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Sample records for korean candu spent

  1. Alternative Concept to Enhance the Disposal Efficiency for CANDU Spent Fuel Disposal System

    International Nuclear Information System (INIS)

    Lee, Jong Youl; Cho, Dong Geun; Kook, Dong Hak; Lee, Min Soo; Choi, Heui Joo

    2011-01-01

    There are two types of nuclear reactors in Korea and they are PWR type and CANDU type. The safe management of the spent fuels from these reactors is very important factor to maintain the sustainable energy supply with nuclear power plant. In Korea, a reference disposal system for the spent fuels has been developed through a study on the direct disposal of the PWR and CANDU spent fuel. Recently, the research on the demonstration and the efficiency analyses of the disposal system has been performed to make the disposal system safer and more economic. PWR spent fuels which include a lot of reusable material can be considered being recycled and a study on the disposal of HLW from this recycling process is being performed. CANDU spent fuels are considered being disposed of directly in deep geological formation, since they have little reusable material. In this study, based on the Korean Reference spent fuel disposal System (KRS) which was to dispose of both PWR type and CANDU type, the more effective CANDU spent fuel disposal systems were developed. To do this, the disposal canister for CANDU spent fuels was modified to hold the storage basket for 60 bundles which is used in nuclear power plant. With these modified disposal canister concepts, the disposal concepts to meet the thermal requirement that the temperature of the buffer materials should not be over 100 .deg. C were developed. These disposal concepts were reviewed and analyzed in terms of disposal effective factors which were thermal effectiveness, U-density, disposal area, excavation volume, material volume etc. and the most effective concept was proposed. The results of this study will be used in the development of various wastes disposal system together with the HLW wastes from the PWR spent fuel recycling process.

  2. Korean experience in CANDU-PHWR operation

    International Nuclear Information System (INIS)

    Sang-kee Park

    1987-01-01

    Among KEPCO's 9 nuclear power units, Korea Nuclear Unit No. 3, the Wolsung Nuclear Power Plant is the only CANDU-PHWR Unit, while the rest of 8 others are PWR units. The unit was designed by Atomic Energy of Canada, Ltd(AECL) of Canada, who also perfomed overall project management for the plant construction under the provisions and arrangement of the relevant contracts. The gross electrical output of the plant is 678.7 MWe and thermal output of the reactor is 2061 MWth. While these figures lead to lower plant eficiency than LWR counterparts, unit energy cost for fuel is more favorable than LWRs because natural uranium is utilized for the fuel bundles, some of which are already being fabricated domestically. Annual capacity factors for 1983 and 1984 could have been improved, if two major planned outages for the modification works on steam generator internals and one major forced outage form the heavy water spill incident could be eliminated. The heavy water spill incident in November, 1984 brought plant staffs many lessons to learn and many things to contemplate. Unique design concepts and features such as on-power refuelling, poison prevent mode, versatile plant control system built around digital computers and power step back/set back logics may be credited for these relatively good plant performances. Human related factors such as staff's technical capabilities and strong will toward good performance were other elements which could not be overlooked

  3. Korean experience in CANDU-PHWR operation

    International Nuclear Information System (INIS)

    Park, S.K.

    1988-01-01

    Among KEPCO's 9 nuclear power units, Korea Nuclear Unit No. 3, the Wolsung Nuclear Power Plant is the only CANDU-PHWR Unit, while the rest of 8 others are PWR units. The unit was designed by Atomic Energy of Canada, Ltd. of Canada, who also performed overall project management for the plant construction under the provisions and arrangement of the relevant contracts. The gross electrical output of the plant is 678.7 MWe and thermal output of the reactor is 2061 MWth. While these figures lead to lower plant efficiency than LWR counterparts, unit energy cost for fuel is more favorable than LWRs because natural uranium is utilized for the fuel bundles, some of which are already being fabricated domestically. Annual capacity factors for 1983 and 1984 could have been improved, if two major planned outages for the modification works on steam generator internals and one major forced outage from the heavy water spill incident could be eliminated. The heavy water spill incident in November, 1984 brought plant staffs many lessons to learn and many things to contemplate. Unique design concepts and features such as on-power refuelling, poison prevent mode, versatile plant control system built around digital computers and power step back/set back logics may be credited for these relatively good plant performances. Human related factors such as staff's technical capabilities and strong will toward good performance were other elements which could not be overlooked

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

    International Nuclear Information System (INIS)

    Villagran, J.E.

    2004-01-01

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

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

    International Nuclear Information System (INIS)

    Hwang, Yongsoo; Miller, Ian

    2009-01-01

    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)

  6. A Preliminary Study on the Reuse of the Recovered Uranium from the Spent CANDU Fuel Using Pyroprocessing

    International Nuclear Information System (INIS)

    Park, C. J.; Na, S. H.; Yang, J. H.; Kang, K. H.; Lee, J. W.

    2009-01-01

    During the pyroprocessing, most of the uranium is gathered in metallic form around a solid cathode during an electro-refining process, which is composed of about 94 weight percent of the spent fuel. In the previous study, a feasibility study has been done to reuse the recovered uranium for the CANDU reactor fuel following the traditional DUPIC (direct use of spent pressurized water reactor fuel into CANDU reactor) fuel fabrication process. However, the weight percent of U-235 in the recovered uranium is about 1 wt% and it is sufficiently re-utilized in a heavy water reactor which uses a natural uranium fuel. The reuse of recovered uranium will bring not only a huge economic profit and saving of uranium resources but also an alleviation of the burden on the management and the disposal of the spent fuel. The research on recycling of recovered uranium was carried out 10 years ago and most of the recovered uranium was assumed to be imported from abroad at that time. The preliminary results showed there is the sufficient possibility to recycle recovered uranium in terms of a reactor's characteristics as well as the fuel performance. However, the spent CANDU fuel is another issue in the storage and disposal problem. At present, most countries are considering that the spent CANDU fuel is disposed directly due to the low enrichment (∼0.5 wt%) of the discharge fissile content and lots of fission products. If mixing the spent CANDU fuel and the spent PWR fuel, the estimated uranium fissile enrichment will be about 0.6 wt% ∼ 1.0 wt% depending on the mixing ratio, which is sufficiently reusable in a CANDU reactor. Therefore, this paper deals with a feasibility study on the recovered uranium of the mixed spent fuel from the pyroprocessing. With the various mixing ratios between the PWR spent fuel and the CANDU spent fuel, a reactor characteristics including the safety parameters of the CANDU reactor was evaluated

  7. Monte Carlo assessment of the dose rates produced by spent fuel from CANDU reactors

    International Nuclear Information System (INIS)

    Pantazi, Doina; Mateescu, Silvia; Stanciu, Marcela

    2003-01-01

    One of the technical measures considered for biological protection is radiation shielding. The implementation process of a spent fuel intermediate storage system at Cernavoda NPP includes an evolution in computation methods related to shielding evaluation: from using simpler computer codes, like MicroShield and QAD, to systems of codes, like SCALE (which contains few independent modules) and the multipurpose and multi-particles transport code MCNP, based on Monte Carlo method. The Monte Carlo assessment of the dose rates produced by CANDU type spent fuel, during its handling for the intermediate storage, is the main objective of this paper. The work had two main features: -establishing of geometrical models according to description mode used in code MCNP, capable to account for the specific characteristics of CANDU nuclear fuel; - confirming the correctness of proposed models, by comparing MCNP results and the related results obtained with other computer codes for shielding evaluation and dose rates calculations. (authors)

  8. Some notes on the Timing of Geological Disposal of CANDU Spent Fuels

    International Nuclear Information System (INIS)

    Choi, Heui Joo; Kook, Dong Hak; Choi, Jong Won

    2010-01-01

    CANDU spent fuel is to be disposed of at repository finally rather than recycled because of its low fissile nuclide concentration. But the difficult situation of finding a repository site can not help introducing a interim storage in the short term. It is required to find an optimum timing of geological disposal of CANDU spent fuels related to the interim storage operation period. The major factors for determining the disposal starting time are considered as safety, economics, and public acceptance. Safety factor is compared in terms of the decay heat and non-proliferation. Economics factor is compared from the point of the operation cost, and public acceptance factor is reviewed from the point of retrievability and inter-generation ethics. This paper recommended the best solution for the disposal starting time by analyzing the above factors. It is concluded that the optimum timing for the CANDU spent fuel disposal is around 2041 and that the sooner disposal time, the better from the point of technical and safety aspects.

  9. An assessment of temperature history on concrete silo dry storage system for CANDU spent fuel

    International Nuclear Information System (INIS)

    Lee, Dong-Gyu; Sung, Nak-Hoon; Park, Jea-Ho; Chung, Sung-Hwan

    2016-01-01

    Highlights: • We performed thermal analysis to predict the temperature distribution in the concrete silo. • Thermal analysis of the concrete silo was based on CFD code. • Temperature distribution and history for storage period was presented. • Thermal analysis results and test results agreed well. • The correlations can predict the maximum fuel temperature over storage period. - Abstract: Concrete silo is a dry storage system for spent fuel generated from CANDU reactors. The silo is designed to remove passively the decay heat from spent fuel, as well as to secure the integrity of spent fuel during storage period. Dominant heat transfer mechanisms must be characterized and validated for the thermal analysis model of the silo, and the temperature history along storage period could be determined by using the validated thermal analysis model. Heat transfer characteristics on the interior and exterior of fuel basket in the silo were assessed to determine the temperature history of silo, which is necessary for evaluating the long-term degradation behavior of CANDU spent fuel stored in the silo. Also a methodology to evaluate the temperature history during dry storage period was proposed in this study. A CFD model of fuel basket including fuel bundles was suggested and temperature difference correlation between fuel bundles and silo’s internal member, as a function of decay heat of fuel basket considering natural convection and radiation heat transfer, was deduced. Temperature difference between silo’s internal cavity and ambient air was determined by using a concept of thermal resistance, which was validated by CFD analysis. Fuel temperature was expressed as a function of ambient temperature and decay heat of fuel basket in the correlation, and fuel temperature along storage period was determined. Therefore, it could be used to assess the degradation behavior of spent fuel by applying the degradation mechanism expressed as a function of spent fuel

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

    International Nuclear Information System (INIS)

    Choi, Hae Yun; Kwon, Jong Soo; Park, Seong Hoon; Kim, Seong Rea; Lee, Gi Won

    1996-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hae Yun; Kwon, Jong Soo; Park, Seong Hoon; Kim, Seong Rea; Lee, Gi Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-01-01

    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.

  12. 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).

  13. Structural design concept and static analysis of CANDU spent fuel compact dry storage system

    International Nuclear Information System (INIS)

    Choi, K. S.; Yang, K. H.; Paek, C. R.; Jung, J. S.; Lee, H. Y.

    2003-01-01

    In this study, an structural design concept on CANDU spent fuel compact dry storage system MACSTOR/KN-400 module has been established with a view to optimally design the structural members of the system. Design loads, loading combination and structural safety criteria of the module were reviewed assuming W olsung Site. The static analysis of the module showed that compressive stress concentration due to dead load and live load occurred around the center of roof slab. Maximum stress resulted from dead load is about twice as much as the stress from live load, and structural behavior of module caused by wind load was not significant. The static analysis results will have influence on the reinforcement bar design of structural members with other structural analyses

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

    International Nuclear Information System (INIS)

    Choi, Hangbok; Ko, Won Il; Yang, Myung Seung

    2001-01-01

    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

  15. An improved system to verify CANDU spent fuel elements in dry storage silos

    International Nuclear Information System (INIS)

    Almeida, Gevaldo L. de; Soares, Milton G.; Filho, Anizio M.; Martorelli, Daniel S.; Fonseca, Manoel

    2000-01-01

    An improved system to verify CANDU spent fuel elements stored in dry storage silos was developed. It is constituted by a mechanical device which moves a semi-conductor detector along a vertical verification pipe incorporated to the silo, and a modified portable multi-channel analyzer. The mechanical device contains a winding drum accommodating a cable hanging the detector, in such a way that the drum rotates as the detector goes down due to its own weight. The detector is coupled to the multi-channel analyzer operating in the multi-scaler mode, generating therefore a spectrum of total counts against time. To assure a linear transformation of time into detector position, the mechanical device dictating the detector speed is controlled by the multi-channel analyzer. This control is performed via a clock type escapement device activated by a solenoid. Whenever the multi-channel analyzer shifts to the next channel, the associated pulse is amplified, powering the solenoid causing the drum to rotate a fixed angle. Spectra taken in laboratory, using radioactive sources, have shown a good reproducibility. This qualify the system to be used as an equipment to get a fingerprint of the overall distribution of the fuel elements along the silo axis, and hence, to verify possible diversion of the nuclear material by comparing spectra taken at consecutive safeguards inspections. All the system is battery operated, being thus capable to operate in the field where no power supply is available. (author)

  16. An improved system to verify CANDU spent fuel elements in dry storage silos

    Energy Technology Data Exchange (ETDEWEB)

    Almeida, Gevaldo L. de; Soares, Milton G.; Filho, Anizio M.; Martorelli, Daniel S.; Fonseca, Manoel [Instituto de Engenharia Nuclear (IEN), Rio de Janeiro, RJ (Brazil)

    2000-07-01

    An improved system to verify CANDU spent fuel elements stored in dry storage silos was developed. It is constituted by a mechanical device which moves a semi-conductor detector along a vertical verification pipe incorporated to the silo, and a modified portable multi-channel analyzer. The mechanical device contains a winding drum accommodating a cable hanging the detector, in such a way that the drum rotates as the detector goes down due to its own weight. The detector is coupled to the multi-channel analyzer operating in the multi-scaler mode, generating therefore a spectrum of total counts against time. To assure a linear transformation of time into detector position, the mechanical device dictating the detector speed is controlled by the multi-channel analyzer. This control is performed via a clock type escapement device activated by a solenoid. Whenever the multi-channel analyzer shifts to the next channel, the associated pulse is amplified, powering the solenoid causing the drum to rotate a fixed angle. Spectra taken in laboratory, using radioactive sources, have shown a good reproducibility. This qualify the system to be used as an equipment to get a fingerprint of the overall distribution of the fuel elements along the silo axis, and hence, to verify possible diversion of the nuclear material by comparing spectra taken at consecutive safeguards inspections. All the system is battery operated, being thus capable to operate in the field where no power supply is available. (author)

  17. Application of a cold spray technique to the fabrication of a copper canister for the geological disposal of CANDU spent fuels

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Heui-Joo, E-mail: hjchoi@kaeri.re.k [Korea Atomic Energy Research Institute, Radioactive Waste Management Technology Development, 150 Dukjin-dong, Yuseong, Daejon, 305-353 (Korea, Republic of); Lee, Minsoo; Lee, Jong Youl [Korea Atomic Energy Research Institute, Radioactive Waste Management Technology Development, 150 Dukjin-dong, Yuseong, Daejon, 305-353 (Korea, Republic of)

    2010-10-15

    A new method was proposed for the manufacture of a copper-cast iron canister for the spent fuel disposal based on the cold spray coating technique. The thickness of a copper shell could be fabricated to be as thin as 10 mm with the new method. Around 6 tons of copper could be saved with a 10 mm thick canister compared with a 50 mm thick canister. The electrochemical properties of the cold sprayed copper layer and forged copper were measured through a polarization test. The two copper layers showed very similar electrochemical properties. The lifetime of a 10 mm copper canister was estimated with a mathematical model based on the mass transport of sulfide ions through the buffer. The results showed that the canister lifetime was more than 140,000 years under the Korean granite groundwater condition. The thermal analysis with a current pre-conceptual design of a CANDU spent fuel canister showed that the maximum temperature between the canister and the saturated buffer was below the thermal criteria, 100 {sup o}C. Finally, the mechanical stability of the copper canister was confirmed with a computer program, ABAQUS, under the rock movement scenario.

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

    International Nuclear Information System (INIS)

    Ryu, Yong Ho

    1992-02-01

    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

  19. Efficiency analyses of the CANDU spent fuel repository using modified disposal canisters for a deep geological disposal system design

    International Nuclear Information System (INIS)

    Lee, J.Y.; Cho, D.K.; Lee, M.S.; Kook, D.H.; Choi, H.J.; Choi, J.W.; Wang, L.M.

    2012-01-01

    Highlights: ► A reference disposal concept for spent nuclear fuels in Korea has been reviewed. ► To enhance the disposal efficiency, alternative disposal concepts were developed. ► Thermal analyses for alternative disposal concepts were performed. ► From the result of the analyses, the disposal efficiency of the concepts was reviewed. ► The most effective concept was suggested. - Abstract: Deep geological disposal concept is considered to be the most preferable for isolating high-level radioactive waste (HLW), including nuclear spent fuels, from the biosphere in a safe manner. The purpose of deep geological disposal of HLW is to isolate radioactive waste and to inhibit its release of for a long time, so that its toxicity does not affect the human beings and the biosphere. One of the most important requirements of HLW repository design for a deep geological disposal system is to keep the buffer temperature below 100 °C in order to maintain the integrity of the engineered barrier system. In this study, a reference disposal concept for spent nuclear fuels in Korea has been reviewed, and based on this concept, efficient alternative concepts that consider modified CANDU spent fuels disposal canister, were developed. To meet the thermal requirement of the disposal system, the spacing of the disposal tunnels and that of the disposal pits for each alternative concept, were drawn following heat transfer analyses. From the result of the thermal analyses, the disposal efficiency of the alternative concepts was reviewed and the most effective concept suggested. The results of these analyses can be used for a deep geological repository design and detailed analyses, based on exact site characteristics data, will reduce the uncertainty of the results.

  20. Radiation shielding at interim storage facility for CANDU-type nuclear spent fuel

    International Nuclear Information System (INIS)

    Mateescu, S.; Radu, M. Pantazi D.; Stanciu, M.

    1997-01-01

    Technical measures in radiological protection are taken in the interim storage facility design to ensure that, during normal operation, exposures of workers and members of public to ionizing radiation are limited to levels lower than regulatory limits. The spent fuel storage design provides for radiation exposure to be as low as reasonable achievable (ALARA principles). The evaluation of radiation shields includes the most conservative provisions: - all locations which may contain spent fuel are full; - the spent fuel has reached the maximum burnup; - the post irradiation cooling period should be the minimum reasonable; - equipment for handling contains the maximum amount of spent fuel. Radiation shields should ensure that external radiation fields do not exceed limits accepted by the Regulatory Body Module. The evaluation has been performed with two computer codes, QAD-5K and MICROSHIELD-4. (authors)

  1. ENVI Model Development for Korean Nuclear Spent Fuel Options Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Sunyoung; Jeong, Yon Hong; Han, Jae-Jun; Lee, Aeri; Hwang, Yong-Soo [Korea Institute of Nuclear Nonproliferation and Control, Daejeon (Korea, Republic of)

    2015-10-15

    The disposal facility of the spent nuclear fuel will be operated from 2051. This paper presents the ENVI code developed by GoldSim Software to simulate options for managing spent nuclear fuel (SNF) in South Korea. The ENVI is a simulator to allow decision-makers to assist to evaluate the performance for spent nuclear fuel management. The multiple options for managing the spent nuclear fuel including the storage and transportation are investigated into interim storage, permanent disposal in geological repositories and overseas and domestic reprocessing. The ENVI code uses the GoldSim software to simulate the logistics of the associated activities. The result by the ENVI model not only produces the total cost to compare among the multiple options but also predict the sizes and timings of different facilities required. In order to decide the policy for spent nuclear management this purpose of this paper is to draw the optimum management plan to solve the nuclear spent fuel issue in the economical aspects. This paper is focused on the development of the ENVI's logic and calculations to simulate four options(No Reprocessing, Overseas Reprocessing, Domestic Reprocessing, and Overseas and Domestic Reprocessing) for managing the spent nuclear fuel in South Korea. The time history of the spent nuclear fuel produced from both the existing and future NPP's can be predicted, based on the Goldsim software made available very user friendly model. The simulation result will be used to suggest the strategic plans for the spent nuclear fuel management.

  2. ENVI Model Development for Korean Nuclear Spent Fuel Options Analysis

    International Nuclear Information System (INIS)

    Chang, Sunyoung; Jeong, Yon Hong; Han, Jae-Jun; Lee, Aeri; Hwang, Yong-Soo

    2015-01-01

    The disposal facility of the spent nuclear fuel will be operated from 2051. This paper presents the ENVI code developed by GoldSim Software to simulate options for managing spent nuclear fuel (SNF) in South Korea. The ENVI is a simulator to allow decision-makers to assist to evaluate the performance for spent nuclear fuel management. The multiple options for managing the spent nuclear fuel including the storage and transportation are investigated into interim storage, permanent disposal in geological repositories and overseas and domestic reprocessing. The ENVI code uses the GoldSim software to simulate the logistics of the associated activities. The result by the ENVI model not only produces the total cost to compare among the multiple options but also predict the sizes and timings of different facilities required. In order to decide the policy for spent nuclear management this purpose of this paper is to draw the optimum management plan to solve the nuclear spent fuel issue in the economical aspects. This paper is focused on the development of the ENVI's logic and calculations to simulate four options(No Reprocessing, Overseas Reprocessing, Domestic Reprocessing, and Overseas and Domestic Reprocessing) for managing the spent nuclear fuel in South Korea. The time history of the spent nuclear fuel produced from both the existing and future NPP's can be predicted, based on the Goldsim software made available very user friendly model. The simulation result will be used to suggest the strategic plans for the spent nuclear fuel management

  3. Application of safeguards design principles to the spent-fuel bundle counters for 600-MW CANDU reactors

    International Nuclear Information System (INIS)

    Stirling, A.J.; Allen, V.H.

    1979-01-01

    The irradiated fuel bundle counters for CANDU 600-MW reactors provide the IAEA with a secure and independent means of estimating the inventory of the spent-fuel storage bay at each inspection. Their function is straightforward - to count the bundles entering the storage area through the normal transfer ports. However, location, reliability, security and operating requirements make them highly ''intelligent'' instruments which have required a major development programme. Moreover, the bundle counters incorporate principles which apply to many unattended safeguards instruments. For example, concealing the operating status from potential diverters eases reliability specifications, continuous self-checking gives the inspector confidence in the readout, independence from continuous station services improves tamper-resistance, and the detailed data display provides tamper indication and a high level of credibility. Each irradiated fuel-bundle counter uses four Geiger counters to detect the passage of fuel bundles as they pass sequentially through the field-of-view. A microprocessor analyses the sequence of the Geiger counter signals and determines the number and direction of bundles transferred. The readout for IAEA inspectors includes both a tally and a printed log. The printer is also used to alert the inspector to abnormal fuel movements, tampering, Geiger counter failures and contamination of the fuel transfer mechanism. (author)

  4. Heat transfer analysis of consolidated dry storage system for CANDU spent fuel considering environmental conditions of Wolsong site

    International Nuclear Information System (INIS)

    Lee, K. H.; Yoon, J. H.; Choi, B. I.; Lee, H. Y.

    2004-01-01

    The purpose of the present paper is to perform heat transfer analysis of the MACSTOR/KN-400 dry storage system for CANDU spent fuel in order to predict maximum concrete temperatures and temperature gradients. This module has twice the capacity of the existing MACSTOR-200, which is in operation at Gentilly-2. In the thermal design of the MACSTOR/KN-400, Thermal Insulation Panels(TIP) were introduced to reduce concrete temperatures and temperature gradients in the module caused by the high fuel heat loads. Environmental factors such as solar heat, daily temperature variations and ambient temperatures in summer and winter at Wolsong site and the assumed presence of hot baskets were taken into consideration in the simulations. Two cases were performed for the MACSTOR/KN-400: Off-normal cases in summer and winter. The maximum local concrete temperatures were predicted to be 63 .deg. C for the off-normal case. The temperature gradients in the concrete walls and roof are predicted to be 28C and 25C for off-normal operation in summer, incorporating a 3C uncertainty. In conclusion, this paper shows that the maximum temperature for the module is expected to meet the temperature limitations of ACI 349

  5. A study on radiation shielding design in MACSTOR-400(CANDU spent fuel storage facility)

    International Nuclear Information System (INIS)

    Lee, Yoon Hee

    2006-02-01

    Since the spent fuel pool will be saturated in the near future, spent fuel storage facilities are urgently needed. Because of high radiation and decay heat, spent fuel management is difficult and important. In this study, the shielding thickness of MACSTOR-400 that satisfies the general surface dose rate limit has been investigated. And the radiation shielding safety at site boundary has also been evaluated. IAEA recommends the safety series as a guideline and the U.S. follows the NUREG guide for spent fuel storage facility design. In Japan, the regulation for internal transfer is applied to the spent fuel storage. In Korea, the ACT notification for radiation protection is considered. As a shielding design requirement, it is stated that the occupational exposure dose rate must not exceed 1 mSv/week. From this value, it is assumed that the surface dose rate limit is 25 μSv/hr. And for multi unit operation in same site, the dose rate limit at the controlled area boundary is 0.25 mSv/yr. MCNP code and Microshield program were used for calculating the surface dose rate and the dose rate at site boundary respectively. The shielding should be at least 90 cm thick except the air inlet to follow the surface dose rate limit. Additional shielding is needed on air inlet because the dose rate on air inlet is higher than the dose rate on concrete surface. Without the shielding structure, the shielding thickness should be at least 127 cm. In order to satisfy the surface dose rate limit with maintaining the same concrete thickness on air inlet, shielding structure is required on air inlet. The optimum shielding structure has been proposed in this study. The allowable number of MACSTORs with considering other nuclear facilities in Wolsung site is calculated at 60. It is expected that the required number of MACSTORs are 28 in order to store the total amount of spent fuel generated during NPP operation in Wolsung. Therefore, it seems to be safe in radiation point at site boundary

  6. A study on radiation shielding design in MACSTOR-400(CANDU spent fuel storage facility)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yoon Hee

    2006-02-15

    Since the spent fuel pool will be saturated in the near future, spent fuel storage facilities are urgently needed. Because of high radiation and decay heat, spent fuel management is difficult and important. In this study, the shielding thickness of MACSTOR-400 that satisfies the general surface dose rate limit has been investigated. And the radiation shielding safety at site boundary has also been evaluated. IAEA recommends the safety series as a guideline and the U.S. follows the NUREG guide for spent fuel storage facility design. In Japan, the regulation for internal transfer is applied to the spent fuel storage. In Korea, the ACT notification for radiation protection is considered. As a shielding design requirement, it is stated that the occupational exposure dose rate must not exceed 1 mSv/week. From this value, it is assumed that the surface dose rate limit is 25 μSv/hr. And for multi unit operation in same site, the dose rate limit at the controlled area boundary is 0.25 mSv/yr. MCNP code and Microshield program were used for calculating the surface dose rate and the dose rate at site boundary respectively. The shielding should be at least 90 cm thick except the air inlet to follow the surface dose rate limit. Additional shielding is needed on air inlet because the dose rate on air inlet is higher than the dose rate on concrete surface. Without the shielding structure, the shielding thickness should be at least 127 cm. In order to satisfy the surface dose rate limit with maintaining the same concrete thickness on air inlet, shielding structure is required on air inlet. The optimum shielding structure has been proposed in this study. The allowable number of MACSTORs with considering other nuclear facilities in Wolsung site is calculated at 60. It is expected that the required number of MACSTORs are 28 in order to store the total amount of spent fuel generated during NPP operation in Wolsung. Therefore, it seems to be safe in radiation point at site boundary

  7. The use of burnup credit in criticality control for the Korean spent fuel management program

    International Nuclear Information System (INIS)

    Koh, Duck Joon; Chon, Je Keun; Park, Chung Ryul; Ji, Pyung Kuk; Kim, Byung Tae; Jo, Chang Keun; Cho, Nam Zin

    1997-01-01

    More than 25% k-eff saving effect is observed in this burnup credit analysis. This mainly comes from the adoption of actinide nuclides and fission products in the criticality analysis. By taking burnup credit, the high capacity of the storage and transportation can be more fully utilized, reducing the space of storage and the number of shipments. Larger storage and fewer shipments for a given inventory of spent fuel result should in remarkable cost savings and more importantly reduce the risks to the public and occupational workers for the Korean Spent Fuel Management Program

  8. Thermal characteristics of spent activated carbon generated from air cleaning units in korean nuclear power plants

    Directory of Open Access Journals (Sweden)

    Ji-Yang So

    2017-06-01

    Full Text Available To identify the feasibility of disposing of spent activated carbon as a clearance level waste, we performed characterization of radioactive pollution for spent activated carbon through radioisotope analysis; results showed that the C-14 concentrations of about half of the spent activated carbon samples taken from Korean NPPs exceeded the clearance level limit. In this situation, we selected thermal treatment technology to remove C-14 and analyzed the moisture content and thermal characteristics. The results of the moisture content analysis showed that the moisture content of the spent activated carbon is in the range of 1.2–23.9 wt% depending on the operation and storage conditions. The results of TGA indicated that most of the spent activated carbon lost weight in 3 temperature ranges. Through py-GC/MS analysis based on the result of TGA, we found that activated carbon loses weight rapidly with moisture desorption reaching to 100°C and desorbs various organic and inorganic carbon compounds reaching to 200°C. The result of pyrolysis analysis showed that the experiment of C-14 desorption using thermal treatment technology requires at least 3 steps of heat treatment, including a heat treatment at high temperature over 850°C, in order to reduce the C-14 concentration below the clearance level.

  9. Thermal characteristics of spent activated carbon generated from air cleaning units in Korean nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Hang Rae; So, Ji Yang [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2017-06-15

    To identify the feasibility of disposing of spent activated carbon as a clearance level waste, we performed characterization of radioactive pollution for spent activated carbon through radioisotope analysis; results showed that the C-14 concentrations of about half of the spent activated carbon samples taken from Korean NPPs exceeded the clearance level limit. In this situation, we selected thermal treatment technology to remove C-14 and analyzed the moisture content and thermal characteristics. The results of the moisture content analysis showed that the moisture content of the spent activated carbon is in the range of 1.2–23.9 wt% depending on the operation and storage conditions. The results of TGA indicated that most of the spent activated carbon lost weight in 3 temperature ranges. Through py-GC/MS analysis based on the result of TGA, we found that activated carbon loses weight rapidly with moisture desorption reaching to 100°C and desorbs various organic and inorganic carbon compounds reaching to 200°C. The result of pyrolysis analysis showed that the experiment of C-14 desorption using thermal treatment technology requires at least 3 steps of heat treatment, including a heat treatment at high temperature over 850°C, in order to reduce the C-14 concentration below the clearance level.

  10. 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).

  11. Use of 'tail' as spent fuel dilution factor of Angra-1 (PWR) for use in the Embalse (Candu) reactor

    International Nuclear Information System (INIS)

    Mai, Luiz Antonio; Maiorino, Jose Rubens

    1995-01-01

    This work purposes a process to use the tail of isotopic enrichment as a factor of dilution (blending) for the burned fuel of Angra-I reactor (PWR) for final utilization in the Embalse (Candu). It was made use of the same technic in previous works that used natural uranium. For this purpose, it was made a tail parametrization inside of the traditional limits of enrichment (between 0.2 and 0.3%). The study showed that the tail utilization represents great savings for the uranium supplies and environment and economic advantages. (author). 8 refs, 4 figs, 11 tabs

  12. Prediction model for exhausted point of ion exchange resin column of moderator purification circuit at Korean CANDU plant

    International Nuclear Information System (INIS)

    Sohn, Wook; Kang, Duck-Won; Ahn, Hyun Kyoung; Rhee, In Hyoung

    2005-01-01

    Most of the carbon-14 produced at CANDU plants are removed by an Ion eXchange (IX) resin column of the moderator purification circuit, and a column is replaced based on an empirical guideline. Since the amount of carbon-14 released from CANDU plants is governed by the performance of a column, optimal operation of IX resin columns through the timely replacement based on an objective criterion is very important. For this, the model for predicting the exhausted point of an IX resin column has been developed based on local chemical equilibrium. The performance evaluation at Wolsong Unit 3 showed that the model was able to simulate the removal of species by an IX resin column to such a high degree that the model could provide an objective criterion to replace an IX resin column timely. The derived maximum service time of a fresh IX resin column was 4,080 h, about twice that of the existing empirical guideline (up to 2,000h). Accordingly, if the maximum service time derived in this paper is applied to Wolsong Unit 3, it is expected to reduce the cost needed for the replacement of IX resin column by about 50%. (author)

  13. Development of the Korean Reference Vertical Disposal System Concept for Spent Fuels

    International Nuclear Information System (INIS)

    Lee, J.Y.; Cho, D.K.; Kim, S.G.; Choi, H.J.; Choi, J.W.; Hahn, P.S.

    2006-01-01

    The development of a deep geologic disposal system for the spent fuel from nuclear power plants has been carried out since this program was launched at 1997 in Korea. In ' this paper, a pre-conceptual design of the Korean Reference HLW Vertical disposal System (KRS-V1) is presented. Though no site for the underground repository has yet been specified in Korea, a generic site with granitic rock is considered for reference HLW repository design. Depth of the repository is assumed to be 500 meters. The repository consists of the disposal area, technical rooms with four shafts to connect them to the ground level in the controlled area and technical rooms with an access tunnel and three shafts to connect them to the ground level in the uncontrolled area. Disposal area consists of disposal tunnels, panel tunnels and a central tunnel. The repository will be excavated, operated and backfilled in several phases including an Underground Research Laboratory (URL) phase. The result of this preliminary conceptual design will be used for an evaluation of the feasibility, analyses of the long term safety, information for public communication and a cost estimation etc. (authors)

  14. Development of geological disposal system for spent fuels and high-level radioactive wastes in Korea

    International Nuclear Information System (INIS)

    Choi, Heui Joo; Lee, Jong Youl; Choi, Jong Won

    2013-01-01

    Two different kinds of nuclear power plants produce a substantial amount of spent fuel annually in Korea. According to the current projection, it is expected that around 60,000 MtU of spent fuel will be produced from 36 PWR and APR reactors and 4 CANDU reactors by the end of 2089. In 2006, KAERI proposed a conceptual design of a geological disposal system (called KRS, Korean Reference disposal System for spent fuel) for PWR and CANDU spent fuel, as a product of a 4-year research project from 2003 to 2006. The major result of the research was that it was feasible to construct a direct disposal system for 20,000 MtU of PWR spent fuels and 16,000 MtU of CANDU spent fuel in the Korean peninsula. Recently, KAERI and MEST launched a project to develop an advanced fuel cycle based on the pyroprocessing of PWR spent fuel to reduce the amount of HLW and reuse the valuable fissile material in PWR spent fuel. Thus, KAERI has developed a geological disposal system for high-level waste from the pyroprocessing of PWR spent fuel since 2007. However, since no decision was made for the CANDU spent fuel, KAERI improved the disposal density of KRS by introducing several improved concepts for the disposal canister. In this paper, the geological disposal systems developed so far are briefly outlined. The amount and characteristics of spent fuel and HLW, 4 kinds of disposal canisters, the characteristics of a buffer with domestic Ca-bentonite, and the results of a thermal design of deposition holes and disposal tunnels are described. The different disposal systems are compared in terms of their disposal density.

  15. DEVELOPMENT OF GEOLOGICAL DISPOSAL SYSTEMS FOR SPENT FUELS AND HIGH-LEVEL RADIOACTIVE WASTES IN KOREA

    Directory of Open Access Journals (Sweden)

    HEUI-JOO CHOI

    2013-02-01

    Full Text Available Two different kinds of nuclear power plants produce a substantial amount of spent fuel annually in Korea. According to the current projection, it is expected that around 60,000 MtU of spent fuel will be produced from 36 PWR and APR reactors and 4 CANDU reactors by the end of 2089. In 2006, KAERI proposed a conceptual design of a geological disposal system (called KRS, Korean Reference disposal System for spent fuel for PWR and CANDU spent fuel, as a product of a 4-year research project from 2003 to 2006. The major result of the research was that it was feasible to construct a direct disposal system for 20,000 MtU of PWR spent fuels and 16,000 MtU of CANDU spent fuel in the Korean peninsula. Recently, KAERI and MEST launched a project to develop an advanced fuel cycle based on the pyroprocessing of PWR spent fuel to reduce the amount of HLW and reuse the valuable fissile material in PWR spent fuel. Thus, KAERI has developed a geological disposal system for high-level waste from the pyroprocessing of PWR spent fuel since 2007. However, since no decision was made for the CANDU spent fuel, KAERI improved the disposal density of KRS by introducing several improved concepts for the disposal canister. In this paper, the geological disposal systems developed so far are briefly outlined. The amount and characteristics of spent fuel and HLW, 4 kinds of disposal canisters, the characteristics of a buffer with domestic Ca-bentonite, and the results of a thermal design of deposition holes and disposal tunnels are described. The different disposal systems are compared in terms of their disposal density.

  16. Enhanced CANDU 6 Reactor

    International Nuclear Information System (INIS)

    Azeez, S.; Alizadeh, A.; Girouard, P.

    2005-01-01

    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)

  17. CANDU 3000

    International Nuclear Information System (INIS)

    Keillor, Mac

    1987-01-01

    In this article, the CANDU 300 design, and the team that designed it, are featured. The CANDU 300 will operate at an energy cost similar to that of the larger CANDU units, but is sized for emerging markets. Ease of construction is an important feature: for example, full 360-degree access is available to each of the five buildings during construction; and the whole plant consists of about 90 modules, which can be built in separate locations, and hoisted into place

  18. Joint studies on large CANDU

    International Nuclear Information System (INIS)

    Lee, Ikhwan; Yu, S. K. W.

    1994-01-01

    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.

  19. Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors - IV: DUPIC Fuel Cycle Cost

    International Nuclear Information System (INIS)

    Ko, Won Il; Choi, Hangbok; Yang, Myung Seung

    2001-01-01

    This study examines the economics of the DUPIC fuel cycle using unit costs of fuel cycle components estimated based on conceptual designs. The fuel cycle cost (FCC) was calculated by a deterministic method in which reference values of fuel cycle components are used. The FCC was then analyzed by a Monte Carlo simulation to get the uncertainty of the FCC associated with the unit costs of the fuel cycle components. From the deterministic analysis on the equilibrium fuel cycle model, the DUPIC FCC was estimated to be 6.21 to 6.34 mills/kW.h for DUPIC fuel options, which is a little smaller than that of the once-through FCC by 0.07 to 0.27 mills/kW.h. Considering the uncertainty (0.40 to 0.44 mills/kW.h) of the FCC estimated by the Monte Carlo simulation method, the cost difference between the DUPIC and once-through fuel cycle is negligible. On the other hand, the material balance calculation has shown that the DUPIC fuel cycle can save natural uranium resources by ∼20% and reduce the spent fuel arising by ∼65% compared with the once-through fuel cycle. In conclusion, the DUPIC fuel cycle is comparable with the once-through fuel cycle from the viewpoint of FCC. In the future, it should be important to consider factors such as the environmental benefit owing to natural uranium savings, the capability of reusing spent pressurized water reactor fuel, and the safeguardability of the fuel cycle when deciding on an advanced nuclear fuel cycle option

  20. CANDU lectures

    International Nuclear Information System (INIS)

    Rouben, B.

    1984-06-01

    This document is a compilation of notes prepared for two lectures given by the author in the winter of 1983 at the Institut de Genie Nucleaire, Ecole Polytechnique, Montreal. The first lecture gives a physical description of the CANDU reactor core: the nuclear lattice, the reactivity mechanisms, their functions and properties. This lecture also covers various aspects of reactor core physics and describes different calculational methods available. The second lecture studies the numerous facets of fuel management in CANDU reactors. The important variables in fuel management, and the rules guiding the refuelling strategy, are presented and illustrated by means of results obtained for the CANDU 600

  1. CANDU fuel

    International Nuclear Information System (INIS)

    MacEwan, J.R.; Notley, M.J.F.; Wood, J.C.; Gacesa, M.

    1982-09-01

    The direction of CANDU fuel development was set in 1957 with the decision to build pressure tube reactors. Short - 50 cm long - rodded bundles of natural UO 2 clad in Zircaloy were adopted to facilitate on-power fuelling to improve uranium utilization. Progressive improvements were made during 25 years of development, involving 650 man years and 180 million dollars. Today's CANDU bundle is based on the knowledge gained from extensive irradiation testing and experience in power reactors. The main thrust of future development is to demonstrate that the present bundle is suitable, with minor modifications, for thorium fuels

  2. CANDU fuel cycle options in Korea

    International Nuclear Information System (INIS)

    Boczar, P. G.; Fehrenbach, P. J.; Meneley, D. A.

    1996-01-01

    There are many reasons for countries embarking on a CANDU R program to start with the natural uranium fuel cycle. Simplicity of fuel design, ease of fabrication, and ready availability of natural uranium all help to localize the technology and to reduce reliance on foreign technology. Nonetheless, at some point, the incentives for using natural uranium fuel may be outweighed by the advantages of alternate fuel cycles. The excellent neutron economy, on-line refuelling, and simple fuel-bundle design provide an unsurpassed degree of fuel-cycle flexibility in CANDU reactors. 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 two- to 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 dose 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

  3. Enhanced candu 6 reactor: status

    International Nuclear Information System (INIS)

    Azeez, S.; Girouard, P.

