuel into CANDU fuel, with no purposeful separation of isotopes from the fuel, and possessing a high degree of proliferation resistance. Between these two extremes are other advanced recycling options that offer particular advantages in exploiting the

  20. CANDU fuel-cycle vision

    International Nuclear Information System (INIS)

    The fuel-cycle path chosen by a particular country will depend on a range of local and global factors. The CANDU reactor provides the fuel-cycle flexibility to enable any country to optimize its fuel-cycle strategy to suit its own needs. AECL has developed the CANFLEX fuel bundle as the near-term carrier of advanced fuel cycles. A demonstration irradiation of 24 CANFLEX bundles in the Point Lepreau power station, and a full-scale critical heat flux (CHF) test in water are planned in 1998, before commercial implementation of CANFLEX fuelling. CANFLEX fuel provides a reduction in peak linear element ratings, and a significant enhancement in thermalhydraulic performance. Whereas natural uranium fuel provides many advantages, the use of slightly enriched uranium (SEU) in CANDU reactors offers even lower fuel-cycle costs and other benefits, such as uprating capability through flattening the channel power distribution across the core. Recycled uranium (RU) from reprocessing spent PWR fuel is a subset of SEU that has significant economic promise. AECL views the use of SEU/RU in the CANFLEX bundle as the first logical step from natural uranium. High neutron economy enables the use of low-fissile fuel in CANDU reactors, which opens up a spectrum of unique fuel-cycle opportunities that exploit the synergism between CANDU reactors and LWRs. At one end of this spectrum is the use of materials from conventional reprocessing: CANDU reactors can utilize the RU directly without reenrichment, the plutonium as conventional mixed-oxide (MOX) fuel, and the actinide waste mixed with plutonium in an inert-matrix carrier. At the other end of the spectrum is the DUPIC cycle, employing only thermal-mechanical processes to convert spent LWR fuel into CANDU fuel, with no purposeful separation of isotopes from the fuel, and possessing a high degree of proliferation resistance. Between these two extremes are other advanced recycling options that offer particular advantages in exploiting the

  1. CANDU fuel cycle options in Korea

    International Nuclear Information System (INIS)

    The easiest first step in CANDU fuel-cycle evolution may be the use of slightly enriched uranium (SEU), including recovered uranium from reprocessed LWR spent fuel. Relatively low enrichment (up to 1.2%) will result in a twoto three-fold reduction in the quantity of spent fuel per unit energy production, reductions in fuel-cycle costs, and greater flexibility in the design of new reactors. The CANFLEX (CANDU FLEXible) fuel bundle would be the optimal fuel carrier. A country that has both CANDU and PWR reactors can exploit the natural synergism between these two reactor types to minimize overall waste production, and maximize energy derived from the fuel. This synergism can be exploited through several different fuel cycles. A high burnup CANDU MOX fuel design could be used to utilize plutonium from conventional reprocessing or more advanced reprocessing options (such as co-processing). DUPIC (Direct Use of Spent PWR Fuel In CANDU) represents a recycle option that has a higher degree of proliferation resistance than does conventional reprocessing, since it uses only dry processes for converting spent PWR fuel into CANDU fuel, without separating the plutonium. Good progress is being made in the current KAERI, AECL, and U.S. Department of State program in demonstrating the technical feasibility of DUPIC. In the longer term, CANDU reactors offer even more dramatic synergistic fuel cycles with PWR or FBR reactors. If the objective of a national fuel-cycle program is the minimization of actinide waste or destruction of long-lived fission products, then studies have shown the superiority of CANDU reactors in meeting this objective. Long-term energy security can be assured either through the thorium cycle or through a CANDU 1 FBR system, in which the FBR would be operated as a 'fuel factory,' providing the fissile material to power a number of lower-cost, high efficiency CANDU reactors. In summary, the CANDU reactor's simple fuel design, high neutron economy, and on

  2. 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).

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

    International Nuclear Information System (INIS)

    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 α-activity and α-, γ-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 ∼ 1/4 lower than that for once-through cycle. (Author)

  4. candu fuel bundle fabrication

    International Nuclear Information System (INIS)

    This paper describes works on CANDU fuel bundle fabrication in the Fuel Fabrication Development and Testing Section (FFDT) of AECL's Chalk River Laboratories. This work does not cover fuel design, pellet manufacturing, Zircaloy material manufacturing, but cover the joining of appendages to sheath tube, endcap preparation and welding, UO2 loading, end plate preparation and welding, and all inspections required in these steps. Materials used in the fabrication of CANDU fuel bundle are: 1)Ceramic UO2 Pellet 2)Zircaloy -4. Fuel Bundle Structural Material 3) Others (Zinc stearate, Colloidal graphite, Beryllium and Heium). Th fabrication of fuel element consist of three process: 1)pellet loading into the sheats, 2) endcap welding, and 3) the element profiling. Endcap welds is tested by metallography and He leak test. The endcaps of the elements are welded to the end plates to form the 37- element bundle assembly

  5. Thorium fuel cycles in CANDU

    International Nuclear Information System (INIS)

    In recent years, Atomic Energy of Canada Limited has been examining in detail the implications of using thorium-based fuels tn the CANDU reactor. Various cycles initiated and enriched either with fissile plutonium or with enriched uranium, and with effective conversion ratios ranging up to 1.0, have been evaluated. We have concluded that: 1. Substantial quantities of uranium can be saved by adoption of the thorium fuel cycle, and the long-term security of fissile supply both for the domestic and overseas market can be considerably enhanced. The amount saved will depend on the details of the fuel cycle and the anticipated growth of nuclear power in Canada. 2. The fuel cycle can be introduced into the basic CANDU design without major modifications and without compromising current safety standards. 3. The economic conditions that make thorium competitive with the once-through natural uranium cycle depend a the price of uranium and on the costs both to fabricate α and γ-emitting fuels and to either enrich uranium or to extract fissile material from spent fuel. While timing is difficult to predict, we believe that competitive economic conditions will prevail toward the end of this century. 4. A twenty-year technological development program will be required to establish commercial confidence in the fuel cycle. (author)

  6. 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)

  7. 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

  8. 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

  9. 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

  10. Improved CANDU fuel performance

    International Nuclear Information System (INIS)

    The fuel defect rate in CANDU power reactors has been very low (0.06 percent) since 1972. Most defects were caused by power ramping. The two measures taken to reduce the defect rate, by about an order of magnitude, were changes in the fuelling schemes and the introduction of thin coatings of graphite on the inside surface of the Zircaloy fuel cladding. Power ramping tests have demonstrated that graphite layers, and also baked poly-dimethyl-siloxane layers, between the UO2 pellets and Zircaloy cladding increase the tolerance of fuel to power ramps. These designs are termed graphite CANLUB and siloxane CANLUB; fuel performance depends on coating parameters such as thickness and wear resistance and on environmental and thermal conditions during the curing of coatings. (author)

  11. 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.)

  12. Overview of activities on CANDU fuel in Argentina

    International Nuclear Information System (INIS)

    This paper gives an outline of activities on CANDU fuel in Argentina. It discusses the nuclear activities and electricity production in Argentina, evolution of the activities in fuel engineering, fuel fabrication, fuel performance at Embalse nuclear power plant and spent fuel storage options.

  13. 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)

  14. The CANDU 9 fuel transfer system

    International Nuclear Information System (INIS)

    The CANDU 9 fuel transfer system is based on the CANDU 6 and the Ontario Hydro Darlington NGD designs, modified to suit the CANDU 9 requirements. The CANDU 9 new fuel transfer system is very similar to the CANDU 6, with modifications to allow new fuel loading from outside containment, similar to Darlington. The CANDU 9 irradiated fuel transfer system is based on the Darlington irradiated fuel transfer system, with modifications to meet the more stringent containment requirements, improve performance, and match station layout. (author). 2 refs., 6 figs

  15. 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)

  16. 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

  17. CANDU: Shortest path to advanced fuel cycles

    International Nuclear Information System (INIS)

    , utilization of alternative fuel sources in the currently operational CANDU reactors without modifications to their existing licensing basis. The use of NUE in CANDU reactors addresses the challenge of dealing with spent LWR fuel that would otherwise require monitored storage or costly re-enrichment and handling of highly radioactive fuels for reuse in LWRs. Using RU in existing nuclear power reactors will improve the utilization rate of NU resources and ultimately improve the sustainability of fuel resources. In addition, NUE reintroduces the fissile content in DU back into the fuel cycle; an effective application of an otherwise limited-use by-product from the uranium enrichment process. There is a wealth of global experience with RU, however, due to previously low economic incentives in LWRs, its utilization in nuclear reactors has been limited. With increased global requirements for uranium resources and fluctuating uranium prices. RU use in CANDU will play a vital role in the nuclear renaissance. Currently AECL and its partners in China, led by Third Qinshan Nuclear Power Company (TQNPC), has undertaken the commercial demonstration of NUE in Qinshan 3 CANDU units. The objective is eventual full core implementation of NUE fuel at a time desired by China. Thorium is available in large quantities in countries such as China and India, both of which have rapidly expanding economies requiring additional power sources driven by sustainable and easily accessible fuel sources. CANDU's excellent neutron economy, resulting from the use of heavy water as both moderator and coolant, provides for the efficient use of an LEU or plutonium driver in thorium fuel cycles. The proven CANDU reactor is able to operate with a full core of thorium fuel without requiring major changes to the existing safety and control systems. AECL has conducted work on thorium fuel cycles, which demonstrated that both plutonium and LEU could be used as driver fuels for thorium. The once-through fuel cycle is

  18. Safeguards seals for underwater spent-fuel storage: the principles and the progress towards a practical system for CANDU reactors

    International Nuclear Information System (INIS)

    While seals are in widespread use for safeguards, the development of a completely satisfactory underwater seal for irradiated fuel has not been straightforward. One aspect of the problem is that, to produce a seal which is (ideally) impossible to duplicate requires a non-reproducible identity. Equally challenging is to provide a simple means of accurately comparing an underwater seal with a reference. The Canadian seal programme is producing seals using two different inspection technologies, ultrasonics and optics. Ultrasonics lends itself to relatively simple measuring tools, but accurate comparative measurements are difficult. The design now proposed for the CANDU ultrasonic cap seal uses a stainless-steel wire coil as the identity element. Tools to install and read the coil seal have been produced and the experimental results obtained with laboratory equipment have confirmed that the seal has real potential for IAEA use. Optical inspection techniques are well developed, but simple tools have yet to be demonstrated for the application. Progress on the optical seal has been in the area of selecting an identity which cannot be duplicated but can be readily examined. Millimetre-sized crystals have been grown in pure zirconium. These are irregular in shape, brightly coloured and the patterns can be readily matched. (author)

  19. Fuel for CANDU pressurized heavy water reactors

    International Nuclear Information System (INIS)

    Unique properties, performance and evolution of CANDU fuel are described. The manufacturing conditions, uranium requirements, and fuel costs are discussed. The in-service performance of the fuel has been excellent and defect mechanisms and operating criterion are described. Evolutionary improvements in CANDU fuel and new fuel cycles such as plutonium and thorium are being explored to insure that the CANDU reactor remains competitive in the future. (author)

  20. The back end of the fuel cycle and CANDU

    International Nuclear Information System (INIS)

    CANDU reactor operators have benefited from several advantages of the CANDU system and from AECL's experience, with regard to spent fuel handling, storage and disposal. AECL has over 20 years experience in development and application of medium-term storage and research and development on the disposal of used fuel. As a result of AECL's experience, short-term and medium-term storage and the associated handling of spent CANDU fuel are well proven and economic, with an extremely high degree of public and environmental protection. In fact, both short-term (water-pool) and medium-term (dry canister) storage of CANDU fuel are comparable or lower in cost per unit of energy than for PWRs. Both pool storage and dry spent fuel storage are fully proven, with many years of successful, safe operating experience. AECL's extensive R and D on the permanent disposal of spent-fuel has resulted in a defined concept for Canadian fuel disposal in crystalline rock. This concept was recently confirmed as ''technically acceptable'' by an independent environmental review panel. Thus, the Canadian program represents an international demonstration of the feasibility and safety of geological disposal of nuclear fuel waste. Much of the technology behind the Canadian concept can be adapted to permanent land-based disposal strategies chosen by other countries. In addition, the Canadian development has established a baseline for CANDU fuel permanent disposal costs. Canadian and international work has shown that the cost of permanent CANDU fuel disposal is similar to the cost of LWR fuel disposal per unit of electricity produced. (author)

  1. 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)

  2. 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

  3. Study on advanced nuclear fuel cycle of PWR/CANDU synergism

    International Nuclear Information System (INIS)

    According to the concrete condition that China has both PWR and CANDU reactors, one of the advanced nuclear fuel cycle strategy of PWR/CANDU synergism ws proposed, i.e. the reprocessed uranium of spent PWR fuel was used in CANDU reactor, which will save the uranium resource, increase the energy output, decrease the quantity of spent fuels to be disposed and lower the cost of nuclear power. Because of the inherent flexibility of nuclear fuel cycle in CANDU reactor, the transition from the natural uranium to the recycled uranium (RU) can be completed without any changes of the structure of reactor core and operation mode. Furthermore, because of the low radiation level of RU, which is acceptable for CANDU reactor fuel fabrication, the present product line of fuel elements of CANDU reactor only need to be shielded slightly, also the conditions of transportation, operation and fuel management need not to be changed. Thus this strategy has significant practical and economical benefit

  4. Monitoring defective CANDU fuel bundles

    International Nuclear Information System (INIS)

    In 2005, it was proposed that a passive substance such as Nanocrystals could be used to monitor and locate defect fuel elements in-core. The experimental goal was to determine if Nanocrystals could be used for this application. Originally nanocrystals tagging was suggested for current operational CANDU-600 fuel. Other methods, including noble gas tagging, are also being investigated. Moreover, the scope of the project has been extended to include the identification of Dysprosium-doped fuel in the new ACR fuel design. The purpose of this paper is to discuss the experimental progress made at RMC on this project. (author)

  5. Spent fuel workshop'2002

    International Nuclear Information System (INIS)

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

  6. 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

  7. 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.

  8. 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

  9. 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)

  10. Advanced fuel cycles for CANDU reactors

    International Nuclear Information System (INIS)

    The current natural uranium-fuelled CANDU system is a world leader, both in terms of overall performance and uranium utilization. Moreover, the CANDU reactor is capable of using many different advanced fuel cycles, with improved uranium utilization relative to the natural uranium one-through cycle. This versatility would enable CANDU to maintain its competitive edge in uranium utilization as improvements are made by the competition. Several CANDU fuel cycles are symbiotic with LWRs, providing an economical vehicle for the recycle of uranium and/or plutonium from discharges LWR fuel. The slightly enriched uranium (SEU) fuel cycle is economically attractive now, and this economic benefit will increase with anticipated increases in the cost of natural uranium, and decreases in the cost of fuel enrichment. The CANFLEX fuel bundle, an advanced 43-element design, will ensure that the full benefits of SEU, and other advanced fuel cycles, can be achieved in the CANDU reactor. 25 refs

  11. Advanced fuel cycles in CANDU reactors

    International Nuclear Information System (INIS)

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

  12. Korea's CANDU fuel R and D program

    International Nuclear Information System (INIS)

    As the first R and D activity led to the nuclear fuel industrialization in Korea, KAERI had successfully developed the CANDU-6 fuel bundle in the period of 1981 to 1986 and has commercially produced more than 35,000 fuel bundles for the use in Wolsong Unit 1 since 1987. The commercial production of the CANDU-6 fuel in KAERI will be terminated on the end of 1997 and KNFC will take over the mission of CANDU-6 fuel production with a capacity of 400 tons of uranium per year form 1998. (author)

  13. SEU43 fuel bundle shielding analysis during spent fuel transport

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-01

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

  14. 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

  15. 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)

  16. Container storage of CANDU fuel

    International Nuclear Information System (INIS)

    Each Ontario Hydro operating plant has bays or water pools large enough to handle about a decade's used fuel. For the long term, however, dry storage is a practical alternative, especially for CANDU fuel with its low decay heat. The economic attractiveness of dry storage has long been recognized. Prospects improve if the same container can be used for transporting the fuel to its final destination without any need for repackaging. There are many engineering, scientific and licensing problems to be dealt with, but Ontario Hydro has made a start with its Dry Storage Demonstration Program. The long term goal of this program is to develop a container that can be used for storage, transportation, and disposal of used fuel. This paper describes the stages in the development of Ontario Hydro's approach to dry storage. Efforts have converged on the concept of a Concrete Integrated Container. This is a self-contained, high-integrity structure designed to be loaded with fuel directly from the station pools, moved to a storage area and eventually transported off-site to a final disposal facility. 5 refs., 4 tabs., 8 figs

  17. CANDU fuel performance and development

    International Nuclear Information System (INIS)

    The fuel defect rate in CANDU (Canada Deuterium Uranium) reactors continues to be very low, 0.06% since 1972. The power ramp defects, which constituted the majority of the early defects, have been virtually eliminated by changed fuelling schemes and through the introduction of graphite CANLUB coatings on the inside of the sheath. Laboratory and loop irradiations have demonstrated that the graphite CANLUB layers increase the tolerance to power ramps, but to obtain the maximum benefit, coating parameters such as thickness, adhesion and wear resistance must be optimized. Siloxane CANLUB coated fuel offers greater tolerance to power ramps than most graphite coatings; quality control appears simpler and no instance of localized sheath hydriding has been seen with cured and irradiated coatings. Limited testing has shown that fuel with graphite discs between fuel pellets also has high tolerance to power ramps, but it is more costly and has lower burnup. The number of defects due to faulty components has been extremely small (0.00014%), but improved quality control and welding procedures can lower this number even further. Defects from causes external to the bundle have also been very few. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-09-15

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

  19. CANFLEX - an advanced fuel bundle for CANDU

    International Nuclear Information System (INIS)

    The performance of CANDU pressurized heavy-water reactors, in terms of lifetime load factors, is excellent. More than 600 000 bundles containing natural-uranium fuel have been irradiated, with a low defect rate; reactor unavailability due to fuel incidents is typically zero. To maintain and improve CANDU's competitive position, Atomic Energy of Canada Limited (AECL) has an ongoing program comprising design, safety and availability improvements, advanced fuel concepts and schemes to reduce construction time. One key finding is that the introduction of slightly-enriched uranium (SEU, less than 1.5 wt% U-235 in U) offers immediate benefits for CANDU, in terms of fuelling and back-end disposal costs. The use of SEU places more demands on the fuel because of extended burnup, and an anticipated capability to load-follow also adds to the performance requirements. To ensure that the duty-cycle targets for SEU and load-following are achieved, AECL is developing a new fuel bundle, termed CANFLEX (CANdu FLEXible), where flexible refers to the versatility of the bundle with respect to operational and fuel-cycle options. Though the initial purpose of the new 43-element bundle is to introduce SEU into CANDU, CANFLEX is extremely versatile in its application, and is compatible with other fuel cycles of interest: natural uranium in existing CANDU reactors, recycled uranium and mixed-oxides from light-water reactors, and thoria-based fuels. Capability with a variety of fuel cycles is the key to future CANDU success in the international market. The improved performance of CANFLEX, particularly at high burnups, will ensure that the full economic benefits of advanced fuels cycles are achieved. A proof-tested CANFLEX bundle design will be available in 1993 for large-scale commercial-reactor demonstration

  20. Canadian CANDU fuel development programs and recent fuel operating experience

    International Nuclear Information System (INIS)