    2006-01-01

    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

  4. CANDU development

    International Nuclear Information System (INIS)

    Brooks, G.L.

    1981-06-01

    Evolution of the 950 MW(e) CANDU reactor is summarized. The design was specifically aimed at the export market. Factors considered in the design were that 900-1000 MW is the maximum practical size for most countries; many countries have warmer condenser cooling water than Canada; the plant may be located on coastal sites; seismic requirements may be more stringent; and the requirements of international, as well as Canadian, standards must be satisfied. These considerations resulted in a 600-channel reactor capable of accepting condenser cooling water at 32 0 C. To satisfy the requirement for a proven design, the 950 MW CANDU draws upon the basic features of the Bruce and Pickering plants which have demonstrated high capacity factors

  5. Thermal safety analysis of a dry storage cask for the Korean standard spent fuel - 16159

    International Nuclear Information System (INIS)

    Cha, Jeonghun; Kim, S.N.; Choi, K.W.

    2009-01-01

    A conceptual dry storage facility, which is based on a commercial dry storage facility, was designed for the Korea standard spent nuclear fuel (SNF) and preliminary thermal safety analysis was performed in this study. To perform the preliminary thermal analysis, a thermal analysis method was proposed. The thermal analysis method consists of 2 parts. By using the method, the surface temperature of the storage canister corresponding to the SNF clad temperature was calculated and the adequate air duct area was decided using the calculation result. The initial temperature of the facility was calculated and the fire condition and half air duct blockage were analyzed. (authors)

  6. Korean interim storage issues and R and D activities on spent fuel management

    International Nuclear Information System (INIS)

    Ji Sup Yoon; Seung-Gy Ro; Hyun-Soo Park

    1999-01-01

    Korea has witnessed over a decade of vicissitudes in the issue of radioactive waste management due mainly to the problem of site acquisition. As the major mission of the nation at radioactive waste management programme was to provide a center for disposal of low-level radwaste and for interim storage of spent nuclear fuel from nuclear power plants, the question of site securing has had a big impact on the implement action of overall programme. The site problem has resulted in, as an aftermath, restructuring of the overall programme for radioactive waste management. Missions of NEMAC (Nuclear Environment Management Center), originally established as a subsidiary of Korea Atomic Energy Research Institute (KAERI), for the national programme was dissolved as of the end of last year. Beginning of this year, a new entity NETEC (Nuclear Environment Technology Center) as a subsidiary of KEPCO (Korea Electric Power Co.) has taken over major tasks of the past NEMAC, while the R and D work associated with the past task of NEMAC is transferred back to KAERI. This paper gives a review on the past background which has driven the radioactive waste management in Korea to the current state of the affairs, with special emphasis on R and D activities associated with spent nuclear fuel transportation, handling, and storage. (author)

  7. CANDU fuel cycle options in Korea

    International Nuclear Information System (INIS)

    Boczar, P.G.; Fehrenbach, P.J.; Meneley, D.A.

    1996-04-01

    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

  8. Supporting CANDU operators-CANDU owners group

    International Nuclear Information System (INIS)

    Collingwood, B.R.

    1997-01-01

    The CANDU Owners Group (COG) was formed in 1984 by the Canadian CANDU owning utilities and Atomic Energy of Canada limited (AECL). Participation was subsequently extended to all CANDU owners world-wide. The mandate of the COG organization is to provide a framework for co-operation, mutual assistance and exchange of information for the successful support, development, operation, maintenance and economics of CANDU nuclear electric generating stations. To meet these objectives COG established co-operative programs in two areas: 1. Station Support. 2. Research and Development. In addition, joint projects are administered by COG on a case by case basis where CANDU owners can benefit from sharing of costs

  9. A study for good regulatin of the CANDU's in Korea. Development of safety regulatory requirement for CANDU nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Se Ki; Shin, Y. K.; Kim, J. S.; Yu, Y. J.; Lee, Y. J. [Ajou Univ., Suwon (Korea, Republic of)

    2001-03-15

    The objective of project is to derive the policy recommendations to improve the efficiency of CANDU plants regulation. These policy recommendations will eventually contribute to the upgrading of Korean nuclear regulatory system and safety enhancement. During the first phase of this 2 years study, following research activities were done. On-site survey and analysis on CANDU plants regulation. Review on CANDU plants regulating experiences and current constraints. Review and analysis on the new Canadian regulatory approach.

  10. CANDU nuclear power system

    International Nuclear Information System (INIS)

    1981-01-01

    This report provides a summary of the components that make up a CANDU reactor. Major emphasis is placed on the CANDU 600 MW(e) design. The reasons for CANDU's performance and the inherent safety of the system are also discussed

  11. CANDU operating experience

    International Nuclear Information System (INIS)

    McConnell, L.G.; Woodhead, L.W.; Fanjoy, G.R.

    1982-03-01

    The CANDU Pressurized Heavy Water (CANDU-PHW) type of nuclear electric generating station has been developed jointly by Atomic Energy of Canada Limited and Ontario Hydro. This paper highlights Ontario Hydro's operating experience using the CANDU-PHW system, with a focus on the operating performance and costs, reliability of system components and nuclear safety considerations both to the workers and the public

  12. Advanced CANDU reactors

    International Nuclear Information System (INIS)

    Dunn, J.T.; Finlay, R.B.; Olmstead, R.A.

    1988-12-01

    AECL has undertaken the design and development of a series of advanced CANDU reactors in the 700-1150 MW(e) size range. These advanced reactor designs are the product of ongoing generic research and development programs on CANDU technology and design studies for advanced CANDU reactors. The prime objective is to create a series of advanced CANDU reactors which are cost competitive with coal-fired plants in the market for large electricity generating stations. Specific plant designs in the advanced CANDU series will be ready for project commitment in the early 1990s and will be capable of further development to remain competitive well into the next century

  13. CANDU fuel-cycle vision

    International Nuclear Information System (INIS)

    Boczar, P.G.

    1999-01-01

    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

  14. CANDU fuel-cycle vision

    International Nuclear Information System (INIS)

    Boczar, P.G

    1998-05-01

    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

  15. CANDU advanced fuel cycles

    International Nuclear Information System (INIS)

    Slater, J.B.

    1986-03-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyun Soo; Jun, Kwan Sik; Nah, Jung Won; Park, Jang Jin; Kim, Jong Hoh; Cho, Yung Hyun; Baek, Seung Woo; Shin, Jin Myung; Yang, Seung Yung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-07-01

    The immobilization materials for radioactive wastes resulting from the DUPIC fuel manufacturing process were selected and their characteristics were evaluated. To predict the trapping behavior of the Ruthenium, a semi-volatile nuclide, its volatility was measured and thermogravimetric analysis were performed with simulated fuel. New Ruthenium trapping material was developed which is deposited on ceramic honey-comb monolith of cordierite. The base glass was manufactured with fly ash added to the borosilicate glass. The composition of the scrap waste was calculated based on the PWR spent fuel which has initial {sup 235}U content of 3.5%, burnup of 35,000 MWD/MTU and cooling time of 10 years. Simulated waste glass was manufactured, and its chemical durability was evaluated by soxhlet leach test. Radioactivity of non-oxidized cladding material were measured. The preliminary design criteria were prepared for off-gas treatment system in IMEF. 31 figs, 42 tabs, 51 refs. (Author).

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

    International Nuclear Information System (INIS)

    Park, Hyun Soo; Jun, Kwan Sik; Nah, Jung Won; Park, Jang Jin; Kim, Jong Hoh; Cho, Yung Hyun; Baek, Seung Woo; Shin, Jin Myung; Yang, Seung Yung

    1994-07-01

    The immobilization materials for radioactive wastes resulting from the DUPIC fuel manufacturing process were selected and their characteristics were evaluated. To predict the trapping behavior of the Ruthenium, a semi-volatile nuclide, its volatility was measured and thermogravimetric analysis were performed with simulated fuel. New Ruthenium trapping material was developed which is deposited on ceramic honey-comb monolith of cordierite. The base glass was manufactured with fly ash added to the borosilicate glass. The composition of the scrap waste was calculated based on the PWR spent fuel which has initial 235 U content of 3.5%, burnup of 35,000 MWD/MTU and cooling time of 10 years. Simulated waste glass was manufactured, and its chemical durability was evaluated by soxhlet leach test. Radioactivity of non-oxidized cladding material were measured. The preliminary design criteria were prepared for off-gas treatment system in IMEF. 31 figs, 42 tabs, 51 refs. (Author)

  18. Economics of CANDU-PHW

    International Nuclear Information System (INIS)

    McConnell, L.G.; Woodhead, L.W.; Fanjoy, G.R.

    1982-03-01

    The CANDU-Pressurized Heavy Water (CANDU-PHW) type of nuclear-electric generating station has been developed jointly by Atomic Energy of Canada Limited and Ontario Hydro. This paper discusses the cost of producing electricity from CANDU, presents actual cost experience of CANDU and coal in Ontario, presents projected CANDU and coal costs in Ontario and compares CANDU and Light Water Reactor cost estimates in Ontario

  19. Economics of CANDU-PHW

    International Nuclear Information System (INIS)

    Jackson, H.A.; Woodhead, L.W.; Fanjoy, G.R.

    1984-03-01

    The CANDU-Pressurized Heavy Water (CANDU-PHW) type of nuclear-electric generating station has been developed jointly by Atomic Energy of Canada Limited and Ontario Hydro. This paper discusses the cost of producing electricity from CANDU, presents actual cost experience of CANDU and coal in Ontario, presents projected CANDU and coal costs in Ontario and compares CANDU and Light Water Reactor cost estimates in Ontario

  20. Economics of CANDU-PHW

    International Nuclear Information System (INIS)

    Jackson, H.A.; Horton, E.P.; Woodhead, L.W.; Fanjoy, G.R.

    1985-03-01

    The CANDU-Pressurized Heavy Water (CANDU-PHW) type of nuclear-electric generating station has been developed jointly by Atomic Energy of Canada Limited and Ontario Hydro. This paper discusses the cost of producing electricity from CANDU, presents actual cost experience of CANDU and coal in Ontario, presents projected CANDU and coal costs in Ontario and compares CANDU and Light Water Reactor cost estimates in Ontario

  1. The CANDU 9

    International Nuclear Information System (INIS)

    Hart, R.S.

    1994-01-01

    The CANDU 9 plants are single unit versions of the Bruce B design, incorporating relevant technical advances made in CANDU 6, and the newer Darlington and CANDU 3 designs. This paper describes the CANDU 9 480/SEU, with an electrical output of about 1050 MW. In this designation, 480 refers to the number of fuel channels, and SEU to slightly enriched uranium. Emphasis is placed on evolutionary design, and the use of well proven design features, to ensure regulatory licensability and reliable operation. Safety is enhanced through simplification and improvement of key systems and components. Relatively low energy costs result from reduced specific capital cost, reduced operating and maintenance cost, and reduced radiation exposure to personnel. Standardization is emphasized inasmuch as all key components (steam generators, heat transport pumps, pressure tubes fuelling machines etc.) ar of the same design as those in operating CANDU stations. Advanced CANDU fuel cycles are readily accommodated. 1 ref., 1 tab., 11 figs

  2. CANDU reg-sign -- A Canadian energy system

    International Nuclear Information System (INIS)

    Sejnoha, R.

    1994-01-01

    Uranium is one of Canada's important natural resources. It is perhaps not surprising that a country with such an abundance of uranium developed its own inexpensive, safe, and environmentally friendly system of energy generation, based on uranium: CANDU reg-sign (CANada-Deuterium-Uranium, a registered trademark of AECL). The objective of this paper is to describe briefly the main features of the CANDU system and explain methods used to assure the compatibility with the requirements for a clean and safe environment. The paper describes the CANDU reactor, and discusses storage of spent fuel, reactor performance, normal operating conditions, safety under accident conditions, and quality assurance in design, manufacturing, and operation

  3. The CANDU 80

    International Nuclear Information System (INIS)

    Hart, R.S.

    1998-01-01

    AECL has completed the conceptual design of a small CANDU plant with an output, in the range of 300 MWth (called the CANDU 80), suitable for a variety of electrical and co-generation applications including desalination, oil sands oil extraction and processing, and the provision of electricity and heat to areas with low demand. This paper provides a brief overview of the CANDU 80, and discusses key features contributing to safety and operational margins

  4. Economics of CANDU

    International Nuclear Information System (INIS)

    McConnell, L.G.; Woodhead, L.W.

    1981-02-01

    The cost of producing electricity from CANDU reactors is discussed. The total unit energy cost of base-load electricity from CANDU reactors is compared with that of coal-fired plants in Ontario. In 1980 nuclear power was 8.41 m$/kW.h less costly for plants of similar size and vintage. Comparison of CANDU with pressurized water reactors indicated that the latter would be about 26 percent more costly in Ontario

  5. Korean Reference HLW Disposal System

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Heui Joo; Lee, J. Y.; Kim, S. S. (and others)

    2008-03-15

    This report outlines the results related to the development of Korean Reference Disposal System for High-level radioactive wastes. The research has been supported around for 10 years through a long-term research plan by MOST. The reference disposal method was selected via the first stage of the research during which the technical guidelines for the geological disposal of HLW were determined too. At the second stage of the research, the conceptual design of the reference disposal system was made. For this purpose the characteristics of the reference spent fuels from PWR and CANDU reactors were specified, and the material and specifications of the canisters were determined in term of structural analysis and manufacturing capability in Korea. Also, the mechanical and chemical characteristics of the domestic Ca-bentonite were analyzed in order to supply the basic design parameters of the buffer. Based on these parameters the thermal and mechanical analysis of the near-field was carried out. Thermal-Hydraulic-Mechanical behavior of the disposal system was analyzed. The reference disposal system was proposed through the second year research. At the final third stage of the research, the Korean Reference disposal System including the engineered barrier, surface facilities, and underground facilities was proposed through the performance analysis of the disposal system.

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

    International Nuclear Information System (INIS)

    Xie Zhongsheng; Huo Xiaodong

    2002-01-01

    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

  7. Spent fuel management in Canada

    International Nuclear Information System (INIS)

    Khan, A.; Pattantyus, P.

    1999-01-01

    The current status of the Canadian spent fuel storage is presented. This includes wet and dry interim storage. Extension of wet interim storage facilities is nor planned, as dry technologies have found wide acceptance. The Canadian nuclear program is sustained by commercial Ontario Hydro CANDU type reactors, since 1971, representing 13600 MW(e) of installed capacity, able to produce 9200 spent fuel bundles (1800 tU) every year, and Hydro Quebec and New Brunswick CANDU reactors each producing 685 MW(e) and about 100 tU of spent fuel annually. The implementation of various interim (wt and dry) storage technologies resulted in simple, dense and low cost systems. Economical factors determined that the open cycle option be adopted for the CANDU type reactors rather that recycling the spent fuel. Research and development activities for immobilization and final disposal of nuclear waste are being undertaken in the Canadian Nuclear Fuel Waste Management Program

  8. Homogeneous Thorium Fuel Cycles in Candu Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hyland, B.; Dyck, G.R.; Edwards, G.W.R.; Magill, M. [Chalk River Laboratories, Atomic Energy of Canada Limited (Canada)

    2009-06-15

    The CANDU{sup R} reactor has an unsurpassed degree of fuel-cycle flexibility, as a consequence of its fuel-channel design, excellent neutron economy, on-power refueling, and simple fuel bundle [1]. These features facilitate the introduction and full exploitation of thorium fuel cycles in Candu reactors in an evolutionary fashion. Because thorium itself does not contain a fissile isotope, neutrons must be provided by adding a fissile material, either within or outside of the thorium-based fuel. Those same Candu features that provide fuel-cycle flexibility also make possible many thorium fuel-cycle options. Various thorium fuel cycles can be categorized by the type and geometry of the added fissile material. The simplest of these fuel cycles are based on homogeneous thorium fuel designs, where the fissile material is mixed uniformly with the fertile thorium. These fuel cycles can be competitive in resource utilization with the best uranium-based fuel cycles, while building up a 'mine' of U-233 in the spent fuel, for possible recycle in thermal reactors. When U-233 is recycled from the spent fuel, thorium-based fuel cycles in Candu reactors can provide substantial improvements in the efficiency of energy production from existing fissile resources. The fissile component driving the initial fuel could be enriched uranium, plutonium, or uranium-233. Many different thorium fuel cycle options have been studied at AECL [2,3]. This paper presents the results of recent homogeneous thorium fuel cycle calculations using plutonium and enriched uranium as driver fuels, with and without U-233 recycle. High and low burnup cases have been investigated for both the once-through and U-233 recycle cases. CANDU{sup R} is a registered trademark of Atomic Energy of Canada Limited (AECL). 1. Boczar, P.G. 'Candu Fuel-Cycle Vision', Presented at IAEA Technical Committee Meeting on 'Fuel Cycle Options for LWRs and HWRs', 1998 April 28 - May 01, also Atomic Energy

  9. CANDU 3 - Modularization

    International Nuclear Information System (INIS)

    McAskie, M.J.

    1991-01-01

    The CANDU 3 Heavy Water Reactor is the newest design developed by AECL CANDU. It has set as a major objective, the achievement of significant reductions in both cost and schedule over previous designs. The basic construction strategy is to incorporate extensive modularization of the plant in order to parallel the civil and mechanical installation works. This results in a target 38 month construction schedule from first concrete to in-service compared to 68 months for the Wolsong-1 CANDU 6 actually achieved and the 54 months envisaged for an improved CANDU 6. This paper describes the module concepts that have been developed and explains how they contribute to the overall construction program and achieve the desired cost and schedule targets set for the CANDU 3. (author). 7 figs, 2 tabs

  10. Advancing CANDU Technology Through R and D

    International Nuclear Information System (INIS)

    Torgerson, David F.

    1993-01-01

    CANDU reactors are evolving to meet future requirements using incremental changes as opposed to revolutionary design changes. The main elements for advancing the technology reducing capital and operating, increasing capacity factors, increasing passive safety, and enhancing fuel/fuel cycle flexibility. These elements are being addressed by carrying out research and development in the areas of safety, plant systems and components, heavy water production, information technology, fuel channels, and fuel/fuel cycle technology. In safety, the focus is on using the inherent features of CANDU to enhance passive or natural safety concepts, such as the use of the moderator as an effective heat sink, and the development of advanced fuels to improve critical heat flux and to reduce source terms. Plant systems and components work includes improvements to plant systems such as steam generators, heat exchangers, pump seals, and advanced control room technology. Heavy water processes are being developed that can be used with existing hydrogen production plants, or that can be used in a stand-alone mode. Information technology is being developed to cover all aspects of CANDU design, construction, and operation. Fuel channel improvements include elucidation and application of basic materials science for life extension, and the development of advanced non-destructive examination methods. Fuel and fuel cycle work is focusing on LWR/CANDU synergy, such as the use of recovered uranium and the direct use of spent PWR fuel in CANDU reactor, advanced fuels to improve burnup and economics (e. g., the joint AECB/KAERI Conflux program), and low void reactivity fuel to enhance passive safety. This paper gives an overview of some of the R and D supporting these activities, with particular emphasis on Alice's vision for advancing CANDU technology over the next 10 years

  11. Next generation CANDU plants

    International Nuclear Information System (INIS)

    Hedges, K.R.; Yu, S.K.W.

    1998-01-01

    Future CANDU designs will continue to meet the emerging design and performance requirements expected by the operating utilities. The next generation CANDU products will integrate new technologies into both the product features as well as into the engineering and construction work processes associated with delivering the products. The timely incorporation of advanced design features is the approach adopted for the development of the next generation of CANDU. AECL's current products consist of 700MW Class CANDU 6 and 900 MW Class CANDU 9. Evolutionary improvements are continuing with our CANDU products to enhance their adaptability to meet customers ever increasing need for higher output. Our key product drivers are for improved safety, environmental protection and improved cost effectiveness. Towards these goals we have made excellent progress in Research and Development and our investments are continuing in areas such as fuel channels and passive safety. Our long term focus is utilizing the fuel cycle flexibility of CANDU reactors as part of the long term energy mix

  12. Localization of CANDU technology

    Energy Technology Data Exchange (ETDEWEB)

    Alizadeh, Ala

    2010-09-15

    The CANDU pressurized heavy water reactor's principal design features suit it particularly well for technology transfer and localization. When the first commercial CANDU reactors of 540 MWe entered service in 1971, Canada's population of less than 24 million supported a 'medium' level of industrial development, lacking the heavy industrial capabilities of larger countries like the USA, Japan and Europe. A key motivation for Canada in developing the CANDU design was to ensure that Canada would have the autonomous capacity to build and operate nuclear power reactors without depending on foreign sources for key components or enriched fuel.

  13. CANDU nuclear reactor technology

    International Nuclear Information System (INIS)

    Kakaria, B. K.

    1994-01-01

    AECL has over 40 years of experience in the nuclear field. Over the past 20 years, this unique Canadian nuclear technology has made a worldwide presence, In addition to 22 CANDU reactors in Canada, there are also two in India, one in Pakistan, one in Argentina, four in Korea and five in Romania. CANDU advancements are based on evolutionary plant improvements. They consist of system performance improvements, design technology improvements and research and development in support of advanced nuclear power. Given the good performance of CANOU plants, it is important that this CANDU operating experience be incorporated into new and repeat designs

  14. Team CANDU : ready for the marketplace

    International Nuclear Information System (INIS)

    Howieson, J.Q.

    2007-01-01

    This paper outlines the partnership between AECL and a number of leading global nuclear suppliers to market the Candu power reactor. The mission of the CANDU team is to develop market opportunities for CANDU technology and deliver successful CANDU projects

  15. CANDU in Romania

    International Nuclear Information System (INIS)

    Keillor, M.

    1990-01-01

    The author, a former journalist, and now manager of media relations at AECL CANDU, visited Romania to get a first-hand account of conditions at the Cernavoda site. He refutes allegations of slave labour, or inhuman conditions

  16. Marketing CANDU internationally

    International Nuclear Information System (INIS)

    Langstaff, J.H.

    1980-06-01

    The market for CANDU reactor sales, both international and domestic, is reviewed. It is reasonable to expect that between five and ten reactors can be sold outside Canada before the end of the centry, and new domestic orders should be forthcoming as well. AECL International has been created to market CANDU, and is working together with the Canadian nuclear industry to promote the reactor and to assemble an attractive package that can be offered abroad. (L.L.)

  17. Privatize Candu (question mark)

    International Nuclear Information System (INIS)

    Kelly, Thomas.

    1981-01-01

    A report sponsored by a group of nuclear suppliers and the Royal Bank suggested that the Candu reactor system would sell better if it were owned by a private company. Licensing of a Candu reactor in the U.S.A. was also suggested. The author of this article agrees with these points, but disagrees with the suggestion that safeguards should be relaxed. He suggests that contracts should stipulate that instrumentation should be supplied as much as possible from Canadian sources

  18. The back end of the fuel cycle and CANDU

    International Nuclear Information System (INIS)

    Allan, C.J.; Dormuth, K.W.

    2001-01-01

    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)

  19. CANDU fuel cycles - present and future

    International Nuclear Information System (INIS)

    Mooradian, A.J.

    1976-05-01

    The present commercially proven Canadian nuclear power system is based on a once-through natural uranium fuel cycle characterized by high uranium utilization and a high conversion efficiency. The cycle closes with secure retrievable storage of spent fuel. This cycle is based on a CANDU reactor concept which is now well understood. Both active and passive fuel storage options have been investigated and will be described in this paper. Future development of the CANDU system is focussed on conservation of uranium by plutonium and thorium recycle. The full exploitation of these options requires continued emphasis on neutron conservation, efficiency of extraction and fuel refabrication processes. The results of recent studies are discussed in this paper. (author)

  20. Effect of DUPIC cycle on CANDU reactor safety parameters

    Energy Technology Data Exchange (ETDEWEB)

    Mohamed, Nader M. A. [Atomic Energy Authority, ETRR-2, Cairo (Egypt); Badawi, Alya [Dept. of Nuclear and Radiation Engineering, Alexandria University, Alexandria (Egypt)

    2016-10-15

    Although, the direct use of spent pressurized water reactor (PWR) fuel in CANda Deuterium Uranium (CANDU) reactors (DUPIC) cycle is still under investigation, DUPIC cycle is a promising method for uranium utilization improvement, for reduction of high level nuclear waste, and for high degree of proliferation resistance. This paper focuses on the effect of DUPIC cycle on CANDU reactor safety parameters. MCNP6 was used for lattice cell simulation of a typical 3,411 MWth PWR fueled by UO{sub 2} enriched to 4.5w/o U-235 to calculate the spent fuel inventories after a burnup of 51.7 MWd/kgU. The code was also used to simulate the lattice cell of CANDU-6 reactor fueled with spent fuel after its fabrication into the standard 37-element fuel bundle. It is assumed a 5-year cooling time between the spent fuel discharges from the PWR to the loading into the CANDU-6. The simulation was carried out to calculate the burnup and the effect of DUPIC fuel on: (1) the power distribution amongst the fuel elements of the bundle; (2) the coolant void reactivity; and (3) the reactor point-kinetics parameters.

  1. Ontario Hydro CANDU operating experience

    International Nuclear Information System (INIS)

    Jackson, H.A.; Woodhead, L.W.; Fanjoy, G.R.

    1984-03-01

    The CANDU Pressurized Heavy Water (CANDU-PHW) type of nuclear-electric generating station has been developed jointly by Atomic Energy of Canada Limited and Ontario Hydro. This report highlights Ontario Hydro's operating experience using the CANDU-PHW system, with a focus on the operating performance and costs, reliability of system components and nuclear safety considerations for the workers and the public

  2. Ontario Hydro CANDU operating experience

    International Nuclear Information System (INIS)

    Bartholomew, R.W.; Woodhead, L.W.; Horton, E.P.; Nichols, M.J.; Daly, I.N.

    1987-01-01

    The CANDU Pressurized Heavy Water (CANDU-PHW) type of nuclear-electric generating station has been developed jointly by Atomic Energy of Canada Limited and Ontario Hydro. This report highlights Ontario Hydro's operating experience using the CANDU-PHW system, with a focus on worker and public safety, operating performance and costs, and reliability of system components

  3. Fuel cycles - a key to future CANDU success

    International Nuclear Information System (INIS)

    Kuran, S.; Hopwood, J.; Hastings, I.J.

    2011-01-01

    Globally, fuel cycles are being evaluated as ways of extending nuclear fuel resources, addressing security of supply and reducing back-end spent-fuel management. Current-technology thermal reactors and future fast reactors are the preferred platform for such fuel cycle applications and as an established thermal reactor with unique fuel-cycle capability, CANDU will play a key role in fulfilling such a vision. The next step in the evolution of CANDU fuel cycles will be the introduction of Recovered Uranium (RU), derived from conventional reprocessing. A low-risk RU option applicable in the short term comprises a combination of RU and Depleted Uranium (DU), both former waste streams, giving a Natural Uranium Equivalent (NUE) fuel. This option has been demonstrated in China, and all test bundles have been removed from the Qinshan 1 reactor. Additionally, work is being done on an NUE full core, a Thorium demonstration irradiation and an Advanced Fuel CANDU Reactor(AFCR). AECL is developing other fuel options for CANDU, including actinide waste burning. AECL has developed the Enhanced CANDU 6 (EC6) reactor, upgraded from its best-performing CANDU 6 design. High neutron economy, on-power refueling and a simple fuel bundle provide the EC6 with the flexibility to accommodate a range of advanced fuels, in addition to its standard natural uranium. (author)

  4. Trends in CANDU licensing

    International Nuclear Information System (INIS)

    Snell, V.G.; Grant, S.D.

    1997-01-01

    Modern utilities view nuclear power more and more as a commodity - it must compete 'today' with current alternatives to attract their investment. With its long construction times and large capital investment, nuclear plants are vulnerable to delays once they have been committed. There are two related issues. Where the purchaser and the regulator are experienced in CANDU, the thrust is a very practical one: to identify and resolve major licensing risks at a very early stage in the project. Thus for a Canadian project, the designer (AECL) and the prospective purchaser would deal directly with the AECB. However CANDU has also been successfully licensed in other countries, including Korea, Romania, Argentina, India and Pakistan. Each of these countries has its own regulatory agency responsible for licensing the plant. In addition, however, the foreign customer and regulator may seek input from the AECB, up to and including a statement of licensability in Canada; this is not normally needed for a ''repeat'' plant and/or if the customer is experienced in CANDU, but can be requested if the plant configuration has been modified significantly from an already-operating CANDU. It is thus the responsibility of the designer to initiate early discussions with the AECB so the foreign CANDU meets the expectations of its customers

  5. An integrated CANDU system

    International Nuclear Information System (INIS)

    Donnelly, J.

    1982-09-01

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

  6. Plutonium Consumption Program, CANDU Reactor Project final report

    Energy Technology Data Exchange (ETDEWEB)

    1994-07-31

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

  7. Plutonium Consumption Program, CANDU Reactor Project final report

    International Nuclear Information System (INIS)

    1994-01-01

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

  8. CANDU: study and review

    Energy Technology Data Exchange (ETDEWEB)

    Morad, César M., E-mail: cesar.morad@usp.br [Universidade de São Paulo (POLI/USP), SP (Brazil). Escola Politécnica; Stefani, Giovanni L. de, E-mail: giovanni.stefani@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil); Santos, Thiago A. dos, E-mail: thiago.santos@ufabc.edu.br [Universidade Federal do ABC (UFABC), Santo André, SP (Brazil)

    2017-07-01

    The CANDU (Canadian Deuterium Uranium) is a nuclear reactor developed by AECL (Atomic Energy of Canada Limited). The first small-scale reactor is known as NPD and was made in 1955 and commenced operation in 1962. It is a pressurized heavy water reactor and uses D2O as moderator and coolant and therefore uses natural uranium as fuel. There have been two major types of CANDU reactors, the original design of around 500 MWe that was intended to be used in multi-reactor installations in large plants, and the rationalized CANDU6 which has units in Argentina, South Korea, Pakistan, Romania and China. Throughout the 1980s and 90s the nuclear power market suffered a major crash, with few new plants being constructed in North America or Europe. Design work continued through, however, and a number of new design concepts were introduced that dramatically improved safety, capital costs, economics and overall performance. These Generation III+ and Generation IV machines became a topic of considerable interest in the early 2000s as it appeared a nuclear renaissance was underway and large numbers of new reactors would be built over the next decade. The present work aims to study the reactors of the CANDU type, exploring from its creation to studies directed to G-III and G-IV reactors. (author)

  9. CANDU: study and review

    International Nuclear Information System (INIS)

    Morad, César M.; Santos, Thiago A. dos

    2017-01-01

    The CANDU (Canadian Deuterium Uranium) is a nuclear reactor developed by AECL (Atomic Energy of Canada Limited). The first small-scale reactor is known as NPD and was made in 1955 and commenced operation in 1962. It is a pressurized heavy water reactor and uses D2O as moderator and coolant and therefore uses natural uranium as fuel. There have been two major types of CANDU reactors, the original design of around 500 MWe that was intended to be used in multi-reactor installations in large plants, and the rationalized CANDU6 which has units in Argentina, South Korea, Pakistan, Romania and China. Throughout the 1980s and 90s the nuclear power market suffered a major crash, with few new plants being constructed in North America or Europe. Design work continued through, however, and a number of new design concepts were introduced that dramatically improved safety, capital costs, economics and overall performance. These Generation III+ and Generation IV machines became a topic of considerable interest in the early 2000s as it appeared a nuclear renaissance was underway and large numbers of new reactors would be built over the next decade. The present work aims to study the reactors of the CANDU type, exploring from its creation to studies directed to G-III and G-IV reactors. (author)

  10. CANDU, building the future

    International Nuclear Information System (INIS)

    Stern, F.

    1997-01-01

    The CEO of Stern Laboratories delivered a speech on the problems and challenges facing the nuclear industry. The CANDU system is looked at as the practical choice for the future of our energy source. The people of the industry must be utilized and respected to deliver to the best of their ability

  11. CANDU market prospects

    International Nuclear Information System (INIS)

    Kakaria, B.K.

    1994-01-01

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

  12. CANDU, building the future

    Energy Technology Data Exchange (ETDEWEB)

    Stern, F. [Stern Laboratories (Canada)

    1997-07-01

    The CEO of Stern Laboratories delivered a speech on the problems and challenges facing the nuclear industry. The CANDU system is looked at as the practical choice for the future of our energy source. The people of the industry must be utilized and respected to deliver to the best of their ability.

  13. CANDU passive shutdown systems

    Energy Technology Data Exchange (ETDEWEB)

    Hart, R S; Olmstead, R A [AECL CANDU, Sheridan Park Research Community, Mississauga, ON (Canada)

    1996-12-01

    CANDU incorporates two diverse, passive shutdown systems, independent of each other and from the reactor regulating system. Both shutdown systems function in the low pressure, low temperature, moderator which surrounds the fuel channels. The shutdown systems are functionally different, physically separate, and passive since the driving force for SDS1 is gravity and the driving force for SDS2 is stored energy. The physics of the reactor core itself ensures a degree of passive safety in that the relatively long prompt neutron generation time inherent in the design of CANDU reactors tend to retard power excursions and reduces the speed required for shutdown action, even for large postulated reactivity increases. All passive systems include a number of active components or initiators. Hence, an important aspect of passive systems is the inclusion of fail safe (activated by active component failure) operation. The mechanisms that achieve the fail safe action should be passive. Consequently the passive performance of the CANDU shutdown systems extends beyond their basic modes of operation to include fail safe operation based on natural phenomenon or stored energy. For example, loss of power to the SDS1 clutches results in the drop of the shutdown rods by gravity, loss of power or instrument air to the injection valves of SDS2 results in valve opening via spring action, and rigorous self checking of logic, data and timing by the shutdown systems computers assures a fail safe reactor trip through the collapse of a fluctuating magnetic field or the discharge of a capacitor. Event statistics from operating CANDU stations indicate a significant decrease in protection system faults that could lead to loss of production and elimination of protection system faults that could lead to loss of protection. This paper provides a comprehensive description of the passive shutdown systems employed by CANDU. (author). 4 figs, 3 tabs.

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

    International Nuclear Information System (INIS)

    2005-01-01

    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

  15. Dimensional measurement of fresh fuel bundle for CANDU reactor

    International Nuclear Information System (INIS)

    Jo, Chang Keun; Cho, Moon Sung; Suk, Ho Chun; Koo, Dae Seo; Jun, Ji Su; Jung, Jong Yeob

    2005-01-01

    This report describes the results of the dimensional measurement of fresh fuel bundles for the CANDU reactor in order to estimate the integrity of fuel bundle in two-phase flow in the CANDU-6 fuel channel. The dimensional measurements of fuel bundles are performed by using the 'CANDU Fuel In-Bay Inspection and Dimensional Measurement System', which was developed by this project. The dimensional measurements are done from February 2004 to March 2004 in the CANDU fuel storage of KNFC for the 36 fresh fuel bundles, which are produced by KNFC and are waiting for the delivery to the Wolsong-3 plant. The detail items of dimensional measurements are included fuel rod and bearing pad profiles of the outer ring in fuel bundle, diameter of fuel bundle, bowing of fuel bundle, fuel rod length, and surface profile of end plate profile. The measurement data will be compared with those of the post-irradiated bundles cooled in Wolsong-3 NPP spent fuel pool by using the same bundles and In-Bay Measurement System. So, this analysis of data will be applied for the evaluation of fuel bundle integrity in two-phase flow of the CANDU-6 fuel channel

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

    International Nuclear Information System (INIS)

    Lau, J.H.

    1997-01-01

    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

  17. 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.

  18. Incentives for improvement of CANDU

    International Nuclear Information System (INIS)

    Hart, R.S.; Dunn, J.T.; Finlay, R.B.

    1988-12-01

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

  19. CANDU in the next century

    International Nuclear Information System (INIS)

    Meneley, D.A.; Torgerson, D.F.

    1997-01-01

    AECL's main product line is available today in two designs, designated as CANDU 6 and CANDU 9. Each of these is based on successfully operating pressurized-heavy-water nuclear plants. Several new CANDU stations are under construction or planned around the world. The author presents plant concepts which may evolve from today's products during the 21st century, indicating the particular development directions which might be followed by AECL product development depending on the future competitive environment, economics, and market circumstances. This study shows that the CANDU energy supply system is sufficiently flexible to be adapted into widely varying circumstances over the next century and beyond

  20. The CANDU 3 containment structure

    International Nuclear Information System (INIS)

    1994-01-01

    The design of the CANDU 3 nuclear power plant is being developed by AECL CANDU's Saskatchewan office. There are 24 CANDU nuclear power units operating in Canada and abroad and eight units are under construction is Romania and South Korea. The design of the CANDU 3 plant has evolved on the basis of the proven CANDU design. The experiences gained during construction, commissioning and operation of the existing CANDU plants are considered in the design. Many technological enhancements have been implemented in the design processes in all areas. The object has been to develop an improved reactor design that is suitable for the current and the future markets worldwide. Throughout the design phase of CANDU 3, emphasis has been placed in reducing the cost and construction schedule of the plant. This has been achieved by implementing design improvements and using new construction techniques. Appropriate changes and improvements to the design to suit new requirements are also adopted. In CANDU plants, the containment structure acts as an ultimate barrier against the leakage of radioactive substances during normal operations and postulated accident conditions. The concept of the structural design of the containment structure has been examined in considerable detail. This has resulted in development of a new conceptual design for the containment structure for CANDU 3. This paper deals with this new design of the containment structure

  1. The travesty of discarding used CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ottensmeyer, P. [Univ. of Toronto, Toronto, Ontario (Canada)

    2016-09-15

    The current plan worldwide for virtually all used nuclear fuels is costly deep burial to attempt to isolate their long-term radiotoxicity permanently. Alternatively Canada's 50,000 tons spent CANDU fuel, of which only 0.74% of the heavy atoms have been fissioned to extract their energy, could supply 130 times more non-carbon energy using proven economical recycling and fast-neutron technologies. The result in this country alone would currently be the creation of $74 trillion of reliable electricity on demand without greenhouse gas emissions. It would avoid adding 475 billion tons CO{sub 2} to the atmosphere compared to the use of coal, to mitigate climate change. Worldwide recycling of stored spent nuclear fuel and replenishing with depleted uranium in fast-neutron reactors could avoid emitting over 20 trillion tons CO{sub 2}, or over six times the current total atmospheric CO{sub 2} content. As added bonus the long-term radiotoxicity of the used CANDU fuel is effectively eliminated, making a long-term deep geological repository unnecessary. Even the shorter-lived radioisotope fission products become valuable stable atoms and minerals that would fetch $3 million per ton. Such an alternative is certainly worth pursuing. (author)

  2. The future for CANDU

    International Nuclear Information System (INIS)

    Foster, J.S.

    1977-06-01

    Canada could have 60,000 MW(e) of installed nuclear-electric generating capacity by the year 2000 and have exported the plan to generate a further 5,000 MW(e). While the CANDU reactor can readily be scaled up to larger unit sizes, its real potential lies in the even greater efficiency that can be obtained by using alternative fuel cycles. The thorium - uranium-233 fuel cycle, for instance, makes it possible to attain a conversion factor of unity, or a little better, on a feed of pure thorium in a substantially unmodified CANDU reactor. Further developments, such as spallation, offer means of converting fertile to fissile material to provide a fissile inventory for an expanding system. The coincidence of expected future shortages of other energy supplies with continuing good experience in the nuclear field should assist in creating a climate that will permit accelerated nuclear power development. (author)

  3. CANDU project development

    Energy Technology Data Exchange (ETDEWEB)

    Hedges, K R [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

    1996-12-31

    Advanced CANDU reactor design strategy follows an evolutionary approach, taking manageable steps in the development of power plants from today`s available designs, and in parallel carrying out longer-term studies to develop future-generation reactor concepts. The major emphasis is on safety, on on reducing cost and schedule. New features are developed and thoroughly proof-tested before introduction into designs, in order to maximize owner confidence. (author). 4 figs.

  4. CANDU project development

    International Nuclear Information System (INIS)

    Hedges, K.R.

    1995-01-01

    Advanced CANDU reactor design strategy follows an evolutionary approach, taking manageable steps in the development of power plants from today's available designs, and in parallel carrying out longer-term studies to develop future-generation reactor concepts. The major emphasis is on safety, on on reducing cost and schedule. New features are developed and thoroughly proof-tested before introduction into designs, in order to maximize owner confidence. (author). 4 figs

  5. CANDU fuel performance

    International Nuclear Information System (INIS)

    Ivanoff, N.V.; Bazeley, E.G.; Hastings, I.J.

    1982-01-01

    CANDU fuel has operated successfully in Ontario Hydro's power reactors since 1962. In the 19 years of experience, about 99.9% of all fuel bundles have performed as designed. Most defects occurred before 1979 and subsequent changes in fuel design, fuel management, reactor control, and manufacturing quality control have reduced the current defect rate to near zero. Loss of power production due to defective fuel has been negligible. The outstanding performance continues while maintaining a low unit energy cost for fuel

  6. Status of advanced technologies for CANDU reactors

    International Nuclear Information System (INIS)

    Lipsett, J.J.

    1989-01-01

    The future development of the CANDU reactor is a continuation of a successful series of reactors, the most recent of which are nine CANDU 6 Mk 1* units and four Darlington units. There are three projects underway that continue the development of the CANDU reactor. These new design projects flow from the original reactor designs and are a natural progression of the CANDU 6 Mk 1, two units of which are operating successfully in Canada, one each in Argentina and Korea, with five more being built in Rumania. These new design projects are known as: CANDU 6 Mk 2, an improved version of CANDU 6 Mk 1; CANDU 3, a small, advanced version of the CANDU 6 Mk 1; CANDU 6 Mk 3, a series of advanced CANDU reactors. A short description of modified versions of CANDU reactors is given in this paper. 5 figs

  7. Plutonium dispositioning in CANDU

    International Nuclear Information System (INIS)

    Boczar, P.G.; Feinroth, H.; Luxat, J.C.

    1995-07-01

    Recently, the U.S. Department of Energy (DOE) sponsored Atomic Energy of Canada Limited (AECL) to evaluate salient technical, strategic, schedule, and cost-related parameters of using CANDU reactors for dispositioning of weapons-grade plutonium in the form of Mixed OXide (MOX) fuel. A study team, consisting of key staff from the CANDU reactor designers and researchers (AECL), operators (Ontario Hydro) and fuel suppliers, analyzed all significant factors involved in such application, with the objective of identifying an arrangement that would permit the burning of MOX in CANDU at the earliest date. One of Ontario Hydro's multi-unit stations, Bruce A nuclear generating station (4x769 MW(e)), was chosen as the reference for the study. The assessment showed that no significant modifications of reactor or process systems are necessary to operate with a full MOX core. Plant modifications would be limited to fuel handling and modifications necessary to accommodate enhanced security and safeguards requirements. No safety limitations were identified

  8. The next generation CANDU 6

    International Nuclear Information System (INIS)

    Hopwood, J.M.

    1999-01-01

    AECL's product line of CANDU 6 and CANDU 9 nuclear power plants are adapted to respond to changing market conditions, experience feedback and technological development by a continuous improvement process of design evolution. The CANDU 6 Nuclear Power Plant design is a successful family of nuclear units, with the first four units entering service in 1983, and the most recent entering service this year. A further four CANDU 6 units are under construction. Starting in 1996, a focused forward-looking development program is under way at AECL to incorporate a series of individual improvements and integrate them into the CANDU 6, leading to the evolutionary development of the next-generation enhanced CANDU 6. The CANDU 6 improvements program includes all aspects of an NPP project, including engineering tools improvements, design for improved constructability, scheduling for faster, more streamlined commissioning, and improved operating performance. This enhanced CANDU 6 product will combine the benefits of design provenness (drawing on the more than 70 reactor-years experience of the seven operating CANDU 6 units), with the advantages of an evolutionary next-generation design. Features of the enhanced CANDU 6 design include: Advanced Human Machine Interface - built around the Advanced CANDU Control Centre; Advanced fuel design - using the newly demonstrated CANFLEX fuel bundle; Improved Efficiency based on improved utilization of waste heat; Streamlined System Design - including simplifications to improve performance and safety system reliability; Advanced Engineering Tools, -- featuring linked electronic databases from 3D CADDS, equipment specification and material management; Advanced Construction Techniques - based on open top equipment installation and the use of small skid mounted modules; Options defined for Passive Heat Sink capability and low-enrichment core optimization. (author)

  9. Reactors based on CANDU technology

    International Nuclear Information System (INIS)

    Bjegun, S.V.; Shirokov, S.V.

    2012-01-01

    The paper analyzes the use CANDU technology in world nuclear energy. Advantages and disadvantages in implementation of this technology are considered in terms of economic and technical aspects. Technological issues related to the use of CANDU reactors and nuclear safety issues are outlined. Risks from implementation of this reactor technology in nuclear energy of Ukraine are determined

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

    International Nuclear Information System (INIS)

    Seok, Ho Cheon; Shim, Ki Seop; Byeon, Taek Sang; Park, Kwang Seok; Kim, Bong Ki; Lee, Yeong Uk; Jeong, Chang Joon; Oh, Deok Joo; Lee, Ui Joo; Park, Joo Hwan; Lee, Sang Yong; Jeong, Beop Dong; Choi, Han Rim; Lee, Yeong Jin; Choi, Cheol Jin; Choi, Jong Ho; Lee, Kwang Won; Cho, Cheon Hyi; On, Myeong Ryong; Kim, Taek Mo; Lim, Hong Sik; Lee, Kang Moon; Lee, Nam Ho; Lee, Kyu Hyeong

    1994-07-01

    It has been projected that a total of 5 pressurized heavy water reactors (PHWR) including Wolsong 1 under operation and Wolsong 2, 3 and 4 under construction will be operated by 2006, and so about 500 ton of natural uranium will be consumed every year and a lot of spent fuels will be generated. Therefore, the ultimate goal of this R and D project is to develop the CANDU advanced fuel having the following capabilities compared with existing standard fuel: (1) To reduce linear heat generation rating by more than 15% (i.e., less than 50 kW/m), (2) To extend fuel burnup by more than 3 times (i.e., higher than 21,000 MWD/MTU), and (3) To increase critical channel power by more than 5%. In accordance, the followings are performed in this fiscal year: (1) Undertake CANFLEX-NU design and thermalmechanical performance analysis, and prepare design documents, (2) Establish reactor physics analysis code system, and investigate the compativility of the CANFLEX-NU fuel with the standard 37-element fuel in the CANDU-6 reactor. (3) Establish safety analysis methodology with the assumption of the CANFLEX-NU loaded CANDU-6 reactor, and perform the preliminary thermalhydraulic and fuel behavior for the selected DBA accidents, (4) Investigate reactor physics analysis code system as pre-study for CANFLEX-SEU loaded reactors

  11. CANDU 9 fuelling machine carriage

    Energy Technology Data Exchange (ETDEWEB)

    Ullrich, D J; Slavik, J F [Atomic Energy of Canada Ltd., Saskatoon, SK (Canada)

    1997-12-31

    Continuous, on-power refuelling is a key feature of all CANDU reactor designs and is essential to maintaining high station capacity factors. The concept of a fuelling machine carriage can be traced to the early CANDU designs, such as the Douglas Point Nuclear Generating Station. In the CANDU 9 480NU unit, the combination of a mobile carriage and a proven fuelling machine head design comprises an effective means of transporting fuel between the reactor and the fuel transfer ports. It is a suitable alternative to the fuelling machine bridge system that has been utilized in the CANDU 6 reactor units. The CANDU 9 480NU fuel handling system successfully combines features that meet the project requirements with respect to fuelling performance, functionality, seismic qualification and the use of proven components. The design incorporates improvements based on experience and applicable current technologies. (author). 4 figs.

  12. CANDU 9 fuelling machine carriage

    International Nuclear Information System (INIS)

    Ullrich, D.J.; Slavik, J.F.

    1996-01-01

    Continuous, on-power refuelling is a key feature of all CANDU reactor designs and is essential to maintaining high station capacity factors. The concept of a fuelling machine carriage can be traced to the early CANDU designs, such as the Douglas Point Nuclear Generating Station. In the CANDU 9 480NU unit, the combination of a mobile carriage and a proven fuelling machine head design comprises an effective means of transporting fuel between the reactor and the fuel transfer ports. It is a suitable alternative to the fuelling machine bridge system that has been utilized in the CANDU 6 reactor units. The CANDU 9 480NU fuel handling system successfully combines features that meet the project requirements with respect to fuelling performance, functionality, seismic qualification and the use of proven components. The design incorporates improvements based on experience and applicable current technologies. (author). 4 figs

  13. CANDU operating experience

    International Nuclear Information System (INIS)

    McConnell, L.G.; Woodhead, L.W.; Fanjoy, G.R.; Thurygill, E.W.

    1980-05-01

    The CANDU-PHW program is based upon 38 years of heavy water reactor experience with 35 years of operating experience. Canada has had 72 reactor years of nuclear-electric operations experience with 10 nuclear units in 4 generating stations during a period of 18 years. All objectives have been met with outstanding performance: worker safety, public safety, environmental emissions, reliable electricity production, and low electricity cost. The achievement has been realized through total teamwork involving all scientific disciplines and all project functions (research, design, manufacturing, construction, and operation). (auth)

  14. The CANDU 6

    International Nuclear Information System (INIS)

    Hopwood, J.M.; Hum, J.

    1999-01-01

    The CANDU 6 is a modem nuclear power plant designed and developed under the aegis of Atomic Energy of Canada, Limited (AECL) for domestic use and for export to other countries. This design has successfully met criteria for operation and redundant safety features required by Canada and by the International Atomic Energy Agency (IAEA) and has an estimable record of performance in all applications to date. Key to this success is a defined program of design enhancement in which changes are made while retaining fundamental features proven by operating experience. Basic design features and progress toward improvements are presented here. (author)

  15. Development of the ENVI simulator to estimate Korean SNF flow and its cost - 16060

    International Nuclear Information System (INIS)

    Hwang, Yongsoo; Miller, Ian

    2009-01-01

    This paper describes an integrated model developed by the Korean Atomic Energy Research Institute (KAERI) to simulate options for managing spent nuclear fuel (SNF) in South Korea. A companion paper (Hwang and Miller, 2009) describes a performance assessment model to address the long-term safety of alternative geological disposal options for different waste streams. The model addresses alternative concepts for storage, transportation, and processing of SNF of different types (Candu, PWR), leading up to permanent disposal in geological repositories. It uses the GoldSim software to simulate the logistics of the associated activities, including the associated capital and operating costs. The model's results allow direct comparison of alternative waste management concepts, and predict the sizes and timings of different facilities required. Future versions of the model will also address the uncertainties associated with the different system components in order to provide risk-based assessments. (authors)

  16. Steps to Advanced CANDU 600

    International Nuclear Information System (INIS)

    Oh, Yongshick; Brooks, G. L.

    1988-01-01

    The CANDU nuclear power system was developed from merging of AECL heavy water reactor technology with Ontario Hydro electrical power station expertise. The original four units of Ontario Hydro's Pickering Generating Station are the first full-scale commercial application of the CANDU system. AECL and Ontario Hydro then moved to the next evolutionary step, a more advanced larger scale design for four units at the Bruce Generating Station. CANDU 600 followed as a single unit nuclear electric power station design derived from an amalgam of features of the multiple unit Pickering and Bruce designs. The design of the CANDU 600 nuclear steam supply system is based on the Pickering design with improvements derived from the Bruce design. For example, most CANDU 600 auxiliary systems are based on Bruce systems, whereas the fuel handling system is based on the Pickering system. Four CANDU 600 units are in operation, and five are under construction in Romania. For the additional four units at Pickering Generating Station 'B', Ontario Hydro selected a replica of the Pickering 'A' design with limited design changes to maintain a high level of standardization across all eight units. Ontario Hydro applied a similar policy for the additional four units at Bruce Generating Station 'B'. For the four unit Darlington station, Ontario Hydro selected a design based on Bruce with improvements derived from operating experience, the CANDU 600 design and development programs

  17. CANDU channel flow verification

    International Nuclear Information System (INIS)

    Mazalu, N.; Negut, Gh.

    1997-01-01

    The purpose of this evaluation was to obtain accurate information on each channel flow that enables us to assess precisely the level of reactor thermal power and, for reasons of safety, to establish which channel is boiling. In order to assess the channel flow parameters, computer simulations were done with the NUCIRC code and the results were checked by measurements. The complete channel flow measurements were made in the zero power cold condition. In hot conditions there were made flow measurements using the Shut Down System 1 (SDS 1) flow devices from 0.1 % F.P. up to 100 % F.P. The NUCIRC prediction for CANDU channel flows and the measurements by Ultrasonic Flow Meter at zero power cold conditions and SDS 1 flow channel measurements at different reactor power levels showed an acceptable agreement. The 100 % F.P. average errors for channel flow of R, shows that suitable NUCIRC flow assessment can be made. So, it can be done a fair prediction of the reactor power distribution. NUCIRC can predict accurately the onset of boiling and helps to warn at the possible power instabilities at high powers or it can detect the flow blockages. The thermal hydraulic analyst has in NUCIRC a suitable tool to do accurate predictions for the thermal hydraulic parameters for different steady state power levels which subsequently leads to an optimal CANDU reactor operation. (authors)

  18. CANDU-BLW-250

    Energy Technology Data Exchange (ETDEWEB)

    Pon, G. A. [Atomic Energy of Canada Ltd, Sheridan Park, ON (Canada)

    1968-04-15

    The plant ''La Centrale nucleaire de Gentiliy'' is located between Montreal and Quebec City on the south shore of the St. Lawrence River. Startup is scheduled for 1971. A CANDU-BLW reactor is the nuclear steam generator. This reactor utilizes a heavy-water moderator, natural uranium oxide fuel, and a boiling light-water coolant. To be economic, this type of plant must have a minimum light-water inventory in the reactor core. A minimum inventory is obtained (a) by reducing the cross-sectional area for coolant flow to a minimum, and (b) by operating at a low coolant density. In CANDU-BLW-250, this is accomplished by operating a closed spaced fuel rod bundle at high steam quality. These features and others in the BLW concept lead to a number of areas of concern and they are summarized below: (i) Heat transfer. It is intended that under normal operating conditions the fuel sheaths will always be wetted with coolant. Some experiments and backup calculations are presented to support this specification. (ii) Hydrodynamic stability. Experiments and analysis indicate that the plant has a considerable over-power capacity before instability is predicted. (iii) Control. This plant does have a positive power coefficient and the transient performance with various disturbances is detailed. (iv) Safety. The positive power coefficient leads to concern over the loss of coolant accident. The results of some accident analyses are presented. (author)

  19. CANDU-BLW-250

    Energy Technology Data Exchange (ETDEWEB)

    Pon, G A

    1967-09-15

    The plant 'La Centrale nucleaire de Gentilly' is located between Montreal and Quebec City on the south shore of the St. Lawrence River and start-up is scheduled for 1971. A CANDU-BLW reactor is the nuclear steam generator. his reactor utilizes a heavy water moderator, natural uranium oxide fuel, and a boiling light water coolant. To be economic, this type of plant must have a minimum light water inventory in the reactor core. A minimum inventory is obtained (a) by reducing the cross-sectional area for coolant flow to a minimum, and (b) by operating at a low-coolant density. In CANDU-BLW-250, this is accomplished by operating a closed spaced fuel rod bundle at high steam quality. These features and others in the BLW concept lead to a number of areas of concern and they are summarized below: (1) Heat Transfer: It is intended that under normal operating conditions the fuel sheaths will always be wetted with coolant. (ii) Hydrodynamic Stability: Experiments and analysis indicate that the plant has a considerable over-power capacity before instability is predicted. (iii) Control: This plant does have a positive power coefficient and the transient performance with various disturbances are detailed. (iv) Safety: The positive power coefficient leads to concern over the loss of coolant accident. The results of some accident analysis are presented. (author)

  20. A study for good regulation of the CANDU's in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Se Ki; Shin, Y. K.; Joe, S. K.; Kim, J. S.; Yu, Y. J.; Lee, Y. J. [Ajou Univ., Suwon (Korea, Republic of)

    2002-03-15

    The objective of project is to derive the policy recommendations to improve the efficiency of CANDU plants regulation. These policy recommendations will eventually contribute to the upgrading of Korean nuclear regulatory system and safety enhancement. During the second phase of this 2 years study, following research activities were done. Review the technical basis and framework of the new Canadian Regulation System and IAEA. Analysis on the interview of Wolsung operation staffs to identify important safety issues and regulation problems experienced at operation. Providing a plan of CANDU regulation system enhancement program.

  1. Cernavoda CANDU severe accident evaluation

    International Nuclear Information System (INIS)

    Negut, G.; Marin, A.

    1997-01-01

    The papers present the activities dedicated to Romania Cernavoda Nuclear Power Plant first CANDU Unit severe accident evaluation. This activity is part of more general PSA assessment activities. CANDU specific safety features are calandria moderator and calandria vault water capabilities to remove the residual heat in the case of severe accidents, when the conventional heat sinks are no more available. Severe accidents evaluation, that is a deterministic thermal hydraulic analysis, assesses the accidents progression and gives the milestones when important events take place. This kind of assessment is important to evaluate to recovery time for the reactor operators that can lead to the accident mitigation. The Cernavoda CANDU unit is modeled for the of all heat sinks accident and results compared with the AECL CANDU 600 assessment. (orig.)

  2. Achieving CANDU excellence through collaboration

    Energy Technology Data Exchange (ETDEWEB)

    Dermarkar, F. [CANDU Owners Group Inc., Toronto, Ontario (Canada)

    2015-07-01

    All Operators of CANDU/PHWR Worldwide, and AECL, are members of Candu Owners Group (COG). COG has evolved to become primarily an Operators Owners Group with annual turnover of $75M. It is all about value to the members providing a diverse offering of services to meet a broad spectrum of member needs, linking our members together prioritizing and organizing to enable members to access what they need, when they need it. Collaboration benefits both COG and EPRI.

  3. Passive heat removal in CANDU

    International Nuclear Information System (INIS)

    Hart, R.S.

    1997-01-01

    CANDU has a tradition of incorporating passive systems and passive components whenever they are shown to offer performance that is equal to or better than that of active systems, and to be economic. Examples include the two independent shutdown systems that employ gravity and stored energy respectively, the dousing subsystem of the CANDU 6 containment system, and the ability of the moderator to cool the fuel in the event that all coolant is lost from the fuel channels. CANDU 9 continues this tradition, incorporating a reserve water system (RWS) that increases the inventory of water in the reactor building and profiles a passive source of makeup water and/or heat sinks to various key process systems. The key component of the CANDU 9 reserve water system is a large (2500 cubic metres) water tank located at a high elevation in the reactor building. The reserve water system, while incorporating the recovery system functions, and the non-dousing functions of the dousing tank in CANDU 6, embraces other key systems to significantly extend the passive makeup/heat sink capability. The capabilities of the reserve water system include makeup to the steam generators secondary side if all other sources of water are lost; makeup to the heat transport system in the event of a leak in excess of the D 2 O makeup system capability; makeup to the moderator in the event of a moderator leak when the moderator heat sink is required; makeup to the emergency core cooling (ECC) system to assure NPSH to the ECC pumps during a loss of coolant accident (LOCA), and provision of a passive heat sink for the shield cooling system. Other passive designs are now being developed by AECL. These will be incorporated in future CANDU plants when their performance has been fully proven. This paper reviews the passive heat removal systems and features of current CANDU plants and the CANDU 9, and briefly reviews some of the passive heat removal concepts now being developed. (author)

  4. Achieving CANDU excellence through collaboration

    International Nuclear Information System (INIS)

    Dermarkar, F.

    2015-01-01

    All Operators of CANDU/PHWR Worldwide, and AECL, are members of Candu Owners Group (COG). COG has evolved to become primarily an Operators Owners Group with annual turnover of $75M. It is all about value to the members providing a diverse offering of services to meet a broad spectrum of member needs, linking our members together prioritizing and organizing to enable members to access what they need, when they need it. Collaboration benefits both COG and EPRI.

  5. Technology transfer in CANDU marketing

    International Nuclear Information System (INIS)

    Pon, G.A.

    1982-06-01

    The author discusses how the CANDU system lends itself to technology transfer, the scope of CANDU technology transfer, and the benefits and problems associated with technology transfer. The establishment of joint ventures between supplier and client nations offers benefits to both parties. Canada can offer varying technology transfer packages, each tailored to a client nation's needs and capabilities. Such a package could include all the hardware and software necessary to develop a self-sufficient nuclear infrastructure in the client nation

  6. The CANDU 9 fuel transfer system

    International Nuclear Information System (INIS)

    Keszthelyi, Z.G.; Morikawa, D.T.

    1996-01-01

    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

  7. The CANDU 9 fuel transfer system

    Energy Technology Data Exchange (ETDEWEB)

    Keszthelyi, Z G [Canadian General Electric Co. Ltd., Peterborough, ON (Canada); Morikawa, D T [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

    1997-12-31

    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.

  8. CANDU safety under severe accidents

    International Nuclear Information System (INIS)

    Snell, V.G.; Howieson, J.Q.; Frescura, G.M.; King, F.; Rogers, J.T.; Tamm, H.

    1988-01-01

    The characteristics of the CANDU reactor relevant to severe accidents are set first by the inherent properties of the design, and second by the Canadian safety/licensing approach. Probabilistic safety assessment studies have been performed on operating CANDU plants, and on the 4 x 880 MW(e) Darlington station now under construction; furthermore a scoping risk assessment has been done for a CANDU 600 plant. They indicate that the summed severe core damage frequency is of the order of 5 x 10 -6 /year. CANDU nuclear plant designers and owner/operators share information and operational experience nationally and internationally through the CANDU Owners' Group (COG). The research program generally emphasizes the unique aspects of the CANDU concept, such as heat removal through the moderator, but it has also contributed significantly to areas generic to most power reactors such as hydrogen combustion, containment failure modes, fission product chemistry, and high temperature fuel behaviour. Abnormal plant operating procedures are aimed at first using event-specific emergency operating procedures, in cases where the event can be diagnosed. If this is not possible, generic procedures are followed to control Critical Safety Parameters and manage the accident. Similarly, the on-site contingency plans include a generic plan covering overall plant response strategy, and a specific plan covering each category of contingency

  9. The CANDU experience in Romania

    International Nuclear Information System (INIS)

    Smith, A.I.

    1984-01-01

    The CANDU program in Romania is now well established. The Cernavoda Nuclear Station presently under construction will consist of 5-CANDU 600 MWE Units and another similar size station is planned to be in operation in the next decade. Progress on the multi-unit station at Cernavoda was stalled for 18 months in 1982/83 as the Canadian Export Development Corporation had suspended their loan disbursements while the Romanian National debt was being rescheduled. Since resumption of the financing in August 1983 contracts worth almost 200M dollars have been placed with Canadian Companies for the supply of major equipment for the first two units. The Canadian design is that which was used in the latest 600 MWE CANDU station at Wolsong, Korea. The vast construction site is now well developed with the cooling water systems/channels and service buildings at an advanced stage of completion. The perimeter walls of the first two reactor buildings are already complete and slip-forming for the 3rd Unit is imminent. Many Romanian organizations are involved in the infrastructure which has been established to handle the design, manufacture, construction and operation of the CANDU stations. The Romanian manufacturing industry has made extensive preparations for the supply of CANDU equipment and components, and although a major portion of the first two units will come from Canada their intentions are to become largely self-supporting for the ensuing CANDU program. Quality assurance programs have been prepared already for many of the facilities

  10. Non-electrical CANDU applications

    International Nuclear Information System (INIS)

    Hopwood, Jerry; Kuran, Sermet; Zhou, Xi; Ivanco, Michael; Rolfe, Brian; Mancuso, Connie; Duffey, Romney

    2005-01-01

    AECL has performed studies to utilize CANDU nuclear energy in areas other than electrical generation. The studies presented in this paper include using CANDU for applications in non-traditional areas which expand the use of zero-greenhouse gas energy source. The Oil sands industry demands significant energy input and the majority of the energy required for bitumen extraction is steam and hot water. As the primary production of a CANDU plant is steam, it can satisfy the steam and hot water requirement without a major modification to the Nuclear Steam Plant (NSP). Reverse Osmosis (RO) has been identified by the IAEA as the most promising method for nuclear desalination. Since the RO desalination efficiency increases as its feedwater temperature rises, using condenser cooling water from a CANDU plant as the feedwater for a RO plant and sharing other facilities between these two plants results in significant benefits in capital and operating costs of a desalination plant. Electrolysis powered by nuclear-generated electricity is the technology currently available to produce hydrogen without greenhouse gas emissions. By using the cheaper electricity available at off-peak periods in an open electricity market, this technology could be economically competitive, improve overall energy system efficiency and reduce overall energy system carbon intensity. The paper summarizes the background, technical approaches, feasibility considerations, along with economic comparisons between CANDU nuclear energy and the traditional energy sources for each study. The results show that the CANDU technology is a promising energy source for various industries. (author)

  11. CANDU severe accident analysis

    International Nuclear Information System (INIS)

    Negut, Gheorghe; Catana, Alexandru; Prisecaru, Ilie; Dupleac, Daniel

    2007-01-01

    Romania is a EU member since January first 2007. This country faces now new challenges which imply also the nuclear power reactors now in operation. Romania operates since 1996 a CANDU nuclear power reactor and soon will start up a second unit. In EU PWR reactors are mostly operated, so that the Romania's reactors have to meet EU standards. Safety analysis guidelines require to model severe accidents for reactors of this type. Starting from previous studies a thermal-hydraulic model for a degraded CANDU core was developed. The initiating event is assumed to be a LOCA with simultaneous loss of moderator and coolant and the failure of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperatures inside a pressure tube reaches 1000 deg. C, a contact between pressure tube and calandria tube occurs and the decay heat is transferred to the moderator. Due to the lack of cooling, the moderator eventually begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) uncover, then disintegrate and fall down to the calandria vessel bottom. All the quantity of calandria moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield water tank surrounding the calandria vessel. The thermal hydraulics phenomena described above are modeled, analyzed and compared with the existing data. (authors)

  12. The Key-Role of shielding analysis in advanced Candu Fuel bundles nuclear safety improvement for some accidental criticality scenarios

    International Nuclear Information System (INIS)

    Margeanu, C.A.; Rizoiu, A.; Olteanu, G.

    2008-01-01

    The paper aims to present the source term and photon dose rates estimation for advanced Candu fuel bundles in some accidental criticality scenarios. As reference, the Candu standard fuel bundle has been used. The scenarios take into account for a very short-time irradiated or spent fuel bundles for some configurations closed to criticality. In order to estimate irradiated fuel characteristic parameters and radiation doses, the ORNL's SCALE 5 codes Origin-S and Monte Carlo MORSE-SGC have been used. The paper includes the irradiated fuel characteristic parameters comparison for the considered Candu fuel bundles, providing also a comparison between the corresponding radiation doses

  13. A proposal for an international program to develop dry recycle of spent nuclear fuel

    International Nuclear Information System (INIS)

    Feinroth, H.

    1999-01-01

    The dry oxidation-reduction process (called OREOX for Oxidation Reduction of Oxide Fuel) being developed by Korea and Canada, in cooperation with IAEA and the US State Department, is limited to recycle of spent LWR fuel into CANDU reactors (DUPIC). When first conceived and demonstrated via irradiation of test elements by Atomics International in 1965, (the process was called AIROX at that time) a wider range of applications was intended, including recycle of spent LWR fuel into LWRs. Studies sponsored by DOE's Idaho Office in 1992 confirmed the applicability of this technology to regions containing LWR's only, and described the potential advantages of such recycle from an environmental, waste management and economic point of view, as compared to the direct disposal option. Recent analyses conducted by the author indicates that such dry recycle may be one of the few acceptable paths remaining for resolution of the US spent fuel storage dilemma that remains consistent with US non-proliferation policy. It is proposed that a new US program be established to develop AIROX dry recycle for use in the US, and this become part of an international cooperative program, including the current Canadian - Korean program, and possibly including participation of other countries wishing to pursue alternatives to the once through cycle, and wet reprocessing. With shared funding of major project elements, such international cooperation would accelerate the demonstration and commercial deployment of dry recycle technology, as compared to separate and independent programs in each country. (author)

  14. Extending the Candu Nuclear Reactor Concept: The Multi-Spectrum Nuclear Reactor

    International Nuclear Information System (INIS)

    Allen, Francis; Bonin, Hugues

    2008-01-01

    The aim of this work is to examine the multi-spectrum nuclear reactor concept as an alternative to fast reactors and accelerator-driven systems for breeding fissile material and reducing the radiotoxicity of spent nuclear fuel. The design characteristics of the CANDU TM nuclear power reactor are shown to provide a basis for a novel approach to this concept. (authors)

  15. Extending the Candu Nuclear Reactor Concept: The Multi-Spectrum Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Allen, Francis [Director General Nuclear Safety, 280 Slater St, Ottawa, K1A OK2 (Canada); Bonin, Hugues [Royal Military College of Canada, 11 General Crerar Cres, Kingston, K7K 7B4 (Canada)

    2008-07-01

    The aim of this work is to examine the multi-spectrum nuclear reactor concept as an alternative to fast reactors and accelerator-driven systems for breeding fissile material and reducing the radiotoxicity of spent nuclear fuel. The design characteristics of the CANDU{sup TM} nuclear power reactor are shown to provide a basis for a novel approach to this concept. (authors)

  16. Quality Products - The CANDU Approach

    International Nuclear Information System (INIS)

    Ingolfsrud, L. John

    1989-01-01

    The prime focus of the CANDU concept (natural uranium fuelled-heavy water moderated reactor) from the beginning has economy, heavy water losses and radiation exposures also were strong incentives for ensuring good design and reliable equipment. It was necessary to depart from previously accepted commercial standards and to adopt those now accepted in industries providing quality products. Also, through feedback from operating experience and specific design and development programs to eliminate problems and improve performance, CANDU has evolved into today's successful product and one from which future products will readily evolve. Many lessons have been learned along the way. On the one hand, short cuts of failures to understand basic requirements have been costly. On the other hand, sound engineering and quality equipment have yielded impressive economic advantages through superior performance and the avoidance of failures and their consequential costs. The achievement of lifetime economical performance demands quality products, good operation and good maintenance. This paper describes some of the basic approaches leading to high CANDU station reliability and overall excellent performance, particularly where difficulties have had to be overcome. Specific improvements in CANDU design and in such CANDU equipment as heat transport pumps, steam generators, valves, the reactor, fuelling machines and station computers, are described. The need for close collaboration among designers, nuclear laboratories, constructors, operators and industry is discussed. This paper has reviewed some of the key components in the CANDU system as a means of indicating the overall effort that is required to provide good designs and highly reliable equipment. This has required a significant investment in people and funding which has handsomely paid off in the excellent performance of CANDU stations. The close collaboration between Atomic Energy of Canada Limited, Canadian industry and the

  17. Evolutionary CANDU 9 plant improvements

    International Nuclear Information System (INIS)

    Yu, S.K.W.

    1999-01-01

    The CANDU 9 is a 935 MW(e) nuclear power plant (NPP) based on the multi-unit Darlington and Bruce B designs with additional enhancements from our ongoing engineering and research programs. Added to the advantages of using proven systems and components, CANDU 9 offers improvement features with enhanced safety, improved operability and maintenance including a control centre with advanced man-machine interface, and improved project delivery in both engineering and construction. The CANDU 9 NPP design incorporated safety enhancements through careful attention to emerging licensing and safety issues. The designers assessed, revised and evolved such systems as the moderator, end shield, containment and emergency core cooling (ECC) systems while providing an integrated final design that is more passive and severe-accident-immune. AECL uses a feedback process to incorporate lessons learned from operating plants, from current projects experiences and from the implementation or construction phase of previous projects. Most of the requirements for design improvements are based on a systematic review of current operating CANDU stations in the areas of design and reliability, operability, and maintainability. The CANDU 9 Control Centre provides plant staff with improved operability and maintainability capabilities due to the combination of systematic design with human factors engineering and enhanced operating and diagnostics features. The use of advanced engineering tools and modem construction methods will reduce project implementation risk on project costs and schedules. (author)

  18. Quantitative and radiological assessment of PYRO-SFR closed fuel cycle against direct disposal of spent nuclear fuel from Korean nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Khan, Muhammad Minhaj; Lee, Suhong; Cheong, Jaehak; Whang, Jooho [Kyunghee University, Seoul (Korea, Republic of)

    2016-12-15

    Calculations have been made to estimate the generation of SNF including Plutonium (Pu) and minor actinides (MAs) from PWR and PHWR NPPs by using the IAEA code Nuclear Fuel Cycle Simulation System (NFCSS) for the period of 2016, 2030 as 12 more NPPs will be added by the year 2029. In order to find the optimize option for SNF management through sustainable use of nuclear energy system and reduce the associated radiological risk, 4 scenarios to burnout the Pu and MAs are analyzed. Estimation for the amount of SNF including major radionuclides has made for the year of 2016, 2030, 2089 and 2109 with an approximate amount of 148,19.65MT, 248,34.6 MT, 41572.23 MT and 61272.24 MT of SNF will be generated respectively. Radioactivity and radiotoxicity is calculated in order to access the radiological risk in terms of ingestion. Four Transmutation strategies is predicted to make Korean NPPs sustainable by incorporating KALIMER-600 (Burner) reactor in Korean nuclear fleet with an approximate transmutation rate of 99.77%, 96.57%, 95.34% and 97.58% for MA only (scenario-1), (MA & Pu) scenario-2, scenario-3 (MA only till the year 2125) and scenario-4 MA & Pu till the year 2148) respectively.

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

    International Nuclear Information System (INIS)

    Margeanu, C.A.; Rizoiu, A.C.

    2009-01-01

    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 ThO 2 -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

  20. Thermosyphoning in the CANDU reactor

    International Nuclear Information System (INIS)

    Spinks, N.J.; Wright, A.C.D.; Caplan, M.Z.; Prawirosoehardjo, S.; Gulshani, P.

    1984-01-01

    Thermosyphoning is defined as the natural convective flow of primary coolant over the boilers. It is the predicted mode of heat transport from core to boilers in many postulated scenarios for CANDU reactor safety analysis. The scenarios encompass a wide range of boundary conditions in reactor power, secondary temperature and primary coolant inventory. Loss of pumping of the primary coolant is a common feature. Thermosyphoning is single or two-phase depending on the boundary conditions. The paper describes the important thermohydraulic characteristics of thermosyphoning in CANDU reactors with emphasis on two-phase thermosyphoning. It utilizes predictions of a transient thermohydraulics computer code and describes experiments done for the purpose of verifying these predictions. Predictions are compared with single-phase thermosyphoning tests done during commissioning of the Gentilly-2 and Point Lepreau CANDU 600 reactors. (orig.)

  1. Delivery improvements for CANDU projects

    International Nuclear Information System (INIS)

    Stephen Yu; Ken Hedges

    1998-01-01

    Future CANDU design will continue to meet emerging design and performance requirements as expected by the operating utilities, and will integrate new technologies into both the product features and work processes. Elements of this risk reduction strategy include feedback of lessons learned from operating plants, project experiences from previous projects, and replication of successful systems and equipment. Project implementation risk is minimized by up-front engineering and licensing prior to contract start. Enhanced competitiveness of the CANDU products is ensured by incorporating improvements based on updated technology. This paper summarizes the strategy used to enhance competitiveness of the CANDU products and the measures introduced to minimize risk during project implementation. This strategy provides a balance between innovation and proven designs; and between the desire for safety and operational improvements and the cost to achieve the improvements

  2. Dynamic analysis of Korean nuclear fuel cycle with fast reactor systems

    International Nuclear Information System (INIS)

    Jeong, Chang Joon

    2004-12-01

    The Korean nuclear fuel cycle scenario was analyzed by the dynamic analysis method, including Pressurized Water Reactor (PWR), Canadian Deuterium Uranium (CANDU) and fast reactor systems. 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 1%. After setting up the once-through fuel cycle model, the Korea Advanced Liquid Metal Reactor (KALIMER) scenario was modeled to investigate the fuel cycle parameters. For the analysis of the fast reactor fuel cycle, both KAILMER-150 and KALIMER-600 reactors were considered. In this analysis, the spent fuel inventory as well as the amount of plutonium, Minor Actinides (MA) and Fission Products (FP) of the recycling fuel cycle was estimated and compared to that of the once-through fuel cycle. Results of the once-through fuel cycle calculation showed that the demand grows up to 64 GWe and total amount of spent fuel would be ∼102 kt in 2100. If the KALIMER scenario is implemented, the total spent fuel inventory can be reduced by ∼80%. However it was found that the KALIMER scenario does not contribute to reduce the amount of MA and FP, which is important when designing a repository. For the further destruction of MA, an actinide burner can be considered in the future nuclear fuel cycle

  3. Build your own Candu reactor

    International Nuclear Information System (INIS)

    Carruthers, J.

    1979-01-01

    The author discusses the marketing of Candu reactors, particularly the export trade. Future sales will probably be of the nuclear side of a station only, thus striking a compromise between licensing and 'turnkey' sales. It is suggested that AECL might have made more money in the past had it not given the right to manufacture Candu fuel away to Canadian industry. Future sales to certain potential customers may be limited by the requirement of strict safeguards, which will almost certainly never be relaxed. (N.D.H.)

  4. Material control and accounting at a CANDU reactor: the instrumented safeguards scheme

    International Nuclear Information System (INIS)

    Stirling, A.J.; Payne, E.

    1985-01-01

    While CANDU reactors differ from LWRs quite markedly in the way they operate, the principles of materials accounting and safeguards are equally applicable. Indeed, since CANDU fuel is not reprocessed, the relatively simple procedure of item accounting is sufficient for CANDUs. However, on-power refueling means that automatic item counting is needed to independently confirm operator records. Surveillance and sealing techniques for spent fuel are needed for a practical system. The equipment developed has allowed the IAEA to apply safeguards at reasonable cost and with minimal interference to the utility operating the station

  5. Improved management of SG BD demineralizer for reduced generation of low-level radioactive spent resin in Korean nuclear power plants

    International Nuclear Information System (INIS)

    Rhee, I.; Cho, D.; Yeon, J.

    2003-01-01

    Most nuclear power plants in Korea have adopted Ethanolamine(ETA) as a secondary pH control agent to increase the pH at the liquid phase, which may reduce the corrosion in steam generator tubes and moisture separator/reheat system. Along with its beneficial effect of SG protection from corrosion and degradation, the replacement of ammonia with ETA causes the increased generation of spent resin and the reduced run time of demineralizer in steam generator blowdown(SG BD) system. The composition ratio of cation- to anion- exchange resin in SG BD mixed bed should be increased in the ETA chemistry environment to meet the ratio of cation to anion in the aqueous solution, which results in the simultaneous exhaustion of cation and anion exchange resins. The utilization rate of mixed bed is greatest at the cation-to-anion ratio of 95:1 on the theoretical equivalent basis in the solution, but practically highest at that of 22:1 due to the possible inhomogeneous distribution of cation and anion exchange resins in SG BD bed. The run time of the bed could be extended by 30% such that, at that much, the purchase cost of new resin is saved and the production rate of spent resin is reduced. The guideline on the replacement of resin in SG BD bed is not necessary to secure the removal of radioactive particles without the leakage of the primary coolant into the secondary side since all the radioactive ions can be eliminated by SG BD bed with the sufficient time. They are retained during more than one month after their ingress into the SG BD bed without leakage. With the reduced replacement, thus, the SG BD spent resin that comprises 65% of low-level radioactive solid waste can be much cut down

  6. Assessment of DUPIC fuel compatibility with CANDU-6

    Energy Technology Data Exchange (ETDEWEB)

    Choi, H B; Roh, G H; Jeong, C J; Rhee, B W; Choi, J W [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-07-01

    The compatibility of DUPIC fuel with the existing CANDU reactor was assessed. The technical issues of DUPIC fuel compatibility were chosen based on the CANDU physics design requirements and inherent characteristics of DUPIC fuel. The compatibility was assessed for the reference DUPIC fuel composition which was determined to reduce the composition heterogeneity and improve the spent PWR fuel utilization. Preliminary studies on a CANDU core loaded with DUPIC fuel have shown that the nominal power distribution is flatter than that of a natural uranium core when a 2-bundle shift refueling scheme is used, which reduces the reactivity worths of devices in the core and, therefore, the performance of reactivity devices was assessed. The safety of the core was assessed by a LOCA simulation and it was found that the power pulse upon LOCA can be maintained below that in the natural uranium core when a poison material is used in the DUPIC fuel. For the feasibility of handling DUPIC fuel in the plant, it will be necessary to introduce new equipment to load the DUPIC fuel in the refueling magazine. The radiation effect of DUPIC fuel on both the reactor hardware and the environment will require a quantitative analysis later. (author).

  7. Preliminary assessment on compatibility of DUPIC fuel with CANDU-6

    International Nuclear Information System (INIS)

    Choi, Hang-Bok; Roh, G.H.; Jeong, C.J.; Rhee, B.W.; Choi, J.W.; Boss, C.R.

    1997-01-01

    The compatibility of DUPIC fuel with the existing CANDU-6 reactor was assessed. The technical issues of DUPIC fuel compatibility were chosen based on the CANDU physics design requirements and inherent characteristics of DUPIC fuel. The compatibility was assessed for the reference DUPIC fuel composition which was determined to reduce the composition heterogeneity and improve the spent PWR fuel utilization. Preliminary studies on a CANDU core loaded with DUPIC fuel have shown that the nominal power distribution is flatter than that of a natural uranium core when a 2-bundle shift refueling scheme is used, which reduces the reactivity worths of devices in the core and, therefore, the performance of reactivity devices was assessed. The safety of the core was assessed by a LOCA simulation and it was found that the power pulse upon LOCA can be maintained below that in the natural uranium core when a poison material is used in the DUPIC fuel. For the feasibility of handling DUPIC fuel in the plant, it will be necessary to introduce new equipment to load the DUPIC fuel in the refueling magazine. The radiation effect of DUPIC fuel on both the reactor hardware and the environment will be qualitatively analyzed later. (author)

  8. CANDU 9 Control Centre Mockup

    International Nuclear Information System (INIS)

    Webster, A.; Macbeth, M.J.