    This paper summarizes the performance and operating experience of CANDU fuel in Canadian CANDU reactors in 1999 and 2000. The extremely low rate of fuel defects continues to demonstrate that CANDU fuel is performing exceptionally well. Over the 2-year period, the fuel bundle defect rate for all bundles irradiated in Canadian CANDU reactors has remained very low, between 0.011% (suspected defects) and 0.007% (confirmed defects). On a fuel element basis, this represents a rate of confirmed defects of about 0.0002%; this rate is approaching 2 defects per million fuel elements! This successful performance is the result of a number of contributing factors, including a simple and robust fuel design with conservative design margins, reliable and specialized manufacturing processes that have been developed over the years, and fuel operations that conform to the fuel operating limits. Strong linkages between plant operation, designers, and Canadian fuel research and development programs also contribute to the high performance of the current CANDU fuel. The Fuel Design and Performance program, funded by the CANDU Owners Group, addresses licensing and operational issues that are common to the Canadian CANDU utilities. In addition, AECL's Fuel and Fuel Cycles working group directs R and D to support evolutionary improvements to the fuel products, as well as longer-term R and D for advanced fuel concepts. This paper describes the development programs in 1999/2000. (author)

  1. Canadian CANDU fuel development programs and recent fuel operating experience

    International Nuclear Information System (INIS)

    This paper summarizes the performance and operating experience of CANDU fuel in Canadian CANDU reactors in 1999 and 2000. The extremely low rate of fuel defects continues to demonstrate that CANDU fuel is performing exceptionally well. Over the 2-year period, the fuel bundle defect rate for all bundles irradiated in Canadian CANDU reactors has remained very low, between 0.011% (suspected defects) and 0.007% (confirmed defects). On a fuel element basis, this represents a rate of confirmed defects of about 0.0002%; this rate is approaching 2 defects per million fuel elements. This successful performance is the result of a number of contributing factors, including a simple and robust fuel design with conservative design margins, reliable and specialized manufacturing processes that have been developed over the years, and fuel operations that conform to the fuel operating limits. Strong linkages between plant operation, designers, and Canadian fuel research and development programs also contribute to the high performance of the current CANDU fuel. The Fuel Design and Performance program, funded by the CANDU Owners Group, addresses licensing and operational issues that are common to the Canadian CANDU utilities. In addition, AECL's Fuel and Fuel Cycles working group directs R and D to support evolutionary improvements to the fuel products, as well as long-term R and D for advanced fuel concepts. This paper describes the development programs in 1999/2000. (author)

  2. Recycled uranium: An advanced fuel for CANDU reactors

    International Nuclear Information System (INIS)

    The use of recycled uranium (RU) fuel offers significant benefits to CANDU reactor operators particularly if used in conjunction with advanced fuel bundle designs that have enhanced performance characteristics. Furthermore, these benefits can be realised using existing fuel production technologies and practices and with almost negligible change to fuel receipt and handling procedures at the reactor. The paper will demonstrate that the supply of RU as a ceramic-grade UO2 powder will increasingly become available as a secure option to virgin natural uranium and slightly enriched uranium(SEU). In the context of RU use in Canadian CANDU reactors, existing national and international transport regulations and arrangements adequately allow all material movements between the reprocessor, RU powder supplier, Canadian CANDU fuel manufacturer and Canadian CANDU reactor operator. Studies have been undertaken of the impact on personnel dose during fuel manufacturing operations from the increased specific activity of the RU compared to natural uranium. These studies have shown that this impact can be readily minimised without significant cost penalty to the acceptable levels recognised in modem standards for fuel manufacturing operations. The successful and extensive use of RU, arising from spent Magnox fuel, in British Energy's Advanced Gas-Cooled reactors is cited as relevant practical commercial scale experience. The CANFLEX fuel bundle design has been developed by AECL (Canada) and KAERI (Korea) to facilitate the achievement of higher bum-ups and greater fuel performance margins necessary if the full economic potential of advanced CANDU fuel cycles are to be achieved. The manufacture of a CANFLEX fuel bundle containing RU pellets derived from irradiated PWR fuel reprocessed in the THORP plant of BNFL is described. This provided a very practical verification of dose modelling calculations and also demonstrated that the increase of external activity is unlikely to require any

  3. Candu 6: versatile and practical fuel technology

    International Nuclear Information System (INIS)

    CANDU reactor technology was originally developed in Canada as part of the original introduction of peaceful nuclear power in the 1960s and has been continuously evolving and improving ever since. The CANDU reactor system was defined with a requirement to be able to efficiently use natural uranium (NU) without the need for enrichment. This led to the adaptation of the pressure tube approach with heavy water coolant and moderator together with on-power fuelling, all of which contribute to excellent neutron efficiency. Since the beginning, CANDU reactors have used [NU] fuel as the fundamental basis of the design. The standard [NU] fuel bundle for CANDU is a very simple design and the simplicity of the fuel design adds to the cost effectiveness of CANDU fuelling because the fuel is relatively straightforward to manufacture and use. These characteristics -- excellent neutron efficiency and simple, readily-manufactured fuel -- together lead to the unique adaptability of CANDU to alternate fuel types, and advancements in fuel cycles. Europe has been an early pioneer in nuclear power; and over the years has accumulated various waste products from reactor fuelling and fuel reprocessing, all being stored safely but which with passing time and ever increasing stockpiles will become issues for both governments and utilities. Several European countries have also pioneered in fuel reprocessing and recycling (UK, France, Russia) in what can be viewed as a good neighbor policy to make most efficient use of fuel. The fact remains that CANDU is the most fuel efficient thermal reactor available today [NU] more efficient in MW per ton of U compared to LWR's and these same features of CANDU (on-power fuelling, D2O, etc) also enable flexibility to adapt to other fuel cycles, particularly recycling. Many years of research (including at ICN Pitesti) have shown CANDU capability: best at Thorium utilization; can use RU without re-enrichment; can readily use MOX. Our premise is that

  4. CANDU fuel bundle skin friction factor

    International Nuclear Information System (INIS)

    Single-phase, incompressible fluid flow skin friction factor correlations, primarily for CANDU 37-rod fuel bundles, were reviewed. The correlations originated from curve-fits to flow test data, mostly with new fuel bundles in new pressure tubes (flow tubes), without internal heating. Skin friction in tubes containing fuel bundles (noncircular flow geometry) was compared to that in equivalent diameter smooth circular tubes. At Reynolds numbers typical of normal flows in CANDU fuel channels, the skin friction in tubes containing bundles is 8 to 15% higher than in equivalent diameter smooth circular tubes. Since the correlations are based on scattered results from measurements, the skin friction with bundles may be even higher than indicated above. The information permits over- or under-prediction of the skin friction, or choosing an intermediate value of friction, with allowance for surface roughnesses, in thermal-hydraulic analyses of CANDU heat transport systems. (author) 9 refs., 2 figs

  5. fuel management in candu reactors: RFSP code

    International Nuclear Information System (INIS)

    The objective of in-core fuel management is to determine the required refuelling strategies for safe and reliable operation of the reactor with minimum total energy cost. CANDU reactors use natural uranium fuel and rely on semi-continuous on-power refuelling. For the purpose of fuel management, the CANDU core with 380 fuel channels is modelled dividing into inner-and outer core. Refuelling rate in the CANDU reactors is evaluated in three periods for the whole operating life: 1)From the initial core to refuelling onset (100-150 EFPD), 2) the intermediate period (400-500 EFPD), and 3)the equilibrium period (approximately 30 years). A channel in the CANDU-6 reactor contains 12 bundles, in the refuelling operation some bundles do not discharged, but are shifted to other place in the same channel. One of the methods used for selection the channel and determination the bundles to be discharged is simulation method one of which is the RFSP (reactor fuelling simulating program). RFSP is a computer programme to do neutronic calculations for CANDU reactors. It can calculate both static and time-dependent neutron flux and power distributions in the core. It is a modular program containing a lot of modules. RFSP can perform fuel-management calculations and simulate a reactor operating history at specified intervals, taking burnup steps and channel refuelling into account

  6. Dynamic Analysis of the Thorium Fuel Cycle in CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Joon; Park, Chang Je

    2006-02-15

    The thorium fuel recycle scenarios through the Canada deuterium uranium (CANDU) reactor have been analyzed for two types of thorium fuel: homogeneous ThO{sub 2}UO{sub 2} and ThO{sub 2}UO{sub 2}-DUPIC fuels. The recycling is performed through the dry process fuel technology which has a proliferation resistance. For the once-through fuel cycle model, the existing nuclear power plant construction plan was considered up to 2016, while the nuclear demand growth rate from the year 2016 was assumed to be 0%. After setting up the once-through fuel cycle model, the thorium fuel CANDU reactor was modeled to investigate the fuel cycle parameters. In this analysis, the spent fuel inventory as well as the amount of plutonium, minor actinides and fission products of the multiple recycling fuel cycle were estimated and compared to those of the once-through fuel cycle. From the analysis results, it was found that the closed or partially closed thorium fuel cycle can be constructed through the dry process technology. Also, it is known that both the homogeneous and heterogeneous thorium fuel cycles can reduce the SF accumulation and save the natural uranium resource compared with the once-through cycle. From the material balance view point, the heterogeneous thorium fuel cycle seems to be more feasible. It is recommended, however, the economic analysis should be performed in future.

  7. 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

  8. Release of segregated nuclides from spent fuel

    International Nuclear Information System (INIS)

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

  9. 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

  10. 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

  11. 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

  12. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

  14. Spent fuel management of NPPs in Argentina

    International Nuclear Information System (INIS)

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

  15. Spent fuel management strategies

    International Nuclear Information System (INIS)

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

  16. 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

  17. The management strategy of spent nuclear fuel

    International Nuclear Information System (INIS)

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

  18. Recent advances in thorium fuel cycles for CANDU reactors

    International Nuclear Information System (INIS)

    The once-through thorium fuel cycle in CANDU reactors provides an evolutionary approach to exploiting the energy potential of thorium. In the 'mixed bundle' strategy, the central 8 elements in a CANFLEX fuel bundle contain thoria, while the outermost 35 elements contain slightly enriched uranium (SEU). Detailed full-core fuel-management simulations have shown that this approach can be successfully implemented in existing CANDU reactors. Uranium requirements are lower than for the natural uranium fuel cycle. Further energy can be derived from the thorium by recycling the irradiated thoria fuel elements, containing 233U, as-is without any processing, into the center of a new mixed bundle. There are several examples of such 'demountable' bundles. Recycle of the central 8 thoria elements results in an additional burnup of 20 MW·d/kgHE from the thoria elements, for each recycle. The reactivity of these thoria elements remains remarkably constant over irradiation for each recycle. The natural uranium requirements for the mixed bundle (which includes the natural uranium feed required for the outer SEU fuel elements), without recycle, is about 10% lower than for the natural uranium fuel cycle. After the first recycle, the uranium requirements are -35% lower than for the natural uranium cycle, and remain fairly constant with further recycling (the total uranium requirement averaged over a number of cycles is 30% lower than a natural uranium fuelled CANDU reactor). This thorium cycle strategy is a cost-effective means of reducing uranium requirements, while producing a stockpile of valuable 233U, safeguarded in the spent fuel, that can be recovered in the future when predicated by economic or resource considerations. (author)

  19. 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

  20. CANDU fuel : safe, reliable and flexible. 12th international conference on CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-07-01

    The Canadian Nuclear Society's 12th International Conference on CANDU Fuel was hosted by Royal Military College of Canada in Kingston, Ontario, Canada on September 15-18, 2013. The theme for the conference was 'CANDU Fuel : Safe, Reliable & Flexible' bringing together international experts of the nuclear fuel industry and academia involved in design, R and D, manufacturing, operation, modeling, safety analysis, and regulations. Over 100 delegates including representatives from other countries, including India, Romania, Argentina, Korea, United States, Austria, and Canada attended this truly successful international event. Although CANDU fuel has performed well a number of the presentations were on a modified design of the standard 37 element bundle called 37M which is now being loaded into the Darlington reactors. The renewed interest in thorium was also the focus of several presentations.

  1. CANDU-PHW fuel channel replacement experience

    International Nuclear Information System (INIS)

    One of the main characteristics of the CANDU pressurized heavy water reactor is the use of pressure tubes rather than one large pressure vessel to contain the fuel and coolant. This provides an inherent design capability to permit their replacement in an expeditious manner, without seriously affecting the high capacity factors of the reactor units. Of th eight Ontario Hydro commercial nuclear generating units, the lifetime performance places seven of them (including two that have had some of their fuel channels replaced), in the top ten positions in the world's large nuclear-electric unit performance ranking. Pressure tube cracks in the rolled joint region have resulted in 70 fuel channels being replaced in three reactor units, the latest being at the Bruce Nuclear Generating Station 'A', Unit 2 in February 1982. The rolled joint design and rolling procedures have been modified to eliminate this problem on CANDU units subsequent to Bruce 'A'. This paper describes the CANDU pressure tube performance history and expectations, and the tooling and procedures used to carry out the fuel channel replacement

  2. 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).

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

    International Nuclear Information System (INIS)

    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)

  4. 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)

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

    International Nuclear Information System (INIS)

    Spent fuel management is an important issue for nuclear power program, requiring careful planning and implementation. With the wait-and-see policy on spent fuel management in Korea, research efforts are directed at KAERI to develop advanced technologies for safer and more efficient management of the accumulating spent fuels. In support of these research perspectives, a test facility of pilot scale is being developed with provisions for integral demonstration of a multitude of technical functions required for spent fuel management. The facility, baptized SMART (Spent fuel MAnagement technology Research and Test facility), is to be capable of handling full size assembly of spent PWR fuel (as well as CANDU fuel) with a maximum capacity of 10 MTU/y (about 24 assemblies of PWR type). Major functions of the facility are consolidation of spent PWR fuel assembly into a half-volume package and optionally transformation of the fuel rod into a fuel of CANDU type (called DUPIC). Objectives of these functions are to demonstrate volume reduction of spent fuel (for either longer-term dry storage or direct disposal ) in the former case and direct refabrication of the spent PWR fuel into CANDU-type DUPIC fuel for reuse in CANDU reactors in the latter case, respectively. In addition to these major functions, there are other associated technologies to be demonstrated : such as waste treatment, remote maintenance, safeguards, etc. As the facility is to demonstrate not only the functional processes but also the safety and efficiency of the test operations, engineering criteria equivalent to industrial standards are incorporated in the design concept. The hot cell structure enclosing the radioactive materials is configured in such way to maximize costs within the given functional and operational requirements. (author). 3 tabs., 4 figs

  6. CANDU fuel behaviour under transient conditions

    International Nuclear Information System (INIS)

    The Canadian R and D program to understand CANDU fuel behaviour under transient conditions is described. Fuel sheath behaviour studies have led to the development of a model of transient plastic strain in inert gas, which integrates the deformation due to several mechanisms. Verification tests demonstrated that on average the model overpredicts strain by 20%. From oxidation kinetics studies a sheath failure embrittlement criterion based on oxygen distribution has been developed. We have also established a rate equation for high-temperature stress-dependent crack formation due to embrittlement of the sheath by beryllium. An electric, simulated fuel element is being used in laboratory tests to characterize the behaviour of fuel in the horizontal. In-reactor, post-dryout tests have been done for several years. There is an axially-segmented, axisymmetric fuel element model in place and a fully two-dimensional code is under development. Laboratory testing of bundles, in its early stages, deals with the effects of geometric distortion and sheath-to-sheath interaction. In-reactor, post-dryout tests of CANDU fuel bundles with extensive central UO2 melting did not result in fuel fragmentation nor damage to the pressure tube. (author)

  7. Slightly enriched uranium in CANDU: An economic first step towards advanced fuel cycles

    International Nuclear Information System (INIS)

    The natural-uranium fuelled Canada Deuterium-Uranium (CANDU) nuclear reactor system has proven to be a safe, reliable and economical producer of electricity for over a quarter of a century. The CANDU system, however, is not restricted to the use of natural-uranium fuel; a wide range of advanced fuel cycles can be accommodated. In the short term, slightly enriched uranium (SEU) is the most promising of these advanced fuel cycles. SEU offers a reduction in the total fuel cycle cost of between 25 and 50% relative to natural-uranium fuel. Uranium consumption is decreased by 30 to 40%. In addition the volume of spent fuel is reduced by a factor of two to three, depending on the enrichment selected. SEU also offers greater flexibility in the design of future CANDU reactors. A variety of fuel management options can be employed in CANDU with slightly enriched fuels. Fuel performance is expected to be good for the burnups of interest, but further fuel testing is planned and is currently in progress in order to confirm this. Programs in place at Atomic Energy of Canada Limited (AECL) will lead to the demonstration and introduction of slightly enriched uranium in CANDU. Ontario Hydro, a Canadian utility with twenty CANDUs operating or under construction, is considering a program which could lead to the implementation of SEU in its nuclear generating stations. (author). 30 refs, 7 figs

  8. 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)

  9. Performance testing of CANDU MOX fuel

    International Nuclear Information System (INIS)

    CANDU fuel bundles containing 0.5 wt % plutonium in natural uranium were fabricated at Chalk River Laboratories and were successfully irradiated in the NRU reactor at powers up to 65 Min and to burnups ranging from 13 to 23 MW·d/kg HE. Two of the bundles experienced power histories that bound the normal powers and burnups of natural UO2 CANDU fuel (2 fuel. Significantly more grain growth was observed than that typically expected for UO2 fuel; however, this increase in grain growth had no apparent effect on the overall performance of the fuel. Pellet-centre columnar grain growth was accompanied by plutonium homogenization. Two other MOX bundles operated to extended burnups of 19 to 23 MW·d/kg HE. Burnup extension above 15 MW·d/kg HE had no apparent effect on sheath strain or grain growth, and only a small effect on FGR and the amount of oxide observed on the inner surface of the sheath. (author)

  10. Performance testing of CANDU MOX fuel

    International Nuclear Information System (INIS)

    CANDU fuel bundles containing 0.5 wt % plutonium in natural uranium were fabricated at Chalk River Laboratories and were successfully irradiated in the NRU reactor at powers up to 65 kW/m and to burnups ranging from 13 to 23 MW·d/kg HE. Two of the bundles experienced power histories that bound the normal powers and burnups of natural UO2 CANDU fuel (2 fuel. Significantly more grain growth was observed than that typically expected for UO2 fuel; however, this increase in grain growth had no apparent effect on the overall performance of the fuel. Pellet-centre columnar grain growth was accompanied by plutonium homogenization. Two other MOX bundles operated to extended burnups of 19 to 23 MW·d/kg HE. Burnup extension above 15 MW·d/kg HE had no apparent effect on sheath strain or grain growth, and only a small effect on FGR and the amount of oxide observed on the inner surface of the sheath. (author)

  11. Decrease of the CANDU spent nuclear waste inventories in fusion-fission (hybrid) reactors

    International Nuclear Information System (INIS)

    The possibility of spent nuclear fuel rejuvenation in fusion reactors is investigated for both (D,T) and catalyzed (D,D) modes. The analysis is conducted for a CANDU spent nuclear fuel which was used up to a total enrichment grade of 0.418%. The behavior of the spent fuel is observed during 48 months for discrete time intervals of Δt = 6 months. The cooling of the fissile fuel zone is considered with three different coolants, notably gas (preferably He), Flibe (Li2BeF4) and natural lithium. A rejuvenation period of 8 months is evaluated for a final fissile fuel enrichment grade of 1% for all coolant types in the fissile zone under a first-wall fusion neutron current load of 1,014 (2.45-MeV n/cm2.s) and 1,014 (14.1-MeV n/cm2.x), corresponding to 2.64 MW/m2 by a plant factor of 75% for the catalyzed (D,D) fusion reactor mode. The rejuvenation period increases to 12 months for the same fissile fuel enrichment grade using the (D,T) fusion reactor mode under a first-wall fusion neutron current load of 1,014 (2.45-MeV n/cm2.s), corresponding to 2.25 MW/m2 by a plant factor of 75%. This enrichment would be sufficient for a re-utilization in a CANDU reactor