    1996-01-01

    This paper provides a summary of the design process being followed, the benefits of applying a systematic design using human factors engineering, presents an overview of the CANDU 9 control centre mockup facility, illustrates the control centre mockup with photographs of the 3D CADD model and the full scale mockup, and provides an update on the current status of the project. (author)

  9. Spent fuel workshop'2002

    International Nuclear Information System (INIS)

    Poinssot, Ch.

    2002-01-01

    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 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 2 dissolution determined from electrochemical experiments with 238 Pu doped UO 2 M. Stroess-Gascoyne (F. King, J.S. Betteridge, F. Garisto), doped UO 2 studies (V. Rondinella), Preliminary results of static and dynamic dissolution tests with α doped UO 2 in Boom clay conditions (K. Lemmens), Studies of the behavior of UO 2 / water interfaces under He 2+ beam (C. Corbel), Alpha and gamma radiolysis effects on UO 2 alteration in water (C. Jegou), Behavior of Pu-doped pellets in brines (M. Kelm), On the potential catalytic behavior of

  10. Physics characteristics of CANDU cores with advanced fuel cycles

    International Nuclear Information System (INIS)

    Garvey, P.M.

    1985-01-01

    The current generation of CANDU reactors, of which some 20 GWE are either in operations or under construction worldwide, have been designed specifically for the natural uranium fuel cycle. The CANDU concept, due to its D 2 O coolant and moderator, on-power refuelling and low absorption structural materials, makes the most effective utilization of mined uranium of all currently commercialized reactors. An economic fuel cycle cost is also achieved through the use of natural uranium and a simple fuel bundle design. Total unit energy costs are achieved that allow this reactor concept to effectively compete with other reactor types and other forms of energy production. There are, however, other fuel cycles that could be introduced into this reactor type. These include the slightly enriched uranium fuel cycle, fuel cycles in which plutonium is recycled with uranium, and the thorium cycle in which U-233 is recycled. There is also a special range of fuel cycles that could utilize the spent fuel from LWR's. Two specific variants are a fuel cycle that only utilizes the spent uranium, and a fuel cycle in which both the uranium and plutonium are recycled into a CANDU. For the main part these fuel cycles are characterized by a higher initial enrichment, and hence discharge burnup, than the natural uranium cycle. For these fuel cycles the main design features of both the reactor and fuel bundle would be retained. Recently a detailed study of the use in a CANDU of mixed plutonium and uranium oxide fuel from an LWR has been undertaken by AECL. This study illustrates many of the generic technical issues associated with the use of Advanced Fuel Cycles. This paper will report the main findings of this evaluation, including the power distribution in the reactor and fuel bundle, the choice of fuel management scheme, and the impact on the control and safety characteristics of the reactor. These studies have not identified any aspects that significantly impact upon the introduction of

  11. Things Korean.

    Science.gov (United States)

    Chang, Roberta

    Presented in this booklet are brief descriptions of items and activities that are symbolic of Korean culture. Some of the items and activities described include traditional Korean clothing and accessories, dolls, fans, a Korean game called "yut," tape recordings of Korean music, a "buhk" (drum), and brass eating utensils. A map of Korea, some…

  12. CANDU steam generator life management

    International Nuclear Information System (INIS)

    Tapping, R.L.; Nickerson, J.; Spekkens, P.; Maruska, C.

    1998-01-01

    Steam generators are a critical component of a nuclear power reactor, and can contribute significantly to station unavailability, as has been amply demonstrated in Pressurized Water Reactors (PWRs). CANDU steam generators are not immune to steam generator degradation, and the variety of CANDU steam generator designs and tube materials has led to some unexpected challenges. However, aggressive remedial actions, and careful proactive maintenance activities, have led to a decrease in steam generator-related station unavailability of Canadian CANDUs. AECL and the CANDU utilities have defined programs that will enable existing or new steam generators to operate effectively for 40 years. Research and development work covers corrosion and mechanical degradation of tube bundles and internals, chemistry, thermalhydraulics, fouling, inspection and cleaning, as well as provision for specially tool development for specific problem solving. A major driving force is development of CANDU-specific fitness-for-service guidelines, including appropriate inspection and monitoring technology to measure steam generator condition. Longer-range work focuses on development of intelligent on-line monitoring for the feedwater system and steam generator. New designs have reduced risk of corrosion and fouling, are more easily inspected and cleaned, and are less susceptible to mechanical damage. The Canadian CANDU utilities have developed programs for remedial actions to combat degradation of performance (Gentilly-2, Point Lepreau, Bruce A/B, Pickering A/B), and have developed strategic plans to ensure that good future operation is ensured. The research and development program, as well as operating experience, has identified where improvements in operating practices and/or designs can be made in order to ensure steam generator design life at an acceptable capacity factory. (author)

  13. CANDU reactor - supporting the nuclear renaissance

    International Nuclear Information System (INIS)

    Oberth, R.

    2010-01-01

    'Full text:' The CANDU reactor has proven to be a strong performer in both the Canada, with 22 units constructed in Ontario, New Brunswick and Quebec, as well as in Argentina, Korea, Romania and China where a further nine units are operating and two in the planning stage. The average lifetime capacity factor of the CANDU reactor fleet is 89%. The last seven CANDU projects in Korea, China, and Romania have been completed on budget and on schedule. CANDU reactors have the highest uranium utilization efficiency measures as electricity output per ton of uranium mined. The CANDU fuel channel design using on-power fuelling and a heavy water moderator enables flexible fueling options - from the current natural uranium option to burning uranium recovered from used LWR reactor fuel and even a thorium-based fuel. AECL and the CANDU reactor are poised to participate in the worldwide construction at least 250 new reactors over the next 20 years. (author)

  14. CANDU plant life management - An integrated approach

    International Nuclear Information System (INIS)

    Hopkins, J.R.

    1998-01-01

    An integrated approach to plant life management has been developed for CANDU reactors. Strategies, methods, and procedures have been developed for assessment of critical systems structures and components and for implementing a reliability centred maintenance program. A Technology Watch program is being implemented to eliminate 'surprises'. Specific work has been identified for 1998. AECL is working on the integrated program with CANDU owners and seeks participation from other CANDU owners

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

    Energy Technology Data Exchange (ETDEWEB)

    Park, S.W.; Lee, H.H.; Lee, J.Y.; Lee, J.S.; Ro, S.G. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    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.

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

    International Nuclear Information System (INIS)

    Park, S.W.; Lee, H.H.; Lee, J.Y.; Lee, J.S.; Ro, S.G.

    1997-01-01

    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

  17. Systems analysis of the CANDU 3 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wolfgong, J.R.; Linn, M.A.; Wright, A.L.; Olszewski, M.; Fontana, M.H. [Oak Ridge National Lab., TN (United States)

    1993-07-01

    This report presents the results of a systems failure analysis study of the CANDU 3 reactor design; the study was performed for the US Nuclear Regulatory Commission. As part of the study a review of the CANDU 3 design documentation was performed, a plant assessment methodology was developed, representative plant initiating events were identified for detailed analysis, and a plant assessment was performed. The results of the plant assessment included classification of the CANDU 3 event sequences that were analyzed, determination of CANDU 3 systems that are ``significant to safety,`` and identification of key operator actions for the analyzed events.

  18. Advancing CANDU technology AECL's Development program

    International Nuclear Information System (INIS)

    Torgerson, D.F.

    1997-01-01

    AECL has a comprehensive product development program that is advancing all aspects of CANDU technology including fuel and fuel cycles, fuel channels, heavy water and tritium technology, safety technology, components and systems, constructability, health and environment, and control and instrumentation. The technology arising from these programs is being incorporated into the CANDU design through an evolutionary process. This evolutionary process is focused on improving economics, enhancing safety and ensuring fuel cycle flexibility to secure fuel supply for the foreseeable future. This strategic thrusts are being used by CANDU designers and researchers to set priorities and goals for AECL's development activities. The goals are part of a 25-year development program that culminates in the 'CANDU X'. The 'CANDU X' is not a specific design - it is a concept that articulates our best extrapolation of what is achievable with the CANDU design over the next 25 years, and includes the advanced features arising from the R and D and engineering to be done over that time. AECL's current product, the 700 MWe class CANDU 6 and the 900 MWe class CANDU 9, both incorporate output from the development programs as the technology become available. A brief description of each development areas is given below. The paper ends with the conclusion that AECL has a clear vision of how CANDU technology and products will evolve over the next several years, and has structured a comprehensive development program to take full advantage of the inherent characteristics of heavy water reactors. (author)

  19. Technology transfer: The CANDU approach

    International Nuclear Information System (INIS)

    Hart, R.S.

    1998-01-01

    The many and diverse technologies necessary for the design, construction licensing and operation of a nuclear power plant can be efficiently assimilated by a recipient country through an effective technology transfer program supported by the firm long term commitment of both the recipient country organizations and the supplier. AECL's experience with nuclear related technology transfer spans four decades and includes the construction and operation of CANDU plants in five countries and four continents. A sixth country will be added to this list with the start of construction of two CANDU 6 plants in China in early 1997. This background provides the basis for addressing the key factors in the successful transfer of nuclear technology, providing insights into the lessons learned and introducing a framework for success. This paper provides an overview of AECL experience relative to the important factors influencing technology transfer, and reviews specific country experiences. (author)

  20. CANDU plant maintenance: Recent developments

    International Nuclear Information System (INIS)

    Charlebois, P.

    2000-01-01

    CANDU units have long been recognized for their exceptional safety and reliability. Continuing development in the maintenance area has played a key role in achieving this performance level. For over two decades, safety system availability has been monitored closely and system maintenance programs adjusted accordingly to maintain high levels of performance. But as the plants approach mid life in a more competitive environment and component aging becomes a concern, new methods and techniques are necessary. As a result, recent developments are moving the maintenance program largely from a corrective and preventive approach to predictive and condition based maintenance. The application of these techniques is also being extended to safety related systems. These recent developments include use of reliability centred methods to define system maintenance requirements and strategies. This approach has been implemented on a number of systems at Canadian CANDU plants with positive results. The pilot projects demonstrated that the overall maintenance effort remained relatively constant while the system performance improved. It was also possible to schedule some of the redundant component maintenance during plant operation without adverse impact on system availability. The probabilistic safety assessment was found to be useful in determining the safety implications of component outages. These new maintenance strategies are now making use of predictive and condition based maintenance techniques to anticipate equipment breakdown and schedule preventive maintenance as the need arises rather than time based. Some of these techniques include valve diagnostics, vibration monitoring, oil analysis, thermography. Of course, these tools and techniques must form part of an overall maintenance management system to ensure that maintenance becomes a living program. To facilitate this process and contain costs, new information technology tools are being introduced to provide system engineers

  1. Advanced fuel cycles in CANDU reactors

    International Nuclear Information System (INIS)

    Green, R.E.; Boczar, P.G.

    1990-04-01

    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

  2. CANDU-PHW fuel management

    International Nuclear Information System (INIS)

    Frescura, G.M.; Wight, A.L.

    1982-01-01

    This report covers the material presented in a series of six lectures at the Winter College on Nuclear Physics and Reactors held at the International Centre for Theoretical Physics in Trieste, Italy, Jan 22 - March 28, 1980. The report deals with fuel management in natural uranium fuelled CANDU-PHW reactors. Assuming that the reader has a basic knowledge of CANDU core physics, some of the reactor systems which are more closely related to fuelling are described. This is followed by a discussion of the methods used to calculate the power distribution and perform fuel management analyses for the equilibrium core. A brief description of some computer codes used in fuel management is given, together with an overview of the calculations required to provide parameters for core design and support the accident analysis. Fuel scheduling during approach to equilibrium and equilibrium is discussed. Fuel management during actual reactor operation is discussed with a review of the operating experience for some of the Ontario Hydro CANDU reactors. (author)

  3. CANDU safety under severe accidents

    International Nuclear Information System (INIS)

    Snell, V.G.; Howieson, J.Q.; Alikhan, S.; Frescura, G.M.; King, F.; Rogers, J.T.; Tamm, H.

    1996-01-01

    The characteristics of the CANDU reactor relevant to severe accidents are set first by the inherent properties of the design, and second by the Canadian safety/licensing approach. The pressure-tube concept allows the separate, low-pressure, heavy-water moderator to act as a backup heat sink even if there is no water in the fuel channels. Should this also fail, the calandria shell itself can contain the debris, with heat being transferred to the water-filled shield tank around the core. Should the severe core damage sequence progress further, the shield tank and the concrete reactor vault significantly delay the challenge to containment. Furthermore, should core melt lead to containment overpressure, the containment behaviour is such that leaks through the concrete containment wall reduce the possibility of catastrophic structural failure. The Canadian licensing philosophy requires that each accident, together with failure of each safety system in turn, be assessed (and specified dose limits met) as part of the design and licensing basis. In response, designers have provided CANDUs with two independent dedicated shutdown systems, and the likelihood of Anticipated Transients Without Scram is negligible. Probabilistic safety assessment studies have been performed on operating CANDU plants, and on the 4 x 880 MW(e) Darlington station now under construction; furthermore a scoping risk assessment has been done for a CANDU 600 plant. They indicate that the summed severe core damage frequency is of the order of 5 x 10 -6 /year. 95 refs, 3 tabs

  4. CANDU safety under severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Snell, V G; Howieson, J Q [Atomic Energy of Canada Ltd. (Canada); Alikhan, S [New Brunswick Electric Power Commission (Canada); Frescura, G M; King, F [Ontario Hydro (Canada); Rogers, J T [Carleton Univ., Ottawa, ON (Canada); Tamm, H [Atomic Energy of Canada Ltd. (Canada). Whiteshell Research Lab.

    1996-12-01

    The characteristics of the CANDU reactor relevant to severe accidents are set first by the inherent properties of the design, and second by the Canadian safety/licensing approach. The pressure-tube concept allows the separate, low-pressure, heavy-water moderator to act as a backup heat sink even if there is no water in the fuel channels. Should this also fail, the calandria shell itself can contain the debris, with heat being transferred to the water-filled shield tank around the core. Should the severe core damage sequence progress further, the shield tank and the concrete reactor vault significantly delay the challenge to containment. Furthermore, should core melt lead to containment overpressure, the containment behaviour is such that leaks through the concrete containment wall reduce the possibility of catastrophic structural failure. The Canadian licensing philosophy requires that each accident, together with failure of each safety system in turn, be assessed (and specified dose limits met) as part of the design and licensing basis. In response, designers have provided CANDUs with two independent dedicated shutdown systems, and the likelihood of Anticipated Transients Without Scram is negligible. Probabilistic safety assessment studies have been performed on operating CANDU plants, and on the 4 x 880 MW(e) Darlington station now under construction; furthermore a scoping risk assessment has been done for a CANDU 600 plant. They indicate that the summed severe core damage frequency is of the order of 5 x 10{sup -6}/year. 95 refs, 3 tabs.

  5. Low Power Shutdown PSA for CANDU Type Plants

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Yeon Kyoung; Kim, Myung Su [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    KHNP also have concentrated on full power PSA. Some recently constructed OPR1000 type plants and APR1400 type plants have performed the low power and shutdown (LPSD) PSA. The purpose of LPSD PSA is to identify the main contributors on the accident sequences of core damage and to find the measure of safety improvement. After the Fukushima accident, Korean regulatory agency required the shutdown severe accident management guidelines (SSAMG) development for safety enhancement. For the reliability of SSAMG, KHNP should develop the LPSD PSA. Especially, the LPSD PSA for CANDU type plant had developed for the first time in Korea. This paper illustrates how the LPSD PSA for CANDU type developed and the core damage frequency (CDF) is different with that of full power PSA. KHNP performed LPSD PSA to develop the SSAMG after the Fukushima accidents. The results show that risk at the specific operation mode during outage is higher than that of full power operation. Also, the results indicated that recovery failure of class 4 power at the POS 5A, 5B contribute dominantly to the total CDF from importances analysis. LPSD PSA results such as CDF with initiating events and POSs, risk results with plant damage state, and containment failure probability and frequency with POSs can be used by inputs for developing the SSAMG.

  6. Release of segregated nuclides from spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, L.H.; Tait, J.C. [Atomic Energy Canada Ltd., Pinawa, MB (Canada). Whiteshell Laboratories

    1997-10-01

    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. 33 refs, 6 figs, 3 tabs.

  7. Spent Fuel Transfer to Dry Storage Using Unattended Monitoring System

    International Nuclear Information System (INIS)

    Park, Jae Hwan; Park, Soo Jin

    2009-01-01

    There are 4 CANDU reactors at Wolsung site together with a spent fuel dry storage associated with unit 1. These CANDU reactors, classified as On-Load Reactor (OLR) for Safeguards application, change 16- 24 fuel bundles with fresh fuel in everyday. Especially, the spent fuel bundles are transferred from spent fuel bays to dry storage throughout a year because of the insufficient capacity of spent fuel pond. Safeguards inspectors verify the spent fuel transfer to meet safeguards purposes according to the safeguards criteria by means of inspector's presence during the transfer campaign. For the verification, 60-80 person-days of inspection (PDIs) are needed during approximately 3 months for each unit. In order to reduce the inspection effort and operators' burden, an Unattended Monitoring System (UMS) was designed and developed by the IAEA for the verification of spent fuel bundles transfers from wet storage to dry storage. Based on the enhanced cooperation of CANDU reactors between the ROK and the IAEA, the IAEA installed the UMS at Wolsung unit 2 in January 2005 at first. After some field trials during the transfer campaign, this system is being replaced the traditional human inspection since September 1, 2006 combined with a Short Notice Inspection (SNI) and a near-real time Mailbox Declaration

  8. Learning from experience: feedback to CANDU design

    International Nuclear Information System (INIS)

    Allen, P.J.; Hopwood, J.M.; Rousseau, G.P.

    1998-01-01

    AECL's main product line is based on two single unit CANDU nuclear power plant designs; CANDU 6 and CANDU 9, each of which is based on successfully operating CANDU plants. AECL's CANDU development program is based upon evolutionary improvement. The evolutionary design approach ensures the maximum degree of operational provenness. It also allows successful features of today's plants to be retained while incorporating improvements as they develop to the appropriate level of design maturity. A key component of this evolutionary development is a formal process of gathering and responding to feedback from: NPP operation, construction and commissioning; regulatory input; equipment supplier input; R and D results; market input. The progresses for gathering and implementing the experience feedback and a number of recent examples of design improvements from this feedback process are described in the paper. (author)

  9. Development of CANDU advanced fuel bundle

    International Nuclear Information System (INIS)

    Suk, H. C.; Hwang, W.; Rhee, B. W.; Jung, S. H.; Chung, C. H.

    1992-05-01

    This research project is underway in cooperation with AECL to develop the CANDU advanced fuel bundle (so-called, CANFLEX) which can enhance reactor safety and fuel economy in comparison with the current CANDU fuel and which can be used with natural uranium, slightly enriched uranium and other advanced fuel cycle. As the final schedule, the advanced fuel will be verified by carrying out a large scale demonstration of the bundle irradiation in a commercial CANDU reactor for 1996 and 1997, and consequently will be used in the existing and future CANDU reactors in Korea. The research activities during this year include the detail design of CANFLEX fuel with natural enriched uranium (CANFLEX-NU). Based on this design, CANFLEX fuel was mocked up. Out-of-pile hydraulic scoping tests were conducted with the fuel in the CANDU Cold Test Loop to investigate the condition under which maximum pressure drop occurs and the maximum value of the bundle pressure drop. (Author)

  10. Spent fuel management of NPPs in Argentina

    International Nuclear Information System (INIS)

    Alvarez, D.E.; Lee Gonzalez, H.M.

    2010-01-01

    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)

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

    International Nuclear Information System (INIS)

    Suk, H.C.; Yang, M.S.; Sim, K-S.; Yoo, K.J.

    1997-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  13. CANDU design options with detritiation

    International Nuclear Information System (INIS)

    Wren, D.J.; Hart, R.S.

    1997-01-01

    CANDU reactors include a number of auxiliary systems to manage the inventory, purification, clean-up and isotopic purity of the heavy water used in the moderator and heat transport system. These systems are designed and installed to treat the moderator and heat transport water in separate parallel systems. One of the reasons for this parallel approach to heavy water management is the tritium inventory in the heavy water. Different levels of tritium accumulate in the moderator and heat transport system during reactor operation, with the moderator water having a much higher tritium concentration. Strict separation of the high- tritium-concentration moderator water from the low-tritium-concentration heat transport system water is an integral component of the CANDU design and operating strategy to limit potential releases of tritium to the containment building atmosphere. AECL is developing a new cost-effective technology for the detritiation of heavy water based on the Combined Electrolysis and Catalytic Exchange (CECE) process. This detritiation technology has the potential to be integrated into the heavy water management systems of a CANDU reactor. On-line detritiation could be used to limit the concentration of tritium in the moderator and also to detritiate any water collected within the containment building from other sources. The availability of economic detritiation technology would provide a flexibility to redesign some of the auxiliary heavy water management systems. In particular, there is potential to eliminate some of the duplication in the current management systems and also reduce costs by reclassifying some reactor systems that would have lower maximum tritium concentrations. This paper discusses some of the advantages of detritiation and some of the conceptual design options that detritiation would provide. The goal would be to lower the overall reactor cost with detritiation, but it is premature to assess whether this goal can be achieved. (author)

  14. Hydrogen in CANDU fuel elements

    International Nuclear Information System (INIS)

    Sejnoha, R.; Manzer, A.M.; Surette, B.A.

    1995-01-01

    Unirradiated and irradiated CANDU fuel cladding was tested to compare the role of stress-corrosion cracking and of hydrogen in the development of fuel defects. The results of the tests are compared with information on fuel performance in-reactor. The role of hydriding (deuteriding) from the coolant and from the fuel element inside is discussed, and the control of 'hydrogen gas' content in the element is confirmed as essential for defect-free fuel performance. Finally, implications for fuel element design are discussed. (author)

  15. CANDU reactors. Experience and innovation

    International Nuclear Information System (INIS)

    Hart, R.S.; Brooks, G.L.

    1989-02-01

    The title of this paper highlights two key considerations which must be properly balanced through good management in the evolution of any engineering product. Excessive reliance on experience will lead to product stagnation; excessive reliance on innovation will often lead to an unsatisfactory product, at least in the first generation of this product. To illustrate this balancing process, the paper reviews CANDU evolution and experience and the balance between proveness and innovation achieved through management of the evolution process from early prototypes to today's large-scale commercial units. A forecast of continuing evolutionary directions is included

  16. Mobile robotics for CANDU maintenance

    International Nuclear Information System (INIS)

    Lipsett, M.G.; Rody, K.H.

    1996-01-01

    Although robotics researchers have been promising that robotics would soon be performing tasks in hazardous environments, the reality has yet to live up to the hype. The presently available crop of robots suitable for deployment in industrial situations are remotely operated, requiring skilled users. This talk describes cases where mobile robots have been used successfully in CANDU stations, discusses the difficulties in using mobile robots for reactor maintenance, and provides near-term goals for achievable improvements in performance and usefulness. (author) 5 refs., 2 ills

  17. CANDU reactor experience: fuel performance

    International Nuclear Information System (INIS)

    Truant, P.T.; Hastings, I.J.

    1985-07-01

    Ontario Hydro has more than 126 reactor-years experience in operating CANDU reactors. Fuel performance has been excellent with 47 000 channel fuelling operations successfully completed and 99.9 percent of the more than 380 000 bundles irradiated operating as designed. Fuel performance limits and fuel defects have had a negligible effect on station safety, reliability, the environment and cost. The actual incapability charged to fuel is less than 0.1 percent over the stations' lifetimes, and more recently has been zero

  18. Candu reactors - experience and innovation

    International Nuclear Information System (INIS)

    Hart, R.S.; Brooks, G.L.

    1989-01-01

    The title of this paper highlights two key considerations which must be properly balanced through good management in the evolution of any engineering product. Excessive reliance on experience will lead to product stagnation; excessive reliance on innovation will often lead to an unsatisfactory product, at least in the first generation of this product. To illustrate this balancing process, the paper reviews CANDU evolution and experience and the balance between proveness and innovation achieved through management of the evolution process from early prototypes to today's large-scale commercial units. A forecast of continuing evolutionary directions is included

  19. Operating Experience of MACSTOR Modules at CANDU 6 Stations

    International Nuclear Information System (INIS)

    Beaudoin, Robert R.

    2005-01-01

    Over the last three decades, Atomic Energy of Canada Limited (AECL) has contributed to the technology development and implementation of dry spent fuel management facilities in Canada, Korea and Romania During that period, AECL has developed a number of concrete canister models and the MACSTOR200 module, a medium size air-cooled vault with a 228 MgU (Mega grams of Uranium) capacity. AECL's dry storage technologies were used for the construction of eight large-scale above ground dry storage facilities for CANDU spent fuel. As of 2005, those facilities have an installed capacity in excess of 5,000 MgU. Since 1995, the two newest dry storage installations built for CANDU 6 reactors at Gentilly 2 (Canada) and Cernavoda (Romania) used the MACSTOR 200 module. Seven such modules have been built at Gentilly 2 during the 1995 to 2004 period and one at Cernavoda in 2003. The construction and operating experience of those modules is reviewed in this paper. The MACSTOR 200 modules were initially designed for a 50-year service life, with recent units at Gentilly 2 licensed for a 100-year service life in a rural (non-maritime) climate. During the 1995-2005 period, six of the eight modules were loaded with fuel. Their operation has brought a significant amount of experience on loading operations, performance of fuel handling equipment, radiation shielding, heat transfer, monitoring of the two confinement boundaries and radiation dose to personnel. Heat dissipation performance of the MACSTOR 200 was initially licensed using values derived from full scale tests made at AECL's Whiteshell Research Laboratories, that were backed-up by temperature measurements made on the first two modules. Results and computer models developed for the MACSTOR 200 module are described. Korea Hydro and Nuclear Power (KHNP) and its subsidiary Nuclear Environment Technology Institute (NETEC), in collaboration with Hyundai Engineering Company Ltd. (HEC) and AECL, are developing a new dry storage module to

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

    International Nuclear Information System (INIS)

    Chung, Jang Hwan; Suk, Ho Cheon; Jeong, Moon Ki; Park, Joo Hwan; Jeong, Heung Joon; Jeon, Ji Soo; Kim, Bok Deuk

    1994-07-01

    This project is underway in cooperation with AECL to develop the CANDU advanced fuel bundle (so-called, CANFLEX) which can enhance reactor safety and fuel economy in comparison with the current CANDU fuel and which can be used with natural uranium, slightly enriched uranium and other advanced fuel cycle. As the final schedule, the advanced fuel will be verified by carrying out a large scale demonstration of the bundle irradiation in a commercial CANDU reactor, and consequently will be used in the existing and future CANDU reactors in Korea. The research activities during this year Out-of-pile hydraulic tests for the prototype of CANFLEX bundle was conducted in the CANDU-hot test loop at KAERI. Thermalhydraulic analysis with the assumption of CANFLEX-NU fuel loaded in Wolsong-1 was performed by using thermalhydraulic code, and the thermal margin and T/H compatibility of CANFLEX bundle with existing fuel for CANDU-6 reactor have been evaluated. (Author)

  1. A CANDU Severe Accident Analysis

    International Nuclear Information System (INIS)

    Negut, Gheorghe; Catana, Alexandru; Prisecaru, Ilie

    2006-01-01

    As interest in severe accident studies has increased in the last years, we have developed a set of simple models to analyze severe accidents for CANDU reactors that should be integrated in the EU codes. The CANDU600 reactor uses natural uranium fuel and heavy water (D2O) as both moderator and coolant, with the moderator and coolant in separate systems. We chose to analyze accident development for a LOCA with simultaneous loss of moderator cooling and the loss of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperatures inside a pressure tube reaches 10000 deg C, a contact between pressure tube and calandria tube occurs and the residual heat is transferred to the moderator. Due to the lack of cooling, the moderator eventually begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) will be uncovered, then will disintegrate and fall down to the calandria vessel bottom. After all the quantity of moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield tank water, which normally surrounds the calandria vessel. The phenomena described above are modelled, analyzed and compared with the existing data. The results are encouraging. (authors)

  2. Evolution of CANDU reactor design

    International Nuclear Information System (INIS)

    Pon, G.A.

    1978-08-01

    The CANDU (CANada Deuterium Uranium) design had its begin-ings in the early 1950's with the preliminary engineering studies that led to the 20 MW(e) NPD (Nuclear Power Demonstration) and the 200 MW(e) Douglas Point station . The next decade saw the first operation of both these stations and the commitment of the 2000 MW(e) Pickering and 3000 MW(e) Bruce plants. The present decade has witnessed the excellent performance of Pickering and Bruce and commitments to construct Gentilly-2, Cordoba, Pt. Lepreau, Wolsung, Pickering B, Bruce B and Darlington. In most cases, successive CANDU designs have meant an increase in plant output. Evolutionary developments have been made to fit the requirements of higher ratings and sizes, new regulations, better reliability and maintainability and lower costs. These changes, which are described system by system, have been introduced in the course of engineering parallel reactor projects with overlapping construction schedules -circumstances which ensure close contact with the practical realities of economics, manufacturing functions, construction activities and performance in commissioning. Features for one project furnished alternative concepts for others still on the drawing board and the experience gained in the first application yielded a sound basis for its re-use in succeeding projects. Thus the experiences gained in NPD, Douglas Point, Gentilly-1 and KANUPP have contributed to Pickering and Bruce, which in turn have contributed to the design of Gentilly-2. (author)

  3. CANDU plant life management - An integrated approach

    International Nuclear Information System (INIS)

    Charlebois, P.; Hart, R.S.; Hopkins, J.R.

    1998-01-01

    Commercial versions of CANDU reactors were put into service starting more than 25 years ago. The first unit of Ontario Hydro's Pickering A station was put into service in 1971, and Bruce A in 1977. Most CANDU reactors, however, are only now approaching their mid-life of 15 to 20 years of operation. In particular, the first series of CANDU 6 plants which entered service in the early 1980's were designed for a 30 year life and are now approaching mid life. The current CANDU 6 design is based on a 40 year life as a result of advancement in design and materials through research and development. In order to assure safe and economic operation of these reactors, a comprehensive CANDU Plant Life Management (PLIM) program is being developed from the knowledge gained during the operation of Ontario Hydro's Pickering, Bruce, and Darlington stations, worldwide information from CANDU 6 stations, CANDU research and development programs, and other national and international sources. This integration began its first phase in 1994, with the identification of most of the critical systems structures and components in these stations, and a preliminary assessment of degradation and mechanisms that could affect their fitness for service for their planned life. Most of these preliminary assessments are now complete, together with the production of the first iteration of Life Management Plans for several of the systems and components. The Generic CANDU 6 PLIM program is now reaching its maturity, with formal processes to systematically identify and evaluate the major CSSCs in the station, and a plan to ensure that the plant surveillance, operation, and maintenance programs monitor and control component degradation well within the original design specifications essential for the plant life attainment. A Technology Watch program is being established to ensure that degradation mechanisms which could impact on plant life are promptly investigated and mitigating programs established. The

  4. Passive heat transport in advanced CANDU containment

    International Nuclear Information System (INIS)

    Krause, M.; Mathew, P.M.

    1994-01-01

    A passive CANDU containment design has been proposed to provide the necessary heat removal following a postulated accident to maintain containment integrity. To study its feasibility and to optimize the design, multi-dimensional containment modelling may be required. This paper presents a comparison of two CFD codes, GOTHIC and PHOENICS, for multi-dimensional containment analysis and gives pressure transient predictions from a lumped-parameter and a three-dimensional GOTHIC model for a modified CANDU-3 containment. GOTHIC proved suitable for multidimensional post-accident containment analysis, as shown by the good agreement with pressure transient predictions from PHOENICS. GOTHIC is, therefore, recommended for passive CANDU containment modelling. (author)

  5. Future CANDU nuclear power plant design requirements document executive summary

    International Nuclear Information System (INIS)

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young; S. A. Usmani

    1996-03-01

    The future CANDU Requirements Document (FCRED) describes a clear and complete statement of utility requirements for the next generation of CANDU nuclear power plants including those in Korea. The requirements are based on proven technology of PHWR experience and are intended to be consistent with those specified in the current international requirement documents. Furthermore, these integrated set of design requirements, incorporate utility input to the extent currently available and assure a simple, robust and more forgiving design that enhances the performance and safety. The FCRED addresses the entire plant, including the nuclear steam supply system and the balance of the plant, up to the interface with the utility grid at the distribution side of the circuit breakers which connect the switchyard to the transmission lines. Requirements for processing of low level radioactive waste at the plant site and spent fuel storage requirements are included in the FCRED. Off-site waste disposal is beyond the scope of the FCRED. 2 tabs., 1 fig. (Author) .new

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

    International Nuclear Information System (INIS)

    Jeong, Chang Joon; Park, Chang Je

    2006-02-01

    The thorium fuel recycle scenarios through the Canada deuterium uranium (CANDU) reactor have been analyzed for two types of thorium fuel: homogeneous ThO 2 UO 2 and ThO 2 UO 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. 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.

  8. Simulation of LOCA power transients of CANDU6 by SCAN/RELAP-CANDU coupled code system

    International Nuclear Information System (INIS)

    Hong, In Seob; Kim, Chang Hyo; Hwang, Su Hyun; Kim, Man Woong; Chung, Bub Dong

    2004-01-01

    As can be seen in the standalone application of RELAP-CANDU for LOCA analysis of CANDU-PHWR, the system thermal-hydraulic code alone cannot predict the transient behavior accurately. Therefore, best estimate neutronics and system thermal-hydraulic coupled code system is necessary to describe the transient behavior with higher accuracy and reliability. The purpose of this research is to develop and test a coupled neutronics and thermal-hydraulics analysis code, SCAN (SNU CANDU-PHWR Neutronics) and RELAP-CANDU, for transient analysis of CANDU-PHWR's. For this purpose, a spatial kinetics calculation module of SCAN, a 3-D CANDU-PHWR neutronics design and analysis code, is dynamically coupled with RELAP-CANDU, the system thermal-hydraulic code for CANDU-PHWR. The performance of the coupled code system is examined by simulation of reactor power transients caused by a hypothetical Loss Of Coolant Accident (LOCA) in Wolsong units, which involves the insertion of positive void reactivity into the core in the course of transients. Specifically, a 40% Reactor Inlet Header (RIH) break LOCA was assumed for the test of the SCAN/RELAP-CANDU coupled code system analysis

  9. Qinshan CANDU project open top construction method

    International Nuclear Information System (INIS)

    Petrunik, K.J.; Wittann, K.; Khan, A.; Ricciuti, R.; Ivanov, A.; Chen, S.

    2003-01-01

    The significant schedule reductions achieved on the Qinshan CANDU Project were due in large part to the incorporation of advanced construction technologies in project design and delivery. For the Qinshan Project, a number of key advantages were realized through the use of the 'Open Top' construction method. This paper discusses the Qinshan Phase III CANDU Project Open Top implementation method. The Open Top method allowed major equipment to be installed simply, via the use of a Very Heavy Lift (VHL) crane and permitted the use of large-scale modularization. The advantages of Open Top construction, such as simplified access, more flexible project scheduling, improved construction safety and quality, and reduced labours are presented. The large-scale modularization of the Reactor Building Dousing System and the Open Top installation method and advantages relative to traditional CANDU 6 construction practices are also presented. Finally, major improvements for future CANDU plant construction using the Open Top method are discussed. (author)

  10. Future fuel cycle development for CANDU reactors

    International Nuclear Information System (INIS)

    Hatcher, S.R.; McDonnell, F.N.; Griffiths, J.; Boczar, P.G.

    1987-01-01

    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)

  11. Mathematical modeling of CANDU-PHWR

    Energy Technology Data Exchange (ETDEWEB)

    Gaber, F.A.; Aly, R.A.; El-Shal, A.O. [Atomic Energy Authority, Cairo (Egypt)

    2003-07-01

    The paper deals with the transient studies of CANDU 600 pressurized Heavy Water Reactor (PHWR). This study involved mathematical modeling of CANDU-PHWR to study its thermodynamic performances. Modeling of CANDU-PHWR was based on lumped parameter technique. The reactor model includes the neutronic, reactivity, and fuel channel heat transfer. The nuclear reactor power was modelled using the point kinetics equations with six groups of delayed neutrons and the reactivity feed back due to the changes in the fuel temperature and coolant temperature. The CANDU-PHWR model was coded in FORTRAN language and solved by using a standard numerical technique. The adequacy of the model was tested by assessing the physical plausibility of the obtained results. (author)

  12. Reactor physics aspects of CANDU reactors

    International Nuclear Information System (INIS)

    Critoph, E.

    1980-01-01

    These four lectures are being given at the Winter Course on Nuclear Physics at Trieste during 1978 February. They constitute part of the third week's lectures in Part II: Reactor Theory and Power Reactors. A physical description of CANDU reactors is given, followed by an overview of CANDU characteristics and some of the design options. Basic lattice physics is discussed in terms of zero energy lattice experiments, irradiation effects and analytical methods. Start-up and commissioning experiments in CANDU reactors are reviewed, and some of the more interesting aspects of operation discussed - fuel management, flux mapping and control of the power distribution. Finally, some of the characteristics of advanced fuel cycles that have been proposed for CANDU reactors are summarized. (author)

  13. Nuclear energy in Canada: the CANDU system

    International Nuclear Information System (INIS)

    Robertson, J.A.L.

    1979-10-01

    Nuclear electricity in Canada is generated by CANDU nuclear power stations. The CANDU reactor - a unique Canadian design - is fuelled by natural uranium and moderated by heavy water. The system has consistently outperformed other comparable nuclear power systems in the western world, and has an outstanding record of reliability, safety and economy. As a source of energy it provides the opportunity for decreasing our dependence on dwindling supplies of conventional fossil fuels. (auth)

  14. Verification tests for CANDU advanced fuel

    International Nuclear Information System (INIS)

    Chung, Chang Hwan; Chang, S.K.; Hong, S.D.

    1997-07-01

    For the development of a CANDU advanced fuel, the CANFLEX-NU fuel bundles were tested under reactor operating conditions at the CANDU-Hot test loop. This report describes test results and test methods in the performance verification tests for the CANFLEX-NU bundle design. The main items described in the report are as follows. - Fuel bundle cross-flow test - Endurance fretting/vibration test - Freon CHF test - Production of technical document. (author). 25 refs., 45 tabs., 46 figs

  15. Passive heat removal in CANDU

    International Nuclear Information System (INIS)

    Hart, R.S.; Snell, V.G.

    1996-01-01

    The Three Mile Island accident spurred a world-wide interest in severe accidents. The initial reaction was to increase the preventative measures in existing designs, followed by development of predictive capabilities to improve the management of severe accidents. Recently, emphasis has been placed in new designs on mitigative measures which slow down or contain the progression of a severe accidents. U.S. requirements for Advanced Light Water Reactor designs must now: provide reactor cavity floor space to enhance debris spreading; provide a means to flood the reactor cavity to assist in the cooling process. The paper describes how CANDU Pressurized Heavy Water Reactors (PHWRs) have severe accident prevention and mitigation inherent in the design; in particular, the U.S. severe accident requirements can be met without significant change to the design of current CANDUs. (author). 32 refs, 7 figs, 1 tab

  16. Economics of CANDU-PHW

    International Nuclear Information System (INIS)

    McConnell, L.G.; Woodhead, L.W.

    1981-01-01

    Nuclear and coal stations are the primary options in Ontario for new power generation over the period 1980-2000. The former are the best for base-load requirements, and the latter for peaking. In 1980 the total unit energy cost for Pickering A was 12.77 mill/kWh, compared with 21.18 mill/kWh for power from the Lambton coal-fired station. With on-power fuelling, CANDU-PHW units have achieved a 77 percent capacity factor since first electricity production and 79 percent since their in-service dates. Assuming a 67 percent capacity factor for PWR performance, the power costs with PWR units would be 26 percent higher. (D.N.)

  17. Passive heat removal in CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Hart, R S; Snell, V G [AECL CANDU, Sheridan Park Research Community, Mississauga, ON (Canada)

    1996-12-01

    The Three Mile Island accident spurred a world-wide interest in severe accidents. The initial reaction was to increase the preventative measures in existing designs, followed by development of predictive capabilities to improve the management of severe accidents. Recently, emphasis has been placed in new designs on mitigative measures which slow down or contain the progression of a severe accidents. U.S. requirements for Advanced Light Water Reactor designs must now: provide reactor cavity floor space to enhance debris spreading; provide a means to flood the reactor cavity to assist in the cooling process. The paper describes how CANDU Pressurized Heavy Water Reactors (PHWRs) have severe accident prevention and mitigation inherent in the design; in particular, the U.S. severe accident requirements can be met without significant change to the design of current CANDUs. (author). 32 refs, 7 figs, 1 tab.

  18. Drift effects in CANDU reactors

    International Nuclear Information System (INIS)

    Koclas, J.; Roy, R.; Marleau, G.

    1993-01-01

    The diffusion equation is an approximation to the transport equation which relies on the validity of Fick's law. Since this law is not explicitly integrated in the transport equation it can only be derived approximately using homogenization theories. However, such homogenization theories state that when the cell is not symmetric Fick's law breaks down due to the presence of an additional term to the neutron current, called the drift term. In fact, this term can be interpreted as a transport correction to Fick's law which tends to increase the neutron current in a direction opposite to that specified by the flux gradient. In this paper, we investigate how the presence of asymmetric liquid zone controllers will modify the flux distribution inside a CANDU core. 5 refs., 2 figs., 1 tab

  19. Absorber materials in CANDU PHWR's

    International Nuclear Information System (INIS)

    Price, E.G.; Boss, C.R.; Novak, W.Z.; Fong, R.W.L.

    1995-03-01

    In a CANDU reactor the fuel channels are arranged on a square lattice in a calandria filled with heavy water moderator. This arrangement allows five types of tubular neutron absorber devices to be located in a relatively benign environment of low pressure, low temperature heavy water between neighbouring rows of columns of fuel channels. This paper will describe the roles of the devices and outline the design requirements of the absorber component from a reactor physics viewpoint. Nuclear heating and activation problems associated with the different absorbers will be briefly discussed. The design and manufacture of the devices will be also discussed. The control rod absorbers and shut off materials are cadmium and stainless steel. In the tubular arrangement, the cadmium is sandwiched between stainless steel tubes. This type of device has functioned well, but there is now concern over the availability and expense of cadmium which is used in two types of CANDU control devices. There are also concerns about the toxicity of cadmium during the fabrication of the absorbers. These concerns are prompting AECL to study alternatives. To minimize design changes, pure boron-10 alloyed in stainless steel is a favoured option. Work is underway to confirm the suitability of the boron-loaded steel and identify other encapsulated absorber materials for practical application. Because the reactivity devices or their guide tubes span the calandria vessel, the long slender components must be sufficiently rigid to resist operational vibration and also be seismically stable. Some of these components are made of Zircaloy to minimize neutron absorption. Slow irradiation growth and creep can reduce the spring tension, and periodic adjustments to the springs are required. Experience with the control absorber devices has generally been good. In one instance liquid zone controllers had a problem of vibration induced fretting but a designed back-fit resolved the problem. (author). 3 refs., 1

  20. MAAP4 CANDU analysis of a generic CANDU-6 plant: preliminary results

    Energy Technology Data Exchange (ETDEWEB)

    Petoukhov, S.M.; Mathew, P.M

    2001-10-01

    To support the generic probabilistic safety analysis (PSA) program at AECL, in particular to conduct Level 2 PSA analysis of a CANDU 6 plant undergoing a postulated severe accident, the capability to conduct severe accident consequence analysis for a CANDU plant is required. For this purpose, AECL selected MAAP4 CANDU from a number of other severe accident codes. The necessary models for a generic CANDU 6 station have been implemented in the code, and the code version 0.2 beta was tested using station data, which were assembled for a generic CANDU 6 station. This paper describes the preliminary results of the consequence analysis using MAAP4 CANDU for a generic CANDU 6 station, when it undergoes a station blackout and a large loss-of-coolant accident scenario. The analysis results show that the plant response is consistent with the physical phenomena modeled and the failure criteria used. The results also confirm that the CANDU design is robust with respect to severe accidents, which is reflected in the calculated long times that are available for administering accident management measures to arrest the accident progression before the calandria vessel or containment become at risk. (author)

  1. Candu

    International Nuclear Information System (INIS)

    Melvin, J.G.

    1983-01-01

    Electricity consumption in Canada over the past twenty years correlates well with the real gross national product (GNP). No effect of electricity conservation measures has yet been discernible because, unlike oil, electricity has never been used very wastefully. A 4% annual expansion of real GNP implies a 5% extrapolated annual growth in electricity demand. To sell reactors, the nuclear industry must promote the use of electricity. Canadian industry generally, even the nuclear industry, has been slow to use electricity for process heating. The advantages include stability of price, good control, and absence of pollution

  2. Recent IAEA activities on CANDU-PHWR fuels and fuel cycles

    International Nuclear Information System (INIS)

    Inozemtsev, V.; Ganguly, C.

    2005-01-01

    Pressurized Heavy Water Reactors (PHWR), widely known as CANDU, are in operation in Argentina, Canada, China, India, Pakistan, Republic of Korea and Romania and account for about 6% of the world's nuclear electricity production. The CANDU reactor and its fuel have several unique features, like horizontal calandria and coolant tubes, on-power fuel loading, thin-walled collapsible clad coated with graphite on the inner surface, very high density (>96%TD) natural uranium oxide fuel and amenability to slightly enriched uranium oxide, mixed uranium plutonium oxide (MOX), mixed thorium plutonium oxide, mixed thorium uranium (U-233) oxide and inert matrix fuels. Several Technical Working Groups (TWG) of IAEA periodically discuss and review CANDU reactors, its fuel and fuel cycle options. These include TWGs on water-cooled nuclear power reactor Fuel Performance and Technology (TWGFPT), on Nuclear Fuel Cycle Options and spent fuel management (TWGNFCO) and on Heavy Water Reactors (TWGHWR). In addition, IAEA-INPRO project also covers Advanced CANDU Reactors (ACR) and DUPIC fuel cycles. The present paper summarises the Agency's activities in CANDU fuel and fuel cycle, highlighting the progress during the last two years. In the past we saw HWR and LWR technologies and fuel cycles separate, but nowadays their interaction is obviously growing, and their mutual influence may have a synergetic character if we look at the world nuclear fuel cycle as at an integrated system where the both are important elements in line with fast neutron, gas cooled and other advanced reactors. As an international organization the IAEA considers this challenge and makes concrete steps to tackle it for the benefit of all Member States. (author)

  3. CANDU radiotoxicity inventories estimation: A calculated experiment cross-check for data verification and validation

    International Nuclear Information System (INIS)

    Pavelescu, Alexandru Octavian; Cepraga, Dan Gabriel

    2007-01-01

    This paper is related to the Clearance Potential Index, Ingestion and Inhalation Hazard Factors of the nuclear spent fuel and radioactive wastes. This study required a complex activity that consisted of various phases such us: the acquisition, setting up, validation and application of procedures, codes and libraries. The paper reflects the validation phase of this study. Its objective was to compare the measured inventories of selected actinide and fission products radionuclides in an element from a Pickering CANDU reactor with inventories predicted using a recent version of the ORIGEN-ARP from SCALE 5 coupled with the time dependent cross sections library, CANDU 28.lib, produced by the sequence SAS2H of SCALE 4.4a. In this way, the procedures, codes and libraries for the characterization of radioactive material in terms of radioactive inventories, clearance, and biological hazard factors are being qualified and validated, in support for the safety management of the radioactive wastes. (authors)

  4. CANDU - Canadian experience and expectations with the heavy-water reactor

    International Nuclear Information System (INIS)

    Foster, J.S.; Russell, S.H.

    1977-05-01

    The paper describes the evolution of the CANDU nuclear-power plants with particular reference to the objectives of safety, reliability and economy; the development of industrial capacity for the supply of fuel, components and heavy water; and the prospective development of advanced fuel cycles and the projected results. It provides data on radiation, releases, and exposures, internal and external to the power plants; plant availability, capacity factors and other performance data; heavy water production data with reference to safety, reliability, and economics; projections of the performance of CANDU reactors operating on a thorium-U-233 cycle and the development required to establish this cycle; and intent with respct to spent-fuel management and radioactive-waste storage. (author)

  5. Feasibility study of CANDU-9 fuel handling system

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Jeong Ki; Jo, C. H.; Kim, H. M.

    1996-12-01

    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.

  6. Variables Related to Career Success: Korean-American Women of Distinction Tell Their Stories.

    Science.gov (United States)

    Rhee, Susan Byungsook

    The goal of this research was to uncover the reasons for the unusual success of a small group of Korean-Americans in the American work arena. All had spent their childhood and formative years in Korea under the sole influence of Korean culture, attending the best high schools and colleges. Korean culture tends not to nurture some of the…

  7. Prospects of Using Reprocessed Uranium in CANDU Reactors, in the U.S. GNEP Program

    International Nuclear Information System (INIS)

    Ellis, Ronald James

    2007-01-01

    Current Global Nuclear Energy Partnership (GNEP) plans envision reprocessing spent fuel (SF) with view to minimizing high-level waste (HLW) repository use and recovering actinides (U, Np, Pu, Am, and Cm) for transmutation in reactors as fuel and targets. The reprocessed uranium (RU), however, is to be disposed of. This paper presents a limited-scope analysis of possible reuse of RU in CANDU (Canada Deuterium Uranium) Reactors, within the context of the US GNEP program. Other papers on this topic submitted to this conference discuss the possibility of RU reuse in light-water reactors (LWRs) (with enrichment) and offer an independent economic analysis of RU reuse. A representative RU uranium 'vector', from reprocessed spent LWR fuel, comprises 98.538 wt% 238U, 0.46 wt% 236 U, 0.986 wt% 235 U, and 0.006 wt% 234 U. After multiple recyclings, the concentration of 234 U can approach 0.02 wt%. The presence of 234 U and 236 U in RU reduces the reactivity and fuel lifetime (exit burnup), which is particularly an issue in LWRs. While in PWR analyses, the burnup penalty caused by the concentration of 236 U in RU needs to be offset by additional 235 U enrichment in the amount of ∼25% to 30% of the weight percentage of the 236 U; however, the effect in CANDU is much smaller. Furthermore, since the 235 U content in RU exceeds that of natural uranium, CANDU offers the advantageous option of uranium recycling without reenrichment. The exit burnup of CANDU RU-derived fuel is considerably larger than that for natural uranium-fueled scenario, despite the presence of 234 U and 236 U.

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

    International Nuclear Information System (INIS)

    Tingle, C.P.; Bonin, H.W.

    1999-01-01

    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 UO 2 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 UO 2 . The model was initially tested and the average discharge burnup for natural UO 2 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

  9. Considerations in selecting tubing materials for CANDU steam generators

    International Nuclear Information System (INIS)

    Hemmings, R.L.

    1978-01-01

    Corrosion resistance is the major consideration in selecting tubing material for CANDU steam generators. Corrosion, and additional considerations, lead to the following steam generator tubing material recommendations: for CANDU-BPHWR's (boiling pressurized heavy water reactors) low-cobalt Incoloy-800; for CANDU-PHWR's (pressurized, non-boiling, heavy water reactors), low-cobalt Monel-400

  10. Approach for seismic risk analysis for CANDU plants in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Lee, B-S; Kim, T; Kang, S-K [Korea Power Engineering Co., Seoul (Korea, Republic of); Hong, S-Y; Roh, S-R [Korea Electric Power Corp., Taejon (Korea, Republic of). Research Centre

    1996-12-31

    A seismic risk analysis for CANDU type plants has never been performed. The study presented here suggested that the approach generally applied to LWR type plants could lead to unacceptable result, if directly applied to CANDU plants. This paper presents a modified approach for the seismic risk analysis of CANDU plants. (author). 5 refs., 2 tabs., 2 figs.

  11. Candu 6: versatile and practical fuel technology

    International Nuclear Information System (INIS)

    Hopwood, J. M.; Saroudis, J.

    2013-01-01

    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, D 2 O, 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

  12. CANDU severe accident management guidance update

    International Nuclear Information System (INIS)

    Jones, L.; Popov, N.; Gilbert, L.; Weed, J.

    2014-01-01

    The CANDU Owners Group (COG) developed a set of generic and initial station-specific Severe Accident Management Guidance (SAMG) documents to mitigate the consequences to the public in the event of a severe accident. The generic portion of the COG SAMG was completed in 2006; the overall project including the station-specific phase was completed in April 2007. Over the years, the CANDU industry and utilities have continuously increased the knowledge base for SAMG and have incorporated various engineered features based on the knowledge obtained. As a result of the event that occurred at the Fukushima Daiiachi nuclear power plant (NPP) in Japan, the Canadian Nuclear Safety Commission (CNSC) established the CNSC Fukushima Task Force. The results of the task force were documented in INFO-0828, CNSC Staff Action Plan on the CNSC Fukushima Task Force Recommendations. Among the recommendation documented in INFO-828 were Fukushima Action Items (FAIs) directed towards the CANDU utilities in Canada; a portion of which are related to SAMG documentation updates and directed at enhancing SAM response. A COG joint project was established to support the closure of the CNSC FAIs and to revise the current CANDU documentation accordingly. This paper provides a high level summary of the COG project scope and results. It also demonstrates that the CANDU SAMG programs in Canada provide robust protection and mitigation of severe accidents. (author)

  13. CANDU severe accident management guidance update

    Energy Technology Data Exchange (ETDEWEB)

    Jones, L., E-mail: lisa.m.jones@opg.com [Ontario Power Generation, Pickering, ON (Canada); Popov, N., E-mail: nik.popov@rogers.com [Candu Owners Group, Toronto, ON (Canada); Gilbert, L., E-mail: lovell.gilbert@brucepower.com [Bruce Power, Tiverton, ON (Canada); Weed, J., E-mail: jeff.weed@candu.gov [Candu Owners Group, Toronto, ON (Canada)

    2014-07-01

    The CANDU Owners Group (COG) developed a set of generic and initial station-specific Severe Accident Management Guidance (SAMG) documents to mitigate the consequences to the public in the event of a severe accident. The generic portion of the COG SAMG was completed in 2006; the overall project including the station-specific phase was completed in April 2007. Over the years, the CANDU industry and utilities have continuously increased the knowledge base for SAMG and have incorporated various engineered features based on the knowledge obtained. As a result of the event that occurred at the Fukushima Daiiachi nuclear power plant (NPP) in Japan, the Canadian Nuclear Safety Commission (CNSC) established the CNSC Fukushima Task Force. The results of the task force were documented in INFO-0828, CNSC Staff Action Plan on the CNSC Fukushima Task Force Recommendations. Among the recommendation documented in INFO-828 were Fukushima Action Items (FAIs) directed towards the CANDU utilities in Canada; a portion of which are related to SAMG documentation updates and directed at enhancing SAM response. A COG joint project was established to support the closure of the CNSC FAIs and to revise the current CANDU documentation accordingly. This paper provides a high level summary of the COG project scope and results. It also demonstrates that the CANDU SAMG programs in Canada provide robust protection and mitigation of severe accidents. (author)

  14. Mathematical modeling of CANDU-PHWR

    Energy Technology Data Exchange (ETDEWEB)

    Gaber, F.A.; Aly, R.A.; El-Shal, A.O. [Atomic Energy Authority, Cairo (Egypt)

    2001-07-01

    The paper deals with the transient studies of CANDU 600 pressurized Heavy Water Reactor (PHWR) system. This study involved mathematical modeling of CANDU PHWR major system components and the developments of software to study the thermodynamic performances. Modeling of CANDU-PHWR was based on lumped parameter technique.The integrated CANDU-PHWR model includes the neutronic, reactivity, fuel channel heat transfer, piping and the preheater type U-tube steam generator (PUTSG). The nuclear reactor power was modelled using the point kinetics equations with six groups of delayed neutrons and reactivity feed back due to the changes in fuel temperature and coolant temperature. The complex operation of the preheater type U-tube steam generator (PUTSG) is represented by a non-linear dynamic model using a state variable, moving boundary and lumped parameter techniques. The secondary side of the PUTSG model has six separate lumps including a preheater region, a lower boiling section, a mixing region, a riser, a chimmeny section, and a down-corner. The tube side of PUTSG has three main thermal zones. The PUTSG model is based on conservation of mass, energy and momentum relation-ships. The CANDU-PHWR integrated model are coded in FORTRAN language and solved by using a standard numerical technique. The adequacy of the model was tested by assessing the physical plausibility of the obtained results. (author)

  15. Corrosion control in CANDU nuclear power reactors

    International Nuclear Information System (INIS)

    Lesurf, J.E.

    1974-01-01

    Corrosion control in CANDU reactors which use pressurized heavy water (PHW) and boiling light water (BLW) coolants is discussed. Discussions are included on pressure tubes, primary water chemistry, fuel sheath oxidation and hydriding, and crud transport. It is noted that corrosion has not been a significant problem in CANDU nuclear power reactors which is a tribute to design, material selection, and chemistry control. This is particularly notable at the Pickering Nuclear Generating Station which will have four CANDU-PHW reactors of 540 MWe each. The net capacity factor for Pickering-I from first full power (May 1971) to March 1972 was 79.5 percent, and for Pickering II (first full power November 1971) to March 1972 was 83.5 percent. Pickering III has just reached full power operation (May 1972) and Pickering IV is still under construction. Gentilly CANDU-BLW reached full power operation in May 1972 after extensive commissioning tests at lower power levels with no major corrosion or chemistry problems appearing. Experience and operating data confirm that the value of careful attention to all aspects of corrosion control and augur well for future CANDU reactors. (U.S.)

  16. Safety research for CANDU reactors

    International Nuclear Information System (INIS)

    Hancox, W.T.

    1982-10-01

    Continuing research to develop and verify computer models of CANDU-PHW reactor process and safety systems is described. It is focussed on loss-of-coolant accidents (LOCAs) because they are the precursors of more serious accidents. Research topics include: (i) fluid-dynamic and heat-transfer processes in the heat transport system during the blowdown and refilling phases of LOCAs; (ii) thermal and mechanical behaviour of fuel elements; (iii) thermal and mechanical behaviour of the fuel and the fuel-channel assembly in situations where the heavy-water moderator is the sink for decay heat produced in the fuel; (iv) chemical behaviour of fission gases that might be released into the reactor coolant and transported to the containment system; and (v) combustion of hydrogen-air-steam mixtures that would be produced if fuel temperatures were sufficiently high to initiate the zirconium-water reaction. The current status of the research on each of these topics is highlighted with particular emphasis on the conclusions reached to date and their impact on the continuing program

  17. Moderator circulation in CANDU reactors

    International Nuclear Information System (INIS)

    Fath, H.E.S.; Hussein, M.A.

    1989-01-01

    A two-dimensional computer code that is capable of predicting the moderator flow and temperature distribution inside CANDU calandria is presented. The code uses a new approach to simulate the calandria tube matrix by blocking the cells containing the tubes in the finite difference mesh. A jet momentum-dominant flow pattern is predicted in the nonisothermal case, and the effect of the buoyancy force, resulting from nuclear heating, is found to enhance the speed of circulation. Hot spots are located in low-velocity areas at the top of the calandria and below the inlet jet level between the fuel channels. A parametric study is carried out to investigate the effect of moderator inlet velocity,moderator inlet nozzle location, and geometric scaling. The results indicate that decreasing the moderator inlet velocity has no significant influence on the general features of the flow pattern (i.e., momentum dominant); however, too many high-temperature hot spots appear within the fuel channels

  18. CANDU fuel performance and development

    International Nuclear Information System (INIS)

    Hardy, D.G.; Wood, J.C.; Bain, A.S.

    1978-12-01

    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)

  19. Fuel deposits, chemistry and CANDU reactor operation

    International Nuclear Information System (INIS)

    Roberts, J.G.

    2013-01-01

    'Hot conditioning' is a process which occurs as part of commissioning and initial start-up of each CANDU reactor, the first being the Nuclear Power Demonstration-2 reactor (NPD). Later, understanding of the cause of the failure of the Pickering Unit 1 G16 fuel channel led to a revised approach to 'hot conditioning', initially demonstrated on Bruce Unit 5, and subsequently utilized for each CANDU unit since. The difference being that during 'hot conditioning' of CANDU heat transport systems fuel was not in-core until Bruce Unit 5. The 'hot conditioning' processes will be briefly described along with the consequences to fuel. (author)

  20. Neutronic parameters calculations of a CANDU reactor

    International Nuclear Information System (INIS)

    Zamonsky, G.

    1991-01-01

    Neutronic calculations that reproduce in a simplified way some aspects of a CANDU reactor design were performed. Starting from some prefixed reactor parameters, cylindrical and uniform iron adjuster rods were designed. An appropriate refueling scheme was established, defininig in a 2 zones model their dimensions and exit burnups. The calculations have been done using the codes WIMS-D4 (cell), SNOD (reactivity device simulations) and PUMA (reactor). Comparing with similar calculations done with codes and models usually employed for CANDU design, it is concluded that the models and methods used are appropriate. (Author) [es

  1. Used fuel packing plant for CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Menzies, I.; Thayer, B.; Bains, N., E-mail: imenzies@atsautomation.com [ATS Automation, Cambridge, ON (Canada); Murchison, A., E-mail: amurchison@nwmo.ca [NWMO, Toronto, ON (Canada)

    2015-07-01

    Large forgings have been selected to containerize Light Water Reactor used nuclear fuel. CANDU fuel, which is significantly smaller in size, allows novel approaches for containerization. For example, by utilizing commercially available extruded ASME pipe a conceptual design of a Used Fuel Packing Plant for containerization of used CANDU fuel in a long lived metallic container has been developed. The design adopts a modular approach with multiple independent work cells to transfer and containerize the used fuel. Based on current technologies and concepts from proven industrial systems, the Used Fuel Packing Plant can assemble twelve used fuel containers per day considering conservative levels of process availability. (author)

  2. Establishment of Safety Analysis System and Technology for CANDU Reactors

    International Nuclear Information System (INIS)

    Park, Joo Hwan; Rhee, B. W.; Min, B. J.; Kim, H. T.; Kim, W. Y.; Yoon, C.; Chun, J. S.; Cho, M. S.; Jeong, J. Y.; Kang, H. S.

    2007-06-01

    The following 4 research items have been studied to establish a CANDU safety analysis system and to develop the relevant elementary technology for CANDU reactors. First, to improve and validate the CANDU design and operational safety analysis codes, the CANDU physics cell code WIMS-CANDU was improved, and validated, and an analysis of the moderator subcooling and pressure tube integrity has been performed for the large break LOCAs without ECCS. Also a CATHENA model and a CFD model for a post-blowdown fuel channel analysis have been developed and validated against two high temperature thermal-chemical experiments, CS28-1 and 2. Second, to improve the integrated operating system of the CANDU safety analysis codes, an extension has been made to them to include the core and fuel accident analyses, and a web-based CANDU database, CANTHIS version 2.0 was completed. Third, to assess the applicability of the ACR-7 safety analysis methodology to CANDU-6 the ACR-7 safety analysis methods were reviewed and the safety analysis methods of ACR-7 applicable to CANDU-6 were recommended. Last, to supplement and improve the existing CANDU safety analysis procedures, detailed analysis procedures have been prepared for individual accident scenarios. The results of this study can be used to resolve the CANDU safety issues, to improve the current design and operational safety analysis codes, and to technically support the Wolsong site to resolve their problems

  3. CANDU 6 - the highly successful medium sized reactor

    International Nuclear Information System (INIS)

    Hedges, K. R.; Allen, P. J.; Hopwood, J. M.

    2000-01-01

    The CANDU 6 Pressurized Heavy Water Reactor system, featuring horizontal fuel channels and heavy water moderator continues to evolve, supported by AECL's strong commitment to comprehensive R and D programs. The initial CANDU 6 design started in the 1970's. The first plants went into service in 1983, and the latest version of the plant is under construction in China. With each plant the technology has evolved giving the dual advantages of proveness and modern technology. CANDU 6 delivers important advantages of the CANDU system with benefit to small and medium-sized grids. This technology has been successfully adopted by, and localized to varying extents in, each of the CANDU 6 markets. For example, all CANDU owners obtain their fuel from domestic suppliers. Progressive CANDU development continues at AECL to enhance this medium size product CANDU 6. There are three key CANDU development strategic thrusts: improved economics, fuel cycle flexibility, and enhanced safety. The CANDU 6 product is also enhanced by incorporating 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. (author)

  4. CANDU flexible and economical fuel technology in China

    Energy Technology Data Exchange (ETDEWEB)

    Mingjun, C. [CNNC Nuclear Power Operation Management Co., Zhejiang (China); Zhenhua, Z.; Zhiliang, M. [CNNC Third Qinshan Nuclear Power Co., Zhejiang (China); Cottrell, C.M.; Kuran, S. [Candu Energy Inc., Mississauga, ON (Canada)

    2014-07-01

    Use in CANDU reactor is one good option of recycled uranium (RU) and thorium (Th) resource. It is also good economy to CANDU fuel. Since 2008 Qinshan CANDU Plant and our partners (Candu Energy and CNNC and NPIC) have made great efforts to develop the engineering technologies of Flexible and Economical Fuel (RU and Th) in CANDU type reactor and finding the CANDU's position in Chinese closed fuel cycle (CFC) system. This paper presents a proposal of developing strategy and implementation plan. Qinshan CANDU reactors will be converted to use recycled and depleted uranium based fuels, a first-of-its-kind. The fuel is composed of both recycled and depleted uranium and simulating natural uranium behavior. This paper discusses its development, design, manufacture and verification tested with success and the full core implementation plan by the end of 2014. (author)

  5. The CANDU 9 distributed control system design process

    International Nuclear Information System (INIS)

    Harber, J.E.; Kattan, M.K.; Macbeth, M.J.

    1997-01-01

    Canadian designed CANDU pressurized heavy water nuclear reactors have been world leaders in electrical power generation. The CANDU 9 project is AECL's next reactor design. Plant control for the CANDU 9 station design is performed by a distributed control system (DCS) as compared to centralized control computers, analog control devices and relay logic used in previous CANDU designs. The selection of a DCS as the platform to perform the process control functions and most of the data acquisition of the plant, is consistent with the evolutionary nature of the CANDU technology. The control strategies for the DCS control programs are based on previous CANDU designs but are implemented on a new hardware platform taking advantage of advances in computer technology. This paper describes the design process for developing the CANDU 9 DCS. Various design activities, prototyping and analyses have been undertaken in order to ensure a safe, functional, and cost-effective design. (author)

  6. CANDU-9/480-SEU fuel handling system assessment document

    International Nuclear Information System (INIS)

    Hwang, Jeong Ki; Jo, C. H.; Kim, H. M.; Morikawa, D. T.

    1996-11-01

    This report summarize the rationale for the CANDU 9 fuel handling system, and the design choices recommended for components of the system. Some of the design requirements applicable to the CANDU 9 480-SEU fuel handling design choices are described. These requirements imposed by the CANDU 9 project. And the design features for the key components of fuel handling system, such as the fuelling machine, the carriage, the new fuel transfer system and the irradiated fuel transfer system, are described. The carriage seismic load evaluations relevant to the design are contained in the appendices. The majority of the carriage components are acceptable, or will likely be acceptable with some redesign. The concept for the CANDU 9 fuel handling system is based on proven CANDU designs, or on improved CANDU technology. Although some development work must be done, the fuel handling concept is judged to be feasible for the CANDU 9 480-SEU reactor. (author). 2 refs

  7. Enhancement of safety analysis reliability for a CANDU-6 reactor using RELAP-CANDU/SCAN coupled code system

    International Nuclear Information System (INIS)

    Kim, Man Woong; Choi, Yong Seog; Sin, Chul; Kim, Hyun Koon; Kim, Hho Jung; Hwang, Su Hyun; Hong, In Seob; Kim, Chang Hyo

    2005-01-01

    In LOCA analysis of the CANDU reactor, the system thermal-hydraulic code, RELAP-CANDU, alone cannot predict the transient behavior accurately. Therefore, the best estimate neutronics and system thermal-hydraulic coupled code system is necessary to describe the transient behavior with higher accuracy and reliability. To perform on-line calculation of safety analysis for CANDU reactor, a coupled thermal hydraulics-neutronics code system was developed in such a way that the best-estimate thermal-hydraulic system code for CANDU reactor, RELAP-CANDU, is coupled with the full three-dimensional reactor core kinetic code

  8. The CANTEACH project: preserving CANDU technical knowledge

    International Nuclear Information System (INIS)

    Garland, B.; Kosarenko, Y.; Meneley, D.

    2003-01-01

    Almost sixty years have passed since the nuclear energy venture began in Canada. Fifty years have passed since the founding of AECL. Tens of thousands of dedicated people have forged a new and successful primary energy supply. CANDU technology is well into its second century. This specialty within the world's fission technology community is quite unique, first because it was established as a separate effort very early in the history of world fission energy, and second because it grew in an isolated environment, with tight security requirements, in its early years. Commercial security rules later sustained a considerable degree of isolation. The pioneers of CANDU development have finished their work. Most of the second generation also has moved on. As yet, we cannot point to a consistent and complete record of this remarkable achievement. We, as a nuclear enterprise, have not captured the design legacy in a form that is readily accessible to the current and future generation of professionals involved with CANDU reactors, be they students, designers, operations staff, regulators, consultants or clients. This is a serious failure. Young people entering our field of study must make do with one or two textbooks and a huge collection of diverse technical papers augmented by limited-scope education and training materials. Those employed in the various parts of the nuclear industry rely mostly on a smaller set of CANDU- related documents available within their own organization; documents that sometimes are rather limited in scope. University professors often have even more limited access to in-depth and up to date information. In fact, they often depend on literature published in other countries when preparing lectures, enhanced by guest lecturers from various parts of the industry. Because CANDU was developed mostly inside Canada, few of these text materials contain useful data describing processes important to the CANDU system. For many years it has been recognized that

  9. Integrated risk assessment for spent fuel transportation using developed software

    International Nuclear Information System (INIS)

    Yun, Mi Rae; Christian, Robby; Kim, Bo Gyung; Almomani, Belal; Ham, Jae Hyun; Kang, Gook Hyun; Lee, Sang hoon

    2016-01-01

    As on-site spent fuel storage meets limitation of their capacity, spent fuel need to be transported to other place. In this research, risk of two ways of transportation method, maritime transportation and on-site transportation, and interim storage facility were analyzed. Easier and integrated risk assessment for spent fuel transportation will be possible by applying this software. Risk assessment for spent fuel transportation has not been researched and this work showed a case for analysis. By using this analysis method and developed software, regulators can get some insights for spent fuel transportation. For example, they can restrict specific region for preventing ocean accident and also they can arrange spend fuel in interim storage facility avoiding most risky region which have high risk from aircraft engine shaft. Finally, they can apply soft material on the floor for specific stage for on-site transportation. In this software, because we targeted Korea, we need to use Korean reference data. However, there were few Korean reference data. Especially, there was no food chain data for Korean ocean. In MARINRAD, they used steady state food chain model, but it is far from reality. Therefore, to get Korean realistic reference data, dynamic food chain model for Korean ocean need to be developed

  10. Integrated risk assessment for spent fuel transportation using developed software

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Mi Rae; Christian, Robby; Kim, Bo Gyung; Almomani, Belal; Ham, Jae Hyun; Kang, Gook Hyun [KAIST, Daejeon (Korea, Republic of); Lee, Sang hoon [Keimyung University, Daegu (Korea, Republic of)

    2016-05-15

    As on-site spent fuel storage meets limitation of their capacity, spent fuel need to be transported to other place. In this research, risk of two ways of transportation method, maritime transportation and on-site transportation, and interim storage facility were analyzed. Easier and integrated risk assessment for spent fuel transportation will be possible by applying this software. Risk assessment for spent fuel transportation has not been researched and this work showed a case for analysis. By using this analysis method and developed software, regulators can get some insights for spent fuel transportation. For example, they can restrict specific region for preventing ocean accident and also they can arrange spend fuel in interim storage facility avoiding most risky region which have high risk from aircraft engine shaft. Finally, they can apply soft material on the floor for specific stage for on-site transportation. In this software, because we targeted Korea, we need to use Korean reference data. However, there were few Korean reference data. Especially, there was no food chain data for Korean ocean. In MARINRAD, they used steady state food chain model, but it is far from reality. Therefore, to get Korean realistic reference data, dynamic food chain model for Korean ocean need to be developed.

  11. Economic case for CANDU life extension projects

    International Nuclear Information System (INIS)

    Qureshi, S.; Tenev, T.; Lewi, M.

    2014-01-01

    As CANDU reactors approach their original end of design life utilities are faced with two options: to extend the operating life of the reactor by undergoing a life extension project (LEP), or to commence decommissioning activities. Recent project experience has shown that there is economic merit in extending the life of the operating reactor. There are many benefits to such a decision, the most obvious being the revenue that will be generated from the additional years of electricity production by the utility. Delays in decommissioning are also advantageous since the large costs associated with such a long-term activity are pushed into the future, thereby decreasing the net present value (NPV) of the investment. In addition, relatively few power reactors have been fully decommissioned to date and deferring this activity transfers the associated risks to others that are currently obligated to undertake decommissioning activities sooner. Candu Energy has been involved with the life extension projects of the following CANDU reactors: Point Lepreau (New Brunswick, Canada), Bruce Unit 1 and Unit 2 (Ontario, Canada), and Wolsong Unit 1 (South Korea). These reactors underwent fuel channel replacement programs in addition to replacement of major reactor components. Most recently, both Ontario Power Generation (OPG) and Nucleoelectrica Argentina Sociedad Anonima (NA-SA) have commenced work on life extension projects at the Darlington (Canada) and Embalse (Argentina) sites respectively. The experience gained from previous LEP projects allows Candu Energy to deliver future projects in a timely, efficient, and cost effective manner. (author)

  12. CANDU 9 operator plant display system

    International Nuclear Information System (INIS)

    Trueman, R.; Webster, A.; MacBeth, M.J.

    1997-01-01

    To meet evolving client and regulatory needs, AECL has adopted an evolutionary approach to the design of the CANDU 9 control centre. That is, the design incorporates feedback from existing stations, reflects the growing diversity in the roles and responsibilities of the operating staff, and reduces costs associated with plant capital and operations, maintenance and administration (OM and A), through the appropriate introduction of new technologies. Underlying this approach is a refined engineering design process that cost-effectively integrates operational feedback and human factors engineering to define the operating staff information and information presentation requirements. Based on this approach, the CANDU 9 control centre will provide utility operating staff with the means to achieve improved operations and reduced OM and A costs. One of the design features that will contribute to the improved operational capabilities of the control centre is a new Plant Display System (PDS) that is separate from the digital control system. The PDS will be used to implement non-safety panel, and console video display systems within the CANDU 9 main control room (MCR). This paper presents a detailed description of the CANDU 9 Plant Display System and features that provide increased operational capabilities. (author)

  13. CANDU-6 fuel optimization for advanced cycles

    Energy Technology Data Exchange (ETDEWEB)

    St-Aubin, Emmanuel, E-mail: emmanuel.st-aubin@polymtl.ca; Marleau, Guy, E-mail: guy.marleau@polymtl.ca

    2015-11-15

    Highlights: • New fuel selection process proposed for advanced CANDU cycles. • Full core time-average CANDU modeling with independent refueling and burnup zones. • New time-average fuel optimization method used for discrete on-power refueling. • Performance metrics evaluated for thorium-uranium and thorium-DUPIC cycles. - Abstract: We implement a selection process based on DRAGON and DONJON simulations to identify interesting thorium fuel cycles driven by low-enriched uranium or DUPIC dioxide fuels for CANDU-6 reactors. We also develop a fuel management optimization method based on the physics of discrete on-power refueling and the time-average approach to maximize the economical advantages of the candidates that have been pre-selected using a corrected infinite lattice model. Credible instantaneous states are also defined using a channel age model and simulated to quantify the hot spots amplitude and the departure from criticality with fixed reactivity devices. For the most promising fuels identified using coarse models, optimized 2D cell and 3D reactivity device supercell DRAGON models are then used to generate accurate reactor databases at low computational cost. The application of the selection process to different cycles demonstrates the efficiency of our procedure in identifying the most interesting fuel compositions and refueling options for a CANDU reactor. The results show that using our optimization method one can obtain fuels that achieve a high average exit burnup while respecting the reference cycle safety limits.

  14. CANDU 6 probabilistic safety study summary

    International Nuclear Information System (INIS)

    1988-07-01

    This report summarizes the methodology, phenomenology and results relevent to the assessment of severe events in a CANDU 6 (formerly designated CANDU 600) station. The station design being analysed is based on a CANDU 6 Mark I currently operating in Canada. This evaluation includes event frequency and fission product release assessments but does not include assessment of radiation dose to the public, so that the information is equivalent to a level 2 Probabilistic Risk Assessment (PRA). The study has shown that the predicted overall average frequency for core melt in a CANDU 6 Mark I is 4.4 x 10 -6 events/year. This low frequency is, in large part due to the heavy water moderator which acts as a heat sink, prevents UO 2 melting and maintains core geometry for many events which could otherwise result in a core melt. The consequences for most core melts will be limited to the release of a fraction of noble gases and organic iodides. Other isotopes will be condensed or dissolved in the containment atmosphere and are ultimately retained in the pool of water in the basement where they are unavailable for release. Most core melts (∼ 90%) can be mitigated by operator action so that there is no danger of consequential damage to the containment structure and leak tightness. The frequency and consequences of less likely, more severe core melt sequences are also discussed in this report and shown to be small contributors to public risk

  15. 3D CAD ON Qinshan CANDU project

    International Nuclear Information System (INIS)

    Goland, D.

    2000-01-01

    This paper briefly describe AECL's work in applying computer-aided engineering tools to the Qinshan CANDU project. The main thrust of this paper is to introduce the major CAD software tools and their use in civil design, process design and EI and C design. Other special software tools and non-CAD tools and their applications are also briefly introduced. (author)

  16. CANDU fuel : design/manufacturing interaction

    International Nuclear Information System (INIS)

    Graham, N.A.

    1999-01-01

    The design of CANDU fuel has been the product of intense cooperation among fuel designers and fuel manufacturers. The developments of some of the novel processes in fuel manufacture are outlined. These include the brazed-split-spacer design, the resistance welded endcap and CANLUB coatings. (author)

  17. Exporting apocalypse: CANDU reactors and nuclear proliferation

    International Nuclear Information System (INIS)

    McKay, Paul.

    The author believes that the peaceful use of nuclear technology leads inevitably to the production of nuclear weapons, and that CANDU reactors are being bought by countries that are likely to build bombs. He states that exports of reactors and nuclear materials cannot be defended and must be stopped

  18. Wet steam turbines for CANDU-Reactors

    International Nuclear Information System (INIS)

    Westmacott, C.H.L.

    1977-01-01

    The technical characteristics of 4 wet steam turbine aggregates used in the Pickering nuclear power station are reported on along with operational experience. So far, the general experience was positive. Furthermore, plans are mentioned to use this type of turbines in other CANDU reactors. (UA) [de

  19. Conditioning experience for spent radium sources

    International Nuclear Information System (INIS)

    Kang, I. S.; Shon, J. S.; Kim, K. J.; Min, D. K.

    2001-01-01

    In order to avoid accidents that could be resulted from improper storage of spent radium sources, it is necessary to condition and store them safely. The program for safe conditioning of spent radium sources by IAEA has been established to assist the developing countries. The main object of this paper is to apply the technology that was adapted by IAEA for the conditioning the national inventory of Ra-226 sources in member states, as a part of IAEA's project with the Korean expert team. This paper is the result that the Korean expert team carried out spent radium conditioning, under the project title 'Radium Conditioning in Myanmar(INT4131-06646C)'. The whole inventory of spent radium sources 1,429.5mCi, was safely conditioned by the Korean expert team according to the manual under the supervision of IAEA's technical officer and the control of Myanmar authority on behalf of Myanmar. These sources were encapsuled and welded into 27 small capsules and 3 large capsules, and conditioned in 3 lead shields, producing 3 concrete-shielded drums. The inventories were distributed into 3 shielding devices, holding 500mCi, 459.5mCi, and 470mCi

  20. A scheme of better utilization of PWR spent fuels

    International Nuclear Information System (INIS)

    Chung, Bum Jin; Kang, Chang Soon

    1991-01-01

    The recycle of PWR spent fuels in a CANDU reactor, so called the tandem fuel cycle is investigated in this study. This scheme of utilizing PWR spent fuels will ease the shortage of spent fuel storage capacity as well as will improve the use of uranium resources. The minimum modification the design of present CANDU reactor is seeked in the recycle. Nine different fuel types are considered in this work and are classified into two categories: refabrication and reconfiguration. For refabrication, PWR spent fuels are processed and refabricated into the present 37 rod lattice structure of fuel bundle, and for reconfiguration, meanwhile, spent fuels are simply disassembled and rods are cut to fit into the present grid configuration of fuel bundle without refabrication. For each fuel option, the neutronics calculation of lattice was conducted to evaluate the allowable burn up and distribution. The fuel cycle cost of each option was also computed to assess the economic justification. The results show that most tandem fuel cycle option considered in this study are technically feasible as well as economically viable. (Author)

  1. The security management of spent filter cartridge in Qinshan phase 3 (heavy water reactor) nuclear power plant

    International Nuclear Information System (INIS)

    Xue Dahai

    2005-01-01

    Qinshan phase 3 nuclear power plant is the first CANDU plant that China fetched in from Canada, and both two units operate under well condition up to now. The radioactive wastes produced during the unit operation mainly include technical waste, spent resin, and spent filter cartridge. The spent filter cartridge is one important part both in the volume and radioactivity of the radioactive waste, and it is the important content of radioactive waste management. Different from PWR, part of high radioactive spent filter in CANDU unit comes from heavy water system such as moderator system. It has to be dried through blowing before replaced from the system. But this working procedure result the filtrate dreg become flexible, and it can bring on the risk of internal or external exposure. It is very important to pay high attention to control the contamination spread during spent filter inside transfer. (authors)

  2. Development situation about the Canadian CANDU Nuclear Power Generating Stations

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Yu Mi; Kim, Yong Hee; Park, Joo Hwan

    2009-07-15

    The CANDU reactor is the most versatile commercial power reactor in the world. The acronym 'CANDU', a registered trademark of Atomic Energy of Canada Limited, stands for 'CANada Deuterium Uranium'. CANDU uses heavy water as moderator and uranium (originally, natural uranium) as fuel. All current power reactors in Canada are of the CANDU type. Canada exports CANDU type reactor in abroad. CANDU type is used as the nuclear power plants to produce electrical. Today, there are 41 CANDU reactors in use around the world, and the design has continuously evolved to maintain into unique technology and performance. The CANDU-6 power reactor offers a combination of proven, superior and state-of-the-art technology. CANDU-6 was designed specifically for electricity production, unlike other major reactor types. One of its characteristics is a very high operating and fuel efficiency. Canada Nuclear Power Generating Stations were succeeded in a commercial reactor of which the successful application of heavy water reactor, natural uranium method and that on-power fuelling could be achieved. It was achieved through the joint development of a major project by strong support of the federal government, public utilities and private enterprises. The potential for customization to any country's needs, with competitive development and within any level of domestic industrial infrastructure, gives CANDU technology strategic importance in the 21st century.