  12. 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

  13. Spent fuel management

    International Nuclear Information System (INIS)

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

  14. Behaviour of CANDU fuel during LOCA

    International Nuclear Information System (INIS)

    A large break loss-of-coolant accident (LOCA) in a CANDU nuclear reactor would result in a rapid increase of fuel and sheath temperatures. The temperature increase would, in turn, increase the gas pressure within the fuel and reduce the strength of the sheath material. Outside the fuel the loss of coolant from the primary heat transport system decreases pressure. The resulting pressure difference would cause deformation of the hot fuel sheath. Under certain circumstances, the deformation could be severe enough to fail the sheath thus releasing the fission products to the primary heat transport system. The computer code ELOCA-A is used to model the transient fuel behaviour following such an accident. ELOCA-A is a modified version of ELOCA.Mk 2 enabling us to consider the effects of axial variations in the microstructure of the sheath material caused by brazing of appendages to the sheath. The ELOCA-A code also features modelling of axial variations in neutron flux, pellet heat generation rate and heat transfer to the coolant. It predicts fuel pellet and sheath temperatures, sheath oxidation, sheath strain and probability of beryllium assisted cracking. A loss-of-coolant accident (LOCA) experiment was jointly sponsored by AECL and Ontario Hydro in the Power Burst Facility (PBF) at Idaho National Engineering Laboratories (INEL). This test was undertaken to provide an all-effects verification of the understanding of CANDU fuel behaviour during LOCA's. An extensive out-pile experimental program had provided single effects data which had been used for modelling such excursions. Integrated out-pile tests have confirmed our understanding of and accurate modelling of fuel under LOCA conditions. The integrated test in PBF provided the final proof that our understanding was complete and provided an experimental database for verification of transient fuel codes (4). The experiment was performed with modified CANDU fuel elements. The post-test measurements are compared with

  15. Canflex: A fuel bundle to facilitate the use of enrichment and fuel cycles in CANDU reactors

    International Nuclear Information System (INIS)

    The neutron economy of the CANDU reactor results in it being an ideal host for a number of resource-conserving fuel cycles, as well as a number of potential ''symbiotic'' fuel cycles, in which fuel discharged from light-water cooled reactors is recycled to extract the maximum energy from the residual fissile material before it is sent for disposal. The resource conserving fuel cycles include the natural-uranium, slightly-enriched-uranium and thorium fuel cycles. The ''LWR-symbiotic'' cycles include recovered uranium and various options for the direct use of spent LWR fuel in CANDU reactors. However, to achieve the maximum economic potential of these fuel-cycle options requires irradiation to burnups higher than that possible with natural uranium. To provide a basis for the economic use of these fuel cycles, a program is underway to develop and demonstrate a CANDU fuel bundle capable of both higher burnups and greater operating margins. This new bundle design is being developed jointly by AECL and KAERI, and 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 allows operation to burnups greater than 21 MWd/KgU. A combination of this lower peak-element rating, plus development work underway at AECL to enhance the thermalhydraulic characteristics of the bundle (including both critical heat flux and bundle pressure drop), provides a greater operating margin for the bundle. This new bundle design is called CANFLEX, and the program for its development in Canada and Korea is described in this paper. (author). 19 refs, 5 figs

  16. 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

  17. 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.)

  18. RU-43 a new uranium fuel bundle design for using in CANDU type reactors

    International Nuclear Information System (INIS)

    A unique feature of the CANDU reactor design is its ability to use alternative fuel cycles other than natural uranium (NU), without requiring major modifications to the basic reactor design. These alternative fuel cycles, which are known as advanced fuel cycles, utilize a variety of fissile materials, including Slightly Enriched Uranium (SEU) from enrichment facilities, and Recovered Uranium (RU) obtained from the reprocessing of the spent fuel of light-water reactors (LWR). 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 sufficient high neutron economy to use RU as fuel. RU from spent LWR fuel can be considered as a lower cost source of enrichment at the optimal enrichment level for CANDU fuel pellets. In Europe the feedstock of RU is approaching thousands tones and would provide sufficient fuel for hundreds CANDU-6 reactors years of operation. The use of RU fuel offers significant benefits to CANDU reactor operators. RU fuels improve fuel cycle economics by increasing the fuel burnup, which enables large cost reductions in fuel consumption and in spent fuel disposal. RU fuel offers enhanced operating margins that can be applied to increase reactor power. These benefits can be realized using existing fuel production technologies and practices, and with almost negligible changes to fuel receipt and handling procedures at the reactor. The application of RU fuel could be an important element in NPP Cernavoda from Romania. For this reason the Institute for Nuclear Research (INR), Pitesti has started a research programme aiming to develop a new fuel bundle RU-43 for extended burnup operation. The current version of the design is the result of a long process of analyses and improvements, in which successive preliminary design versions have been evaluated. The most relevant calculations performed on this fuel element design version are presented. Also, the stages of an experimental

  19. Development of fabrication technology for CANDU advanced fuel -Development of the advanced CANDU technology-

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Chang Beom; Kim, Hyeong Soo; Kim, Sang Won; Seok, Ho Cheon; Shim, Ki Seop; Byeon, Taek Sang; Jang, Ho Il; Kim, Sang Sik; Choi, Il Kwon; Cho, Dae Sik; Sheo, Seung Won; Lee, Soo Cheol; Kim, Yoon Hoi; Park, Choon Ho; Jeong, Seong Hoon; Kang, Myeong Soo; Park, Kwang Seok; Oh, Hee Kwan; Jang, Hong Seop; Kim, Yang Kon; Shin, Won Cheol; Lee, Do Yeon; Beon, Yeong Cheol; Lee, Sang Uh; Sho, Dal Yeong; Han, Eun Deok; Kim, Bong Soon; Park, Cheol Joo; Lee, Kyu Am; Yeon, Jin Yeong; Choi, Seok Mo; Shon, Jae Moon [Korea Atomic Energy Res. Inst., Taejon (Korea, Republic of)

    1994-07-01

    The present study is to develop the advanced CANDU fuel fabrication technologies by means of applying the R and D results and experiences gained from localization of mass production technologies of CANDU fuels. The annual portion of this year study includes following: 1. manufacturing of demo-fuel bundles for out-of-pile testing 2. development of technologies for the fabrication and inspection of advanced fuels 3. design and munufacturing of fuel fabrication facilities 4. performance of fundamental studies related to the development of advanced fuel fabrication technology.

  20. Study on spent fuel rejuvenation in PROMETHEUS fusion reactor

    International Nuclear Information System (INIS)

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

  1. 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)

  2. 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)

  3. Fabrication of PWR fuel assembly and CANDU fuel bundle

    International Nuclear Information System (INIS)

    For the project of localization of nuclear fuel fabrication, the R and D to establish the fabrication technology of CANDU fuel bundle as well as PWR fuel assembly was carried out. The suitable boss height and the prober Beryllium coating thickness to get good brazing condition of appendage were studied in the fabrication process of CANDU fuel rod. Basic Studies on CANLUB coating method also were performed. Problems in each fabrication process step and process flow between steps were reviewed and modified. The welding conditions for top and bottom nozzles, guide tube, seal and thimble screw pin were established in the fabrication processes of PWR fuel assembly. Additionally, some researches for a part of PWR grid brazing problems are also carried out

  4. 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)

  5. CANDU fuel compression tests at elevated temperatures

    International Nuclear Information System (INIS)

    An inlet header large break loss of coolant accident (LOCA) in CANDU reactors with fuelling against flow can cause the fuel to shift in the channels with a consequent reactivity insertion. This results in an increased fuel power transient, and a potential increase in the mialyzed consequences for such events. As the reactor's age and the channel axial gaps increase, the magnitude of the predicted power u-dmient increases. A design solution to reduce the power transient is to limit the amount of fuel movement by reducing the channel axial gap. This solution was implemented into Ontario Hydro's Bruce B and Darlington reactors. A consequence of a reduced channel axial gap is the potential for the fuel column axial expansion to become constrained by the channel end components in large break LOCAs. This experimental program investigated the effects of pellet cracking and elevated sheath temperatures on the ability of the fuel elements, of the 37-element bundle design, to sustain axial loads. The unirradiated fuel elements tested were either in the as-received condition or with the U02 fuel pellets cracked in a mechanical process to simulate the effect of inufflation. The load deformation characteristics demonstrated that, for a given amount of axial compression. the loads sustainable by the elements at elevated sheath temperatures were low. As a result. excess axial expansion would be easily accommodated without further challenge to pressure tube integrity. (author)

  6. 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

  7. Thorium fuel-cycle studies for CANDU reactors

    International Nuclear Information System (INIS)

    The high neutron economy of the CANDU reactor, its ability to be refuelled while operating at full power, its fuel channel design, and its simple fuel bundle provide an evolutionary path for allowing full exploitation of the energy potential of thorium fuel cycles in existing reactors. AECL has done considerable work on many aspects of thorium fuel cycles, including fuel-cycle analysis, reactor physics measurements and analysis, fuel fabrication, irradiation and PIE studies, and waste management studies. Use of the thorium fuel cycle in CANDU reactors ensures long-term supplies of nuclear fuel, using a proven, reliable reactor technology. (author)

  8. The behaviour of transuranic mixed oxide fuel in a Candu-900 reactor

    International Nuclear Information System (INIS)

    The production of transuranic actinide fuels for use in current thermal reactors provides a useful intermediary step in closing the nuclear fuel cycle. Extraction of actinides reduces the longevity, radiation and heat loads of spent material. The burning of transuranic fuels in current reactors for a limited amount of cycles reduces the infrastructure demand for fast reactors and provides an effective synergy that can result in a reduction of as much as 95% of spent fuel waste while reducing the fast reactor infrastructure needed by a factor of almost 13.5 [1]. This paper examines the features of actinide mixed oxide fuel, TRUMOX, in a CANDUR* nuclear reactor. The actinide concentrations used were based on extraction from 30 year cooled spent fuel and mixed with natural uranium in 3.1 wt% actinide MOX fuel. Full lattice cell modeling was performed using the WIMS-AECL code, super-cell calculations were analyzed in DRAGON and full core analysis was executed in the RFSP 2-group diffusion code. A time-average full core model was produced and analyzed for reactor coefficients, reactivity device worth and online fuelling impacts. The standard CANDU operational limits were maintained throughout operations. The TRUMOX fuel design achieved a burnup of 27.36 MWd/kg HE. A full TRUMOX fuelled CANDU was shown to operate within acceptable limits and provided a viable intermediary step for burning actinides. The recycling, reprocessing and reuse of spent fuels produces a much more sustainable and efficient nuclear fuel cycle. (authors)

  9. 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)

  10. Spent fuel pyroprocessing demonstration

    International Nuclear Information System (INIS)

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

  11. 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

  12. R and D activities at INR pitesti related to safety and reliability of CANDU type fuel

    International Nuclear Information System (INIS)

    The focus of Nuclear Fuel R and D Program of Institute for Nuclear Research (INR) Pitesti is to maintain and improve the reliability, economics and safety of 37-element natural uranium CANDU fuel bundles in Cernavoda Nuclear Generating Station (CNGS). The second requirement is to improve the CANDU fuel design and to develop 43-element advanced fuel bundle that will reduce capital and fuelling cost, increase the operating and safety margins, improve natural - uranium utilization, and provide synergy with other reactor systems to improve resource utilization and spent fuel management. An experimental database of fuel behaviour parameters including fission - gas release, sheath strain, power - burnup history etc. has been obtained using in-pile measurements and PIE results of CANDU fuel elements irradiated in the TRIGA Material Testing Reactor (MTR) of INR Pitesti. In last time the data base was updated to include the results of Power Pulse Tests performed in TRIGA - Annular Core Pulse Reactor (ACPR) of INR Pitesti. One of the current research objective of our fuel bahaviour studies is to investigate the reliability behaviour of CANDU type fuel during power cycling operation condition. The INR research programme also include the out pile separate effects experiments to evaluate properties of the UO2 and cladding and development of computer models to describe sheath deformation and gas release processes. A program for LOCA simulating in-reactor tests is in progress at INR Pitesti to provide a database for verification of transient fuel performance codes and demonstrate that the significant fuel behaviour phenomena have all been included in the models.This data base is used extensively for the validation of the fuel behaviour codes. This paper summarizes R and D activities of INR Pitesti, related to safety and reliability of CANDU type fuel and presents some of the recent results obtained from in reactor tests. (author)

  13. Reducing the impact of used fuel by transmuting actinides in a CANDU reactor

    International Nuclear Information System (INIS)

    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 in CANDU reactors to reduce the decay heat period. Many of the design features of the CANDU reactor make it uniquely adaptable to actinide transmutation. The small, simple fuel bundle facilitates the fabrication and handling of active fuels. Online refueling 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 in CANDU reactors, including both recent and past activities. The transmutation schemes that are presented reflect several different partitioning schemes and include both 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. (author)

  14. 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)

  15. 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

  16. Fission Product Inventory in CANDU Fuel

    International Nuclear Information System (INIS)

    When the reactor is operated at power, fuel composition changes continuously. The fission reaction produces a large variety of fission fragments which are radioactive and decay into other isotopic species. For different accident analyses or operational events, detailed calculations of the fuel radioactive inventory (fission products and actinides) are needed. The present paper reviews two types of radioactive inventory calculations performed at Cernavoda NPP: one for determining the whole core inventory and one for determining the evolution of the inventory within fuel bundles stored in the Spent Fuel Bay. Two computer codes are currently used for radioactive inventory calculations: ORIGEN-S and ELESTRES-IST. The whole core inventory calculation was performed with both codes, the comparison showing that ELESTRES-IST gives a more conservative result. One of the challenges met during the analysis was to set a credible, yet conservative “image” of the in core fuel power/burnup distribution. Consequently, a statistical analysis was performed to find the best estimate plus uncertainties map for the power/burnup distribution of all in core fuel elements. For each power/burnup in the map, the fission product inventory was computed using a scaled irradiation history based on the Limiting Overpower Envelope. After the Fukushima accident, the problem of assessing the consequences of a loss of cooling event at the Spent Fuel Bay was raised. In order to estimate its impact, a calculation for determining the fission products inventory and decay heat evolution within the spent fuel bundles stored in the bay was performed. The calculation was done for a bay filled with fuel bundles up to its maximum capacity. The results obtained have provided a conservative estimation of the decay heat released and the expected evolution of the water temperature in the bay. This provided a technical basis for selecting the emergency actions required to cope with such events. (author)

  17. 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

  18. The behaviour of transuranic mixed oxide fuel in a Candu-900 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Morreale, A. C.; Ball, M. R.; Novog, D. R.; Luxat, J. C. [Dept. of Engineering Physics, McMaster Univ., 1280 Main St. W, Hamilton, ON (Canada)

    2012-07-01

    The production of transuranic actinide fuels for use in current thermal reactors provides a useful intermediary step in closing the nuclear fuel cycle. Extraction of actinides reduces the longevity, radiation and heat loads of spent material. The burning of transuranic fuels in current reactors for a limited amount of cycles reduces the infrastructure demand for fast reactors and provides an effective synergy that can result in a reduction of as much as 95% of spent fuel waste while reducing the fast reactor infrastructure needed by a factor of almost 13.5 [1]. This paper examines the features of actinide mixed oxide fuel, TRUMOX, in a CANDU{sup R}* nuclear reactor. The actinide concentrations used were based on extraction from 30 year cooled spent fuel and mixed with natural uranium in 3.1 wt% actinide MOX fuel. Full lattice cell modeling was performed using the WIMS-AECL code, super-cell calculations were analyzed in DRAGON and full core analysis was executed in the RFSP 2-group diffusion code. A time-average full core model was produced and analyzed for reactor coefficients, reactivity device worth and online fuelling impacts. The standard CANDU operational limits were maintained throughout operations. The TRUMOX fuel design achieved a burnup of 27.36 MWd/kg HE. A full TRUMOX fuelled CANDU was shown to operate within acceptable limits and provided a viable intermediary step for burning actinides. The recycling, reprocessing and reuse of spent fuels produces a much more sustainable and efficient nuclear fuel cycle. (authors)

  19. Studies at INR-Pitesti for developing fuels of high burnup suitable to CANDU 6 reactor

    International Nuclear Information System (INIS)

    Increasing burnup allows the utility to get the same kWh output with a diminished tonnage of fissile material and provides a saving in the cost of fuel manufacturing as well as of spent fuel disposal. The RU, SEU, MOX, DUPIC fuel cycles and CANFLEX fuel bundles concept compatible with CANDU 6 reactor are presented. INR projects for developing SEU 43 fuel bundles supported by IAEA-Vienna are also presented. Particularly, one gives an overlook of standard CANDU and advanced SEU 43 nuclear fuel cycles. The paper presents also the current and future directions of studies implied by the research program in the nuclear fuel field of RAAN (The Autonomous Authority for Nuclear Activities). Among these, mentioned are: working out of the manual of physics of CANDU core with slightly enriched uranium; technological studies aiming at reducing the effects of limiting factors of the fuel lifetime and at burnup extension; obtaining new fuels as vectors of advanced cycles; off reactor tests of SEU 43 clusters; in-reactor tests of SEU 43 experimental fuel elements; developing computer codes for analysis of SEU, MOX and DUPIC fuel behavior; in-reactor tests of experimental MOX and DUPIC elements

  20. Study of advanced nuclear fuel cycles in Candu type power reactors

    International Nuclear Information System (INIS)

    The fuel burn up can be increased to a large extent, up to 14, 0000 MWD/te, by using the slightly enriched uranium or Pu mixed fuel in CANDU type power reactors. In the present study, the previous work was extended to compare the isotopic inventories and corresponding activities of important nuclides for different fuel cycles of a CANDU 600 type power reactor. The detail can be found in our studies. The calculations were performed using the computer code WIMSD4. The isotopic inventories and corresponding activities were calculated versus the fuel burn-up for the natural UO/sub 2/ fuel, 1.2 % enriched UO/sub 2/ fuel and 0.45 % PuO/sub 2/-UO/sub 2/ fuel. It was found that 1.2 % enriched uranium fuel has the lowest activity as compared to other two fuel cycles. It means that improvement in the fuel cycle technology of CANDU type power reactors can lead to high burn up which results in the reduction of actinide content in the spent fuel, and hence has a good environmental impact. (orig./A.B.)

  1. 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)

  2. Spent fuel management in Switzerland

    International Nuclear Information System (INIS)

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

  3. 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

  4. 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

  5. Economic and system aspects of CANDU advanced fuel cycles

    International Nuclear Information System (INIS)

    It is somewhat a paradox that Canada, which ranks as one of the world's leading uranium producers and has large economic uranium resources, should also have developed the CANDU reactor. This reactor is the most fuel efficient of all reactor types which are commercially available at the present time. The explanation of the paradox is that the design basis of the CANDU was established three decades ago when the full extent of Canadian uranium resources was unknown, and an early transition to recycle fuelling was anticipated as being necessary to sustain a growing power generation system. Consequently, the objectives of fuel efficiency and flexibility in using a variety of uranium, plutonium and thorium fuels were established at an early stage. One result of this is the ability to use the current design of CANDU in an advanced converter role with very little change in reactor design or operating procedures. As a result, in projections of future power costs, all major uncertainty is focused on fuel cycle parameters since the capital and operating costs are well defined by current commercial experience. The paper will examine the economic and resource characteristics of CANDU in an advanced converter role, both in terms of stand-alone technology and as a partner in a CANDU-light-water-reactor and in a CANDU-fast-breeder-reactor system. The use of results to establish cost targets to guide the current research and development program will be discussed, together with considerations of deployment strategy. (author)

  6. CANDU fuel elements behaviour in the load following tests

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, Grigore [Instiute for Nuclear Research, Pitesti (Romania). Nuclear Fuel Engineering Lab.; Palleck, Steve [Sheridan Park Research-AECL, Mississauga, ON (Canada). Fuel Deisgn Branch

    2011-08-15

    Two load following (LF) tests on CANDU type fuel elements were performed in TRIGA Research Reactor of INR Pitesti. In the first LF test the 78R fuel element has successfully experienced 367 power cycles, mostly between 23 and 56 kW/m average linear power. In the second LF test, the fuel element 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 test period. New LF tests are planed 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 TRIGA Research Reactor and their relation to CANDU fuel performance in LF conditions. (orig.)

  7. Selection criteria for spent fuel storage technologies

    International Nuclear Information System (INIS)

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

  8. 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

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

    International Nuclear Information System (INIS)

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

  10. Modified ADS molten salt processes for back-end fuel cycle of PWR spent fuel

    International Nuclear Information System (INIS)

    The back-end fuel cycle concept for PWR spent fuel is explained. This concept is adequate for Korea, which has operated both PWR and CANDU reactors. Molten salt processes for accelerator driven system (ADS) were modified both for the transmutation of long-lived radioisotopes and for the utilisation of the remained fissile uranium in PWR spent fuels. Prior to applying molten salt processes to PWR fuel, hydrofluorination and fluorination processes are applied to obtain uranium hexafluoride from the spent fuel pellet. It is converted to uranium dioxide and fabricated into CANDU fuel. From the remained fluoride compounds, transuranium elements can be separated by the molten salt technology such as electrowinning and reductive extraction processes for transmutation purpose without weakening the proliferation resistance of molten salt technology. The proposed fuel cycle concept using fluorination processes is thought to be adequate for our nuclear program and can replace DUPIC (Direct Use of spent PWR fuel in CANDU reactor) fuel cycle. Each process for the proposed fuel cycle concept was evaluated in detail

  11. Improving the service life and performance of CANDU fuel channels

    International Nuclear Information System (INIS)

    The development objective for CANDU fuel channels is to produce a design that can operate for 40 years at 90% capacity. Steady progress toward this objective is being made. The factors that determine the life of a CANDU fuel channel are reviewed and the processes necessary to achieve the objectives are identified. Performance of future fuel channels will be enhanced by reduced operating costs and increased safety margins to postulated accident conditions compared with those for current channels. The approaches to these issues are discussed briefly in this report. (author)

  12. 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)

  13. 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)

  14. Thermal-hydraulics analysis for advanced fuel to be used in Candu 600 nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Catana, Alexandru [RAAN, Institute for Nuclear Research, Str. Campului Nr. 1, Pitesti, Arges (Romania); Danila, Nicolae; Prisecaru, Ilie; Dupleac, Daniel [University POLITEHNICA of Bucharest (Romania)

    2008-07-01

    Two Candu 600 pressure tube nuclear reactors cover about 17% of Romania's electricity demand. These nuclear reactors are moderated/cooled with D{sub 2}O, fuelled on-power with Natural Uranium (NU) dioxide encapsulated in a standard (STD37) fuel bundle. High neutron economy is achieved using D{sub 2}O as moderator and coolant in separated systems. To reduce fuel cycle costs, programs were initiated in Canada, S.Korea, Argentina and Romania for the design and build new fuel bundles able to accommodate different fuel compositions. Candu core structure and modular fuel bundles, permits flexible fuel cycles. The main expected achievements are: reduced fuel cycle costs, increased discharge burn-up, plutonium and minor actinides management, thorium cycle, use of recycled PWR and in the same time waste minimization and operating cost reduction. These new fuel bundles are to be used in already operated Candu reactors. Advanced fuel bundle were proposed: CANFLEX bundle (Canada, S-Korea); the Romanian 'SEU43' bundle (Fig 1). In this paper thermal-hydraulic analysis in sub-channel approach is presented for SEU43. Comparisons with standard (STD37) fuel bundles are made using SEU-NU for NU fuel composition and SEU-0.96, for recycled uranium (RU) fuel with 0.96% U-235. Extended and comprehensive analysis must be made in order to assess the TH behaviour of SEU43. In this paper, considering STD37, SEU43-NU and SEU43-0.96 fuel bundles, main TH parameters were analysed: pressure drop, fuel highest temperatures, coolant density, critical heat flux. Differences between these fuel types are outlined. Benefits are: fuel costs reduction, spent fuel waste minimization, increase in competitiveness of nuclear power. Safety margins must be, at least, conserved. (authors)

  15. 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)

  16. Characteristics of used CANDU fuel relevant to the Canadian nuclear fuel waste management program

    International Nuclear Information System (INIS)

    Literature data on the characteristics of used CANDU power reactor fuel that are relevant to its performance as a waste form have been compiled in a convenient handbook. Information about the quantities of used fuel generated, burnup, radionuclide inventories, fission gas release, void volume and surface area, fuel microstructure, fuel cladding properties, changes in fuel bundle properties due to immobilization processes, radiation fields, decay heat and future trends is presented for various CANDU fuel designs. (author). 199 refs., 39 tabs., 100 figs

  17. Spent fuel storage at KURRI

    International Nuclear Information System (INIS)

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

  18. Spent fuel management in Switzerland

    International Nuclear Information System (INIS)

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

  19. 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)

  20. 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

  1. 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)

  2. Thorium fuel studies for CANDU reactors

    International Nuclear Information System (INIS)

    Applying the once-through Thorium (OTT) cycle in existing and advanced CANDU reactors might be seen as an evolved concept for the sustainable development both from the economic and waste management points of view. 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 higher burnup, lesser actinide inventory and proliferation resistance. The calculations were performed using the lattice codes WIMS and DRAGON (together with the corresponding nuclear data library based on ENDF/B-VII). (authors)

  3. WWER spent fuel storage

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  5. Spent fuel management in Japan

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  7. Spent fuel storage facility

    International Nuclear Information System (INIS)

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

  8. Seismic qualification of spent fuel storage stacks

    International Nuclear Information System (INIS)

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

  9. Remotely operated inspection equipment for the Candu fuel channels

    International Nuclear Information System (INIS)

    Equipment is described which has been successfully used for the nondestructive inspection of fuel channel components within Ontario Hydro's CANDU nuclear reactors. By the use of automated systems, significant savings in personnel radiation exposure and unit outage duration have been realized, with improved quality and quantity of nondestructive examination information. (author)

  10. 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)

  11. Effect of Candu Fuel Bundle Modeling on Sever Accident Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Dupleac, D.; Prisecaru, I. [Power Plant Engineering Faculty, Politehnica University, 313 Splaiul Independentei, 060042, sect. 6, Bucharest (Romania); Mladin, M. [Institute for Nuclear Research, Pitesti-Mioveni, 115400 (Romania)

    2009-06-15

    In a Candu 6 nuclear power reactor fuel bundles are located in horizontal Zircaloy pressure tubes through which the heavy-water coolant flows. Each pressure tube is surrounded by a concentric calandria tube. Outside the calandria tubes is the heavy-water moderator contained in the calandria itself. The moderator is maintained at a temperature of 70 deg. C by a separate cooling circuit. The moderator surrounding the calandria tubes provides a potential heat sink following a loss of core heat removal. The calandria vessel is in turn contained within a shield tank (or reactor vault), which provides biological shielding during normal operation and maintenance. It is a large concrete tank filled with ordinary water. During normal operation, about 0.4% of the core's thermal output is deposited in the shield tank and end shields, through heat transfer from the calandria structure and fission heating. In a severe accident scenario, the shield tank could provide an external calandria vessel cooling which can be maintained until the shield tank water level drops below the debris level. The Candu system design has specific features which are important to severe accidents progression and requires selective consideration of models, methods and techniques of severe accident evaluation. Moreover, it should be noted that the mechanistic models for severe accident in Candu system are largely less well validated and as the result the level of uncertainty remains high in many instances. Unlike the light water reactors, for which are several developed computer codes to analyze severe accidents, for Candu severe accidents analysis two codes were developed: MAAP4-Candu and ISAAC. However, both codes started by using MAAP4/PWR as reference code and implemented Candu 6 specific models. Thus, these two codes had many common features. Recently, a joint project involving Romanian nuclear organizations and coordinated by Politehnica University of Bucharest has been started. The purpose

  12. 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)

  13. 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

  14. Spent fuel treatment in Japan

    International Nuclear Information System (INIS)

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

  15. 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.

  16. IAEA spent fuel storage glossary

    International Nuclear Information System (INIS)

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

  17. 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

  18. 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.

  19. 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.

  20. 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)

  1. Spent fuel reprocessing. Main operations

    International Nuclear Information System (INIS)

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

  2. 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)

  3. 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.

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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

  9. Transportation of spent MTR fuels

    Energy Technology Data Exchange (ETDEWEB)

    Raisonnier, D.

    1997-08-01

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

  10. Technical feasibility of using RU-43 fuel in the CANDU-6 reactors of the Cernavoda NPP

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, G.; Patrulescu, I. [Inst. for Nuclear Research, Pitesti (Romania). Physics

    2008-03-15

    Recovered uranium (RU) is a by-product of many light-water reactor (LWR) fuel recycling programs. 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. The Institute for Nuclear Research (INR) Pitesti has analyzed the feasibility of using RU fuel with 0.9-1.1 w% {sup 235}U in the CANDU-6 reactors of the Cernavoda Nuclear Power Plant (Cernavoda NPP). Using RU fuel would produce a significant increase in the fuel discharge burnup, from 170 MWh/kgU currently achieves with natural-uranium (NU) fuel to about 355 MWh/kgU. This would lead to reduced fuel-cycle cost and a large reduction in spent-fuel volume per full-power-year of operation. The RU fuel bundle design with recovered uranium fuel, known as RU-43, is being developed by the INR Pitesti and is now at the stage of final design verification. Early work has been concentrated on RU-43 fuel bundle design optimization, safety and reactor physics assessment. The changes in fuel element and fuel bundle design contribute to the many advantages offered by the RU-43 bundle. Verification of the design of the RU-43 fuel bundle is performed in a way that shows that design criteria are met, and is mostly covered by proof tests such as flow and irradiation tests. The most relevant calculations performed on this fuel bundle design version are presented. Also, the stages of an experimental program aiming to verify the operating performance are briefly described in this paper. (orig.)

  11. Spent Fuel Management in Slovakia

    International Nuclear Information System (INIS)

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

  12. 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

  13. 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)

  14. Post-irradiation neutron emissions from CANDU fuels and their nuclear forensics applications

    International Nuclear Information System (INIS)

    Fissile materials within a fuel pellet can be discerned rapidly and non-destructively via the analysis of post-irradiation delayed neutron emissions. The delayed neutron counting technique is well established within the Canadian nuclear industry, and uses include the detection of defective CANDU fuel. This work discusses these neutron emissions from CANDU fuel in the context of detection and attribution. Monte Carlo simulations of these emissions from current and proposed CANDU fuels (thoria and mix oxide based) have been performed. These simulations are compared to measurements of delayed neutron emissions from 233U and 235U, and the feasibility of CANDU fuel characterization is discussed. (author)

  15. Spent Nuclear Fuel project, project management plan

    International Nuclear Information System (INIS)

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

  16. Spent Nuclear Fuel project, project management plan

    Energy Technology Data Exchange (ETDEWEB)

    Fuquay, B.J.

    1995-10-25

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

  17. Rejuvenation of spent fuel in the force-free helical reactor (FFHR)

    International Nuclear Information System (INIS)

    This study considers spent fuel rejuvenation potential of the force-free helical reactor (FFHR), which is a relevant heliotron-type D-T fusion reactor. For this purpose, three different spent fuels were selected, i.e.: (1) CANDU (2) PWR-UO2, and (3) PWR-MOX spent fuels. The spent fuel (volume fraction of 60%), spherically prepared and cladded with SiC (volume fraction of 10%), was located in the fuel zone (FZ) in the blanket of the FFHR. The FZ was cooled with high-pressured helium gas (volume fraction of 30%) for the nuclear heat transfer. The neutronic calculations were performed by solving the Boltzmann transport equation with the help of the neutron transport code XSDRNPM-S/SCALE4.3. The calculations of the time dependent atomic densities of the isotopes were performed for an operation period (OP) of up to 4 years by 75% plant factor (η) under a first-wall neutron load (P) of 1.5 MW/m2. The temporal variations of the atomic densities of the isotopes in the spent fuel composition and other physical parameters were calculated for a discrete time interval (Δt) of 1/12 year (one month), by using the interface program (code). In all investigated spent fuel cases, the tritium self-sufficiency is maintained for DT fusion driver along the OP. The CANDU spent fuel becomes usable in a conventional CANDU reactor after a regeneration time of ∼6 months. The CFFE value approaches 3.5% in the blanket fuelled with the PWR-UO2 and -MOX spent fuels after 40 and 34 months, respectively. The plutonium component can never reach a nuclear weapon grade quality during spent fuel rejuvenation. Consequently, the blanket has high neutronic performance for the rejuvenation of the spent fuels

  18. Romanian progress in the advanced CANDU fuel manufacturing

    Energy Technology Data Exchange (ETDEWEB)

    Ohai, D.; Benga, D. [RAAN, Inst. for Nuclear Research, Pitesti- Mioveni (Romania)]. E-mail: dohai@nuclear.ro

    2005-07-01

    The initial concept in developing an advanced fuel compatible with CANDU 6 Reactor, using part of Nuclear Fuel Plant (FCN) Pitesti facilities [1] should be revised. New aspects were considered: working within FCN area, a technological transfer suspicion appears (inobservance of AECL-FCN confidentiality agreement), and the enriched Uranium use on FCN area is prohibited (IAEA requirement). Under these conditions, the Institute for Nuclear Research (ICN) decided to develop or modernize its own facilities for nuclear fuel (CANDU type) manufacturing. The intention was to cover the main technological steps in fuel manufacturing, beginning with powder manufacturing and ending up with fuel bundle assembling. The development or modernization of own facilities for the nuclear fuel manufacturing open the possibilities for the collaboration with other entities interested in advanced fuel development. Having a Research Reactor for material testing and a Post Irradiation+ Facility, ICN can complete the irradiation and post-irradiation services with experimental fuel elements manufacturing, the services being completed. This can be a possibility to eliminate the interstates transport of nuclear materials. The new international requirements for the transport of the nuclear materials are drastic and need a lot of time and money for obtaining authorizations and for transport. It is financially advantageous to manufacture experimental fuel elements on the same site with the irradiation and post-irradiation facilities. (author)

  19. Spent fuel management in Japan

    International Nuclear Information System (INIS)

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

  20. Status and prospects for spent fuel management in China

    International Nuclear Information System (INIS)

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

  1. Nanocrystal and noble gas tagging for monitoring defective CANDU fuel bundles

    International Nuclear Information System (INIS)

    The purpose of this paper is to discuss two possible defective fuel bundle tagging techniques that have been suggested for CANDU-6 nuclear reactors. The general design of a CANDU-6 reactor and fuel bundle is reviewed. Nanocrystal tagging is introduced. A current production method for CdTe nanocrystals and future experimental goals are outlined and noble gas tagging is reviewed. Considerations for the future implementation of these tagging methods for fuel in a CANDU-6 reactor is also discussed. (author)

  2. Advances in fuel channel technology for CANDU reactors

    International Nuclear Information System (INIS)

    The components of the CANDU fuel channels are being developed to have service lives of over 30 years with large margins of safety. Information from research programs and the examination of components removed from reactors has enable improvements to be made to pressure tubes, spacers, calandria tubes and end fittings. Improvements have also been made to the channel design to facilitate planned retubing. (author). 22 refs., 5 tabs., 31 figs

  3. The effects of actinide based fuels on incremental cross sections in a Candu reactor

    International Nuclear Information System (INIS)

    The reprocessing of spent fuel such as the extraction of actinide materials for use in mixed oxide fuels is a key component of reducing the end waste from nuclear power plant operations. Using recycled spent fuels in current reactors is becoming a popular option to help close the fuel cycle. In order to ensure safe and consistent operations in existing facilities, the properties of these fuels must be compatible with current reactor designs. This paper examines the features of actinide mixed oxide fuel, TRUMOX, in a CANDU reactor. Specifically, the effect of this fuel design on the incremental cross sections related to the use of adjuster rods is investigated. The actinide concentrations studied in this work were based on extraction from thirty year cooled spent fuel and mixed with natural uranium to yield a MOX fuel of 4.75% actinide by weight. The incremental cross sections were calculated using the DRAGON neutron transport code. The results for the actinide fuel were compared to those for standard natural uranium fuel and for a slightly enriched (1% U-235) fuel designed to reduce void reactivity. Adjuster reactivity effect calculations and void reactivity simulations were also performed. The impact of the adjuster on reactivity decreased by as much as 56% with TRUMOX fuel while the CVR was reduced by 71% due to the addition of central burnable poison. The incremental cross sections were largely affected by the use of the TRUMOX fuel primarily due to its increased level of fissile material (five times that of NU). The largest effects are in the thermal neutron group where the ΣT value is increased by 46.7%, the Σny) values increased by 13.0% and 9.9%. The value associated with thermal fission, υΣf, increased by 496.6% over regular natural uranium which is expected due to the much higher reactivity of the fuel. (author)

  4. Thorium utilization in ACR (Advanced CANDU) and CANDU-6 reactors

    International Nuclear Information System (INIS)

    It is the main objective of this study to investigate fuel composition options for CANDU type of reactors that are capable of using a mixture of U-Th as fuel. A homogenous mixture of (U-Th)O2 was used in all elements of fuel bundles. The core of CANDU-6 and ACR (Advanced CANDU) were modeled using MCNP5. In equilibrium core, using MONTEBURNS2 code (coupled with MCNP5 and ORIGENS) for once-through uranium and once-through uranium-thorium fuel cycle of CANDU-6 and ACR, discharge burnups and spent fuel compositions were computed. For various enrichments of uranium and different fractions of thorium in a uranium-thorium fuel mixture, performing burnup calculations, relevant relations were derived; in addition, conversion ratio, fuel requirement, uranium resource utilization, and natural uranium savings were determined, and their changes with burnup were observed. Appropriate fuel compositions were discussed.

  5. Luncheon address: Early days of CANDU fuel

    International Nuclear Information System (INIS)

    I will briefly describe how the original dimensions of the fuel bundle were defined and how that early designs of fuel evolved. I will also touch on some of the historical events of the materials and experiments which effected the fuel programme. Also how I became with Canada's Nuclear Fuel programme. (author)

  6. 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

  7. 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

  8. 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)

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

    International Nuclear Information System (INIS)

    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 UO2, thermal stabilities of Zr based metallic glasses, were curried out. 51 figs, 22 tabs, 42 refs. (Author)

  10. 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).