  3. Development situation about the Canadian CANDU Nuclear Power Generating Stations

    International Nuclear Information System (INIS)

    Jeon, Yu Mi; Kim, Yong Hee; Park, Joo Hwan

    2009-07-01

    The CANDU reactor is the most versatile commercial power reactor in the world. The acronym 'CANDU', a registered trademark of Atomic Energy of Canada Limited, stands for 'CANada Deuterium Uranium'. CANDU uses heavy water as moderator and uranium (originally, natural uranium) as fuel. All current power reactors in Canada are of the CANDU type. Canada exports CANDU type reactor in abroad. CANDU type is used as the nuclear power plants to produce electrical. Today, there are 41 CANDU reactors in use around the world, and the design has continuously evolved to maintain into unique technology and performance. The CANDU-6 power reactor offers a combination of proven, superior and state-of-the-art technology. CANDU-6 was designed specifically for electricity production, unlike other major reactor types. One of its characteristics is a very high operating and fuel efficiency. Canada Nuclear Power Generating Stations were succeeded in a commercial reactor of which the successful application of heavy water reactor, natural uranium method and that on-power fuelling could be achieved. It was achieved through the joint development of a major project by strong support of the federal government, public utilities and private enterprises. The potential for customization to any country's needs, with competitive development and within any level of domestic industrial infrastructure, gives CANDU technology strategic importance in the 21st century

  4. AECL's advanced CANDU reactor - the ACR

    International Nuclear Information System (INIS)

    Alizadeh, Ala; Allsop, Peter; Hedges, Ken; Hopwood, Jerry; Yu, Stephen

    2003-01-01

    The ACR, the next generation CANDU design, represents the next step in development of the CANDU family of designs. AECL has achieved significant incremental improvements to the mid-size CANDU 6 nuclear power plant through successive projects, both in design and in project delivery. Building on this knowledge base, AECL is continuing to adapt the CANDU design to develop the ACR. This paper summarizes the ACR design features, which include major improvements in economics, inherent safety characteristics, performance and construction methods. Aimed at producing electrical power at a capital cost significantly less than that of the current reactor designs, the ACR is an evolutionary design based on the very successful CANDU 6 reactor. The new ACR product is specifically designed to produce power at a cost competitive with other forms of power generation while achieving short construction times, improved safety, international licensability, high investor returns, and low investor risk. It achieves these targets by taking advantage of the latest advances in both pressure-tube and pressure-vessel reactor technologies and experience. The flexibility and development potential of the fuel channel approach also enables designs to be developed that address priorities identified in international long-term specification programs such as the US Department of Energy (DOE) sponsored Generation IV program and IAEA hosted INPRO program. ACR-700 can be built in 36 months with a 48 month project duration, and deliver a lifetime capacity factor in excess of 90%. Overall, the ACR design represents a balance of proven design basis and innovations to give step improvements in safety, reliability and economics. The ACR development program, now well into the detail design stage, includes parallel formal licensing in the USA and Canada. Based on the status of the ACR design and AECL's on-going experience delivering reactor projects on-time and on-budget, the first ACR could be in service by

  5. CANDU fuel behaviour under LOCA conditions

    International Nuclear Information System (INIS)

    Kohn, E.

    1989-07-01

    This report summarizes the current understanding of CANDU fuel-element behaviour under loss-of-coolant (LOCA) accidents. It focuses on a key in-reactor verification experiment conducted at Idaho National Engineering Laboratory (INEL) and on three Canadian in-reactor tests. The in-reactor data, and the considerable body of supporting information developed from out-reactor tests, support the general conclusion that CANDU fuel behaviour during LOCA transients is well understood. Four elements of 37-element CANDU fuel-bundle design were tested under conditions typical of a large-break LOCA blowdown in a CANDU reactor. The purpose of the test was to confirm our current understanding of fuel behaviour under loss-of-coolant accident blowdown conditions. The test also provided data for comparison with predictions made with the steady-state and transient fuel-element performance codes ELESIM and ELOCA. Key components of typical LOCA transients were incorporated in the test: namely, a rapid depressurization rate of the hot coolant, a simultaneous power increase before decreasing to decay values (a power pulse), and prototype fuel element under pre-transient power and burnup conditions. The test was successfully completed in the Power Burst Facility (PBF) reactor at INEL under contract to Ontario Hydro and AECL. The three CANDU Owners Group LOCA tests performed at Chalk River Nuclear Laboratories measured both the thermal-mechanical response and fission-gas release resulting from exposure to a LOCA transient. Results from these three tests provided further confirmation that the behaviour of the fuel under LOCA conditions is understood

  6. Candu reactors with thorium fuel cycles

    International Nuclear Information System (INIS)

    Hopwood, J.M.; Fehrenbach, P.; Duffey, R.; Kuran, S.; Ivanco, M.; Dyck, G.R.; Chan, P.S.W.; Tyagi, A.K.; Mancuso, C.

    2006-01-01

    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 550 0 C. 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

  7. CANDU 9 nuclear power plant simulator

    International Nuclear Information System (INIS)

    Kattan, M.; MacBeth, M.J.; Lam, K.

    1995-01-01

    Simulators are playing, an important role in the design and operations of CANDU reactors. They are used to analyze operating procedures under standard and upset conditions. The CANDU 9 nuclear power plant simulator is a low fidelity, near full scope capability simulator. It is designed to play an integral part in the design and verification of the control centre mock-up located in the AECL design office. It will also provide CANDU plant process dynamic data to the plant display system (PDS), distributed control system (DCS) and to the mock-up panel devices. The simulator model employs dynamic mathematical models of the various process and control components that make up a nuclear power plant. It provides the flexibility to add, remove or update user supplied component models. A block oriented process input is provided with the simulator. Individual blocks which represent independent algorithms of the model are linked together to generate the required overall plant model. As a design tool the simulator will be used for control strategy development, human factors studies (information access, readability, graphical display design, operability), analysis of overall plant control performance, tuning estimates for major control loops and commissioning strategy development. As a design evaluation tool, the simulator will be used to perform routine and non-routine procedures, practice 'what if' scenarios for operational strategy development, practice malfunction recovery procedures and verify human factors activities. This paper will describe the CANDU 9 plant simulator and demonstrate its implementation and proposed utility as a tool in the control system and control centre design of a CANDU 9 nuclear power plant. (author). 2 figs

  8. Role of operator response guidelines in CANDU 9 design program

    International Nuclear Information System (INIS)

    Jaitly, R.K.

    2000-01-01

    The CANDU 9 is a large version of the CANDU Pressurized Heavy Water Reactor (PHWR) system developed in Canada. With an electrical output of approximately 935 MWe, the CANDU 9 complements the established mid-size CANDU 6 (700 MWe) and makes use of proven technology updated with state of the art features resulting from ongoing development. The CANDU 9 builds on the reactor and process system designs of the operating Darlington and Bruce B plants, and incorporates a modified CANDU 6 station layout, as well as improved construction methods and operational features. A high level of standardization has always been a feature of CANDU reactors. This theme is emphasized in the CANDU 9; all key components (reactor core, steam generators, coolant pumps, pressure tubes, etc.) are of the same design as those proven in service in the operating CANDU power stations. Including Probabilistic Safety Assessment (PSA) as part of the CANDU 9 design process from the outset of the program was seen as key to ensuring completeness of safety related requirements. The PSA work provided an in-depth understanding of the plant response to various postulated accidents. As well, the time frame for recovery and the related operator actions were identified. This information together with AECL's experience in supporting the development of Emergency Operating Procedures (EOPs) for the operating CANDU reactors are the basis for preparation of CANDU 9 Operator Response Guidelines (ORGs). Technical content, format and human factors considerations adopted for the ORGs are such that these can be readily converted to EOPs. The scope of ORGs includes generic as well as event specific ORGs. This dual approach is required to provide defense-in-depth. This paper describes the process used to prepare ORGs for the CANDU 9 reactor and discusses important benefits gained from the application of ORGs as input to the control center design and future preparation of the EOPs. (author)

  9. CANDU technology for generation III + AND IV reactors

    International Nuclear Information System (INIS)

    Torgerson, D.F.

    2005-01-01

    Atomic Energy of Canada Limited (AECL) is the original developer of the CANDU?reactor, one of the three major commercial power reactor designs now used throughout the world. For over 60 years, AECL has continued to evolve the CANDU design from the CANDU prototypes in the 1950s and 1960s through to the second generation reactors now in operation, including the Generation II+ CANDU 6. The next phase of this evolution, the Generation III+ Advanced CANDU ReactorTM (ACRTM), continues the strategy of basing next generation technology on existing CANDU reactors. Beyond the ACR, AECL is developing the Generation IV CANDU Super Critical Water Reactor. Owing to the evolutionary nature of these advanced reactors, advanced technology from the development programs is also being applied to operating CANDU plants, for both refurbishments and upgrading of existing systems and components. In addition, AECL is developing advanced technology that covers the entire life cycle of the CANDU plant, including waste management and decommissioning. Thus, AECL maintains state-of-the-art expertise and technology to support both operating and future CANDU plants. This paper outlines the scale of the current core knowledge base that is the foundation for advancement and support of CANDU technology. The knowledge base includes advancements in materials, fuel, safety, plant operations, components and systems, environmental technology, waste management, and construction. Our approach in each of these areas is to develop the underlying science, carry out integrated engineering scale tests, and perform large-scale demonstration testing. AECL has comprehensive R and D and engineering development programs to cover all of these elements. The paper will show how the ongoing expansion of the CANDU knowledge base has led to the development of the Advanced CANDU Reactor. The ACR is a Generation III+ reactor with substantially reduced costs, faster construction, and enhanced passive safety and operating

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

    International Nuclear Information System (INIS)

    Wati, Nurokhim

    2008-01-01

    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)

  11. Subchannel analysis code development for CANDU fuel channel

    International Nuclear Information System (INIS)

    Park, J. H.; Suk, H. C.; Jun, J. S.; Oh, D. J.; Hwang, D. H.; Yoo, Y. J.

    1998-07-01

    Since there are several subchannel codes such as COBRA and TORC codes for a PWR fuel channel but not for a CANDU fuel channel in our country, the subchannel analysis code for a CANDU fuel channel was developed for the prediction of flow conditions on the subchannels, for the accurate assessment of the thermal margin, the effect of appendages, and radial/axial power profile of fuel bundles on flow conditions and CHF and so on. In order to develop the subchannel analysis code for a CANDU fuel channel, subchannel analysis methodology and its applicability/pertinence for a fuel channel were reviewed from the CANDU fuel channel point of view. Several thermalhydraulic and numerical models for the subchannel analysis on a CANDU fuel channel were developed. The experimental data of the CANDU fuel channel were collected, analyzed and used for validation of a subchannel analysis code developed in this work. (author). 11 refs., 3 tabs., 50 figs

  12. Technologies in support of CANDU development

    International Nuclear Information System (INIS)

    Turner, C.; Tapping, B.

    2005-01-01

    Atomic Energy of Canada, Ltd. (AECL) has significant research and development (R and D) programs designed to meet the needs of both existing CANDU reactors and new and evolving CANDU plant designs. These R and D programs cover a wide range of technology, from chemistry and materials support through to inspection and life management tools. Emphasis is placed on effective technology development programs for fuel channels, feeders and steam generators to ensure their operation through design life, and beyond. This paper specifically addresses how the R and D has been applied in the production of longer-lived pressure tubes for the most recent CANDU 6 reactors, and how this technology forms the basis for the pressure tubes of the Advanced CANDU Reactor (ACR). Similarly, AECL has developed solutions for other critical components such as calandria tubes, feeder pipe and steam generators. The paper also discusses how the R and D knowledge has been integrated into aging management databases and health monitoring tools. Since 1997, AECL has been working with CANDU utilities on comprehensive and integrated CANDU Plant Life Management (PLiM) programs for successful and reliable plant operation through design life and beyond. AECL has developed and implemented an advanced chemistry monitoring and diagnostic system, called ChemAND which allows on-line access by the operators to current and past chemistry conditions enabling appropriate responses and facilitating planning of shutdown maintenance actions. An equivalent tool for monitoring, trending and diagnosing thermal and mechanical data has also been developed; this tool is called ThermAND. AECL is developing the Maintenance Information, Monitoring, and Control (MIMC) system, which provide information to the user for condition-based decision-making in maintenance. To enable more effective inspections, surveillance and data collection, AECL has developed unique one-off tooling to carry out unanticipated inspection and repair

  13. Candu 600 fuelling machine testing, the romanian experience

    International Nuclear Information System (INIS)

    Valeca, S.; Doca, C.; Iorga, C.

    2013-01-01

    The Candu 600 Fuelling Machine is a complex mechanism which must run in safety conditions and with high reliability in the Candu Reactor. The testing and commissioning process of this nuclear equipment meets the high standards of NPPs requirements using special technological facilities, modern measurement instruments as well the appropriate IT resources for data acquisition and processing. The paper presents the experience of the Institute for Nuclear Research Pitesti, Romania, in testing Candu 600 Fuelling Machines, inclusive the implied facilities, and in development of four simulators: two dedicated for the training of the Candu 600 Fuelling Machine Operators, and another two to simulate some process signals and actions. (authors)

  14. Validation of WIMS-CANDU using Pin-Cell Lattices

    International Nuclear Information System (INIS)

    Kim, Won Young; Min, Byung Joo; Park, Joo Hwan

    2006-01-01

    The WIMS-CANDU is a lattice code which has a depletion capability for the analysis of reactor physics problems related to a design and safety. The WIMS-CANDU code has been developed from the WIMSD5B, a version of the WIMS code released from the OECD/NEA data bank in 1998. The lattice code POWDERPUFS-V (PPV) has been used for the physics design and analysis of a natural uranium fuel for the CANDU reactor. However since the application of PPV is limited to a fresh fuel due to its empirical correlations, the WIMS-AECL code has been developed by AECL to substitute the PPV. Also, the WIMS-CANDU code is being developed to perform the physics analysis of the present operating CANDU reactors as a replacement of PPV. As one of the developing work of WIMS-CANDU, the U 238 absorption cross-section in the nuclear data library of WIMS-CANDU was updated and WIMS-CANDU was validated using the benchmark problems for pin-cell lattices such as TRX-1, TRX-2, Bapl-1, Bapl-2 and Bapl-3. The results by the WIMS-CANDU and the WIMS-AECL were compared with the experimental data

  15. Development of CANDU fuel in-bay inspection and dimensional measurement system

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Chang Keun; Cho, Moon Sung; Suk, Ho Chun; Koo, Dae Seo; Park, Kwang June; Jun, Ji Su; Jung, Jong Yeob

    2004-09-01

    In this report, we describe a spent fuel inspection and dimensional measuring system for CANDU fuel bundles in atmosphere or water. The submissible camera with radiation resistance was used to inspect a spent fuel in water. The design criteria of the dimensional measuring system was set up {+-}0.01mm(10{mu}m) for measuring accuracy. The LVDT (Linear Variable Differential Transformer) was used as measuring sensor in this dimensional measuring system. An LVDT calibration equipment was developed in order to satisfy the required accuracy of the system. Also, aluminum master and CANDU Master with same dimension of fuel bundle was made to calibrate the mechanical error of the dimensional measuring system. The accuracy of the fuel inspection system was examined. The results show that the accuracy in dimensional measurement of fuel rod profile and bearing pad profile, diameter of fuel bundle, fuel rod length, and end plate profile using standard test equipment satisfies the design criteria, i.e., maximum measurement error of {+-}0.01mm(10{mu}m)

  16. 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)

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

    International Nuclear Information System (INIS)

    Zhang, D.

    2009-01-01

    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. Heavy water cycle in the CANDU reactor

    International Nuclear Information System (INIS)

    Nanis, R.

    2000-01-01

    Hydrogen atom has two isotopes: deuterium 1 H 2 and tritium 1 H 3 . The deuterium oxide D 2 O is called heavy water due to its density of 1105.2 Kg/m 3 . Another important physical property of the heavy water is the low neutron capture section, suitable to moderate the neutrons into natural uranium fission reactor as CANDU. Due to the fact that into this reactor the fuel is cooled into the pressure tubes surrounded by a moderator, the usage of D 2 O as primary heat transport (PHT) agent is mandatory. Therefore a large amount of heavy water (approx. 500 tons) is used in a CANDU reactor. Being a costly resource - it represents 20% of the initial plant capital cost, D 2 O management is required to preserve it. (author)

  19. Role of computers in CANDU safety systems

    International Nuclear Information System (INIS)

    Hepburn, G.A.; Gilbert, R.S.; Ichiyen, N.M.

    1985-01-01

    Small digital computers are playing an expanding role in the safety systems of CANDU nuclear generating stations, both as active components in the trip logic, and as monitoring and testing systems. The paper describes three recent applications: (i) A programmable controller was retro-fitted to Bruce ''A'' Nuclear Generating Station to handle trip setpoint modification as a function of booster rod insertion. (ii) A centralized monitoring computer to monitor both shutdown systems and the Emergency Coolant Injection system, is currently being retro-fitted to Bruce ''A''. (iii) The implementation of process trips on the CANDU 600 design using microcomputers. While not truly a retrofit, this feature was added very late in the design cycle to increase the margin against spurious trips, and has now seen about 4 unit-years of service at three separate sites. Committed future applications of computers in special safety systems are also described. (author)

  20. Moderator heat recovery of CANDU reactors

    International Nuclear Information System (INIS)

    Fath, H.E.S.; Ahmed, S.T.

    1986-01-01

    A moderator heat recovery scheme is proposed for CANDU reactors. The proposed circuit utilizes all the moderator heat to the first stages of the plant feedwater heating system. CANDU-600 reactors are considered with moderator heat load varying from 120 to 160 MWsub(th), and moderator outlet temperature (from calandria) varying from 80 to 100 0 C. The steam saved from the turbine extraction system was found to produce an additional electric power ranging from 5 to 11 MW. This additional power represents a 0.7-1.7% increase in the plant electric output power and a 0.2-0.7% increase in the plant thermal efficiency. The outstanding features and advantages of the proposed scheme are presented. (author)

  1. Technical specifications and performance of CANDU fuel

    International Nuclear Information System (INIS)

    Sejnoha, R.

    1997-01-01

    The relations between Technical Specifications and fuel performance are discussed in terms of design limits and margins. The excellent performance record of CANDU reactor fuel demonstrates that the fuel design defined in the Technical Specifications (and with it other components of the procurement cycle, such as fuel manufacturing), satisfy the requirements. New requirements, changing conditions of fuel application and accumulating experience make periodic updates of the Technical Specifications necessary. Under the CANDU Owners Group (COG) Working Party 9, a Work Package has been conducted to support the review of the Specifications and the documentation of the rationales for their requirements. So far, the review has been completed for 4 Specifications: 1 for Zircaloy tubing, and 3 for uranium dioxide powder. It is planned to complete the review of all 11 currently used specifications by 1999. The paper summarizes the results achieved to mid 1997. (author)

  2. Transmutation of DUPIC spent fuel in the hyper system

    International Nuclear Information System (INIS)

    Kim, Y.H.; Song, T.Y.

    2005-01-01

    In this paper, the transmutation of TRUs of the DUPIC (Direct Use of Spent PWR Fuel in CANDU) spent fuel has been studied with the HYPER system, which is an LBE-cooled ADS. The DUPIC concept is a synergistic combination of PWRs and CANDUs, in which PWR spent fuels are directly re-utilized in CANDU reactors after a very simple re-fabrication process. In the DUPIC-HYPER fuel cycle, TRUs are recovered by using a pyro-technology and they are incinerated in a metallic fuel form of U-TRU-Zr. The objective of this study is to investigate the TRU transmutation potential of the HYPER core for the DUPIC-HYPER fuel cycle. All the previously-developed HYPER core design concepts were retained except that fuel is composed of TRU from the DUPIC spent fuel. In order to reduce the burnup reactivity swing, a B 4 C burnable absorber is used. The HYPER core characteristics have been analyzed with the REBUS-3/DIF3D code system. (authors)

  3. Operating performance of CANDU pressure tubes

    International Nuclear Information System (INIS)

    Cheadle, B.A.; Price, E.G.

    1989-04-01

    The performance of Zircaloy-2 and Zr-2.5 Nb pressure tubes in CANDU reactors is reviewed. The accelerated hydriding of Zircaloy-2 in reducing water chemistries can lower the toughness of this material and it is essential that defect-initiating phenomena, such as hydride blister formation from pressure tube to calandria tube contact, be prevented. Zr-2.5 Nb pressure tubes are performing well with low rates of hydrogen pick-up and good retention of material properties

  4. Candu technology: the next generation now

    International Nuclear Information System (INIS)

    Hopwood, J.M.; Duffey, R.B.; Torgerson, D.F.

    2001-01-01

    We describe the development philosophy, direction and concepts that are being utilized by AECL to refine the CANDU reactor to meet the needs of current and future competitive energy markets. The technology development path for CANDU reactors is based on the optimization of the pressure tube concept. Because of the inherent modularity and flexibility of this basis for the core design, it is possible to provide a seamless and continuous evolution of the reactor design and performance. There is no need for a drastic shift in concept, in technology or in fuel. By continual refinement of the flow and materials conditions in the channels, the basic reactor can be thermally and operationally efficient, highly competitive and economic, and highly flexible in application. Thus, the design can build on the successful construction and operating experience of the existing plants, and no step changes in development direction are needed. This approach minimizes investor, operator and development risk but still provides technological, safety and performance advances. In today's world energy markets, major drivers for the technology development are: (a) reduced capital cost; (b) improved operation; (c) enhanced safety; and (d) fuel cycle flexibility. The drivers provide specific numerical targets. Meeting these drivers ensures that the concept meets and exceeds the customer economic, performance, safety and resource use goals and requirements, including the suitable national and international standards. This logical development of the CANDU concept leads naturally to the 'Next Generation' of CANDU reactors. The major features under development include an optimized lattice for SEU (slightly enriched uranium) fuel, light water cooling coupled with heavy water moderation, advanced fuel channels and CANFLEX fuel, optimization of plant performance, enhanced thermal and BOP (balance of plant) efficiency, and the adoption of layout and construction technology adapted from successful on

  5. Conditioning technology of spent radium sources

    International Nuclear Information System (INIS)

    Kang, Il Sik; Kim, K. J.; Jang, K. D.

    2001-03-01

    In order to avoid accidents that could be resulted from improper storage of spent radium sources, it is necessary to condition and store them safely. The program for safe conditioning of spent radium sources by IAEA has been established to assist the developing countries. The main object of this report is to understand well and apply the technology that was applied in conditioning the national inventory of Ra-226 sources in Myanmar, as a part of IAEA's project by the Korean expert team. The report is the result that the Korean expert team carried out in Myanmar under the project title 'Radium Conditioning Service in Myanmar(INT4131-06646C)'. As a result of the mission, a whole inventory, 1,429.5 mCi of spent radium sources was safely conditioned by the Korean expert team according to the manual under the supervision of IAEA's technical officer, Mr. Al-Mughrabi, and under the control of DAE authority. These sources were encapsuled in 27 small capsules and 3 large capsules, and conditioned in 3 lead shields, producing 3 packages. The inventories were distributed into 3 shielding devices, holding 500, 459.5, and 470 mCi

  6. The design of the DUPIC spent fuel bundle counter

    International Nuclear Information System (INIS)

    Menlove, H.O.; Rinard, P.M.; Kroncke, K.E.; Lee, Y.G.

    1997-05-01

    A neutron coincidence detector had been designed to measure the amount of curium in the fuel bundles and associated process samples used in the direct use of plutonium in Canadian deuterium-uranium (CANDU) fuel cycle. All of the sample categories are highly radioactive from the fission products contained in the pressurized water reactor (PWR) spent fuel feed stock. Substantial shielding is required to protect the He-3 detectors from the intense gamma rays. The Monte Carlo neutron and photon calculational code has been used to design the counter with a uniform response profile along the length of the CANDU-type fuel bundle. Other samples, including cut PWR rods, process powder, waste, and finished rods, can be measured in the system. This report describes the performance characteristics of the counter and support electronics. 3 refs., 23 figs., 6 tabs

  7. The water chemistry of CANDU PHW reactors

    International Nuclear Information System (INIS)

    LeSurf, J.E.

    1978-01-01

    This review will discuss the chemistry of the three major water circuits in a CANDU-PHW reactor, viz., the Primary Heat Transport (PHT) water, the moderator and the boiler water. An important consideration for the PHT chemistry is the control of corrosion and of the transport of corrosion products to minimize the growth of radiation fields. In new reactors the PHT will be allowed to boil, requiring reconsideration of the methods used to radiolytic oxygen and elevate the pH. Separation of the moderator from the PHT in the pressure-tubed CANDU design permits better optimization of the chemistry of each system, avoiding the compromises necessary when the same water serves both functions. Major objectives in moderator chemistry are to control (a) the radiolytic decomposition of D 2 0; (b) the concentration of soluble neutron poisons added to adjust reactivity; and (c) the chemistry of shutdown systems. The boiler water and its feed water are treated to avoid boiler tube corrosion, both during normal operation and when perturbations are caused to the feed by, for example, leaks in the condenser tubes which permit ingress of untreated condenser cooling water. Development of a system for automatic analysis and control of feed water to give rapid, reliable response to abnormal conditions is a novel feature which has been developed for incorporation in future CANDU-PHW reactors. (author)

  8. Luncheon address: Development of the CANDU reactor

    International Nuclear Information System (INIS)

    Bain, A.S.

    1997-01-01

    The paper is a highlight of the some of the achievements in the development of the CANDU Reactor, taken from the book C anada Enters the Nuclear Age . The CANDU reactor is one of Canada's greatest scientific/engineering achievements, that started in the 1940's and bore fruit with the reactors of the 60's, 70's, and 80's. The Government decided in the 1950's to proceed with a demonstration nuclear power reactor (NPD), AECL invited 7 Canadian corporations to bid on a contract to design and construct the NPD plant. General Electric was selected. A utility was also essential for participation and Ontario Hydro was chosen. In May 1957 it was concluded that the minimum commercial size would be about 200MWe and it should use horizontal pressure tubes to contain the fuel and pressurized heavy water coolant. The book also talks of standard out-reactor components such as pumps, valves, steam generators and piping. A major in-reactor component of interest was the fuel, fuel channels and pressure tubes. A very high level of cooperation was required for the success of the CANDU program

  9. Tritium inventory prediction in a CANDU plant

    International Nuclear Information System (INIS)

    Song, M.J.; Son, S.H.; Jang, C.H.

    1995-01-01

    The flow of tritium in a CANDU nuclear power plant was modeled to predict tritium activity build-up. Predictions were generally in good agreement with field measurements for the period 1983--1994. Fractional contributions of coolant and moderator systems to the environmental tritium release were calculated by least square analysis using field data from the Wolsong plant. From the analysis, it was found that: (1) about 94% of tritiated heavy water loss came from the coolant system; (2) however, about 64% of environmental tritium release came from the moderator system. Predictions of environmental tritium release were also in good agreement with field data from a few other CANDU plants. The model was used to calculate future tritium build-up and environmental tritium release at Wolsong site, Korea, where one unit is operating and three more units are under construction. The model predicts the tritium inventory at Wolsong site to increase steadily until it reaches the maximum of 66.3 MCi in the year 2026. The model also predicts the tritium release rate to reach a maximum of 79 KCi/yr in the year 2012. To reduce the tritium inventory at Wolsong site, construction of a tritium removal facility (TRF) is under consideration. The maximum needed TRF capacity of 8.7 MCi/yr was calculated to maintain tritium concentration effectively in CANDU reactors

  10. Coolability of severely degraded CANDU cores

    International Nuclear Information System (INIS)

    Meneley, D.A.; Blahnik, C.; Rogers, J.T.; Snell, V.G.; Mijhawan, S.

    1995-07-01

    Analytical and experimental studies have shown that the separately cooled moderator in a CANDU reactor provides an effective heat sink in the event of a loss-of-coolant accident (LOCA) accompanied by total failure of the emergency core cooling system (ECCS). The moderator heat sink prevents fuel melting and maintains the integrity of the fuel channels, therefore terminating this severe accident short of severe core damage. Nevertheless, there is a probability, however low, that the moderator heat sink could fail in such an accident. The pioneering work of Rogers (1984) for such a severe accident using simplified models showed that the fuel channels would fail and a bed of dry, solid debris would be formed at the bottom of the calandria which would heat up and eventually melt. However, the molten pool of core material would be retained in the calandria vessel, cooled by the independently cooled shield-tank water, and would eventually re solidify. Thus, the calandria vessel would act inherently as a core-catcher as long as the shield tank integrity is maintained. The present paper reviews subsequent work on the damage to a CANDU core under severe accident conditions and describes an empirically based mechanistic model of this process. It is shown that, for such severe accident sequences in a CANDU reactor, the end state following core disassembly consists of a porous bed of dry solid, coarse debris, irrespective of the initiating event and the core disassembly process. (author). 48 refs., 3 tabs., 18 figs

  11. Coolability of severely degraded CANDU cores. Revised

    International Nuclear Information System (INIS)

    Meneley, D.A.; Blahnik, C.; Rogers, J.T.; Snell, V.G.; Nijhawan, S.

    1996-01-01

    Analytical and experimental studies have shown that the separately cooled moderator in a CANDU reactor provides an effective heat sink in the event of a loss-of-coolant accident (LOCA) accompanied by total failure of the emergency core cooling system (ECCS). The moderator heat sink prevents fuel melting and maintains the integrity of the fuel channels, therefore terminating this severe accident short of severe core damage. Nevertheless, there is a probability, however low, that the moderator heat sink could fail in such an accident. The pioneering work of Rogers (1984) for such a severe accident using simplified models showed that the fuel channels would fail and a bed of dry, solid debris would be formed at the bottom of the calandria which would heat up and eventually melt. However, the molten pool of core material would be retained in the calandria vessel, cooled by the independently cooled shield-tank water, and would eventually resolidify. Thus, the calandria vessel would act inherently as a 'core-catcher' as long as the shield tank integrity is maintained. The present paper reviews subsequent work on the damage to a CANDU core under severe accident conditions and describes an empirically based mechanistic model of this process. It is shown that, for such severe accident sequences in a CANDU reactor, the end state following core disassembly consists of a porous bed of dry solid, coarse debris, irrespective of the initiating event and the core disassembly process. (author)

  12. CANDU reactors and greenhouse gas emissions

    International Nuclear Information System (INIS)

    Andseta, S.; Thompson, M.J.; Jarrell, J.P.; Pendergast, D.R.

    1998-01-01

    This paper was originally presented at the 11th Pacific Basin Nuclear Conference, Banff, Alberta, Canada, May 3-7, 1998. It has been updated to include additional lifecycle data on chemical releases from ore treatment and CANDU fuel fabrication. It is sometimes stated that nuclear power plants can supply electricity with zero emissions of greenhouse gases. In fact, consideration of the entire fuel cycle indicates that some greenhouse gases are generated during their construction and decommissioning and by the preparation of fuel and other materials required for their operation. This follows from the use of fossil fuels in the preparation of materials and during the construction and decommissioning of the plants. This paper reviews life cycle studies of several different kinds of power plants. Greenhouse gases generated by fossil fuels during the preparation of fuel and heavy water used by operating CANDU power plants are estimated. The total greenhouse gas emissions from CANDU nuclear plants, per unit of electricity ultimately produced, are very small in comparison with emissions from most other types of power plants. (author)

  13. New flux detectors for CANDU 6 reactors

    International Nuclear Information System (INIS)

    Cuttler, J.M.; Medak, N.

    1992-06-01

    CANDU reactors utilize large numbers of in-core self-powered detectors for control and protection. In the original design, the detectors (coaxial cables) were wound on carrier tubes and immersed in the heavy water moderator. Failures occurred due to corrosion and other factors, and replacement was very costly because the assemblies were not designed with maintenance in mind. A new design was conceived based on straight detectors, of larger diameter, in a sealed package of individual 'well' tubes. This protected the detectors from hostile environments and enabled individual failed sensors to be replaced by inserting spares in vacant neighbouring tubes. The new design was made retrofittable to older CANDU reactors. Provision was made for on-line scanning of the core with a miniature fission chamber. The modified detectors were tested in a lengthy development program and found to exhibit superior performance to that of the original detectors. Most of the CANDU reactors have now adopted the new design. In the case of the Gentilly-2 and Point Lepreau reactors, advantage was taken of the opportunity to redesign the detector layout (using better codes and the increased flexibility in positioning detectors) to achieve better coverage of abnormal events, leading to higher trip setpoints and wider operating margins

  14. CANDU reactors and greenhouse gas emissions

    International Nuclear Information System (INIS)

    Andseta, S.; Thompson, M.J.; Jarrell, J.P.; Pendergast, D.R.

    1999-01-01

    This paper was originally presented at the 11th Pacific Basin Nuclear Conference, Banff, Alberta, Canada, May 3-7, 1998. It has been updated to include additional lifecycle data on chemical releases from ore treatment and CANDU fuel fabrication. It is sometimes stated that nuclear power plants can supply electricity with zero emissions of greenhouse gases. In fact, consideration of the entire fuel cycle indicates that some greenhouse gases are generated during their construction and decommissioning and by the preparation of fuel and other materials required for their operation. This follows from the use of fossil fuels in the preparation of materials and during the construction and decommissioning of the plants. This paper reviews life cycle studies of several different kinds of power plants. Greenhouse gases generated by fossil fuels during the preparation of fuel and heavy water used by operating CANDU power plants are estimated. The total greenhouse gas emissions from CANDU nuclear plants, per unit of electricity ultimately produced, are very small in comparison with emissions from most other types of power plants. (author)

  15. CANDU-PHW fuel channel replacement experience

    International Nuclear Information System (INIS)

    Dunn, J.T.; Kakaria, B.K.

    1982-09-01

    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

  16. 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.

  17. SEU43 fuel bundles in CANDU 600

    International Nuclear Information System (INIS)

    Catana, Alexandru; Prodea, Iosif; Danila, Nicolae; Prisecaru, Ilie; Dupleac, Daniel

    2008-01-01

    Cernavoda Unit 1 and Unit 2 are pressure tube 650 MWe nuclear stations moderated and cooled with heavy water, of Canada design, located in Romania. Fuelling is on-power and the plant is currently fuelled with natural uranium dioxide. Fuel is encapsulated in a 37 fuel rod assembly having a specific standard geometry (STD37). In order to reduce fuel cycle costs programs were initiated in Canada, South Korea and at SCN Pitesti, Romania for design and build of a new, improved geometry fuel bundle and some fuel compositions. Among fuel compositions, which are considered, is the slightly enriched uranium (SEU) fuel (0.96 w% U-235) with an associated burn-up increase from ∼7900 MWd/tU up to ∼15000 MWd/tU. Neutron analysis showed that the Canadian-Korean fuel bundle geometry with 43 rods called SEU (SEU43) can be used in already operated reactors. A new fuel bundle resulted. Extended, comprehensive analysis must be conducted in order to assess the TH behavior of SEU43 besides the neutron, mechanical (drag force, etc) analyses. In this paper, using the sub-channel approach, main thermal-hydraulic parameters were analyzed: pressure drop; fuel, sheath and coolant temperatures; coolant density; critical heat flux. Some significant differences versus standard fuel are outlined in the paper and some conclusions are drawn. While, by using this new fuel, there are many benefits to be attained like: fuel costs reduction, spent fuel waste minimization, increase in competitiveness of nuclear power generation against other sources of generation, etc., the safety margins must be, at least, conserved. The introduction of a new fuel bundle type, different in geometry and fuel composition, requires a detailed preparation, a testing program and a series of neutron and thermal-hydraulic analysis. The results reported by this paper is part of this effort. The feasibility to increase the enrichment from 0.71% U-235 (NU) to 0.96% U-235, with an estimated burn-up increase up to 14000 MWd

  18. Cobalt-60 production in CANDU power reactors

    International Nuclear Information System (INIS)

    Malkoske, G.R.; Norton, J.L.; Slack, J.

    2002-01-01

    MDS Nordion has been supplying cobalt-60 sources to industry for industrial and medical purposes since 1946. These cobalt-60 sources are used in many market and product segments, but are primarily used to sterilize single-use medical products including; surgical kits, gloves, gowns, drapes, and cotton swabs. Other applications include sanitization of cosmetics, microbial reduction of pharmaceutical raw materials, and food irradiation. The technology for producing the cobalt-60 isotope was developed by MDS Nordion and Atomic Energy of Canada Limited (AECL) almost 55 years ago using research reactors at the AECL Chalk River Laboratories in Ontario, Canada. The first cobalt-60 source produced for medical applications was manufactured by MDS Nordion and used in cancer therapy. The benefits of cobalt-60 as applied to medical product manufacturing, were quickly realized and the demand for this radioisotope quickly grew. The same technology for producing cobalt-60 in research reactors was then designed and packaged such that it could be conveniently transferred to a utility/power reactor. In the early 1970's, in co-operation with Ontario Power Generation (formerly Ontario Hydro), bulk cobalt-60 production for industrial irradiation applications was initiated in the four Pickering A CANDU reactors. As the demand and acceptance of sterilization of medical products grew, MDS Nordion expanded its bulk supply by installing the proprietary Canadian technology for producing cobalt-60 in additional CANDU reactors. CANDU is unique among the power reactors of the world, being heavy water moderated and fuelled with natural uranium. They are also designed and supplied with stainless steel adjusters, the primary function of which is to shape the neutron flux to optimize reactor power and fuel bum-up, and to provide excess reactivity needed to overcome xenon-135 poisoning following a reduction of power. The reactor is designed to develop full power output with all of the adjuster

  19. Cobalt-60 production in CANDU power reactors

    International Nuclear Information System (INIS)

    Slack, J.; Norton, J.L.; Malkoske, G.R.

    2003-01-01

    MDS Nordion has been supplying cobalt-60 sources to industry for industrial and medical purposes since 1946. These cobalt-60 sources are used in many market and product segments. The major application is in the health care industry where irradiators are used to sterilize single use medical products. These irradiators are designed and built by MDS Nordion and are used by manufacturers of surgical kits, gloves, gowns, drapes and other medical products. The irradiator is a large shielded room with a storage pool for the cobalt-60 sources. The medical products are circulated through the shielded room and exposed to the cobalt-60 sources. This treatment sterilizes the medical products which can then be shipped to hospitals for immediate use. Other applications for this irradiation technology include sanitisation of cosmetics, microbial reduction of pharmaceutical raw materials and food irradiation. The cobalt-60 sources are manufactured by MDS Nordion in their Cobalt Operations Facility in Kanata. More than 75,000 cobalt-60 sources for use in irradiators have been manufactured by MDS Nordion. The cobalt-60 sources are double encapsulated in stainless steel capsules, seal welded and helium leak tested. Each source may contain up to 14,000 curies. These sources are shipped to over 170 industrial irradiators around the world. This paper will focus on the MDS Nordion proprietary technology used to produce the cobalt-60 isotope in CANDU reactors. Almost 55 years ago MDS Nordion and Atomic Energy of Canada developed the process for manufacturing cobalt-60 at the Chalk River Labs, in Ontario, Canada. A cobalt-59 target was introduced into a research reactor where the cobalt-59 atom absorbed one neutron to become cobalt-60. Once the cobalt-60 material was removed from the research reactor it was encapsulated in stainless steel and seal welded using a Tungsten Inert Gas weld. The first cobalt-60 sources manufactured using material from the Chalk River Labs were used in cancer

  20. Fouling issues within CANDU fuel bay purification systems

    International Nuclear Information System (INIS)

    Angell, P.; Murray, J.