  11. 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

  12. Overview of spent fuel management

    International Nuclear Information System (INIS)

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

  13. Advanced operator interface design for CANDU-3 fuel handling system

    International Nuclear Information System (INIS)

    The Operator Interface for the CANDU 3 Fuel Handling (F/H) System incorporates several improvements over the existing designs. A functionally independent sit-down CRT (cathode-ray tube) based Control Console is provided for the Fuel Handling Operator in the Main Control Room. The Display System makes use of current technology and provides a user friendly operator interface. Regular and emergency control operations can be carried out from this control console. A stand-up control panel is provided as a back-up with limited functionality adequate to put the F/H System in a safe state in case of an unlikely non-availability of the Plant Display System or the F/H Control System'. The system design philosophy, hardware configuration and the advanced display system features are described in this paper The F/H Operator Interface System developed for CANDU 3 can be adapted to CANDU 9 as well as to the existing stations. (author)

  14. 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

  15. 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)

  16. Developments in CANDU MOX fuel fabrication

    International Nuclear Information System (INIS)

    As a strategic component of its advanced fuel cycle program, AECL continues to implement the MOX fuel development program involving MOX fuel fabrication and characterization, irradiation testing, post-irradiation examination, as well as reactor physics and fuel management studies. AECL performs its MOX fuel fabrication activities in the Recycle Fuel Fabrication Laboratories (RFFL) located at the Chalk River site. The RFFL facility is designed to handle alpha-active fuel material and produce experimental quantities of MOX fuel for reactor physics tests and demonstration irradiations. From 1979 to 1988, several fabrication campaigns were conducted in the RFFL, producing close to two tonnes of MOX fuel with various compositions. RFFL operations were suspended from 1989 until 1994, at which time the facility was needed to fabricate MOX fuel for physics testing in the ZED-2 reactor. After completion of an extensive rehabilitation and re-commissioning of the RFFL, MOX operations were resumed in the facility in August 1996. An up-to-date description of the facility, including the fabrication process and the associated equipment, as well as the upgraded safety systems and laboratory services, is presented. Since the resumption of MOX operations in the RFFL in 1996, several MOX fuel fabrication campaigns have been conducted in the facility; increasing the total amount of MOX fuel fabricated to-date in the RFFL to about three tonnes of MOX fuel. An overview of each of the fabrication campaigns is discussed. The fabrication processes used to manufacture the fuel from the starting powders to the finished elements are summarized. The various fabrication campaigns involved different technical requirements mainly due to the different intended uses of the fuel, i.e., test irradiations in NRU, physics tests in ZED-2, and dissolution experiments in support of the waste management program. Fabrication data including production throughputs and typical inspection results are discussed

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

    International Nuclear Information System (INIS)

    Spent fuel management is an important issue for nuclear power programs, requiring careful planning and implementation. With the wait-and-see policy on spent fuel management in Korea, research efforts at KAERI are directed to developing advanced technologies for the safer and more efficient management of the accumulating spent fuels. In support of these research perspectives, a test facility on a pilot scale is being developed with provisions for integral demonstration of the multitude of technical functions required for advanced spent fuel management. The facility, named SMATER, is to be capable of handling full size assemblies of spent PWR fuel (as well as CANDU fuel) with a maximum capacity of 20 MTU/year (about 48 assemblies of PWR type). Major functions of the facility are consolidation of spent PWR fuel assembly into a half volume package and optionally, the transformation of the fuel rod into a fuel of CANDU-type (called DUPIC). Objectives of these functions are to demonstrate volume reduction of spent fuel (for either longer term dry storage or direct disposal) in the former case and direct refabrication of the spent PWR fuel into DUPIC fuel for reuse in CANDU reactors in the latter case. In addition to these major functions, there are other associated technologies to be demonstrated: such as waste treatment, remote maintenance, safeguards, etc. As the facility is to demonstrate not only the functional processes but also the safety and efficiency of the test operations, engineering criteria equivalent to industrial standards are incorporated in the design concept. The hot cell structure enclosing the radioactive material is configured in such a way as to maximize space utilization and ergonometrics, and to minimize costs within the given functional and operational requirements. (author)

  18. Criticality of spent reactor fuel

    International Nuclear Information System (INIS)

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

  19. Improving the service life and performance of CANDU fuel channels

    International Nuclear Information System (INIS)

    The development objective for CANDU fuel channels is to produce a design that can operate for 40 years at 90% capacity. Steady progress toward this objective is being made. The factors that determine the life of the channel are reviewed and the processes necessary to achieve the objectives identified. Performance of future fuel channels will be enhanced by reduced operating costs, increased safety margins to postulated accident conditions, and reduced retubing costs compared to current channels. The approaches to these issues are discussed briefly in the paper. (author)

  20. 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

  1. Thermalhydraulic analysis of spent fuel baskets

    International Nuclear Information System (INIS)

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

  2. Extended storage of spent fuel

    International Nuclear Information System (INIS)

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

  3. 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

  4. 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.

  5. 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

  6. 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)

  7. Spent fuel management in Poland

    International Nuclear Information System (INIS)

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

  8. Considerations in recycling used natural uranium fuel from CANDU reactors in Canada

    International Nuclear Information System (INIS)

    This paper identifies the key factors that would affect the recycling of used natural uranium (NU) fuel from CANDU reactors which are in operation in Canada and in several other countries. There has been little analysis of those considerations over the past 25 years and this paper provides a framework for such analysis. In particular, the large energy potential of the plutonium in used CANDU NU fuel provides a driver for consideration of used-fuel recycling. There would be a long lead-time (at least 30 years) and a large investment required for establishing the infrastructure for used-fuel recycling. While this paper does not promote the recycling of used CANDU NU fuel in Canadian CANDU reactors, it does suggest that it is timely to start the analysis and to consider the key factors or circumstances that warrant the recycling of used CANDU NU fuel. (author)

  9. 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

  10. Spent fuel integrity during transportation

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-01-01

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

  11. Improved CANDU fuel performance. A summary of previous AECL publications

    International Nuclear Information System (INIS)

    The fuel defect rate in CANDU power reactors has been very low (0.06%) since 1972. Most defects were caused by power ramping. The two measures taken to reduce the defect rate, by about an order of magnitude, were changes in the fuelling schemes and the introduction of thin coatings of graphite on the inside surface of the Zircaloy fuel cladding. Power ramping tests have demonstrated that graphite layers, and also baked poly-dimethyl-siloxane layers, between the UO2 pellets and Zircaloy cladding, increase the tolerance of fuel to power ramps. These designs are termed graphite CANLUB and siloxane CANLUB; fuel performance depends on coating parameters such as thickness, wear resistance and on environmental and thermal conditions during the curing of coatings. (author)

  12. Worldwide spent fuel transportation logistics

    International Nuclear Information System (INIS)

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

  13. 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

  14. Economic potential of advanced fuel cycles in CANDU

    International Nuclear Information System (INIS)

    Advanced fuel cycles in CANDU offer the potential of a many-fold increase in energy yield over that which can be obtained from uranium resources using the current once-through natural uranium cycle. This paper examines the associated economics of alternative once-through and recycle fuelling. Results indicate that these cycles will limit the impact of higher uranium prices and offer the potential of a period of stable constant-dollar generating costs that are only approximately 20% higher than current levels

  15. 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)

  16. 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)

  17. 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)

  18. 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.

  19. Spent fuel bundle counter sequence error manual - DARLINGTON NGS

    International Nuclear Information System (INIS)

    The Spent Fuel Bundle Counter (SFBC) is used to count the number and type of spent fuel transfers that occur into or out of controlled areas at CANDU reactor sites. However if the transfers are executed in a non-standard manner or the SFBC is malfunctioning, the transfers are recorded as sequence errors. Each sequence error message typically contains adequate information to determine the cause of the message. This manual provides a guide to interpret the various sequence error messages that can occur and suggests probable cause or causes of the sequence errors. Each likely sequence error is presented on a 'card' in Appendix A. Note that it would be impractical to generate a sequence error card file with entries for all possible combinations of faults. Therefore the card file contains sequences with only one fault at a time. Some exceptions have been included however where experience has indicated that several faults can occur simultaneously

  20. Spent fuel storage process equipment development

    International Nuclear Information System (INIS)

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

  1. 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)

  2. The R and D program in support of advanced fuel cycles for CANDU

    International Nuclear Information System (INIS)

    Advanced fuel cycles for CANDU reactors are well on their way to being implemented. The first step is slightly enriched uranium (SEU) which is economical today. A new fuel bundle is seen as the vehicle for all fuels in CANDU. CANDU fuel fabricated from uranium recovered from fuel discharged from light-water reactors (LWR) is also economical today and readily achievable technically. Future fuel cycles would utilize plutonium recovered from light-water reactors or CANDU's and eventually thorium. R and D in support of these cycles focuses on those topics that require a high degree of confidence in their implementation such as fuel fabrication and defect-free performance to high burnup. Reactor physics codes and nuclear data for advanced fuel cycles will be validated against experiments. (author). 8 refs

  3. CANDU improvement

    International Nuclear Information System (INIS)

    The evolution of the CANDU family of nuclear power plants is based on a continuous product development approach. Proven equipment and system concepts from operating stations are standardized and used in new products. Due to the modular nature of the CANDU reactor concept, product features developed for CANDU 9 can easily be incorporated in other CANDU products such as CANDU 6. Design concepts are being developed for advanced CANDU 6 or larger advanced CANDU, depending on the number of fuel channels and the fuel cycle selected. This paper provides a description of the design improvements being incorporated in CANDU 9 and further design enhancements being studied for future incorporation in CANDU 6 or larger advanced CANDU meeting the requirements of future CANDU owners. The design enhancement objectives are: To improve operational simplicity by applying modern information technology; to improve safety in a cost effective way; to improve system and component reliability and to increase plant life; to improve economics and to reduce owners' risks during all phases of a project using up-front licensing, an improved engineering process and project tools during design, construction and operation; to continue to exploit the neutron economy of CANDU with the development of advanced fuels and fuel cycles. (author)

  4. Spent fuel storage criticality safety

    International Nuclear Information System (INIS)

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

  5. 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

  6. Assessment of CANDU-6 reactivity devices for DUPIC fuel

    International Nuclear Information System (INIS)

    Reactivity device characteristics for a CANDU 6 reactor loaded with DUPIC fuel have been assessed. The lattice parameters were generated by WIMS-AECL code and the core calculations were performed by RFSP code with a 3-dimensional full core model. The reactivity devices studied are the zone controller, adjusters, mechanical control absorber and shutoff rods. For the zone controller system, damping capability for spatial oscillation was investigated. For the adjusters, the restart capability was investigated. For the adjusters, the restart capability was investigated. The shin operation and power stepback calculation were also performed to confirm the compatibility of the current adjuster system. The mechanical control absorber was assessed for the function of compensating temperature reactivity feedback following a power reduction. And shutoff rods were also assessed to investigate the following a power reduction. And shutoff rods were also assessed to investigate the static reactivity worth. This study has shown that the current reactivity device system of CANDU-6 core with the DUPIC fuel. (author). 9 refs., 17 tabs., 7 figs

  7. 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)

  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. 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)

  10. Regulatory review of the CANDU fuel modification program in Canada

    International Nuclear Information System (INIS)

    Aging of a CANDU nuclear power plant affects various safety margins of the plant. Margin to fuel sheath dryout is one of the safety margins that have been detrimentally affected, leading to a reduced margin to dryout with time. If no proactive actions are taken, the plant will have to de-rate its operation at an earlier time. To postpone the de-rating, the Canadian nuclear Industry has taken multi-initiatives to restore, or partially restore the safety margins that have been eroded due to plant aging. One of the initiatives is modification/re-optimization of the current fuel design, in order to improve the fuel thermalhydraulic performance, i.e., to suppress fuel sheath dryout, whereby offset partially the erosion of margin to fuel sheath dryout. Several Canadian utilities have already proceeded with the fuel modification program and requested approval of the Canadian Nuclear Safety Commission (CNSC) to load the modified fuel into their reactors. This paper summarizes the CNSC requirements, the review processes, the current status, and the technical challenges associated with licensing review of the fuel modification in Canada. (author)

  11. 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.)

  12. 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

  13. Near surface spent fuel storage: environmental issues

    International Nuclear Information System (INIS)

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

  14. 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)

  15. 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.

  16. Spent Fuel Working Group Report

    International Nuclear Information System (INIS)

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

  17. Spent nuclear fuel sampling strategy

    International Nuclear Information System (INIS)

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

  18. Spent fuel storage requirements 1987

    International Nuclear Information System (INIS)

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

  19. 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

  20. Sealed can of spent fuel

    International Nuclear Information System (INIS)

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

  1. Spent fuel management in France

    International Nuclear Information System (INIS)

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

  2. Spent fuel management in the Slovak Republic

    International Nuclear Information System (INIS)

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

  3. 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.

  4. Safeguards approach for spent fuel transfers to dry storage

    International Nuclear Information System (INIS)

    , permanent surveillance measures are applied to the spent fuel pond area at power reactors to maintain continuity of knowledge of the verified spent fuel. The surveillance is evaluated quarterly to meet the 3- month timeliness goal for spent fuel (which is categorized as 'irradiated direct use material'). When the spent fuel is transferred to dry storage where it becomes difficult to access, the inspectors verify the spent fuel prior to transfer and maintain continuity of knowledge by observing the loading and transfer activities or applying temporary C/S measures until permanent C/S is applied. In most cases, the spent fuel is then designated as 'difficult-to access' and Dual C/S is applied. The defect test level under the traditional safeguards depends on whether the fuel assemblies can be dismantled in the storage pond. For fuel assembly types that cannot be easily dismantled (e.g., CANDU bundles and welded LWR assemblies), a gross defect test is sufficient. For fuel assembly types that can be easily dismantled (e.g., non-welded LWR assemblies), a partial defect test should be applied to protect against pin diversion, using the best available method approved for inspection use. Between the time of final verification and placement in dry storage under C/S, continuity of knowledge should be maintained by inspector presence or C/S measures. Summary of Integrated Safeguards Approach When spent fuel has been previously verified (the verification requirements would be the same as specified in traditional requirements) and has remained under successful C/S measures, verification of spent fuel to be loaded into dry storage containers is not required before that spent fuel is transferred to dry storage. Between the time of final verification and placement in dry storage under permanent C/S, continuity of knowledge should be maintained by continuous inspector presence, temporary C/S measures, an unannounced inspection programme, or some combination of those measures. Unannounced

  5. Systems for transporting used CANDU fuel by road, rail and water

    International Nuclear Information System (INIS)

    Ontario Hydro's CANDU nuclear power stations are situated on the shores of the Great Lakes and are accessible by road, rail and water. For the off-site shipment of used CANDU fuel from the stations to a disposal, reprocessing or central storage facility, all three modes are being considered. This paper presents Ontario Hydro's 'reference transportation systems' for the shipment of used CANDU fuel as developed for the Nuclear Fuel Waste Management Program (NFWMP) in Canada. These are workable systems developed by Ontario Hydro for the purpose of showing modal feasibility. The systems have not yet been optimized

  6. Safety assessment to support NUE fuel full core implementation in CANDU reactors

    International Nuclear Information System (INIS)

    The Natural Uranium Equivalent (NUE) fuel contains a combination of recycled uranium and depleted uranium, in such a manner that the resulting mixture is similar to the natural uranium currently used in CANDU® reactors. Based on successful preliminary results of 24 bundles of NUE fuel demonstration irradiation in Qinshan CANDU 6 Unit 1, the NUE full core implementation program has been developed in cooperation with the Third Qinshan Nuclear Power Company and Candu Energy Inc, which has recently received Chinese government policy and funding support from their National-Level Energy Innovation program. This paper presents the safety assessment results to technically support NUE fuel full core implementation in CANDU reactors. (author)

  7. Radioactive decay properties of CANDU fuel. Volume 1: the natural uranium fuel cycle

    International Nuclear Information System (INIS)

    The computer code CANIGEN was used to obtain the mass, activity, decay heat and toxicity of CANDU fuel and its component isotopes. Data are also presented on gamma spectra and neutron emissions. Part 1 presents these data for unirradiated fuel, uranium ore and uranium mill tailings. In Part 2 they have been computed for fuel irradiated to levels of burnup ranging from 140 GJ/kg U to 1150 GJ/kg U. (author)

  8. Fuel management optimization in CANDU reactors cooled with light water

    International Nuclear Information System (INIS)

    This research has two main goals. First, we wanted to introduce optimization tools in the diffusion code DONJON, mostly for fuel management. The second objective is more practical. The optimization capabilities are applied to the fuel management problem for different CANDU reactors at refueling equilibrium state. Two kinds of approaches are considered and implemented in this study to solve optimization problems in the code DONJON. The first methods are based on gradients and on the quasi-linear mathematical programming. The method initially developed in the code OPTEX is implemented as a reference approach for the gradient based methods. However, this approach has a major drawback. Indeed, the starting point has to be a feasible point. Then, several approaches have been developed to be more general and not limited by the initial point choice. Among the different methods we developed, two were found to be very efficient: the multi-step method and the mixte method. The second kind of approach are the meta-heuristic methods. We implemented the tabu search method. Initially, it was designed to optimize combinatory variable problems. However, we successfully used it for continuous variables. The major advantage of the tabu method over the gradient methods is the capability to exit from local minima. Optimisation of the average exit burnup has been performed for CANDU-6 and ACR-700 reactors. The fresh fuel enrichment has also been optimized for ACR-700. Results match very well what the reactor physics can predict. Moreover, a comparison of the two totally different types of optimization methods validated the results we obtained. (author)

  9. 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

  10. 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)

  11. 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)

  12. The relationship between natural uranium and advanced fuel cycles in CANDU reactors

    International Nuclear Information System (INIS)

    CANDU is the most uranium-economic type of thermal power reactor, and is the only type used in Canada. CANDU reactors consume approximately 15% of Canadian uranium production and support a fuel service industry valued at ∼$250 M/a. In addition to their once-through, natural-uranium fuel cycle, CANDU reactors are capable of operating with slightly-enriched uranium (SEU), uranium-plutonium and thorium cycles, more efficiently than other reactors. Only SEU is economically attractive in Canada now, but the other cycles are of interest to countries without indigenous fuel resources. A program is underway to establish the fuel technologies necessary for the use of SEU and the other fuel cycles in CANDU reactors

  13. The relationship between natural uranium and advanced fuel cycles in CANDU reactors

    International Nuclear Information System (INIS)

    CANDU is the most uranium-economic type of thermal power reactor, and is the only type used in Canada. CANDU reactors consume approximately 15% of Canadian uranium production and support a fuel service industry valued at ∼$250 M/a. In addition to their once-through, natural-uranium fuel cycle, CANDU reactors are capable of operating with slightly-enriched uranium (SEU), uranium-plutonium and thorium cycles, more efficiently than other reactors. Only SEU is economically attractive in Canada now, but the other cycles are of interest to countries without indigenous fuel resources. A program is underway to establish the fuel technologies necessary for the use of SEU and the other fuel cycles in CANDU reactors. 22 refs

  14. Spent fuel storage in India

    International Nuclear Information System (INIS)

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

  15. Five years of successful CANDU-6 fuel manufacturing in Romania

    International Nuclear Information System (INIS)

    This paper describes the evolution of CANDU-6 nuclear fuel manufacturing in Romania at FCN Pitesti, after the completion of the qualification in 1994. Commercial production was resumed early 1995 and fuel bundles produced were entirely delivered to Cernavoda Plant and charged in the reactor. More than 12,000 fuel bundles have been produced in the last five years and the fuel behaved very well. Defective bundles represents less than 0.06% from the total irradiated fuel, and the most defects are associated to the highest power positions. After qualification, FCN focused the effort to improve braze quality and also to maintain a low residual hydrogen content in graphite coated sheaths. The production capacity was increased especially for component manufacturing, appendages tack welding and brazing. A new graphite baking furnace with increased capacity, is under design. In the pelleting area, a rotating press will replace the older hydraulic presses used for pelleting. Plant development taken inter consideration the future demands for Cernavoda Unit 2. (author)

  16. 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

  17. CANDU advanced fuel cycles: key to energy sustainability

    International Nuclear Information System (INIS)

    In the fast-growing economies of the Pacific Basin region, sustainability is an important requisite for new energy development. Many countries in this region have seen, and continue to see, very large increases in energy and electricity demand. The investment in any nuclear technology is large. Countries making that investment want to ensure that the technology can be sustained and that it can evolve in an ever-changing environment. Three key aspects in ensuring a sustainable energy future are: technological sustainability; economic sustainability; and environmental sustainability (including resource utilization). The fuel-cycle flexibility of the CANDU reactor provides a ready path to sustainable energy development in both the short and the long term. (author). 23 refs

  18. Spent fuel management in Belgium

    International Nuclear Information System (INIS)

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

  19. Spent fuel management in Switzerland

    International Nuclear Information System (INIS)

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

  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. Spent fuel management in the Slovak Republic

    International Nuclear Information System (INIS)

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

  2. A central spent fuel storage in Sweden

    International Nuclear Information System (INIS)

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

  3. Transportation accident scenarios for commercial spent fuel

    International Nuclear Information System (INIS)

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

  4. 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

  5. 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)

  6. 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.