    2003-01-01

    Similar to all nuclear power stations, CANDU stations store their spent fuel in large open fuel bays to allow the fuel bundles to cool down. Over the years there have been a number of minor issues relating to the operation of the fuel bays, and the annual operational and maintenance costs to the stations are in the order of $1-2M. Ontario Power Generation (OPG) stations have been treating the fuel bays of their CANDU stations with hydrogen peroxide. While this treatment has been effective for the control of microbial corrosion and algal problems, an ongoing problem with the rapid plugging of the purification system filters and the presence of a thin translucent film on the surface of the water has been reported. Analysis suggested that the translucent film was of biological origin and initially it was suspected as being responsible for the plugging of the filters. Samples of the bay water, translucent film and filters were collected for microbial analysis. From all samples a number of bacterial species was isolated, all of which had the enzyme catalase that is responsible for the microbial breakdown of hydrogen peroxide, and many of the isolates were found to be capable of producing extracellular polymeric substance (slime). Observation of the filter revealed that no slime layer was present on the outer surface of the filter. This suggested that the plugging was not the result of removal of slime from the bay. Rather it suggested that the plugging was the result of microbial colonization and growth within the matrix of the filter. The large surface area of the filter matrix provides ideal conditions for biofilm growth and the flow of water provides a ready supply of nutrients. Based on a careful review of the chemistry specifications and requirements for the fuel bay water it was suggested that the filters, as currently used, were unnecessary and could be either removed or valved out, thereby eliminating the high costs associated with their operation and

  1. On the feasibility of a CANDU PHWR actinide burner

    International Nuclear Information System (INIS)

    Anton, V.

    1995-01-01

    In this work a review of the current solutions to burn the actinide i.e. the spallation method, LWR, FBR, Siemens proposal and inert matrix is presented. Finally, a proposal is made to use the CANDU PHWR for this purpose, taking into account the techniques envisaged for LWR and the prospect of the advanced fuel cycle in CANDU system. (Author) 5 Refs

  2. Valve maintainability in CANDU-PHW nuclear generating stations

    International Nuclear Information System (INIS)

    Pothier, N.E.; Crago, W.A.

    1977-09-01

    Design, application, layout and administrative factors which affect valve maintainability in CANDU-PHW power reactors are identified and discussed. Some of these are illustrated by examples based on prototype reactor operation experience. Valve maintainability improvements resulting from laboratory development and maintainability analysis, have been incorporated in commercial CANDU-PHW nuclear generating stations. These, also, are discussed and illustrated. (author)

  3. Modularized construction, structural design and analysis of CANDU 3 plant

    Energy Technology Data Exchange (ETDEWEB)

    Biswas, J K; Wollin, S; Selvadurai, S; Saudy, A M [Atomic Energy of Canada Ltd., Saskatoon, SK (Canada)

    1996-12-31

    CANDU 3 is rated at 450 MW electric, and is a smaller and advanced version of CANDU reactors successfully operating in Canada and abroad. The design uses modularization to minimize the construction schedule and thereby reduce cost. The paper (which is published only as a long summary), deals with the concept of modularization, and with stress analysis of the various civil structures.

  4. An approach to neutronics analysis of candu reactors

    International Nuclear Information System (INIS)

    Gul, S.; Arshad, M.

    1982-12-01

    An attempt is made to tackle the problem of neutronics analysis of CANDU reactors. Until now CANDU reactors have been analysed by the methods developed at AECL and CGE using mainly receipe methods. Relying on multigroup transport codes GAM-GATHER in combination with diffusion code CITATION a package of codes is established to use it for survey as well as production purposes. (authors)

  5. Advanced instrumentation and control systems for CANDU refurbishment

    International Nuclear Information System (INIS)

    Sklyar, V.; Bakhmach, I.; Kharchenko, V.; Andrashov, A.; Baranova, O.

    2011-01-01

    The purpose of the work is to discuss opportunities to modernize I and C systems of CANDU reactors on the base of Radiy's digital safety platform. This paper discusses the following topics: a business model for CANDU, I and C systems refurbishment, FPGA technology issues, comparison of different approaches to refurbish obsolete I and C systems. (author)

  6. Modularized construction, structural design and analysis of CANDU 3 plant

    International Nuclear Information System (INIS)

    Biswas, J.K.; Wollin, S.; Selvadurai, S.; Saudy, A.M.

    1995-01-01

    CANDU 3 is rated at 450 MW electric, and is a smaller and advanced version of CANDU reactors successfully operating in Canada and abroad. The design uses modularization to minimize the construction schedule and thereby reduce cost. The paper (which is published only as a long summary), deals with the concept of modularization, and with stress analysis of the various civil structures

  7. Numerical simulator of the CANDU fueling machine driving desk

    International Nuclear Information System (INIS)

    Doca, Cezar

    2008-01-01

    As a national and European premiere, in the 2003 - 2005 period, at the Institute for Nuclear Research Pitesti two CANDU fueling machine heads, no.4 and no.5, for the Nuclear Power Plant Cernavoda - Unit 2 were successfully tested. To perform the tests of these machines, a special CANDU fueling machine testing rig was built and was (and is) available for this goal. The design of the CANDU fueling machine test rig from the Institute for Nuclear Research Pitesti is a replica of the similar equipment operating in CANDU 6 type nuclear power plants. High technical level of the CANDU fueling machine tests required the using of an efficient data acquisition and processing Computer Control System. The challenging goal was to build a computer system (hardware and software) designed and engineered to control the test and calibration process of these fuel handling machines. The design takes care both of the functionality required to correctly control the CANDU fueling machine and of the additional functionality required to assist the testing process. Both the fueling machine testing rig and staff had successfully assessed by the AECL representatives during two missions. At same the time, at the Institute for Nuclear Research Pitesti was/is developed a numerical simulator for the CANDU fueling machine operators training. The paper presents the numerical simulator - a special PC program (software) which simulates the graphics and the functions and the operations at the main desk of the computer control system. The simulator permits 'to drive' a CANDU fueling machine in two manners: manual or automatic. The numerical simulator is dedicated to the training of operators who operate the CANDU fueling machine in a nuclear power plant with CANDU reactor. (author)

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

    International Nuclear Information System (INIS)

    Choi, In-Kyu; Yeon, Jei-Won; Kim, Won-Ho

    2002-01-01

    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

  9. Coolant Void Reactivity Analysis of CANDU Lattice

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin Su; Lee, Hyun Suk; Tak, Tae Woo; Lee, Deok Jung [UNIST, Ulsan (Korea, Republic of)

    2016-05-15

    Models of CANDU-6 and ACR-700 fuel lattices were constructed for a single bundle and 2 by 2 checkerboard to understand the physics related to CVR. Also, a familiar four factor formula was used to predict the specific contributions to reactivity change in order to achieve an understanding of the physics issues related to the CVR. At the same time, because the situation of coolant voiding should bring about a change of neutron behavior, the spectral changes and neutron current were also analyzed. The models of the CANDU- 6 and ACR-700 fuel lattices were constructed using the Monte Carlo code MCNP6 using the ENDF/B-VII.0 continuous energy cross section library based on the specification from AECL. The CANDU fuel lattice was searched through sensitivity studies of each design parameter such as fuel enrichment, fuel pitch, and types of burnable absorber for obtaining better behavior in terms of CVR. Unlike the single channel coolant voiding, the ACR-700 bundle has a positive reactivity change upon 2x2 checkerboard coolant voiding. Because of the new path for neutron moderation, the neutrons from the voided channel move to the no-void channel where they lose energy and come back to the voided channel as thermal neutrons. This phenomenon causes the positive CVR when checkerboard voiding occurs. The sensitivity study revealed the effects of the moderator to fuel volume ratio, fuel enrichment, and burnable absorber on the CVR. A fuel bundle with low moderator to fuel volume ratio and high fuel enrichment can help achieve negative CVR.

  10. Qinshan CANDU commissioning - a successful partnership

    International Nuclear Information System (INIS)

    Alikhan, S.; Thomson, J.; Jun, G.; Guoyuan, J.

    2004-01-01

    The Qinshan CANDU Nuclear Power Plant consists of 2 x 728 MWe CANDU 6 units, built in Zhejiang Province, China, by the Third Qinshan Nuclear Power Company (TQNPC) as the owner and Atomic Energy of Canada Limited (AECL) as the main contractor. The Contract between China National Nuclear Corporation (CNNC) and AECL was signed in November 1996 and became effective on February 12, 1997 with scheduled completion dates of February 12, 2003 for Unit 1 and November 12, 2003 for Unit 2. Unit 1 was declared in-service on December 31, 2002, 43 days ahead of schedule and Unit 2 was declared in service on July 20, 2003, 115 days ahead of schedule. The successful partnership between AECL, Bechtel, Hitachi and TQNPC working as a team is the key to this success. Total commissioning period from first energization of the system service transformer to in-service for both units was 20.7 months, which is significantly better than the experience at other comparable CANDU 6 units. It has clearly demonstrated the benefits of building two units together, about 6 months apart, to achieve optimum utilization of resources already mobilized for the first unit; the second unit is commissioned with less than 40% of the effort required for the first unit. Since in-service to the end of March 2004, Unit 1 has operated at a gross capacity factor of 93% and Unit 2 at 82.5%, including loss of production for one month in August 2003 to repair the failure of turbine LP blades tie-wire. (author)

  11. Structural integrity evaluations of CANDU pressure tubes

    International Nuclear Information System (INIS)

    Radu, Vasile

    2003-01-01

    The core of a CANDU-6 pressurized heavy water reactor consists of some hundred horizontal pressure tubes that are manufactured from a Zr-2.5%Nb alloy and which contain the fuel bundles. These tubes are susceptible to a damaging phenomenon known as Delayed Hydride Cracking (DHC). The Zr-2.5%Nb alloy is susceptible to DHC phenomenon when there is diffusion of hydrogen atoms to a service-induced flaws, followed by the hydride platelets formation on the certain crystallographic planes in the matrix material. Finally, the development of hydride regions at the flaw-tip will happened. These hydride regions are able to fracture under stress-temperature conditions (DHC initiation) and the cracks can extend and grow by DHC mechanism. Some studies have been focused on the potential to initiate DHC at the blunt flaws in a CANDU reactor pressure tube and a methodology for structural integrity evaluation was developed. The methodology based on the Failure Assessment Diagrams (FAD's) consists in an integrated graphical plot, where the fracture failure and plastic collapse are simultaneously evaluated by means of two non-dimensional variables (K r and L r ). These two variables represent the ratio of the applied value of either stress or stress intensity factor and the resistance parameter of corresponding magnitude (yield stress or fracture toughness, respectively). Once the plotting plane is determined by the variables K r and L r , the procedure defines a critical failure line that establishes the safe area. The paper will demonstrate the possibility to perform structural integrity evaluations by means of Failure Assessment Diagrams for flaws occurring in CANDU pressure tubes. (author)

  12. New developments in dry spent fuel storage

    International Nuclear Information System (INIS)

    Bonnet, C.; Chevalier, Ph.

    2001-01-01

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

  13. Remote handling equipment for CANDU retubing

    International Nuclear Information System (INIS)

    Crawford, G.S.; Lowe, H.

    1993-01-01

    Numet Engineering Ltd. has designed and supplied remote handling equipment for Ontario Hydro's retubing operation of its CANDU reactors at the Bruce Nuclear Generating Station. This equipment consists of ''Retubing Tool Carriers'' an'' Worktables'' which operate remotely or manually at the reactor face. Together they function to transport tooling to and from the reactor face, to position and support tooling during retubing operations, and to deliver and retrieve fuel channels and channel components. This paper presents the fundamentals of the process and discusses the equipment supplied in terms of its design, manufacturing, components and controls, to meet the functional and quality requirements of Ontario Hydro's retubing process. (author)

  14. Proof testing of CANDU concrete containment structures

    International Nuclear Information System (INIS)

    Pandey, M.D.

    1996-05-01

    Prior to commissioning of a CANDU reactor, a proof pressure test is required to demonstrate the structural integrity of the containment envelope. The test pressure specified by AECB Regulatory Document R-7 (1991) was selected without a rigorous consideration of uncertainties associated with estimates of accident pressure and conatinment resistance. This study was undertaken to develop a reliability-based philosophy for defining proof testing requirements that are consistent with the current limit states design code for concrete containments (CSA N287.3).It was shown that the upodated probability of failure after a successful test is always less than the original estimate

  15. Leak before break experience in CANDU reactors

    International Nuclear Information System (INIS)

    Price, E.G.; Moan, G.D.; Coleman, C.E.

    1988-04-01

    The paper describes how the requirements for Leak-Before-Break are met in CANDU reactors. The requirements are based on operational and laboratory experience. After the onset of leakage in a fuel channel from a delayed hydride crack, time is available to the operator to take action before the crack grows to an unstable length. The time available is calculated using different models which use crack growth data from small specimen tests. When the results from crack growth behaviour experiments, carried out on components removed from reactor are used in the model, the time available for operator response is about 100 hours

  16. A generalized perturbation program for CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do Heon; Kim, Jong Kyung [Hanyang University, Seoul (Korea, Republic of); Choi, Hang Bok; Roh, Gyu Hong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Yang, Won Sik [Chosun University, Kwangju (Korea, Republic of)

    1999-12-31

    A generalized perturbation program has been developed for the purpose of estimating zonal power variation of a CANDU reactor upon refueling operation. The forward and adjoint calculation modules of RFSP code were used to construct the generalized perturbation program. The numerical algorithm for the generalized adjoint flux calculation was verified by comparing the zone power estimates upon refueling with those of forward calculation. It was, however, noticed that the truncation error from the iteration process of the generalized adjoint flux is not negligible. 2 refs., 1 figs., 1 tab. (Author)

  17. Improved containment isolation for CANDU plants

    Energy Technology Data Exchange (ETDEWEB)

    Stretch, A H [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

    1996-12-31

    The publication of Regulatory Policy Statement R- 7 in February 1991 by the Atomic Energy Control Board imposes new requirement for the isolation of fluid piping penetrating the containment boundary. The Appendix of R-7 describes the detailed requirements for metal extensions of the containment envelope, including the code classification qf the pressure retaining portions and isolation requirements for fluid piping and tubing.The application of these new requirements to the existing CANDU 6 design resulted in a number of design changes, including the addition of manual and automatic isolation valves, changes in code classification, and leakage monitoring considerations. (author). 3 refs., 3 figs.

  18. A generalized perturbation program for CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do Heon; Kim, Jong Kyung [Hanyang University, Seoul (Korea, Republic of); Choi, Hang Bok; Roh, Gyu Hong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Yang, Won Sik [Chosun University, Kwangju (Korea, Republic of)

    1998-12-31

    A generalized perturbation program has been developed for the purpose of estimating zonal power variation of a CANDU reactor upon refueling operation. The forward and adjoint calculation modules of RFSP code were used to construct the generalized perturbation program. The numerical algorithm for the generalized adjoint flux calculation was verified by comparing the zone power estimates upon refueling with those of forward calculation. It was, however, noticed that the truncation error from the iteration process of the generalized adjoint flux is not negligible. 2 refs., 1 figs., 1 tab. (Author)

  19. Development of the advanced CANDU technology

    International Nuclear Information System (INIS)

    Suk, Soo Dong; Min, Byung Joo; Na, Y. H.; Lee, S. Y.; Choi, J. H.; Lee, B. C.; Kim, S. N.; Jo, C. H.; Paik, J. S.; On, M. R.; Park, H. S.; Kim, S. R.

    1997-07-01

    The purpose of this study is to develop the advanced design technology to improve safety, operability and economy and to develop and advanced safety evaluation system. More realistic and reasonable methodology and modeling was employed to improve safety margin in containment analysis. Various efforts have been made to verify the CATHENA code which is the major safety analysis code for CANDU PHWR system. Fully computerized prototype ECCS was developed. The feasibility study and conceptual design of the distributed digital control system have been performed as well. The core characteristics of advanced fuel cycle, fuel management and power upgrade have been studied to determine the advanced core. (author). 77 refs., 51 tabs., 108 figs

  20. Decommissioning of the CANDU-PHW reactor

    International Nuclear Information System (INIS)

    Unsworth, G.N.

    1977-04-01

    This report contains the results of a study of various aspects of decommissioning of reactors. The study places in perspective the size of the job, the hazards involved, the cost and the environmental impact. The three internationally agreed ''stages'' of decommissioning, namely, mothballing, entombment, and dismantling are defined and discussed. The single unit 600 MW(e) CANDU is chosen as the type of reactor on which the discussion is focussed but the conclusions reached will provide a basis for judgement of the costs and problems associated with decommissioning reactors of other sizes and types. (author)

  1. Development of the advanced CANDU technology

    Energy Technology Data Exchange (ETDEWEB)

    Suk, Soo Dong; Min, Byung Joo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Na, Y H; Lee, S Y; Choi, J H; Lee, B C; Kim, S N; Jo, C H; Paik, J S; On, M R; Park, H S; Kim, S R [Korea Electric Power Co., Taejon (Korea, Republic of)

    1997-07-01

    The purpose of this study is to develop the advanced design technology to improve safety, operability and economy and to develop and advanced safety evaluation system. More realistic and reasonable methodology and modeling was employed to improve safety margin in containment analysis. Various efforts have been made to verify the CATHENA code which is the major safety analysis code for CANDU PHWR system. Fully computerized prototype ECCS was developed. The feasibility study and conceptual design of the distributed digital control system have been performed as well. The core characteristics of advanced fuel cycle, fuel management and power upgrade have been studied to determine the advanced core. (author). 77 refs., 51 tabs., 108 figs.

  2. Decontamination of CANDU primary coolant system

    International Nuclear Information System (INIS)

    Pettit, P.J.

    1975-01-01

    Decontamination of radioactive systems is necessary to reduce personnel radiation exposures and also to reduce exposure during special work. Mechanical decontamination methods are sometimes useful, but most contaminated surfaces are inaccessible, so chemical decontamination often is preferred. The A-P Citrox method will remove most contaminants from CANDU systems, but is costly and long, damages components, and produces large quantities of radioactive liquid waste. The Redox cycling process is fast and inexpensive, produces only solid wastes, but removes small quantities of deposit from Monel only. The CAN-DECON process removes deposits from most materials including fuel cladding and has many other advantages. (author)

  3. 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

  4. 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)

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

    International Nuclear Information System (INIS)

    Catana, Alexandru; Danila, Nicolae; Prisecaru, Ilie; Dupleac, Daniel

    2008-01-01

    Two Candu 600 pressure tube nuclear reactors cover about 17% of Romania's electricity demand. These nuclear reactors are moderated/cooled with D 2 O, fuelled on-power with Natural Uranium (NU) dioxide encapsulated in a standard (STD37) fuel bundle. High neutron economy is achieved using D 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)

  6. Development of thermal hydraulic evaluation code for CANDU reactors

    International Nuclear Information System (INIS)

    Kim, Man Woong; Yu, Seon Oh; Choi, Yong Seog; Shin, Chull; Hwang, Soo Hyun

    2004-02-01

    To enhance the safety of operating CANDU reactors, the establishment of the safety analysis codes system for CANDU reactors is in progress. As for the development of thermal-hydraulic analysis code for CANDU system, the studies for improvement of evaluation model inside RELAP/CANDU code and the development of safety assessment methodology for GSI (Generic Safety Issues) are in progress as a part of establishment of CANDU safety assessment system. To develop the 3-D thermal-hydraulic analysis code for moderator system, the CFD models for analyzing the CANDU-6 moderator circulation are developed. One model uses a structured grid system with the porous media approach for the 380 Calandria tubes in the core region. The other uses a unstructured grid system on the real geometry of 380 Calandria tubes, so that the detailed fluid flow between the Calandria tubes can be observed. As to the development of thermal-hydraulic analysis code for containment, the study on the applicability of CONTAIN 2.0 code to a CANDU containment was conducted and a simulation of the thermal-hydraulic phenomena during the accident was performed. Besides, the model comparison of ESFs (Engineered Safety Features) inside CONTAIN 2.0 code and PRESCON code has also conducted

  7. Trends in the capital costs of CANDU generating stations

    International Nuclear Information System (INIS)

    Yu, A.M.

    1982-09-01

    This paper consolidates the actual cost experience gained by Atomic Energy of Canada Limited, Ontario Hydro, and other Canadian electric utlities in the planning, design and construction of CANDU-PHWR (CANada Deuterium Uranium-Pressurized Heavy Water Reactor) generating stations over the past 30 years. For each of the major CANDU-PHWR generating stations in operation and under construction in Canada, an analysis is made to trace the evolution of the capital cost estimates. Major technical, economic and other parameters that affect the cost trends of CANDU-PHWR generating stations are identified and their impacts assessed. An analysis of the real cost of CANDU generating stations is made by eliminating interest during construction and escalation, and the effects of planned deferment of in-service dates. An historical trend in the increase in the real cost of CANDU power plants is established. Based on the cost experience gained in the design and construction of CANDU-PHWR units in Canada, as well as on the assessment of parameters that influence the costs of such projects, the future costs of CANDU-PHWRs are presented

  8. 9th International conference on CANDU maintenance

    International Nuclear Information System (INIS)

    2011-01-01

    The 9th International Conference on CANDU Maintenance was held in Toronto, Ontario, Canada on December 4-6, 2011. The conference focused on Nuclear plant reliability and maintenance. Equipment reliability is a critical factor in achieving safe and reliable Nuclear Power Plant operations for many reasons. For one it reduces the challenges upon the operating staff and allows station personnel to 'go on the offence' instead of having to play defense in a reactive mode. Plant reliability ensures that there is time to study the issues in detail and develop solutions for long-term success. Let us not forget that the owner carries the burden of demonstrating the return on investment, and plant reliability goes a long way in helping to make the case for plant life extension. A good reputation for plant reliability provides the public capital necessary to build confidence and facilitate the licensing process. The proceedings papers and presentations given at the 9th International Conference on Candu Maintenance covered topics that include: Managing Worker; Radiation Dose; Full Life Cycle Management; Managing Maintenance Refurbishment; Designing for Maintainability; Inspection Techniques; and, Mitigating Degradation.

  9. Advanced CANDU reactor design for operability

    International Nuclear Information System (INIS)

    Hopwood, J.M.; Lalonde, R.; Soulard, M.

    2003-01-01

    This paper outlines design features and engineering processes in the ACR TM development program which contribute to excellence in performance and low operating cost. AECL recognizes that future plant owners will place a high priority in these operational characteristics. A successful next generation plant will have a best-in-class capability, both in its design characteristics, in the engineering philosophy and program adopted during the product development, and in the vendor's approach to operating station support. The ACR program addresses each of these drivers. Operability considerations are built-in to the design at an overall, plant wide level. For example, based on the strong CANDU 6 operating record, targets for standard outage duration, time between outages and component durability are set, while the design engineering is managed to achieve these targets. The ultimate maintenance target for the ACR, once initial operating experience has been gained, is to operate with a 21-day standard maintenance outage at an interval of once every three years. At the detailed design level, close attention is paid to space allocation, to enable good maintenance access. Selection of components also places emphasis on maintainability based on the extensive and current experience with CANDU projects. (author)

  10. Thermal hydraulics of CANDU severe accident analysis

    International Nuclear Information System (INIS)

    Negut, Gheorghe; Catana, Alexandru; Prisecaru, Ilie; Dupleac, Daniel

    2007-01-01

    As interest in severe accident studies has increased in the last years, we have developed a set of simple models to analyze severe accidents in CANDU reactors that should be integrated in the EU codes. The CANDU600 reactor uses natural uranium fuel and heavy water (D 2 O) as both moderator and coolant, with the moderator and coolant in separate systems. We chose to analyze accident development for a LOCA with simultaneous loss of moderator cooling and the loss of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperatures inside a pressure tube reaches 1000 deg. C, a contact between pressure tube and calandria tube occurs and the residual heat is transferred to the moderator. Due to the lack of cooling, the moderator eventually begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) will be uncovered, then will disintegrate and fall down to the calandria vessel bottom. After all the quantity of moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield tank water, which normally surrounds the calandria vessel. The phenomena described above are modelled, analyzed and compared with the available data. The results are encouraging. (authors)

  11. CANDU fuel behaviour under transient conditions

    International Nuclear Information System (INIS)

    Segel, A.W.L.

    1979-04-01

    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 UO 2 melting did not result in fuel fragmentation nor damage to the pressure tube. (author)

  12. ROP design for Enhanced CANDU 6 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hu, J.; Scherbakova, D; Kastanya, D.; Ovanes, M. [Candu Energy Inc., Mississauga, Ontario (Canada)

    2011-07-01

    The Enhanced CANDU 6 (EC6) nuclear power plant is a mid-sized pressurized heavy water reactor design, based on the highly successful CANDU 6 (C6) family of power plants, upgraded to meet today's Canadian and international safety requirements and to satisfy Generation III expectations. The EC6 reactor is equipped with two independent Regional Overpower Protection (ROP) systems to prevent overpowers in the reactor fuel. The ROP system design, retaining the traditional C6 methodology, is determined to cover the End-of-Life (EOL) reactor core condition since the reactor operating/thermal margin gradually decreases as plant equipment ages. Several design changes have been incorporated into the reference C6 plant to mitigate the ageing effect on the ROP trip margin. This paper outlines the basis for the EC6 ROP physics design and presents the ROP related improvements made in the EC6 design to ensure that full power operation is not limited by the ROP throughout the entire life of the reactor. (author)

  13. Seismic fragility analysis of a CANDU containment structure for near-fault ground motions

    International Nuclear Information System (INIS)

    Choi, In Kil; Choun, Young Sun; Seo, Jeong Moon; Ahn, Seong Moon

    2005-01-01

    The R. G. 1.60 spectrum used for the seismic design of Korean nuclear power plants provides a generally conservative design basis due to its broadband nature. A survey on some of the Quaternary fault segments near Korean nuclear power plants is ongoing. It is likely that these faults will be identified as active ones. If the faults are confirmed as active ones, it will be necessary to reevaluate the seismic safety of the nuclear power plants located near these faults. The probability based scenario earthquakes were identified as near-field earthquakes. In general, the near-fault ground motion records exhibit a distinctive long period pulse like time history with very high peak velocities. These features are induced by the slip of the earthquake fault. Near-fault ground motions, which have caused much of the damage in recent major earthquakes, can be characterized by a pulse-like motion that exposes the structure to a high input energy at the beginning of the motion. It is necessary to estimate the near-fault ground motion effects on the nuclear power plant structures and components located near the faults. In this study, the seismic fragility analysis of a CANDU containment structure was performed based on the results of nonlinear dynamic time-history analyses

  14. Nonlinear seismic behavior of a CANDU containment building subjected to near-field ground motions

    International Nuclear Information System (INIS)

    Choi, In Kil; Ahn, Seong Moon; Choun, Young Sun; Seo, Jeong Moon

    2004-01-01

    The standard response spectrum proposed by US NRC has been used as a design earthquake for the design of Korean nuclear power plant structures. A survey on some of the Quaternary fault segments near Korean nuclear power plants is ongoing. It is likely that these faults will be identified as active ones. If the faults are confirmed as active ones, it will be necessary to reevaluate the seismic safety of the nuclear power plants located near the fault. Near-fault ground motions are the ground motions that occur near an earthquake fault. In general, the near-fault ground motion records exhibit a distinctive long period pulse like time history with very high peak velocities. These features are induced by the slip of the earthquake fault. Near-fault ground motions, which have caused much of the damage in recent major earthquakes, can be characterized by a pulse-like motion that exposes the structure to a high input energy at the beginning of the motion. In this study, nonlinear dynamic time-history analyses were performed to investigate the seismic behavior of a CANDU containment structure subjected to various earthquake ground motions including the near-field ground motions

  15. Distributed control system for CANDU 9 nuclear power plant

    International Nuclear Information System (INIS)

    Harber, J.E.; Kattan, M.K.; Macbeth, M.J.

    1996-01-01

    Canadian designed CANDU pressurized heavy water nuclear reactors have been world leaders in electrical power generation. The CANDU 9 project is AECL's next reactor design. The CANDU 9 plant monitoring, annunciation, and control functions are implemented in two evolutionary systems; the distributed control system (DCS) and the plant display system (PDS). The CDS implements most of the plant control functions in a single hardware platform. The DCS communicates with the PDS to provide the main operator interface and annunciation capabilities of the previous control computer designs along with human interface enhancements required in a modern control system. (author)

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

    International Nuclear Information System (INIS)

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

    1989-02-01

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

  17. Management of spent fuel in Republic of Korea

    International Nuclear Information System (INIS)

    Pak, Hyun-Soo; Seo, In-Seok; Pak, Sang-Ki.

    1989-01-01

    At present in Republic of Korea, 8 PWR and 1 CANDU power plants are in operation or under construction, and the total capacity of power generation facilities has become 7.6 GWe. In addition, two PWRs of more than 900 MWe each are expected to be constructed by mid 1990s. More than 50 % of the electric power demand was supplied by nuclear power generation since 1987, but the spent fuel generated in nuclear power plants is stored in storage water tanks in respective reactor sites. The total capacity of spent fuel to be stored in the AR facilities of 9 nuclear power plants is about 2730 MTU, and the spent fuel released from these reactors since 1980 is about 810 MTU. The present capacity of AR storage pools seems to be used up by mid 1990s. According to the revised Atomic Energy Acts in May, 1986, the government is to take the responsibility of spent fuel management, and the policy of constructing the storage facilities outside reactor sites by the end of 1997 was established by the Atomic Energy Commission. The responsibility of the management of spent fuel that exceeds the present capacity of AR pools is to be taken by KEPCO, therefore the preliminary analysis of the feasible option on the extension of AR facilities and the comprehensive management plan for spent fuel placing emphasis on the research and development of away-from-reactor storage were decided. (Kako, I.)

  18. Development of advanced CANDU PHWR -Development of the advanced CANDU technology-

    International Nuclear Information System (INIS)

    Seok, Ho Cheon; Na, Yeong Hwan; Seok, Soo Dong; Lee, Bo Uk; Kwak, Ho Sang; Kim, Bong Ki; Kim, Seok Nam; Min, Byeong Joo; Park, Jong Ryunl; Shin, Jeong Cheol; Lee, Kyeong Ho; Lee, Dae Hee; Lee, Deuk Soo; Lee, Yeong Uk; Lee, Jeong Yang; Jwon, Jong Seon; Jwon, Chang Joon; Ji, Yong Kwan; Han, Ki Nam; Kim, Kang Soo; Kim, Dae Jin; Kim, Seon Cheol; Kim, Seong Hak; Kim, Yeon Seung; Kim, Yun Jae; Kim, Jeong Kyu; Kim, Jeong Taek; Kim, Hang Bae; Na, Bok Kyun; Namgung, In; Moon, Ki Hwan; Park, Keun Ok; Shon, Ki Chang; Song, In Ho; Shin, Ji Tae; Yeo, Ji Won; Oh, In Seok; Jang, Ik Ho; Jeong, Dae Won; Jeong, Yong Hwan; Ha, Jae Hong; Ha, Jeong Koo; Hong, Hyeong Pyo; Hwang, Jeong Ki

    1994-07-01

    The target of this project is to assess the feasibility of improving PHWR and to establish the parameter of the improved concept and requirements for developing it. To set up the requirements for the Improved Pressurized Heavy Water Reactor: (1) Design requirements of PHWR main systems and Safety Design Regulatory Requirements for Safety Related System i.e. Reactor Shutdown System, Emergency Core Cooling System and Containment System were prepared. (2) Licensing Basis Documents were summarized and Safety Analysis Regulatory. Requirements were reviewed and analyzed. To estimate the feasibility of improving PHWR and to establish the main parameters of the concept of new PHWR in future: (1) technical level/developing trend of PHWR in Korea through Wolsong 2, 3 and 4 design experience and Technical Transfer Program was investigated to analyze the state of basic technology and PHWR improvement potential. (2) CANDU 6 design improvement tendency, CANDU 3 design concept and CANDU 9 development state in other country was analyzed. (3) design improvement items to apply to the reactors after Wolsong 2, 3 and 4 were selected and Plant Design Requirements and Conceptual Design Description were prepared and the viability of improved HWR was estimated by analyzing economics, performance and safety. (4) PHWR technology improving research and development plan was established and international joint study initiated for main design improvement items

  19. History of Korean Neurosurgery.

    Science.gov (United States)

    Hwang, Sung-nam

    2015-08-01

    The year 2012 was the 50th anniversary of the Korean Neurosurgical Society, and in 2013, the 15th World Congress of Neurosurgery took place in Seoul, Korea. Thus, it is an appropriate occasion to introduce the world to the history of the Korean Neurosurgical Society and the foundation, development, and growth of Korean neurosurgery. Historical materials and pictures were collected and reviewed from the history book and photo albums of the Korean Neurosurgical Society. During the last 50 years, the Korean Neurosurgical Society and Korean neurosurgery have developed and grown enormously not only in quantity but also in quality. In every aspect, the turning point from the old to the new era of the Korean Neurosurgical Society and Korean neurosurgery was the year 1980. Copyright © 2015. Published by Elsevier Inc.

  20. Towards more efficient safeguards approach of transfer campaign in Wolsong CANDU reactor

    International Nuclear Information System (INIS)

    Park, S. K.; Na, W. W.; Park, W. S.; Jung, S. T.; Park, S. J.

    2002-01-01

    Due to the unique character of the CANDU reactor, a transfer campaign has been carried out every year for 2 months or so in Wolsong unit 1. It is expected to require performance of the transfer campaign for 4 units of Wolsong site in 2006 and more than 50% of the ROK safeguards inspection efforts. The IAEA and TCNC have gained several years of experience in safeguards approach during transfer campaign. Occasionally the deterrence and interference of operator transfer work have been occurred in order to attain safeguards inspection goal. These could be minimized using optimum C/S applications, NDA and RDT techniques. This paper shows the interrelationship and combinations of these resources and proposes new safeguards approach to maintain the continuity of knowledge from fuel loading in the spent fuel pond to the point of canister loading and closure during transfer campaign

  1. Advanced CANDU reactor pre-licensing progress

    International Nuclear Information System (INIS)

    Popov, N.K.; West, J.; Snell, V.G.; Ion, R.; Archinoff, G.; Xu, C.

    2005-01-01

    The Advanced CANDU Reactor (ACR) is an evolutionary advancement of the current CANDU 6 reactor, aimed at producing electrical power for a capital cost and at a unit-energy cost significantly less than that of the current reactor designs. The Canadian Nuclear Safety Commission (CNSC) staff are currently reviewing the ACR design to determine whether, in their opinion, there are any fundamental barriers that would prevent the licensing of the design in Canada. This CNSC licensability review will not constitute a licence, but is expected to reduce regulatory risk. The CNSC pre-licensing review started in September 2003, and was focused on identifying topics and issues for ACR-700 that will require a more detailed review. CNSC staff reviewed about 120 reports, and issued to AECL 65 packages of questions and comments. Currently CNSC staff is reviewing AECL responses to all packages of comments. AECL has recently refocused the design efforts to the ACR-1000, which is a larger version of the ACR design. During the remainder of the pre-licensing review, the CNSC review will be focused on the ACR-1000. AECL Technologies Inc. (AECLT), a wholly-owned US subsidiary of AECL, is engaged in a pre-application process for the ACR-700 with the US Nuclear Regulatory Commission (USNRC) to identify and resolve major issues prior to entering a formal process to obtain standard design certification. To date, the USNRC has produced a Pre-Application Safety Assessment Report (PASAR), which contains their reviews of key focus topics. During the remainder of the pre-application phase, AECLT will address the issues identified in the PASAR. Pursuant to the bilateral agreement between AECL and the Chinese nuclear regulator, the National Nuclear Safety Administration (NNSA) and its Nuclear Safety Center (NSC), NNSA/NSC are reviewing the ACR in seven focus areas. The review started in September 2004, and will take three years. The main objective of the review is to determine how the ACR complies

  2. Maintenance, rehabilitation, long life-the CANDU potential

    International Nuclear Information System (INIS)

    Torgerson, D.F.; Charlebois, P.; Hopkins, J.

    1998-01-01

    Plant life extension beyond the original design life is becoming an attractive economic consideration in the nuclear industry. Plant Life Management and life extension considerations have been built into the complete life cycle of the CANDU plant. The plant life management studies demonstrate that life extension for operating plants beyond 30 years is economically viable. The new CANDU designs benefit from this experience feedback and as a result, the plant design basis is now 40 years or better with potential for economical life extension. AECL is therefore confident that the new CANDU designs will exceed the performance record of the first generation CANDU 6 units and is committed to providing continued support and services during the operating life of the plant

  3. Development of a new bundle welding technology for CANDU fuels

    International Nuclear Information System (INIS)

    Kim, Soo Sung; Lee, D. Y.; Goo, D. S.

    2010-01-01

    The new technology of welding process for fuel bundle of CANDU nuclear fuels is considered important in respect to the soundness of weldments and the improvement of the performance of nuclear fuels during the operation in reactor. The probability of leakage of the fission products is mostly apt to occur at the weldments of fuel bundles, and it is connected directly with the safety and life prediction of the nuclear reactor in operation. The fuel bundles of CANDU nuclear fuels are welded by the electrical resistance method, connecting the endplates and endcaps with fuel rods. Therefore, the purpose of this study of the 2nd year is to select the proper welding parameters and to investigate the characteristics of the full-sized samples using the projection endplates and make some prototype samples for the endplate welding of CANDU nuclear fuels. This study will be also provide the fundamental data for the new design and fabrications of CANDU nuclear fuel bundles

  4. Periodic inspection of CANDU nuclear power plant containment components

    International Nuclear Information System (INIS)

    1989-09-01

    This Standard is one in a series intended to provide uniform requirements for CANDU nuclear power plants. It provides requirements for the periodic inspection of containment components including the containment pressure suppression systems

  5. Qinshan Phase III (CANDU) nuclear power project quality assurance

    International Nuclear Information System (INIS)

    Wang Lingen; Du Jinxiang

    2001-01-01

    The completion and implementation of quality assurance system of Qinshan Phase III (CANDU) nuclear power project are presented. Some comments and understanding with consideration of the project characteristics are put forward

  6. Manufacturing opportunities in the Canadian CANDU and heavy water programs

    International Nuclear Information System (INIS)

    Reny, J.P.

    The volume of business available to Canadian manufacturers of CANDU power plant and heavy water plant components is analyzed over about the next 10 years. Implications of exported nuclear technology and plants are explored. (E.C.B.)

  7. MATLAB/SIMULINK model of CANDU reactor for control studies

    International Nuclear Information System (INIS)

    Javidnia, H.; Jiang, J.

    2006-01-01

    In this paper a MATLAB/SIMULINK model is developed for a CANDU type reactor. The data for the reactor are taken from an Indian PHWR, which is very similar to CANDU in its design. Among the different feedback mechanisms in the core of the reactor, only xenon has been considered which plays an important role in spatial oscillations. The model is verified under closed loop scenarios with simple PI controller. The results of the simulation show that this model can be used for controller design and simulation of the reactor systems. Adding models of the other components of a CANDU reactor would ultimately result in a complete model of CANDU plant in MATLAB/SIMULINK. (author)

  8. Key thrusts in next generation CANDU. Annex 10

    International Nuclear Information System (INIS)

    Shalaby, B.A.; Torgerson, D.F.; Duffey, R.B.

    2002-01-01

    Current electricity markets and the competitiveness of other generation options such as CCGT have influenced the directions of future nuclear generation. The next generation CANDU has used its key characteristics as the basis to leap frog into a new design featuring improved economics, enhanced passive safety, enhanced operability and demonstrated fuel cycle flexibility. Many enabling technologies spinning of current CANDU design features are used in the next generation design. Some of these technologies have been developed in support of existing plants and near term designs while others will need to be developed and tested. This paper will discuss the key principles driving the next generation CANDU design and the fuel cycle flexibility of the CANDU system which provide synergism with the PWR fuel cycle. (author)

  9. Diagnostic Technology Development for Core Internal Structure in CANDU reactor

    International Nuclear Information System (INIS)

    Jung, Hyun Kyu; Cheong, Y. M.; Lee, Y. S. and others

    2005-04-01

    Degradation of critical components of nuclear power plants has become important as the operating years of plants increase. The necessity of degradation study including measurement and monitoring technology has increased continuously. Because the fuel channels and the neighboring sensing tubes and control rods 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 CANDU reactor safety improvement. To ensure the core structure integrity in CANDU nuclear plant, the following 2 research tasks were performed: Development of NDE technologies for the gap measurement between the fuel channels and LIN tubes. Development of vibration monitoring technology of the fuel channels and sensing tubes. The technologies developed in this study could contribute to the nuclear safety and estimation of the remaining life of operating CANDU nuclear power plants

  10. The ageing of CANDU steam generator due to localized corrosion

    International Nuclear Information System (INIS)

    Lucan, D.; Fulger, M.; Jinescu, Ghe.

    2001-01-01

    The principal types of corrosion are presented which can occur in CANDU steam generator. There are also presented the operation conditions, the specifications referring to the water chemistry and the construction materials of Steam Generator, the factors that have a great influence on the corrosion behaviour during the whole exploitation period of this equipment. The most important elements of CANDU Steam Generator ageing management program are also discussed. (R. P.)