  7. 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

  8. CANDU-6 fuel bundle fabrication and advanced fuels development in China

    International Nuclear Information System (INIS)

    In recent years, China North Nuclear Fuel Corporation (CNNFC) has introduced several modifications to the manufacturing processes and the production line equipment. This has been beneficial in achieving a very high level of quality in the production of fuel bundles. Since 2008 CNNFC has participated in a multi party project with the goal of developing advanced fuels for use in CANDU reactors. Other project team members include the Nuclear Power Institute of China (NPIC), Third Qinshan Nuclear Power Company (TQNPC) and Atomic Energy of Canada Ltd (AECL). This paper will present the improvements developed during the manufacture of natural fuel bundles and advanced fuels. (author)

  9. The R and D program in support of advanced fuel cycles for CANDU

    International Nuclear Information System (INIS)

    Advanced fuel cycles for CANDU reactors are well on their way to being implemented. The first step is slightly enriched uranium (SEU) which is economical today. A new fuel bundle is seen as the vehicle for all fuels in CANDU. CANDU fuel fabricated from uranium recovered from fuel discharged from light-water reactors (LWR) is also economical today and readily achievable technically. Future fuel cycles would utilize plutonium recovered from light-water reactors or CANDUs and eventually thorium. R and D in support of these cycles focuses on those topics that require a high degree of confidence in their implementation such as fuel fabrication and defect-free performance to high burnup. Reactor physics codes and nuclear data for advanced fuel cycles will be validated against experiments. (author).

  10. 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.)

  11. 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

  12. Electrochemical methods for leaching of spent fuel

    International Nuclear Information System (INIS)

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

  13. Method of processing spent fuel

    International Nuclear Information System (INIS)

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

  14. CANDU 300

    International Nuclear Information System (INIS)

    The CANDU nuclear power system is under continuous review by AECL in order to advance the CANDU concept in a manner that will assure competitiveness in both current and future markets. Over the past three years development effort has featured the CANDU 300, a CANDU nuclear generating station with a net output in the range of 320 MW9e) to 380 MW(e). At the outset AECL recognized that coal-fired power plants would be the primary competition for the CANDU advantages such as the use of natural uranium fuel and on-power refuelling, while enhancing capacity factor, reducing man-rem exposure, reducing capital cost, and minimizing construction schedules. AECL believes that the resulting CANDU 300 nuclear generating station will have substantial appeal to many utilities, in both developed and developing countries. The key features of the CANDU 300 are presented here, with particular attention to the station layout, construction methods, and construction schedules

  15. 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)

  16. The next generation of CANDU technologies: profiling the potential for hydrogen fuel

    International Nuclear Information System (INIS)

    This report discusses the Next-generation CANDU Power Reactor technologies currently under development at AECL. The innovations introduced into proven CANDU technologies include a compact reactor core design, which reduces the size by a factor of one third for the same power output; improved thermal efficiency through higher-pressure steam turbines; reduced use of heavy water (one quarter of the heavy water required for existing plants), thus reducing the cost and eliminating many material handling concerns; use of slightly enriched uranium to extend fuel life to three times that of existing natural uranium fuel and additions to CANDU's inherent passive safety. With these advanced features, the capital cost of constructing the plant can be reduced by up to 40 per cent compared to existing designs. The clean, affordable CANDU-generated electricity can be used to produce hydrogen for fuel cells for the transportation sector, thereby reducing emissions from the transportation sector

  17. 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)

  18. 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.

  19. Computer simulation of the behaviour and performance of a CANDU fuel rod

    International Nuclear Information System (INIS)

    At the Argentine Atomic Energy Commission (Comision Nacional de Energia Atomica, CNEA) the BACO code (for 'BArra COmbustible', fuel rod) was developed. It allows the simulation of the thermo-mechanical performance of a cylindrical fuel rod in a Pressurized Heavy Water Reactor (PHWR). The standard present version of the code (2.30), is a powerful tool for a relatively easy and complete evaluation of fuel behaviour predictions. Input parameters and, therefore, output ones may include statistical dispersion. As a demonstration of BACO capabilities we include a review of CANDU fuel applications, and the calculation and a parametric analysis of a characteristic CANDU fuel. (author)

  20. Development of the advanced PHWR technology -Verification tests for CANDU advanced fuel-

    International Nuclear Information System (INIS)

    This is final report of the CANDU advanced fuel (CANFLEX fuel) verification test project. This report describes performance verification tests performed for the development of the CANFLEX-NU bundle. The test items described in the report are as follows. - Fuel channel pressure drop test, -Fuel strength tests, - Fuel impact test, - Fuel endurance test (vibration test), - Compatibility test with fueling machine, - Critical heat flux test. 58 tabs., 60 figs., 32 refs. (Author)

  1. 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

  2. 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.

  3. 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

  4. A feasible approach to implement a commercial scale CANDU fuel manufacturing plant in Egypt

    International Nuclear Information System (INIS)

    Many planning scenarios have been examined to assess and evaluate the economic estimates for implementing a commercial scale CANDU fuel manufacturing plant in Egypt. The cost estimates indicated strong influence of the annual capital costs on total fuel manufacturing cost; this is particularly evident in a small initial plant where the proposed design output is only sufficient to supply reload fuel for a single CANDU-6 reactor. A modular approach is investigated as a possible way, to reduce the capital costs for a small initial fuel plant. In this approach the plant would do fuel assembly operations only and the remainder of a plant would be constructed and equipped in the stages when high production volumes can justify the capital expenses. Such approach seems economically feasible for implementing a small scale CANDU fuel manufacturing plant in developing countries such as Egypt and further improvement could be achieved over the years of operation. (author)

  5. 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

  6. 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)

  7. Evaluation of radiological impact due to direct exposure to a radiological dispersal device using spent fuel assembly

    International Nuclear Information System (INIS)

    Nowadays, one of the most serious terrorist threats implies radiological dispersal devices (RDDs), so-called 'dirty bombs', which combine a conventional explosive surrounded by an inflammatory material (like Thermit) with radioactive material. The paper objective is to assess the radiological impact due to direct exposure to a RDD using a spent nuclear fuel assembly as radioactive material. Simulations were performed for TRIGA LEU fuel assembly, CANDU standard and CANDU SEU fuel bundles. The source term for fuel burnup simulation in specific conditions was obtained by means of ORIGEN-S code, included in ORNL's SCALE6 programs package. The contamination potential level and radiation exposure due to radioactive material spreading from RDD was modelled by means of LLNL's HOTSPOT 2.06 computer code. The worst case scenario has been considered, simulations being performed for General radioactive Plume and General Explosion dispersion models. The specific parameters (total effective dose equivalent, time-integrated air concentration, ground surface deposition, ground shine dose rates and time of radioactive plume arrival in a specific location) evolution as a function of distance from the radiation source has been investigated. Comparisons between fuel projects and radioactive material dispersion models have been also performed. The most drastic effects on population and environment characterize TRIGA LEU spent fuel assembly use in RDD. Whereas in CANDU spent fuel the total inventory of fission products, at the end of irradiation, is much higher than the TRIGA spent fuel one, the volatile fission gases inventory (Kr, Xe, I) is smaller. (authors)

  8. 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

  9. 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

  10. Spent Nuclear Fuel (SNF) Project Execution Plan

    Energy Technology Data Exchange (ETDEWEB)

    LEROY, P.G.

    2000-11-03

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

  11. 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

  12. Fuel Management in Candu Reactors Using Tabu Search

    International Nuclear Information System (INIS)

    Meta-heuristic methods are perfectly suited to solve fuel management optimization problem in LWR. Indeed, they are originally designed for combinatorial or integer parameter problems which can represent the reloading pattern of the assemblies. For the Candu reactors the problem is however completely different. Indeed, this type of reactor is refueled online. Thus, for their design at fuel reloading equilibrium, the parameter to optimize is the average exit burnup of each fuel channel (which is related to the frequency at which each channel has to be reloaded). It is then a continuous variable that we have to deal with. Originally, this problem was solved using gradient methods. However, their major drawback is the potential local optimum into which they can be trapped. This makes the meta-heuristic methods interesting. In this paper, we have successfully implemented the Tabu Search (TS) method in the reactor diffusion code DONJON. The case of an ACR-700 using 7 burnup zones has been tested. The results have been compared to those we obtained previously with gradient methods. Both methods give equivalent results. This validates them both. The TS has however a major drawback concerning the computation time. A problem with the enrichment as an additional parameter has been tested. In this case, the feasible domain is very narrow, and the optimization process has encountered limitations. Actually, the TS method may not be suitable to find the exact solution of the fuel management problem, but it may be used in a hybrid method such as a TS to find the global optimum region coupled with a gradient method to converge faster on the exact solution. (authors)

  13. Spent nuclear fuel integrity during dry storage

    International Nuclear Information System (INIS)

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

  14. Diagnostic technology for degradation of feeder pipes and fuel channels in CANDU reactor

    International Nuclear Information System (INIS)

    Degradation of critical components of nuclear power plants has become important as the operating years of plants increase. The necessity of degradation study including detection and monitoring technology has raised its head. Because the feeder pipes and the fuel channels are particularly one of the critical components in CANDU nuclear plant, they are treated as a major research target in order to counteract the possible problems and establish the counterplan for the improvement of CANDU reactor safety. To ensure the integrity of feeder pipes and fuel channels in CANDU nuclear plant, the following 3 research tasks were performed in the first stage. - Development of a model for prediction of feeder wall thinning - Development of RFEC detection technology - Development of ICFD noise signal analysis. The technologies developed in this study could contribute to the nuclear safety and estimation of the remaining life of operating CANDU nuclear power plants

  15. 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

  16. 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

  17. System study of CANDU/LWR synergy in advanced nuclear fuel cycles

    International Nuclear Information System (INIS)

    This report proposes a study that will evaluate the effects of advanced nuclear fuel cycles on resource utilisation, repository capacity, waste streams, economics, and proliferation resistance. The proposed fuel cycles are designed to exploit the unique synergy that exists between light water and CANDU reactors. Also, several fuel cycle simulation codes have been proposed to be used. (author)

  18. 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

  19. Burn up Analysis for Fuel Assembly Unit i n a Pressurized Heavy Water CANDU Reactor

    International Nuclear Information System (INIS)

    MCNPX code has been used for modeling a nd simulation of an assembly of CANDU Fuel bundle . The assembly is composed of a heterogeneous lattice of 37-element natural Uranium fuel, heavy water moderator and coolant. The fuel bundle is burnt in normal operation conditions of CANDU reactors. The effective multiplication factor (Keff ) of the bundle is calculated as a function of fuel burnup. The flux and power distribution are determined. Comparing t he concentrations of both Uranium and Plutonium isotopes are analyzed in the bundle. The results of the present model with the results of a benchmark problem, a good agreement was found PWR

  20. Development of the advanced PHWR technology -Verification tests for CANDU advanced fuel-

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Jang Hwan; Suk, Hoh Chun; Jung, Moon Kee; Oh, Duk Joo; Park, Joo Hwan; Shim, Kee Sub; Jang, Suk Kyoo; Jung, Heung Joon; Park, Jin Suk; Jung, Seung Hoh; Jun, Ji Soo; Lee, Yung Wook; Jung, Chang Joon; Byun, Taek Sang; Park, Kwang Suk; Kim, Bok Deuk; Min, Kyung Hoh [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This is the `94 annual report of the CANDU advanced fuel verification test project. This report describes the out-of pile hydraulic tests at CANDU-hot test loop for verification of CANFLEX fuel bundle. It is also describes the reactor thermal-hydraulic analysis for thermal margin and flow stability. The contents in this report are as follows; (1) Out-of pile hydraulic tests for verification of CANFLEX fuel bundle. (a) Pressure drop tests at reactor operation condition (b) Strength test during reload at static condition (c) Impact test during reload at impact load condition (d) Endurance test for verification of fuel integrity during life time (2) Reactor thermal-hydraulic analysis with CANFLEX fuel bundle. (a) Critical channel power sensitivity analysis (b) CANDU-6 channel flow analysis (c) Flow instability analysis. 61 figs, 29 tabs, 21 refs. (Author).

  1. The Canadian research reactor spent fuel situation

    International Nuclear Information System (INIS)

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

  2. Proceedings of the third spent fuel workshop

    International Nuclear Information System (INIS)

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

  3. 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

  4. German Approach to Spent Fuel Management

    International Nuclear Information System (INIS)

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

  5. Study of the end flux peaking for the Candu fuel bundle types by transport methods

    International Nuclear Information System (INIS)

    The region separating the Candu fuel in two adjoining bundles in a channel is called the end region. The end of the last pellet in the fuel stack adjacent to the end region is called the fuel end. In the end region of the bundle the thermal neutron flux is higher than at the axial mid-point, because the end region of the bundle is made up of very low neutron absorption material: coolant and Zircaloy-4. For accurate evaluation of fuel performance, it is important to have capability to calculate the three dimensional spatial flux distributions in the fuel bundle, including the end region. The work reported here had two objectives. First, calculation of the flux distributions (axial and radial) and the end flux peaking factors for some Candu fuel bundles. Second objective is a comparative analysis of the obtained results. The Candu fuel bundles considered in this paper are NU37 (Natural Uranium, 37 elements) and SEU43 (Slightly Enriched Uranium, 43 elements, with 1.1wt% enrichment). For realization of the proposed objectives, a methodology based on WIMS, PIJXYZ and LEGENTR codes is used in this paper. WIMS is a standard lattice-cell code, based on transport theory and it is used for producing fuel cell multigroup macroscopic cross sections. For obtaining the flux distribution in Candu fuel bundles it is used PIJXYZ and LEGENTR respectively codes. These codes are consistent with WIMS lattice-cell calculations and allow a good geometrical representation of the Candu bundle in three dimensions. PIJXYZ is a 3D integral transport code using the first collision probability method and it has been developed for Candu cell geometry. LEGENTR is a 3D SN transport code based on projectors technique and can be used for 3D cell and 3D core calculations. (author)

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

    International Nuclear Information System (INIS)

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

  7. 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

  8. Status report on spent fuel management

    International Nuclear Information System (INIS)

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

  9. Safeguards for spent fuels: Verification problems

    International Nuclear Information System (INIS)

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

  10. Modeling the highway transportation of spent fuel

    International Nuclear Information System (INIS)

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

  11. The application of the goal programming to CANDU fuel management optimization

    International Nuclear Information System (INIS)

    A Goal Programming formulation to CANDU fuel management optimization is proposed. Four objectives are considered, respectively : the feed rate, CPPF (Channel Power Peaking Factor), BPPF (Bundle Power Peaking Factor) and CPR (Critical Power Ratio). This problem is investigated using a numerical approach to optimization established on STepMethod and the use of loss matrix. The optimization technique developed is more adequate for fuel management analysis for fissile enriched fuel cycles, in which cases the relative importance of the objectives could be modified. Numerical results are presented for 0.93% SEU fuelled CANDU 6Mk1 core and weapons-grade plutonium burning in CANDU 6Mk1 core, using standard 37 rod fuel bundle. (author). 12 refs., 3 tabs., 7 figs

  12. The evolution of Candu fuel cycles and their potential contribution to world peace

    International Nuclear Information System (INIS)

    The Candu(r) reactor is the most versatile commercial power reactor in the world. It has the potential to extend resource utilization significantly, to allow countries with developing industrial infrastructures access to clean and abundant energy, and to destroy long-lived nuclear waste or surplus weapons plutonium. These benefits are available by choosing from an array of possible fuel cycles. Several factors, including Canada's early focus on heavy-water technology, limited heavy-industry infrastructure at the time, and a desire for both technological autonomy and energy self-sufficiency, contributed to the creation of the first Candu reactor in 1962. With the maturation of the CANDU industry, the unique design features of the now-familiar product - on-power refuelling, high neutron economy, and simple fuel design - make possible the realization of its potential fuel-cycle versatility. Several fuel-cycle options currently under development are described. (authors)

  13. 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

  14. 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

  15. 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)

  16. Spent Nuclear Fuel Project dose management plan

    International Nuclear Information System (INIS)

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

  17. 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

  18. 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

  19. 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

  20. Enhanced CANDU 6 Reactor

    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 much easier to operate with minimal operator intervention. Improvements to the fire protection system and enhanced security features will further protect the assets. And given the CANDU 6's 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. (authors)

  1. Programme on spent fuel management in India

    International Nuclear Information System (INIS)

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

  2. 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

  3. Development of a remote monitoring system for spent fuel storage silo

    International Nuclear Information System (INIS)

    A remote monitoring system has been developed to verify CANDU spent fuels in dry storage casks and to provide a COK(Continuous of Knowledge) signature of the spent fuels in the silo. The system is composed of an optical fiber scintillator as a radiation sensor, a preamplifier, a high speed counter module, a wireless transmitter and a main computer. Due to site condition and constraints such as limited space and existing hardware, monitoring system was developed to adopt radio frequency to ease any conflict of installation of system. Also, because of the limited memory size of main computer, we developed the high-speed count module (comparator) for calculating the radiation volume rapidly. The purpose of comparator compares the actual value to threshold value before A/D (Analog to Digital). The effectiveness of the comparator reduces the storage data and the unnecessary data was deleted before A/D by way of hardware. Its performance on high radiation, temperature, reproducibility, reparability of gamma and neutron, powers was investigated and verified the data using Mini-MCA. It shows that it has high potential to be used for spent fuel verification in the view point of safeguard monitoring for verifying spent fuel at CANDU reactor

  4. INR Recent Contributions to Thorium-Based Fuel Using in CANDU Reactors

    International Nuclear Information System (INIS)

    The paper summarizes INR Pitesti contributions and latest developments to the Thorium-based fuel (TF) using in present CANDU nuclear reactors. Earlier studies performed in INR Pitesti revealed the CANDU design potential to use Recovered Uranium (RU) and Slightly Enriched Uranium (SEU) as alternative fuels in PHWRs. In this paper, we performed both lattice and CANDU core calculations using TF, revealing the main neutron physics parameters of interest: k-infinity, coolant void reactivity (CVR), channel and bundle power distributions over a CANDU 6 reactor core similar to that of Cernavoda, Unit 1. We modelled the so called Once Through Thorium (OTT) fuel cycle, using the 3D finite-differences DIREN code, developed in INR. The INR flexible SEU-43 bundle design was the candidate for TF carrying. Preliminary analysis regarding TF burning in CANDU reactors has been performed using the finite differences 3D code DIREN. TFs showed safety features improvement regarding lower CVRs in the case of fresh fuel use. Improvements added to the INR ELESIMTORIU- 1 computer code give the possibility to fairly simulate irradiation experiments in INR TRIGA research reactor. Efforts are still needed in order to get better accuracy and agreement of simulations to the experimental results. (author)

  5. 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

  6. MACSTOR trademark: Dry spent fuel storage for the nuclear power industry

    International Nuclear Information System (INIS)

    Safe storage of spent fuel has long been an area of critical concern for the nuclear power industry. As fuel pools fill up and re-racking possibilities become exhausted, power plant operators will find that they must ship spent fuel assemblies off-site or develop new on-site storage options. Many utility companies are turning to dry storage for their spent fuel assemblies. The MACSTOR (Modular Air-cooled Canister STORage) concept was developed with this in mind. Derived from AECL's successful vertical loading, concrete silo program for storing CANDU nuclear spent fuel, MACSTOR was developed for light water reactor spent fuel and was subjected to full scale thermal testing. The MACSTOR Module is a monolithic, shielded concrete vault structure than can accommodate up to 24 spent fuel canisters. Each canister holds 12 PWR or 32 PWR previously cooled spent fuel assemblies with burn-up rates as high as 45,000 MWD/MTU. The structure is passively cooled by natural convection through an array of inlet and outlet gratings and galleries serving a central plenum where the (vertically) stored canisters are located. The canisters are continuously monitored by means of a pressure monitoring system developed by TNI. The MACSTOR system includes the storage module(s), an overhead gantry system for cask handling, a transfer cask for moving fuel from wet to dry storage and a cask transporter. The canister and transfer cask designs are based on Transnuclear transport cask designs and proven hot cell transfer cask technology, adapted to requirements for on-site spent fuel storage. This Modular Air Cooled System has a number of inherent advantages: efficient use of construction materials and site space; cooling is virtually impossible to impede; has the ability to monitor fuel confinement boundary integrity during storage; the fuel canisters may be used for both storage and transport and canisters utilize a flanged, ASME-III closure system that allows for easy inspection

  7. Assessment of neutron transport codes for application to CANDU fuel lattices analysis

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Gyu Hong; Choi, Hang Bok

    1999-08-01

    In order to assess the applicability of WIMS-AECL and HELIOS code to the CANDU fuel lattice analysis, the physics calculations has been carried out for the standard CANDU fuel and DUPIC fuel lattices, and the results were compared with those of Monte Carlo code MCNP-4B. In this study, in order to consider the full isotopic composition and the temperature effect, new MCNP libraries have been generated from ENDF/B-VI release 3 and validated for typical benchmark problems. The TRX-1,2,BAPL-1,2,3 pin -cell lattices and KENO criticality safety benchmark calculations have been performed for the new MCNP libraries, and the results have shown that the new MCNP library has sufficient accuracy to be used for physics calculation. Then, the lattice codes have been benchmarked by the MCNP code for the major physics parameters such as the burnup reactivity, void reactivity, relative pin power and Doppler coefficient, etc. for the standard CANDU fuel and DUPIC fuel lattices. For the standard CANDU fuel lattice, it was found that the results of WIMS-AECL calculations are consistent with those of MCNP. For the DUPIC fuel lattice, however, the results of WIMS-AECL calculations with ENDF/B-V library have shown that the discrepancy from the results of MCNP calculations increases when the fuel burnup is relatively high. The burnup reactivities of WIMS-ACEL calculations with ENDF/B-VI library have shown excellent agreements with those of MCNP calculation for both the standard CANDU and DUPIC fuel lattices. However, the Doppler coefficient have relatively large discrepancies compared with MCNP calculations, and the difference increases as the fuel burns. On the other hand, the results of HELIOS calculation are consistent with those of MCNP even though the discrepancy is slightly larger compared with the case of the standard CANDU fuel lattice. this study has shown that the WIMS-AECL products reliable results for the natural uranium fuel. However, it is recommended that the WIMS

  8. 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)

  9. 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)

  10. 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.

  11. 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)

  12. Choosing a spent fuel interim storage system

    International Nuclear Information System (INIS)

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

  13. TRIGA Mark II Ljubljana - spent fuel transportation

    International Nuclear Information System (INIS)

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

  14. Burnup credit demands for spent fuel management in Ukraine

    International Nuclear Information System (INIS)

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

  15. 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)

  16. 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.

  17. 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

  18. 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)

  19. 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)

  20. An intelligent spent fuel database for BWR fuels

    International Nuclear Information System (INIS)

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

  1. 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...

  2. 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...

  3. Status on spent fuel management in Spain

    International Nuclear Information System (INIS)

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

  4. Probable leaching mechanisms for spent fuel

    International Nuclear Information System (INIS)

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

  5. Optimized CANDU-6 cell and reactivity device supercell models for advanced fuels reactor database generation

    International Nuclear Information System (INIS)

    Highlights: • Propose an optimize 2-D model for CANDU lattice cell. • Propose a new 3-D simulation model for CANDU reactivity devices. • Implement other acceleration techniques for reactivity device simulations. • Reactivity device incremental cross sections for advanced CANDU fuels with thorium. - Abstract: Several 2D cell and 3D supercell models for reactivity device simulation have been proposed along the years for CANDU-6 reactors to generate 2-group cross section databases for finite core calculations in diffusion. Although these models are appropriate for natural uranium fuel, they are either too approximate or too expensive in terms of computer time to be used for optimization studies of advanced fuel cycles. Here we present a method to optimize the 2D spatial mesh to be used for a collision probability solution of the transport equation for CANDU cells. We also propose a technique to improve the modeling and accelerate the evaluation, in deterministic transport theory, of the incremental cross sections and diffusion coefficients associated with reactivity devices required for reactor calculations

  6. ORIGEN-S cross section libraries for CANDU used-fuel characterization

    International Nuclear Information System (INIS)

    A code system for producing burn-up dependent cross-section libraries for CANDU used-fuel characterization for use with the ORIGEN-S isotope generation and depletion code system is described. Benchmark results against experimental isotopic data for three CANDU-PHW reactor stations are presented. The code system couples the WIMS-AECL reactor physics analysis code with an ORIGEN-S depletion analysis to produce application-specific libraries that can be used in subsequent used-fuel analyses. 11 refs., 1 fig., 3 tabs

  7. Next generation CANDU plants

    International Nuclear Information System (INIS)

    The CANDU Pressurized Heavy Water Reactors systems featuring horizontal fuel channels and heavy water moderator will continue to evolve, supported by AECL's strong commitment to comprehensive R and D programs. There are three key CANDU development strategic thrusts: improved economics, fuel cycle flexibility, and enhanced safety operation based on design feedback. Therefore, CANDU reactor products will continue to evolve by incorporating further improvements and advanced features that will be arising from our CANDU Technology R and D programs in areas such as heavy water and tritium, control and instrumentation, fuel and fuel cycles, systems and equipment and safety and constructability. Progressive CANDU development will continue in AECL to enhance the medium size product - CANDU 6, and to evolve the larger size product - CANDU 9. The development of features for CANDU 6 and CANDU 9 is carried out in parallel. Developments completed for one reactor size can then be applied to the other design with minimum costs and risk. (author)

  8. Application of Sipping and Visual Inspection Systems for the Evaluation of Spent Fuel Bundle Integrity

    International Nuclear Information System (INIS)

    When CANDU reactor has defective fuel bundle during its operation, then the defective fuel bundle should be discharged by 2(two) fuel bundles at a time from the corresponding fuel channel until the failed fuel bundle is found. Existing fuel failure detection system GFP(Gaseous Fission Product) & DN(Delayed Neutron) Monitoring System can’t exactly distinguish fuel elements failure from each fuel bundle. Because of fuelling machine mechanism and discharge procedure, always two fuel bundles at a time are being inspected. In case visual inspection is available for inspecting fuel elements and suppose that there are no defects and damaged marks on the surface of outer fuel elements, 2(two) defective fuel bundles should be canned and kept in the separate region of spent fuel storage pool. Therefore, the purpose of this study was to develop a system which is capable of inspecting whether each fuel bundle is failed or not. KNF (KEPCO Nuclear Fuel Co. Ltd) developed two evaluation systems to investigate the integrity of CANDU spent fuel bundle. The first one is a sipping system that detects fission gases leaked from fuel element. The second one is a visual inspection system with radiation resistant underwater camera and remotely controlled devices. The sipping technology enables to analyze the leakage of fission products not only in gaseous state but also liquid state. The performance of developed systems was successfully demonstrated at Wolsong power plant this year. This paper describes the results of the development of the failed fuel detection technology and its application. (author)

  9. 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)

  10. 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).

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

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, G. [Inst. for Nuclear Research (INR), Pitesti (Romania); Palleck, S. [Atomic Energy of Canada Limited., Mississauga, Ontario (Canada); Ionescu, D. [Inst. for Nuclear Research (INR), Pitesti (Romania)

    2010-07-01

    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)

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

    International Nuclear Information System (INIS)

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

  13. Spent fuel storage and transportation - ANSTO experience

    International Nuclear Information System (INIS)

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

  14. Fact sheet on spent fuel management

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

    2013-01-17

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

  16. CANDU fuel cycle economic efficiency assessments using the IAEA-MESSAGE-V code

    International Nuclear Information System (INIS)

    The main goal of the paper is to evaluate different electricity generation costs in a CANDU Nuclear Power Plant (NPP) using different nuclear fuel cycles. The IAEA-MESSAGE code (Model for Energy Supply Strategy Alternatives and their General Environmental Impacts) will be used to accomplish these assessments. This complex tool was supplied by International Atomic Energy Agency (IAEA) in 2002 at 'IAEA-Regional Training Course on Development and Evaluation of Alternative Energy Strategies in Support of Sustainable Development' held in Institute for Nuclear Research Pitesti. It is worthy to remind that the sustainable development requires satisfying the energy demand of present generations without compromising the possibility of future generations to meet their own needs. Based on the latest public information in the next 10-15 years four CANDU-6 based NPP could be in operation in Romania. Two of them will have some enhancements not clearly specified, yet. Therefore we consider being necessary to investigate possibility to enhance the economic efficiency of existing in-service CANDU-6 power reactors. The MESSAGE program can satisfy these requirements if appropriate input models will be built. As it is mentioned in the dedicated issues, a major inherent feature of CANDU is its fuel cycle flexibility. Keeping this in mind, some proposed CANDU fuel cycles will be analyzed in the paper: Natural Uranium (NU), Slightly Enriched Uranium (SEU), Recovered Uranium (RU) with and without reprocessing. Finally, based on optimization of the MESSAGE objective function an economic hierarchy of CANDU fuel cycles will be proposed. The authors used mainly public information on different costs required by analysis. (authors)

  17. 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

  18. 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.

  19. 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.

  20. Spent Nuclear Fuel Project Technical Databook

    International Nuclear Information System (INIS)

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

  1. Dissolution studies of spent nuclear fuels

    International Nuclear Information System (INIS)

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

  2. 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)

  3. 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

  4. 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

  5. Thermo-mechanical analysis of SEU 43 fuel element in CANDU reactor normal operational conditions

    International Nuclear Information System (INIS)

    The main direction of developing the CANDU type fuel is designing a new type of fuel cluster with the number of elements increased from 37 to 43. This work presents results of the research done in INR Pitesti on the new concept of SEU 43 fuel cluster designed for burnups as high as 25 Mw·day/kgU using slightly enriched uranium (up to 1.1% U235). By using ROFEM 1.0 code the behaviour of two types of SEU 43 fuel was analyzed in normal conditions of CANDU 6 reactor operation. The main performance parameters of SEU fuel were analyzed. These are: temperature distribution; the volume and pressure of fission gases; stresses in the fuel can; can deformations. Comparisons with the standard CANDU fuel are done. The results show the adequacy of the design solutions implemented for the SEU 43 fuel. The power-burnup history required by the ROFEM computations was obtained from the overpower envelope of the fuel cluster, with the radial power distribution on cluster taken into account. Evolution of the main performance parameters during irradiation is given

  6. Status and prospects for spent fuel management in France

    International Nuclear Information System (INIS)

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

  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. 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)

  9. The parallex project: CANDU MOX fuel testing with weapons-derived plutonium

    International Nuclear Information System (INIS)

    The Parallex Project consists of a parallel experiment in which weapons-derived plutonium (WPu) from the United States and from the Russian Federation will be tested as mixed-oxide (MOX) CANDU fuel in the National Research Universal (NRU) reactor at the Chalk River Laboratories in Canada. Plutonium derived from excess weapons will be fabricated into CANDU MOX fuel at the A.A. Bochvar Institute in Moscow and at the Los Alamos National Laboratory in the United States. The MOX fuel will be transported to CRL, where it will be characterized, assembled into fuel bundles and then irradiated in the NRU reactor. Following irradiation, the fuel will be examined in hot cells to assess its irradiation performance. This paper describes the scope, rationale and current status of the Parallex Project. (author)

  10. 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

  11. The dependence of the global neutronic parameters on the fuel burnup for CANDU SEU43 core

    Energy Technology Data Exchange (ETDEWEB)

    Balaceanu, V. [Institute for Nuclear Research, Pitesti (Romania); Pavelescu, M. [Academy of Romanian Scientists, Bucharest (Romania)

    2010-05-15

    In order to reduce the total fuel costs for the CANDU reactors, mainly by extending the fuel burnup limits, some fuel bundle concepts have been developed in different CANDU owner countries. Therefore, in our Institute the SEU43 (Slightly Enriched Uranium with 43 fuel elements) project was started in early '90s. The neutronic behavior analysis of the CANDU core with SEU43 fuel was an important step in our project design. The objective of this paper is to highline an analysis of the neutronic behavior of the CANDU SEU43 core with the fuel burnup. More exactly, the study refers to the dependence of some global neutronic parameters, mainly the reactivity, on the fuel burnup. Two types of CANDU core were taken into consideration: reference core (without any reactivity devices) and perturbed core (with a strong reactivity system inserted). The considered reactivity system is the Mechanical Control Absorber (MCA) one. The performed parameters are: k{sub eff.} values, the MCA reactivity worth and flux distributions. The fuel bundles in the core are SEU43, with the fuel enrichment in U{sup 235} of 0.96% and at nominal power. For the fuel burnup the values are: 0.00 GWd/tU (fresh fuel); 8.00 GWd/tU and 25.00 GWd/tU. For reaching this objective, a global neutronic calculation system named WIMSPIJXYZ LEGENTR is used. Starting from a 69-groups ENDF/B-V based library, this system uses three transport codes: (1) the standard lattice-cell code WIMS, for generating macroscopic cross sections in supercell option and also for burnup calculations; (2) the PIJXYZ code for 3D simulation of the MCA reactivity devices and the 3D correction of the macroscopic cross sections; (3) the LEGENTR 3D transport code for estimating global neutronic parameters (CANDU core). The analysis of the neutronic parameters consists of comparing the obtained results with the similar results calculated with the DRAGON and DIREN codes. This comparison shows a good agreement between these results. (orig.)

  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. R and D activities on CANDU-type fuel in Indonesia

    Energy Technology Data Exchange (ETDEWEB)

    Suripto, A.; Badruzzaman, M.; Latief, A. [Nuclear Fuel Element Centre, National Atomic Energy Agency of Indonesia (BATAN), Puspiptek, Serpong (Indonesia)

    1997-07-01

    The status of R and D activities in Indonesia with respect of CANDU-type fuel development is presented. The activities have been started since the first feasibility study to introduce nuclear power plants was carried out in 1970s. The early research comprised the in-situ pilot production of yellow-cake in Kalimantan (Borneo) experimental mining site, uranium purification and pellet preparation. This program continued to gain a full support from the Government which culminated in the realisation of the construction by BATAN of a large fuel development laboratory in Serpong, starting from 1984 in co-operation with NIRA Ansaldo of Italy. The laboratory, which is called the Power Reactor Experimental Fuel Element Installation (EFEI) was originally designed as an experimental facility to integrate the acquired domestic R and D results gained so far on the CANDU-type fuel technology and the additional know-how received from NIRA Ansaldo which at that time was engaged, in developing a CANDU-type fuel, called the CIRENE fuel design. In the present days the facility houses the power reactor fuel development activities carried out to build up the national capability on power reactor fuel fabrication technology in anticipation to embark upon the nuclear energy era in the near future. (author)

  14. 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

  15. Spent fuel management for research reactors

    International Nuclear Information System (INIS)

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

  16. Laser surveillance system for spent fuel

    International Nuclear Information System (INIS)

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

  17. Spent Nuclear Fuel Alternative Technology Decision Analysis

    International Nuclear Information System (INIS)

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

  18. 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.

  19. Taipower's spent fuel interim storage program

    International Nuclear Information System (INIS)

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

  20. 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

  1. 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

  2. Spent fuel storage system for LMFBR fuel experiments

    International Nuclear Information System (INIS)

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

  3. 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

  4. 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

  5. International collaboration to study the feasibility of implementing the use of slightly enriched uranium fuel in the Embalse CANDU reactor

    International Nuclear Information System (INIS)

    In the last few years, Nucleoelectrica Argentina S.A. and Atomic Energy of Canada Limited have collaborated on a study of the technical feasibility of implementing Slightly Enriched Uranium (SEU) fuel in the Embalse CANDU reactor in Argentina. The successful conversion to SEU fuel of the other Argentine heavy-water reactor, Atucha 1, served as a good example. SEU presents an attractive incentive from the point of view of fuel utilization: if fuel enriched to 0.9% 235U were used in Embalse instead of natural uranium, the average fuel discharge burnup would increase significantly (by a factor of about 2), with consequent reduction in fuel requirements, leading to lower fuel-cycle costs and a large reduction in spent-fuel volume per unit energy produced. Another advantage is the change in the axial power shape: with SEU fuel, the maximum bundle power in a channel decreases and shifts towards the coolant inlet end, consequently increasing the thermalhydraulics safety margin. Two SEU fuel carriers, the traditional 37-element bundle and the 43-element CANFLEX bundle, which has enhanced thermalhydraulic characteristics as well as lower peak linear element ratings, have been examined. The feasibility study gave the organizations an excellent opportunity to perform cooperatively a large number of analyses, e.g., in reactor physics, thermalhydraulics, fuel performance, and safety. A Draft Plan for a Demonstration Irradiation of SEU fuel in Embalse was prepared. Safety analyses have been performed for a number of hypothetical accidents, such as Large Loss of Coolant, Loss of Reactivity Control, and an off-normal condition corresponding to introducing 8 SEU bundles in a channel (instead of 2 or 4 bundles). There are concrete safety improvements which result from the reduced maximum bundle powers and their shift towards the inlet end of the fuel channel. Further improvements in safety margins would accrue with CANFLEX. In conclusion, the analyses identified no issues that would

  6. 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)

  7. 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

  8. Status of the parallex project--testing CANDU MOX fuel with weapons-derived plutonium

    International Nuclear Information System (INIS)

    The Parallex project is a parallel experiment, the purpose of which is to demonstrate the use of weapons-derived plutonium (WPu) from the United States (U.S.) and the Russian Federation (R.F.) in CANDU mixed-oxide (MOX) fuel elements. The scope of the project includes the fabrication of CANDU MOX fuel in the R.F. and the U.S., the transporting of the fuel to Canada, and the testing of the fuel in the NRU research reactor at the Chalk River Laboratories (CRL). A significant milestone in the project was achieved earlier this year, with the start of the irradiation testing of the MOX elements received from both the U.S. and the R.F. This paper presents the status of the project, and highlights the major activities that lead to the commencement of this historic experiment. (author)

  9. Analytical assessment for stress corrosion fatigue of CANDU fuel elements under load following conditions

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, Grigore; Ionescu, Dragos; Pauna, Eduard [Institute for Nuclear Research (INR), Pitesti (Romania)

    2011-07-01

    Two load following (LF) tests on CANDU type fuel elements were performed in TRIGA Research Reactor of INR Pitesti. In the first LF test the 78R fuel element has successfully experienced 367 power cycles, mostly between 23 and 56 kW/m average linear power. In the second LF test, the ME01 fuel element 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. Both LF tests were simulated with finite elements computer codes in order to evaluate Stress Corrosion Fatigue (SCF) of the sheath arising from expansion and contraction of the pellets. This paper presents the analytical assessment for SCF conditions and their relation to CANDU fuel performance in LF conditions. (orig.)

  10. Analytical assessment for stress corrosion fatigue of CANDU fuel elements under load following conditions

    International Nuclear Information System (INIS)

    Two load following (LF) tests on CANDU type fuel elements were performed in TRIGA Research Reactor of INR Pitesti. In the first LF test the 78R fuel element has successfully experienced 367 power cycles, mostly between 23 and 56 kW/m average linear power. In the second LF test, the ME01 fuel element 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. Both LF tests were simulated with finite elements computer codes in order to evaluate Stress Corrosion Fatigue (SCF) of the sheath arising from expansion and contraction of the pellets. This paper presents the analytical assessment for SCF conditions and their relation to CANDU fuel performance in LF conditions. (orig.)

  11. Hanford spent fuel inventory baseline

    International Nuclear Information System (INIS)

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

  12. 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.

  13. 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

  14. Dry spent fuel transfer system design

    International Nuclear Information System (INIS)

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

  15. Integrated model of Korean spent fuel and high level waste disposal options - 16091

    International Nuclear Information System (INIS)

    This paper describes an integrated model developed by the Korean Atomic Energy Research Institute (KAERI) to simulate options for disposal of spent nuclear fuel (SNF) and reprocessing products in South Korea. A companion paper (Hwang and Miller, 2009) describes a systems-level model of Korean options for spent nuclear fuel (SNF) management in the 21. century. The model addresses alternative design concepts for disposal of SNF of different types (Candu, PWR), high level waste, and fission products arising from a variety of alternative fuel cycle back ends. It uses the GoldSim software to simulate the engineered system, near-field and far-field geosphere, and biosphere, resulting in long-term dose predictions for a variety of receptor groups. The model's results allow direct comparison of alternative repository design concepts, and identification of key parameter uncertainties and contributors to receptor doses. (authors)

  16. 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)

  17. 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)

  18. Thermal-hydraulic modeling of reracked spent fuel pool

    International Nuclear Information System (INIS)

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

  19. 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

  20. Spent Fuel Storage Operation - Lessons Learned

    International Nuclear Information System (INIS)

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

  1. Experience in the manufacture and performance of CANDU fuel for KANUPP

    International Nuclear Information System (INIS)

    Karachi Nuclear Power Plant (KANUPP) a 137 MWe CANDU unit is In operation since 1971. Initially, it was fueled with Canadian fuel bundles. In July 1980 Pakistani manufactured fuel was introduced in the reactor core, irradiated to a burnup of about 7500 MWd-teU-1 and successfully discharged in May 1984. The core was progressively fuelled with Pakistani fuel and in August 1990 the reactor core contained all Pakistani made fuel. As of the present, 3 core equivalent Pakistani fuel bundles have been successfully discharged at an average bumup of 6500 MWd-teU-1. with a maximum burnup of ∼ 10,200 MWd-teU-1. No fuel failure of Pakistani bundles has been observed so far. This paper presents the indigenous efforts towards manufacture and operational aspects of KANUPP fuel and compares its behaviour with that of Canadian supplied fuel. The Pakistani fuel has performed well and is as good as the Canadian fuel. (author)

  2. 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

  3. 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.

  4. 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.

  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. Spent fuel treatment at ANL-West

    International Nuclear Information System (INIS)

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

  7. BR-100 spent fuel shipping cask development

    International Nuclear Information System (INIS)

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

  8. 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)

  9. Fuel bundle loss of cooling inside the fuelling machine at CANDU6 PHWR

    International Nuclear Information System (INIS)

    This article describes the that loss of forced circulation cooling flow of induce spent fuel bundle loss of cooling and fission product releasing, analyzes the effect of reactor building and environment due to the fuel bundle rupturing, discusses the countermeasure to deal with the event of loss of cooling of spent fuel bundle. (authors)

  10. Development of Romanian SEU-43 fuel bundle for CANDU type reactors

    International Nuclear Information System (INIS)

    SEU-43 fuel bundle is a CANDU type fuel consisting of two element sizes, to reduce element ratings, while maintaining the same bundle power, and an uranium content very close to the uranium content of a standard 37-element bundle. In order to reduce the detrimental effects of the life limiting factors at extended burnup a set of solution have been adopted for fuel element design. As a part of the design verification program, experimental bundles have been fabricated and utilized in typical out of reactor tests conducted at the laboratories of INR, Pitesti. These tests simulated current CANDU-6 reactor normal operating conditions of flow, temperature and pressure. The results are in accordance with the specified acceptance criteria. (author)

  11. Validation of the ORIGEN-S code for predicting radionuclide inventories in used CANDU fuel

    International Nuclear Information System (INIS)

    The safety assessment being conducted by AECL Research for the concept of deep geological disposal of used CANDU UO2 fuel requires the calculation of radionuclide inventories in the fuel to provide source terms for radionuclide release. This report discusses the validation of selected actinide and fission-product inventories calculated using the ORIGEN-S code coupled with the WIMS-AECL lattice code, using data from analytical measurements of radioisotope inventories in Pickering CANDU reactor fuel. The recent processing of new ENDF/B-VI cross-section data has allowed the ORIGEN-S calculations to be performed using the most up-to-date nuclear data available. The results indicate that the code is reliably predicting actinide and the majority of fission-product inventories to within the analytical uncertainty. ((orig.))

  12. Spent nuclear fuel project integrated schedule plan

    International Nuclear Information System (INIS)

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

  13. 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)

  14. 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

  15. Spent nuclear fuel project product specification

    International Nuclear Information System (INIS)

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

  16. 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

  17. Total quality in spent fuel pool reracking

    International Nuclear Information System (INIS)

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

  18. 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

  19. 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)

  20. Spent Nuclear Fuel (SNF) Project Product Specification

    Energy Technology Data Exchange (ETDEWEB)

    PAJUNEN, A.L.

    2000-01-20

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

  1. 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)

  2. The cascad spent fuel dry storage facility

    International Nuclear Information System (INIS)

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

  3. Reactor BN-350 spent fuel handling

    International Nuclear Information System (INIS)

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

  4. Spent fuel management strategy in Italy

    International Nuclear Information System (INIS)

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

  5. Fuel cost analysis of CANDU-PHWR Wolsung Nuclear Power Plant unit 1

    International Nuclear Information System (INIS)

    Being based on the Segal method, calculation was carried out for the natural uranium nuclear fuel cost with Zircaloy-4 cladding having design parameters of Wolsung Nuclear Power Plant, CANDU-PHWR (Unit 1), currently under construction in Korea aiming at its completion in 1982. An attempt was also made for the sensitivity analysis of each fuel component; i.e., depreciation of fuel manufacturing plant caused by its life time, its load factor, production scale expansion of plant facilities, variations of construction and operating costs of fuel manufacturing plant, fluctuation of interest rates, extent of uranium ore price increases and effect of learning factor. (author)

  6. The psychosocial consequences of spent fuel disposal

    International Nuclear Information System (INIS)

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

  7. Numerical Estimation of the Spent Fuel Ratio

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-01-01

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

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

    International Nuclear Information System (INIS)

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

  9. 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)

  10. Reprocessing of spent nuclear fuel

    International Nuclear Information System (INIS)

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

  11. 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

  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. LWR spent fuel management in the Federal Republic of Germany

    International Nuclear Information System (INIS)

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

  14. 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)

  15. Spent fuel assembly source term parameters

    International Nuclear Information System (INIS)

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

  16. Development of advanced spent fuel management process

    International Nuclear Information System (INIS)

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

  17. Characteristics of Loviisa NPP spent fuel

    International Nuclear Information System (INIS)

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

  18. 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

  19. Post irradiation tests on CANDU fuel irradiated in power ramp conditions

    International Nuclear Information System (INIS)

    Nuclear Research Branch Pitesti disposes of facilities, which allow the testing, manipulation and examination of nuclear fuel and of irradiated structure materials in CANDU reactor from Cernavoda NPP. These facilities imply the materials testing reactor TRIGA and the Post-Irradiation Examination Laboratory (PIEL). The purpose of this work is to determine by post-irradiate examination, the behavior of CANDU indigenous fuel, irradiated in 14 MW(th) TRIGA reactor into a multiple/various power ramp tests. The results of post-irradiate examination consist of: - 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 and tomography; - measurement of pressure, volume and isotopic composition of fission gas; - Microstructural characterization by metallographic and ceramographic analyzes; - Isotopic composition and burn-up determination; - Mechanical properties determination. The obtained data from the post-irradiate examinations are used, on one hand, in order to confirm the security, reliability and nuclear fuel performance, and on the other hand, for further optimization of the CANDU fuel. (authors)

  20. Spent fuel bundle counter sequence error manual - KANUPP (125 MW) NGS

    International Nuclear Information System (INIS)

    The Spent Fuel Bundle Counter (SFBC) is used to count the number and type of spent fuel transfers that occur into or out of controlled areas at CANDU reactor sites. However if the transfers are executed in a non-standard manner or the SFBC is malfunctioning, the transfers are recorded as sequence errors. Each sequence error message may contain adequate information to determine the cause of the message. This manual provides a guide to interpret the various sequence error messages that can occur and suggests probable cause or causes of the sequence errors. Each likely sequence error is presented on a 'card' in Appendix A. Note that it would be impractical to generate a sequence error card file with entries for all possible combinations of faults. Therefore the card file contains sequences with only one fault at a time. Some exceptions have been included however where experience has indicated that several faults can occur simultaneously

  1. Spent fuel bundle counter sequence error manual - RAPPS (200 MW) NGS

    International Nuclear Information System (INIS)

    The Spent Fuel Bundle Counter (SFBC) is used to count the number and type of spent fuel transfers that occur into or out of controlled areas at CANDU reactor sites. However if the transfers are executed in a non-standard manner or the SFBC is malfunctioning, the transfers are recorded as sequence errors. Each sequence error message typically contains adequate information to determine the cause of the message. This manual provides a guide to interpret the various sequence error messages that can occur and suggests probable cause or causes of the sequence errors. Each likely sequence error is presented on a 'card' in Appendix A. Note that it would be impractical to generate a sequence error card file with entries for all possible combinations of faults. Therefore the card file contains sequences with only one fault at a time. Some exceptions have been included however where experience has indicated that several faults can occur simultaneously

  2. 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.)

  3. Spent Fuel Performance and Research in Germany

    International Nuclear Information System (INIS)

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

  4. 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)

  5. Factors affecting the differences in reactivity and dissolution rates between UO2 and spent nuclear fuel

    International Nuclear Information System (INIS)

    Strategies for the permanent disposal of spent nuclear fuel are being investigated by the U.S. Department of Energy at the Yucca Mountain site and by Atomic Energy of Canada Limited (AECL) in plutonic rock formations in the Canadian Shield. Uranium dioxide is the primary constituent of spent nuclear fuel and dissolution of the matrix is regarded as a necessary step for the release of radionuclides to repository groundwaters. In order to develop models to describe the dissolution of the U02 fuel matrix and subsequent release of radionuclides, it is necessary to understand both chemical and oxidative dissolution processes and how they can be affected by parameters such as groundwater composition, pH, temperature, surface area, radiolysis and redox potential. This report summarizes both published and on-going dissolution studies of U02 and both LWR and CANDU spent fuels being conducted at the Pacific Northwest Laboratory, Lawrence Livermore National Laboratory and Lawrence Berkeley Laboratory in the U.S. and at AECL's Whiteshell Laboratories in Canada. The studies include both dissolution tests and electrochemical experiments to measure uranium dissolution rates. The report focuses on identifying differences in reactivity towards aqueous dissolution between U02 and spent fuel samples as well as estimating bounding values for uranium dissolution rates. This review also outlines the basic tenets for the development of a dissolution model that is based on electrochemical principles. (author). 49 refs., 2 tabs., 11 figs

  6. 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...

  7. Neutron measurements for CANDU-type fuel characterization at TRIGA-ACPR

    International Nuclear Information System (INIS)

    In order to measure the parameters for a CANDU cell it is important to determine thermal flux distribution in the cell, spectral indices (absolute values and distributions). If it has to be done in a critical assembly, the small flux value poses problems to the measurements. Also the measurements performed with detectors placed into the bundle have to be treated with care. In order to test the methods related to such measurements we decided to perform the most important of them using a CANDU bundle placed on the experiment loading tube of TRIGA-ACPR. The reactor was operated in stationary regime to give the necessary thermal flux at the experiment position. A set of activation and fissionable foil detectors was used and measurements for absolute reaction rate determinations, thermal flux distributions and spectral indices absolute values were performed. Also the sensitivity to heterogeneous poisoning of the bundle was measured in the same configuration. The sensitivity of the ACPR to heterogeneous poisoning of a CANDU - bundle placed in central hole is also determined. In conclusion a set of methods for neutronic measurements on CANDU type fuel were tested in TRIGA-ACPR, in spectral conditions which can be considered worse than in a D2O lattice. The source of error was investigated in detail. One can conclude that these methods will work in measurements upon a D2O - natural uranium lattice

  8. 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)

  9. 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)

  10. Spent fuel management system in Slovak Republic

    International Nuclear Information System (INIS)

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

  11. Measurement of gap and grain-boundary inventories of 129I in used CANDU fuels

    International Nuclear Information System (INIS)

    Combined gap and grain-boundary inventories of 129I in 14 used CANDU fuel elements were measured by crushing and simultaneously leaching fuel segments for 4 h in a solution containing KI carrier. From analogy with previous work a near one-to-one correlation was anticipated between the amount of stable Xe and the amount of 128I in the combined gap and grain-boundary regions of the fuel. However, the results showed that such a correlation was only apparent for low linear power rating (LLPR) fuels with an average linear power rating of 44 kW/m), the 129I values were considerably smaller than expected. The combined gap and grain-boundary inventories of 129I in the 14 fuels tested varied from 1.8 to 11.0%, with an average value of 3.6 ± 2.4% which suggests that the average value of 8.1 ± 1% used in safety assessment calculations overestimates the instant release fraction for 129I. Segments of used CANDU fuels were leached for 92 d (samples taken at 5, 28 and 92 d) to determine the kinetics of 129I release. Results could be fitted tentatively to half-order reaction kinetics, implying that 129I release is a diffusion-controlled process for LLPR fuels, and also for HLPR fuels, once the gap inventory has been leached. However, more data are needed over longer leaching periods to gain more understanding of the processes that control grain-boundary release of 129I from used CANDU fuel

  12. Analytical and experimental assessment of CANDU fuel sheath integrity under post dryout conditions

    International Nuclear Information System (INIS)

    The experiments that investigated the CANDU fuel sheath behavior under different pressures, temperatures, oxidizing environment, material structure (as-received or thermally treated to attach appendages), and heating rates were reviewed and assessed to determine the limits of post-dryout duration, sheath temperature, and pressure difference across the sheath required to ensure the fuel sheath integrity. A number of burst curves at different heating rates were studied. Time-at- temperature fuel sheath failure maps were developed based on temperature ramp and isothermal experiments for the 28-element fuel bundle. Analytical time-at-temperature fuel sheath failure maps were also developed for both of 28- and 37-element fuel bundles using the ELOCA fuel analysis computer code and were compared to the experimental time-at-temperature sheath failure maps. Time-at-temperature sheath failure maps could be used as a simple and effective screening tool to demonstrate fuel sheath integrity during postulated design basis accident. (author)

  13. Fuel management simulations for 0.9% SEU in CANDU 6 reactors

    International Nuclear Information System (INIS)

    Slightly Enriched Uranium (SEU) of 0.9 weight % 235U enrichment is a promising fuel cycle option for CANDU reactors. An important component of the investigation of this option is the demonstration of the feasibility of on-line refuelling with this fuel type in reactor physics fuel-management simulations. Two fuel-management schemes have been investigated in detail during 500-day core-follow simulations, these were a 2-bundle-shift and a 4-bundle-shift axial refuelling scheme. The 43-element CANFLEX fuel design has been used in these studies because of its improved fuel performance characteristics in this application. The results of the studies are discussed in detail in this paper. The most significant conclusion of this study was that both 2- and 4-bundle-shift refuelling schemes with CANFLEX fuel result in bundle power and bundle power boost envelopes that meet current fuel-performance requirements. (author)

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

    International Nuclear Information System (INIS)

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

  15. 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)

  16. 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)

  17. 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)

  18. A finite element model for static strength analysis of CANDU fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, G.; Ionescu, D.V. [Institute for Nuclear Research, Pitesti (Romania)

    2006-08-15

    A static strength analysis finite-element model has been developed using the ANSYS computer code in order to simulate the axial compression in CANDU type fuel bundle subject to hydraulic drag loads, deflection of fuel elements and stresses and displacements in the end plates. The validation of the finite-element model has been done by comparison with the out-reactor strength test results. Comparison of model predictions with the experimental results showed very good agreement. The comparative assessment reveals that SEU43 and SEU43L fuel bundles are able to withstand high flow rate without showing a significant geometric instability. (orig.)

  19. 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.

  20. 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

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

    International Nuclear Information System (INIS)

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

  2. 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

  3. Investigation of the Ru-43LV fuel behaviour under LOCA conditions in a CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, G.; Serbanel, M. [Institute for Nuclear Research, Pitesti (Romania); Diaconu, C.

    2012-11-15

    Presently, INR Pitesti is developing an advanced fuel design RU-43LV (recovered uranium fuel bundle with 43 elements and low void reactivity feature) based on recovered uranium from LWR. Compared with the current design of 37 natural uranium element (NU-37) fuel bundle, RU-43LV will have higher power capability and higher burn-up potential in CANDU reactors of Cernavoda-Romania Nuclear Power Plant (NPP). Fuel burn-up of RU-43LV fuel will be about two times the burn-up usually achieved in CANDU reactors fuelled with natural uranium fuel. The effect of the design changes of RU-43LV bundle on the reactor safety has been analyzed and the results are presented in this paper. As part of the conceptual design study, the performance of the RU-43LV fuelled core during a large loss-of-coolant accident (LLOCA) was assessed with the use of several computer codes. The most relevant calculations performed regarding RU-43 RV fuel safety are presented in this paper. Also, the stages of an experimental program aiming to study RU-43LV fuel behaviour in high temperature transients are briefly described. (orig.)

  4. 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

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

    International Nuclear Information System (INIS)

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

  6. The licensing process of Cernavoda NPP interim spent fuel dry storage

    International Nuclear Information System (INIS)

    The first 12,000 spent fuel bundles capacity of a new modular interim spent fuel dry storage facility is in operation at the Cernavoda NPP site, since May 2003. The facility consists of a preparation station, a shielded transfer flask and a concrete monolith module of the MACSTOR type, a system designed by Atomic Energy of Canada Limited. The capacity can be extended to accommodate 300 000 spent fuel bundles, which represent the entire production of spent fuel during the lifetime of two CANDU-6 units. The project was implemented by Nuclearelectrica based on the licenses and permits granted by the National Commission for Nuclear Activities Control (CNCAN) for each step: the siting, construction, commissioning and operation. According to the specific Romanian regulations, the interim dry storage facility is also subject to the licensing process by the Environmental and Public Health authorities. The public involvement has been an important step of the environmental licensing procedure. Cernavoda NPP used different legal procedures for the public debate, including announcements in local and national newspapers and public hearings. No objections against the storage facility have been raised. (author)

  7. 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)

  8. Research reactor spent fuel management in Argentina

    International Nuclear Information System (INIS)

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

  9. 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

  10. 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

  11. 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

  12. 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)

  13. 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.

  14. 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

  15. 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)

  16. 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

  17. Development of spent fuel examination technology

    International Nuclear Information System (INIS)

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

  18. 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)

  19. Treatment strategies for spent nuclear fuel

    International Nuclear Information System (INIS)

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

  20. Retrievability of spent nuclear fuel canisters

    International Nuclear Information System (INIS)

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