  11. Controllability studies for an advanced CANDU boiling light water reactor

    International Nuclear Information System (INIS)

    Lepp, R.M.; Hinds, H.W.

    1976-12-01

    Bulk controllability studies carried out as part of a conceptual design study of a 1200 MWe CANDU boiling-light-water reactor fuelled with U 235 - or Pu-enriched uranium oxide are outlined. The concept, the various models developed for its simulation on a hybrid computer and the perturbations used to test system controllability, are described. The results show that this concept will have better bulk controllability than similar CANDU-BLW reactors fuelled with natural uranium. (author)

  12. Thermal-hydraulic interfacing code modules for CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Liu, W.S.; Gold, M.; Sills, H. [Ontario Hydro Nuclear, Toronto (Canada)] [and others

    1997-07-01

    The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis.

  13. Results of the CANDU 3 probabilistic safety assessment

    International Nuclear Information System (INIS)

    Jaitly, R.K.

    1995-01-01

    The purpose of the Conceptual Probabilistic Safety Assessment (PSA) of the CANDU 3 reactor was to provide safety assistance in the early stages of design to ensure that the design included adequate redundancy and functional separation of the mitigating systems; the final design should therefore give better results, particularly after modifications involving control, electrical power, instrument air, and service water. The initial PSA gave a total CANDU 3 core damage frequency of 7.8 x 10 -6 /year. 4 refs., 1 fig

  14. Thermal-hydraulic interfacing code modules for CANDU reactors

    International Nuclear Information System (INIS)

    Liu, W.S.; Gold, M.; Sills, H.

    1997-01-01

    The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis

  15. Highlights of the metallurgical behaviour of CANDU pressure tubes

    International Nuclear Information System (INIS)

    Price, E.G.

    1984-10-01

    This paper is an overview of the service induced metallurgical changes that take place in Zircaloy-2 and Zr-2.5 wt. percent Nb pressure tubes in CANDU reactors. It incorporates the findings of an evaluation program, that followed a significant pressure tube failure at Ontario Hydro's Pickering Nuclear Generating Station, and also provides valid reasons for continued confidence in the current CANDU design

  16. Leak detection capability in CANDU reactors

    International Nuclear Information System (INIS)

    Azer, N.; Barber, D.H.; Boucher, P.J.

    1997-01-01

    This paper addresses the moisture leak detection capability of Ontario Hydro CANDU reactors which has been demonstrated by performing tests on the reactor. The tests confirmed the response of the annulus gas system (AGS) to the presence of moisture injected to simulate a pressure tube leak and also confirmed the dew point response assumed in leak before break assessments. The tests were performed on Bruce A Unit 4 by injecting known and controlled rates of heavy water vapor. To avoid condensation during test conditions, the amount of moisture which could be injected was small (2-3.5 g/hr). The test response demonstrated that the AGS is capable of detecting and annunciating small leaks. Thus confidence is provided that it would alarm for a growing pressure tube leak where the leak rate is expected to increase to kg/hr rapidly. The measured dew point response was close to that predicted by analysis

  17. CANDU Digital Control Computer upgrade options

    International Nuclear Information System (INIS)

    De Jong, M.S.; De Grosbois, J.; Qian, T.

    1997-01-01

    This paper reviews the evolution of Digital Control Computers (DCC) in CANDU power plants to the present day. Much of this evolution has been to meeting changing control or display requirements as well as the replacement of obsolete, or old and less reliable technology with better equipment that is now available. The current work at AECL and Canadian utilities to investigate DCC upgrade options, alternatives, and strategies are examined. The dependence of a particular upgrade strategy on the overall plant refurbishment plans are also discussed. Presently, the upgrade options range from replacement of individual obsolete system components, to replacement of the entire DCC hardware without changing the software, to complete replacement of the DCCs with a functionally equivalent system using new control computer equipment and software. Key issues, constraints and objectives associated with these DCC upgrade options are highlighted. (author)

  18. Tests on CANDU fuel elements sheath samples

    International Nuclear Information System (INIS)

    Ionescu, S.; Uta, O.; Mincu, M.; Prisecaru, I.

    2016-01-01

    This work is a study of the behavior of CANDU fuel elements after irradiation. The tests are made on ring samples taken from fuel cladding in INR Pitesti. This paper presents the results of examinations performed in the Post Irradiation Examination Laboratory. By metallographic and ceramographic examination we determinate that the hydride precipitates are orientated parallel to the cladding surface. A content of hydrogen of about 120 ppm was estimated. After the preliminary tests, ring samples were cut from the fuel rod, and were subject of tensile test on an INSTRON 5569 model machine in order to evaluate the changes of their mechanical properties as consequence of irradiation. Scanning electron microscopy was performed on a microscope model TESCAN MIRA II LMU CS with Schottky FE emitter and variable pressure. The analysis shows that the central zone has deeper dimples, whereas on the outer zone, the dimples are tilted and smaller. (authors)

  19. Hydrogen distribution studies relevant to CANDU containments

    International Nuclear Information System (INIS)

    Krause, M.; Whitehouse, D.R.; Chan, C.K.; Jones, S.C.A.

    1995-01-01

    Following a loss of coolant accident with coincident loss of emergency core cooling, hydrogen may be produced in a CANDU reactor from the in-core Zircaloy-steam reaction, and released into containment. To meet the requirements for predicting containment hydrogen distribution, and to support measures for mitigation, a computer code GOTHIC is used. Simulations of gas mixing were performed using simple well defined experiments in a small-scale compartment, helium being substituted for hydrogen. At the time of the conference, results indicated that GOTHIC could quantitatively predict the stratified gas distribution resulting from buoyant gas injection near the bottom of an unobstructed compartment. When gas was injected near the top, GOTHIC underpredicted maximum gas concentration at the top, and overpredicted mixing. These errors arise from the finite-volume approximation. 2 refs., 11 figs

  20. Leak detection capability in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Azer, N.; Barber, D.H.; Boucher, P.J. [and others

    1997-04-01

    This paper addresses the moisture leak detection capability of Ontario Hydro CANDU reactors which has been demonstrated by performing tests on the reactor. The tests confirmed the response of the annulus gas system (AGS) to the presence of moisture injected to simulate a pressure tube leak and also confirmed the dew point response assumed in leak before break assessments. The tests were performed on Bruce A Unit 4 by injecting known and controlled rates of heavy water vapor. To avoid condensation during test conditions, the amount of moisture which could be injected was small (2-3.5 g/hr). The test response demonstrated that the AGS is capable of detecting and annunciating small leaks. Thus confidence is provided that it would alarm for a growing pressure tube leak where the leak rate is expected to increase to kg/hr rapidly. The measured dew point response was close to that predicted by analysis.

  1. Nondestructive examination techniques on Candu fuel elements

    International Nuclear Information System (INIS)

    Gheorghe, G.; Man, I.

    2013-01-01

    During irradiation in nuclear reactor, fuel elements undergo dimensional and structural changes, and changes of surface conditions sheath as well, which can lead to damages and even loss of integrity. Visual examination and photography of Candu fuel elements are among the non-destructive examination techniques, next to dimensional measurements that include profiling (diameter, bending, camber) and length, sheath integrity control with eddy currents, measurement of the oxide layer thickness by eddy current techniques. Unirradiated Zircaloy-4 tubes were used for calibration purposes, whereas irradiated Zircaloy-4 tubes were actually subjected to visual inspection and dimensional measurements. We present results of measurements done by eddy current techniques on Zircaloy- 4 tubes, unirradiated, but oxidized in an autoclave prior to examinations. The purpose of these nondestructive examination techniques is to determine those parameters that characterize the behavior and performance of nuclear fuel operation. (authors)

  2. Inspection of Candu Nuclear Reactor Fuel Channels

    International Nuclear Information System (INIS)

    Baron, J.; Jarvis, G.N.; Dolbey, M.P.; Hayter, D.M.

    1986-01-01

    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) [pt

  3. The future role of thorium in assuring CANDU fuel supplies

    International Nuclear Information System (INIS)

    Slater, J.B.

    1985-01-01

    Atomic Energy of Canada Limited (AECL), in partnership with Canadian industry and power utilities, has developed the CANDU reactor as a safe, reliable and economic means of transforming nuclear fuel into useable power. The use of thorium/uranium-233 recycle gives the possibility of a many-fold increase in energy yield over that which can be obtained from the use of uranium in once-through cycles. The neutronic properties of uranium-233 combine with the inherent neutron economy of the CANDU reactor to offer the possibility of near-breeder cycles in which there is no net consumption of fissile material under equilibrium fuelling conditions. Use of thorium cycles in CANDU will limit the impact of higher uranium prices. When combined with the potential for significant reductions in CANDU capital costs, then the long-term prospect is for generating costs near to current levels. Development of thorium cycles in CANDU will safeguard against possible uranium shortages in the next century, and will maintain and continue the commercial viability of CANDU as a long-term energy technology. (author)

  4. Qinshan CANDU NPP outage performance improvement through benchmarking

    International Nuclear Information System (INIS)

    Jiang Fuming

    2005-01-01

    With the increasingly fierce competition in the deregulated Energy Market, the optimization of outage duration has become one of the focal points for the Nuclear Power Plant owners around the world. People are seeking various ways to shorten the outage duration of NPP. Great efforts have been made in the Light Water Reactor (LWR) family with the concept of benchmarking and evaluation, which great reduced the outage duration and improved outage performance. The average capacity factor of LWRs has been greatly improved over the last three decades, which now is close to 90%. CANDU (Pressurized Heavy Water Reactor) stations, with its unique feature of on power refueling, of nuclear fuel remaining in the reactor all through the planned outage, have given raise to more stringent safety requirements during planned outage. In addition, the above feature gives more variations to the critical path of planned outage in different station. In order to benchmarking again the best practices in the CANDU stations, Third Qinshan Nuclear Power Company (TQNPC) have initiated the benchmarking program among the CANDU stations aiming to standardize the outage maintenance windows and optimize the outage duration. The initial benchmarking has resulted the optimization of outage duration in Qinshan CANDU NPP and the formulation of its first long-term outage plan. This paper describes the benchmarking works that have been proven to be useful for optimizing outage duration in Qinshan CANDU NPP, and the vision of further optimize the duration with joint effort from the CANDU community. (authors)

  5. Proceedings of the international conference on CANDU fuel

    International Nuclear Information System (INIS)

    Hastings, I.J.

    1986-01-01

    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

  6. An update on CANDU fuel channel performance

    International Nuclear Information System (INIS)

    Price, E.G.; Holt, R.A; Wong, H.; Gautheir, P.; Ellis, P.J.; Slade, J.

    1998-01-01

    Fuel channel components, particularly pressure tubes, are monitored to establish changes in their condition expected from their service environment. Such monitoring takes the form of the in-service NDE, in-service sampling and periodic tube removal for destructive examination. The pressure tubes in early CANDU 6 stations have now reached mid-life. The expected trends in behaviour of the various properties of the pressure tubes are being confirmed. Diametral deformation potentially limits the life of some pressure tubes, and fuel developments, such as the CANFLEX fuel bundle, will provide additional CHF margins. Corrosion and the associated deuterium pick-up are monitored to assess if any change in rates have occurred. Both have historically continued on an essentially linear trend with time. Recent data indicate that the rates of pick-up could be gradually increasing. Changes in mechanical properties, particularly toughness, have reached saturation levels. In the early reactors, the toughness is acceptable, assuring leak-before-break. The pressure tubes in later reactors have very low concentrations of chlorine and phosphorous and the level of hydrogen are also low, and these tubes are predicted to retain relativity high fracture toughness in-service. Debris fretting wear has occurred in CANDU reactors. Laboratory testing and experience show that no cracks have initiated from debris frets to date. Further debris fretting will have to be avoided by care in maintenance procedures, and existing defects must be monitored in the future as hydrogen concentration levels increase. Spacer location and repositioning (SLAR) has been necessary in earlier reactors with loose spacers. Procedures and equipment for SLAR have been developed and successfully applied. The efficiency of spacer relocation operations has improved with experience. (authors)

  7. TEPC performance in the CANDU workplace

    International Nuclear Information System (INIS)

    Waker, A.J.; Szornel, K.; Nunes, J.

    1997-01-01

    Tissue-equivalent proportional counters (TEPCs) have a number of features that make them an attractive option for neutron monitoring around power reactors. These features are principally the ability of the TEPC to operate in a mixed field environment, the direct determination of the dose equivalent from first principles and a reasonably well-understood response to the wide range of neutron energy encountered. A spherical TEPC from Far West Technology (5'' in diameter) and a commercial, TEPC-based, neutron monitor (REM 500) have been used to map the neutron fields at different locations in a CANDU 6 (Canadian Deuterium Uranium) power reactor operated by New Brunswick Power. Neutron ambient dose equivalent rates ranged between 8 μSv.h -1 and 500 μSv.h -1 as measured with a multisphere spectrometer and photon dose equivalent rates from 40 μSv.h -1 to 1.2 mSv.H -1 as determined with a TEPC. It is shown that, for the CANDU workplace, the energy response of TEPCs is still not known accurately enough to enable field correction factors to be derived from neutron fluence measurements or to avoid the the need for more appropriate calibration procedures based on simulated 'workplace' neutron fields. The sensitivity of TEPCs is sufficient for workplace monitoring, but needs consideration when counters of 2'' diameter or smaller are proposed for use. For mixed-field work, the use of a TEPC for photon dosimetry as opposed to photon discrimination requires attention to count rates and instrument dead time. (author)

  8. Life extension of CANDU reactor cores

    International Nuclear Information System (INIS)

    Millard, J.; Kerker, J.; Albert, M.

    2011-01-01

    Candu Energy (formerly AECL), in partnership with station operators, has developed a robust methodology for demonstrating the fitness of reactor core structures, and associated reactivity control devices, as an essential element in conducting a station life extension project. The ageing of reactors is affected by ageing mechanisms impacted by operational history and design related factors such as materials, chemistries and stress distributions. The methodology of this life extension work is based on the IAEA TECDOC 1197; which documents practices for ageing management in CANDU reactors. This paper uses the work in Bruce Units 1 and 2, conducted from 2007 through to 2011, to explain the methodology. The work started with analysis of historical operational conditions and identification of the forms of degradation that could have occurred. The assessment and related inspections considered the safety and pressure boundary significance of each item, as well as its failure modes and margins. It then moved through both general and local inspection, focused mainly inside the calandria vessel once the calandria tubes were removed. The inspection found the bulk of the hardware to be in good condition, with a small number of remediation opportunities. In the course of that remediation some foreign material was sampled and removed. The minor remediation was successful and the work was completed through formal documentation of the fitness for extended life. It has been demonstrated through these analyses and visual inspections that the reactor structures and components inspected are free of indications and active degradation mechanisms that would prevent the safe and reliable operation of Bruce A Units 1 and 2 through its next 25 years of life. (author)

  9. Cobalt-60 production in CANDU reactors

    International Nuclear Information System (INIS)

    Ross, Michel; Lemire, Christian

    2002-01-01

    CANDU reactors can produce cobalt-60 very efficiently and with an interesting return on investment. This paper discusses what is needed to convert a CANDU reactor into a cobalt-60 producer: what are the different phases, the safety studies required, the physical modifications needed, and what is the minimum involvement of the utility owning the plant. The past ten years of experience of Hydro-Quebec as a cobalt-60 producer will be reviewed, including the management of the risk of both incident and electricity generation loss, and including the benefits for the utility and its personnel. Originally a simple metal used for centuries as a pigment, cobalt-59 today is transformed into cobalt-60, a radioactive element of unprecedented value. Well known in medicine for cancer treatment, cobalt-60 is also used to sterilize a wide range of disposable medical products used in hospitals and to sanitize pharmaceutical and cosmetic products. Cobalt-60 is proving to be a new and effective solution, in the food sector, for preserving harvests and controlling food-borne diseases, or to advantageously replace certain gases and chemical products which are suspected of being harmful or carcinogenic. There are also other applications, such as: hardening of some plastics, treatment of sewage sludge and elimination of harmful insect populations. With a half-life of 5,3 years, cobalt-60 is a metal not found in nature. It is a radioactive isotope produced by exposing stable nuclei of cobalt-59 to neutrons. One of the best places to find such an important neutron source is a nuclear reactor. High energy gamma rays are then emitted during the process of radioactive decay, where cobalt-60 seeks again its stable state

  10. Safety Parameters for the Recycled Uranium Loaded into a CANDU Reactor

    International Nuclear Information System (INIS)

    Park, Chang Je; Kang, Kweon Ho; Na, Sang Ho; Kim, Young Hee; Ryu, Ho Jin; Park, Geun Il; Song, Kee Chan

    2008-01-01

    In order to recover uranium and TRU from spent nuclear fuels, a pyroprocessing has been developed through a dry and metallurgical reprocess technology using a series of electrolyses such as an electro-reduction, an electro-refining, and an electro-winning. When the spent fuel is being fed into the pyroprocess, most of the uranium is gathered in metallic form around a solid cathode during an electro-refining process. It is expected that the recovered uranium will be sent to a spent fuel storage site after converting it into a metal ingot form to reduce its storage space and transportation burden. However, the weight percent of U-235 in the recovered uranium is about 0.9 wt% and it is sufficiently re-utilized in a heavy water reactor which uses a natural uranium fuel. The reuse of recovered uranium will bring not only a huge economical profit and save of uranium resources but also an alleviation of burden on the management and disposal of the spent fuel. A previous research on recycling of recovered uranium was carried out and most of the recovered uranium was assumed to be imported from abroad at that time. The preliminary results showed there is a sufficient possibility to recycle recovered uranium in terms of a reactor's characteristics as well as the fuel performance. And the DUPIC (direct use of spent pressurized water reactor fuel into CANDU reactor) program has also been performed and demonstrated the fundamental technologies. The recovered uranium from a pyroprocess contains some TRU as an impurity and it will exhibit a slightly different behavior from the previous recycling options. In this paper, the reactor's characteristics including safety parameters are investigated based on the lattice calculations which are performed for the CANFELX bundle

  11. Applying operating experience to design the CANDU 3 process

    International Nuclear Information System (INIS)

    Harris, D.S.; Hinchley, E.M.; Pauksens, J.; Snell, V.; Yu, S.K.W.

    1991-01-01

    The CANDU 3 is an advanced, smaller (450 MWe), standardized version of the CANDU now being designed for service later in the decade and beyond. The design of this evolutionary nuclear power plant has been carefully planned and organized to gain maximum benefits from new technologies and from world experience to date in designing, building, commissioning and operating nuclear power stations. The good performance record of existing CANDU reactors makes consideration of operating experience from these plants a particularly vital component of the design process. Since the completion of the first four CANDU 6 stations in the early 1980s, and with the continuing evolution of the multi-unit CANDU station designs since then, AECL CANDU has devised several processes to ensure that such feedback is made available to designers. An important step was made in 1986 when a task force was set up to review and process ideas arising from the commissioning and early operation of the CANDU 6 reactors which were, by that time, operating successfully in Argentina and Korea, as well as the Canadian provinces of Quebec and New Brunswick. The task force issued a comprehensive report which, although aimed at the design of an improved CANDU 6 station, was made available to the CANDU 3 team. By that time also, the Institute of Power Operations (INPO) in the U.S., of which AECL is a Supplier Participant member, was starting to publish Good Practices and Guidelines related to the review and the use of operating experiences. In addition, details of significant events were being made available via the INPO SEE-IN (Significant Event Evaluation and Information Network) Program, and subsequently the CANNET network of the CANDU Owners' Group (COG). Systematic review was thus possible by designers of operations reports, significant event reports, and related documents in a continuing program of design improvement. Another method of incorporating operations feedback is to involve experienced utility

  12. Applying operating experience to design the CANDU 3 process

    Energy Technology Data Exchange (ETDEWEB)

    Harris, D S; Hinchley, E M; Pauksens, J; Snell, V; Yu, S K.W. [AECL-CANDU, Ontario (Canada)

    1991-04-01

    The CANDU 3 is an advanced, smaller (450 MWe), standardized version of the CANDU now being designed for service later in the decade and beyond. The design of this evolutionary nuclear power plant has been carefully planned and organized to gain maximum benefits from new technologies and from world experience to date in designing, building, commissioning and operating nuclear power stations. The good performance record of existing CANDU reactors makes consideration of operating experience from these plants a particularly vital component of the design process. Since the completion of the first four CANDU 6 stations in the early 1980s, and with the continuing evolution of the multi-unit CANDU station designs since then, AECL CANDU has devised several processes to ensure that such feedback is made available to designers. An important step was made in 1986 when a task force was set up to review and process ideas arising from the commissioning and early operation of the CANDU 6 reactors which were, by that time, operating successfully in Argentina and Korea, as well as the Canadian provinces of Quebec and New Brunswick. The task force issued a comprehensive report which, although aimed at the design of an improved CANDU 6 station, was made available to the CANDU 3 team. By that time also, the Institute of Power Operations (INPO) in the U.S., of which AECL is a Supplier Participant member, was starting to publish Good Practices and Guidelines related to the review and the use of operating experiences. In addition, details of significant events were being made available via the INPO SEE-IN (Significant Event Evaluation and Information Network) Program, and subsequently the CANNET network of the CANDU Owners' Group (COG). Systematic review was thus possible by designers of operations reports, significant event reports, and related documents in a continuing program of design improvement. Another method of incorporating operations feedback is to involve experienced utility

  13. Main attributes influencing spent nuclear fuel management

    International Nuclear Information System (INIS)

    Andreescu, N.; Ohai, D.

    1997-01-01

    All activities regarding nuclear fuel, following its discharge from the NPP, constitute the spent fuel management and are grouped in two possible back end variants, namely reprocessing (including HLW vitrification and geological disposal) and direct disposal of spent fuel. In order to select the appropriate variant it is necessary to analyse the aggregate fulfillment of the imposed requirements, particularly of the derived attributes, defined as distinguishing characteristics of the factors used in the decision making process. The main identified attributes are the following: - environmental impact, - availability of suitable sites, - non-proliferation degree, -strategy of energy, - technological complexity and technical maturity, -possible further technical improvements, - size of nuclear programme, - total costs, - public acceptance, - peculiarity of CANDU fuel. The significance of the attributes in the Romanian case, taking into consideration the present situation, as a low scenario and a high scenario corresponding to an important development of the nuclear power, after the year 2010, is presented. According to their importance the ranking of attributes is proposed . Subsequently, the ranking could be used for adequate weighing of attributes in order to realize a multi-criteria analysis and a relevant comparison of back end variants. (authors)

  14. Muon Tomography for Imaging and Verification of Spent Fuel

    International Nuclear Information System (INIS)

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

    2010-01-01

    This paper explores the use of cosmic ray muons to image the content of, and to detect high-Z Special Fissionable Material inside, shielded containers. 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 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 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 to perform material accountancy, in the case of the DSC, exceeds the IAEA detection target for nuclear material accountancy. (author)

  15. Spent fuel bundle counter sequence error manual - BRUCE NGS

    International Nuclear Information System (INIS)

    Nicholson, L.E.

    1992-01-01

    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

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

    International Nuclear Information System (INIS)

    Nicholson, L.E.

    1992-01-01

    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

  17. Proceedings of the Korean Radioactive Waste Society Spring 2010

    International Nuclear Information System (INIS)

    2010-01-01

    This proceedings contains articles of the Korean Radioactive Waste Society Spring 2010. It was held on May 13-14, 2010 in Yesan, Korea. The main topics are as follows: Radioactive wastes policy and decontamination and decommissioning, Radioactive waste treatment, Radioactive waste disposal and site selection, Spent fuel and fuel cycle and Radiation safety and environment. (Yi, J. H.)

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

    International Nuclear Information System (INIS)

    Yilmaz, S.

    2000-01-01

    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

  19. Cost and schedule reduction for next-generation Candu

    International Nuclear Information System (INIS)

    Hopwood, J.M.; Yu, S.; Pakan, M.; Soulard, M.

    2002-01-01

    AECL has developed a suite of technologies for Candu R reactors that enable the next step in the evolution of the Candu family of heavy-water-moderated fuel-channel reactors. These technologies have been combined in the design for the Advanced Candu Reactor TM1 (ACRTM), AECL's next generation Candu power plant. The ACR design builds extensively on the existing Candu experience base, but includes innovations, in design and in delivery technology, that provide very substantial reductions in capital cost and in project schedules. In this paper, main features of next generation design and delivery are summarized, to provide the background basis for the cost and schedule reductions that have been achieved. In particular the paper outlines the impact of the innovative design steps for ACR: - Selection of slightly enriched fuel bundle design; - Use of light water coolant in place of traditional Candu heavy water coolant; - Compact core design with unique reactor physics benefits; - Optimized coolant and turbine system conditions. In addition to the direct cost benefits arising from efficiency improvement, and from the reduction in heavy water, the next generation Candu configuration results in numerous additional indirect cost benefits, including: - Reduction in number and complexity of reactivity mechanisms; - Reduction in number of heavy water auxiliary systems; - Simplification in heat transport and its support systems; - Simplified human-machine interface. The paper also describes the ACR approach to design for constructability. The application of module assembly and open-top construction techniques, based on Candu and other worldwide experience, has been proven to generate savings in both schedule durations and overall project cost, by reducing premium on-site activities, and by improving efficiency of system and subsystem assembly. AECL's up-to-date experience in the use of 3-D CADDS and related engineering tools has also been proven to reduce both engineering and

  20. Design of a PWR for long cycle and direct recycling of spent fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-15

    Highlights: • Single-batch loading PWR with a new fuel assembly for 36 calendar months cycle was designed. • The new fuel assembly is constructed from a number of CANDU fuel bundles. • This design enables to recycle the spent fuel directly in CANDU reactors for high burnup. • Around 56 MWd/kgU burnup is achieved from fuel that has average enrichment of 4.8 w/o U-235 using this strategy. • Safety parameters such as the power distribution and CANDU coolant void reactivity were considered. - Abstract: In a previous work, a new design was proposed for the Pressurized Water Reactor (PWR) fuel assembly for direct use of the PWR spent fuel without processing. The proposed assembly has four zircaloy-4 tubes contains a number of 61-element CANDU fuel bundles (8 bundles per tube) stacked end to end. The space between the tubes contains 44 lower enriched UO{sub 2} fuel rods and 12 guide tubes. In this paper, this assembly is used to build a single batch loading 36-month PWR and the spent CANDU bundles are recycled in the on power refueling CANDU reactors. The Advanced PWR (APWR) is considered as a reference design. The average enrichment in the core is 4.76%w U-235. IFBA and Gd{sub 2}O{sub 3} as burnable poisons are used for controlling the excess reactivity and to flatten the power distribution. The calculations using MCNPX showed that the PWR will discharge the fuel with average burnup of 31.8 MWd/kgU after 1000 effective full power days. Assuming a 95 days plant outage, 36 calendar months can be achieved with a capacity factor of 91.3%. Good power distribution in the core is obtained during the cycle and the required critical boron concentration is less than 1750 ppm. Recycling of the discharged CANDU fuel bundles that represents 85% of the fuel in the assembly, in CANDU-6 or in 700 MWe Advanced CANDU Reactor (ACR-700), an additional burnup of about 31 or 26 MWd/kgU burnup can be achieved, respectively. Averaging the fuel burnup on the all fuel in the PWR

  1. Passive safety features for next generation CANDU power plants

    International Nuclear Information System (INIS)

    Natalizio, A.; Hart, R.S.; Lipsett, J.J.; Soedijono, P.; Dick, J.E.

    1989-01-01

    CANDU offers an evolutionary approach to simpler and safer reactors. The CANDU 3, an advanced CANDU, currently in the detailed design stage, offers significant improvements in the areas of safety, design simplicity, constructibility, operability, maintainability, schedule and cost. These are being accomplished by retaining all of the well known CANDU benefits, and by relying on the use of proven components and technologies. A major safety benefit of CANDU is the moderator system which is separate from the coolant. The presence of a cold moderator reduces the consequences arising from a LOCA or loss of heat sink event. In existing CANDU plants even the severe accident - LOCA with failure of the emergency core cooling system - is a design basis event. Further advances toward a simpler and more passively safe reactor will be made using the same evolutionary approach. Building on the strength of the moderator system to mitigate against severe accidents, a passive moderator cooling system, depending only on the law of gravity to perform its function, will be the next step of development. AECL is currently investigating a number of other features that could be incorporated in future evolutionary CANDU designs to enhance protection against accidents, and to limit off-site consequences to an acceptable level, for even the worst event. The additional features being investigated include passive decay heat removal from the heat transport system, a simpler emergency core cooling system and a containment pressure suppression/venting capability for beyond design basis events. Central to these passive decay heat removal schemes is the availability of a short-term heat sink to provide a decay heat removal capability of at least three days, without any station services. Preliminary results from these investigations confirm the feasibility of these schemes. (author)

  2. Hyperfine 3D neutronic calculations in CANDU supercells

    International Nuclear Information System (INIS)

    Balaceanu, V.; Aioanei, L.; Pavelescu, M.

    2010-01-01

    For an accurate evaluation of the fuel performances, it is very important to have capability to calculate the three dimensional spatial flux distributions in the fuel bundle. According this issue, in our Institute, a multigroup calculation methodology named WIMS-PIJXYZ was especially developed for estimating the local neutronic parameters in CANDU cell/supercells. The objective of this paper is to present this calculation methodology and to use it in performing some hyperfine neutronic calculations in CANDU type supercells. More exactly, after a short description for the WIMS-PIJXYZ methodology, the end effect for some CANDU fuel bundles is estimated. The WIMS-PIJXYZ methodology is based on WIMS and PIJXYZ transport codes. WIMS is a standard lattice-cell code and it is used for generating the multigroup macroscopic cross sections for the materials in the fuel cells. For obtaining the flux and power distributions in CANDU fuel bundles the PIJXYZ code is used. This code is consistent with WIMS lattice-cell calculations and allows a good geometrical representation of the CANDU bundle in three dimensions. The end effect consists in the increasing of the thermal neutron flux in the end region and the increasing of power in the end of the fuel rod. The region separating the CANDU fuel in two adjoining bundles in a channel is called the 'end region' and the end of the last pellet in the fuel stack adjacent to the end region is called the 'fuel end'. The end effect appears because the end region of the bundle is made up of coolant and Zircaloy-4, a very low neutron absorption material. To estimate the end effect, the flux peaking factors and the power peaking factors are calculated. It was taken in consideration CANDU Standard (Natural Uranium, with 37 elements) fuel bundles. In the end of the paper, the results obtained by WIMS-PIJXYZ methodology with the similar LEGENTR results are compared. The comparative analysis shows a good agreement. (authors)

  3. CANDU in Korea, present and future - Operating performance of Wolsong 1 and construction of Wolsong 2, 3 and 4

    International Nuclear Information System (INIS)

    Hong, Joo Bo

    1994-01-01

    The excellent performance of Wolsong-1 reactor, attributable to the excellence of CANDU features and facilities, and to traditional Korean diligence and dedication, encouraged the purchase of Units 2, 3, and 4. Units 2 and 3 were under construction at the time of the conference, while Unit 4 was about to be started; some information on contracts and scheduling is provided. Modifications to the design of Wolsong-1 included an additional separator in the steam generators, and conversion of the raw water service system and the recirculated cooling water system to dual loop design. Concerns with regard to maintenance of Wolsong-1 included pressure tube integrity, delayed neutron tube maintenance, and inadequate supply of spare parts and system components (particularly for the plant control computer and the fuelling machines). The expectation was that these problems would be remedied in time for the later units. 9 tabs., 2 figs

  4. Fission products distributions in Candu primary heat transport and Candu containment systems during a severe accident

    International Nuclear Information System (INIS)

    Constantin, Marin; Rizoiu, Andrei

    2005-01-01

    The paper is intended to analyse the distribution of the fission products (FPs) in CANDU Primary Heat Transport (PHT) and CANDU Containment Systems by using the ASTEC code (Accident Source Term Evaluation Code). The complexity of the data required by ASTEC and the complexity both of CANDU PHT and Containment System were strong motivations to begin with a simplified geometry in order to avoid the introducing of unmanageable errors at the level of input deck. Thus only 1/4 of the PHT circuit was simulated and a simplified FPs inventory, some simplifications in the feeders geometry and containment were used. The circuit consists of 95 horizontal fuel channels connected to 95 horizontal out-feeders, then through vertical feeders to the outlet-header (a big pipe that collects the water from feeders); the circuit continues from the outlet-header with a riser and then with the steam generator and a pump. After this pump, the circuit was broken; in this point the FPs are transferred to the containment. The containment model consists of 4 rooms connected between by 6 links. The data related to the nodes' definitions, temperatures and pressure conditions were chosen as possible as real data from CANDU NPP loss of coolant accident sequence. Temperature and pressure conditions in the time of the accident were calculated by the CATHENA code and the source term of FPs introduced into the PHT was estimated by the ORIGEN code. The FPs distribution in the nodes of the circuit and the FPs mass transfer per isotope and chemical species are obtained by using SOPHAEROS module of ASTEC code. The distributions into the containment are obtained by the CPA module of ASTEC code (thermalhydraulics calculations in the containment and FPs aerosol transport). The results consist of mass distributions in the nodes of the circuit and the transferred mass to the containment through the break for different species (FPs and chemical species) and mass distributions in the different parts and

  5. Value added services to CANDU plants

    International Nuclear Information System (INIS)

    Kakaria, B.K.

    2003-01-01

    Over the last decade or so, nuclear power plants, just like other types of electricity generating plants, have been facing a number of challenges. Depending on the operating environment of the utility, these challenges are forcing plant owners to examine all facets of the operating costs. Privatization, deregulation and economics of alternative electricity generation methods are exerting enormous pressure on nuclear power plants to streamline costs and improve their operational performance. CANDU reactors are no exception to these forces and face similar pressures. In particular, operating plants that are contemplating plant life extensions are being required to clearly demonstrate the economics of continued operation over other forms of power generation available to the utility. Improvement of capacity factors has the effect of increasing the revenues from the plant and as these revenues increase, the fixed portion of the plant costs including OM and A costs become a smaller percentage of the total revenues. Similar results can be achieved by aiming to reduce the plant OM and A costs. In reality, most well-planned intervention schemes directed at reducing OM and A costs tend to also increase the plant availability. Following plant turnover after commissioning, AECL has been supporting the CANDU owners and utilities with an assortment of products and services dealing with plant operations and outage management issues. AECL has taken the lead in arranging specialized resources, products and services by teaming with other complementary organizations to provide a complete suite of services. Recent examples of such support to operating CANDU plants will be described in the paper. AECL is responding to this changing business environment in two important ways. First, AECL is changing from simply providing a service to its clients towards providing value, something much more important. To this end, AECL is looking to other organizations to form alliances, partnerships and

  6. Development of a computer program for the cost analysis of spent fuel management

    International Nuclear Information System (INIS)

    Choi, Heui Joo; Lee, Jong Youl; Choi, Jong Won; Cha, Jeong Hun; Whang, Joo Ho

    2009-01-01

    So far, a substantial amount of spent fuels have been generated from the PWR and CANDU reactors. They are being temporarily stored at the nuclear power plant sites. It is expected that the temporary storage facility will be full of spent fuels by around 2016. The government plans to solve the problem by constructing an interim storage facility soon. The radioactive management act was enacted in 2008 to manage the spent fuels safety in Korea. According to the act, the radioactive waste management fund which will be used for the transportation, interim storage, and the final disposal of spent fuels has been established. The cost for the management of spent fuels is surprisingly high and could include a lot of uncertainty. KAERI and Kyunghee University have developed cost estimation tools to evaluate the cost for a spent fuel management based on an engineering design and calculation. It is not easy to develop a tool for a cost estimation under the situation that the national policy on a spent fuel management has not yet been fixed at all. Thus, the current version of the computer program is based on the current conceptual design of each management system. The main purpose of this paper is to introduce the computer program developed for the cost analysis of a spent fuel management. In order to show the application of the program, a spent fuel management scenario is prepared, and the cost for the scenario is estimated

  7. 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

  8. Basic research and industrialization of CANDU advanced fuel

    International Nuclear Information System (INIS)

    Chun, Suk Ho; Park, Joo Hwan; Jun, Ji Su

    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

  9. The North Korean nuclear dilemma

    International Nuclear Information System (INIS)

    Hecker, Siegfried S.

    2004-01-01

    The current nuclear crisis, the second one in ten years, erupted when North Korea expelled international nuclear inspectors in December 2002, then withdrew from the Nuclear Nonproliferation Treaty (NPT), and claimed to be building more nuclear weapons with the plutonium extracted from the spent fuel rods heretofore stored under international inspection. These actions were triggered by a disagreement over U.S. assertions that North Korea had violated the Agreed Framework (which froze the plutonium path to nuclear weapons to end the first crisis in 1994) by clandestinely developing uranium enrichment capabilities providing an alternative path to nuclear weapons. With Stanford University Professor John Lewis and three other Americans, I was allowed to visit the Yongbyon Nuclear Center on Jan. 8, 2004. We toured the 5 MWe reactor, the 50 MWe reactor construction site, the spent fuel pool storage building, and the radiochemical laboratory. We concluded that North Korea has restarted its 5 MWe reactor (which produces roughly 6 kg of plutonium annually), it removed the 8000 spent fuel rods that were previously stored under IAEA safeguards from the spent fuel pool, and that it most likely extracted the 25 to 30 kg of plutonium contained in these fuel rods. Although North Korean officials showed us what they claimed was their plutonium metal product from this reprocessing campaign, we were not able to conclude definitively that it was in fact plutonium metal and that it came from the most recent reprocessing campaign. Nevertheless, our North Korean hosts demonstrated that they had the capability, the facility and requisite capacity, and the technical expertise to produce plutonium metal. On the basis of our visit, we were not able to address the issue of whether or not North Korea had a 'deterrent' as claimed - that is, we were not able to conclude that North Korea can build a nuclear device and that it can integrate nuclear devices into suitable delivery systems. However

  10. CANDU 9 Design improvements based on experience feedback

    International Nuclear Information System (INIS)

    Yu, S. K. W.; Bonechi, M.; Snell, V. G.

    2000-01-01

    An evolutionary approach utilizing advance technologies has been implenented for the enhancement introduced in the CANDU 9 Nuclear Power Plant (NPP) design. The design of these systems and associated equipment has also benfited from experience feedback from operating CANDU stations and from including advanced products from CANDU engineering and research programs. This paper highlights the design features that contribute to the safety improvements of the CANDU 9 design, summarizes the analysis results which demonstrate the improved performance and also emphasizes design features which reduce operation and maintenance (Q and M) costs. The safety design features highlighted include the increased use of passive devices and heat sinks to achieve extensive system simplification; this also improves reliability and reduces maintenance workloads. System features that contribute to improved operability are also described. The CANDU 9 Control Center provides plant staff with enhanced operating, maintenance and diagnostics features which significantly improve operability, testing and maintainability due to the integration of human factors engineering with a systematic design process. (author)

  11. Advancing the CANDU reactor: From generation to generation

    International Nuclear Information System (INIS)

    Hopwood, Jerry; Duffey, Romney B.; Yu, Steven; Torgerson, Dave F.

    2006-01-01

    Emphasizing safety, reliability and economics, the CANDU reactor development strategy is one of continuous improvement, offering value and assured support to customers worldwide. The Advanced CANDU Reactor (ACR-1000) generation, designed by Atomic Energy of Canada Limited (AECL), meets the new economic expectation for low-cost power generation with high capacity factors. The ACR is designed to meet customer needs for reduced capital cost, shorter construction schedule, high plant capacity factor, low operating cost, increased operating life, simple component replacement, enhanced safety features, and low environmental impact. The ACR-1000 design evolved from the internationally successful medium-sized pressure tube reactor (PTR) CANDU 6 and incorporates operational feedback from eight utilities that operate 31 CANDU units. This technical paper provides a brief description of the main features of the ACR-1000, and its major role in the development path of the generations of the pressure tube reactor concept. The motivation, philosophy and design approach being taken for future generation of CANDU pressure tube reactors are described

  12. Canadian CANDU fuel development program and recent fuel operating experience

    International Nuclear Information System (INIS)

    Lau, J.H.K.; Inch, W.W.R.; Cox, D.S.; Steed, R.G.; Kohn, E.; Macici, N.N.

    1999-01-01

    This paper reviews the performance of the CANDU fuel in the Canadian CANDU reactors in 1997 and 1998. The operating experience demonstrates that the CANDU fuel has performed very well. Over the 2-year period, the fuel-bundle defect rate for all bundles irradiated in the Canadian CANDU reactors has remained very low, at between 0.006% to 0.016%. On a fuel element basis, this represents an element defect rate of less than about 0.0005%. One of the reasons for the good fuel performance is the support provided by the Canadian fuel research and development programs. These programs address operational issues and provide evolutionary improvements to the fuel products. The programs consist of the Fuel Technology Program, funded by the CANDU Owners Group, and the Advanced Fuel and Fuel Cycles Technology Program, funded by Atomic Energy of Canada Ltd. These 2 programs, which have been in place for many years, complement each other by sharing expert resources and experimental facilities. This paper describes the programs in 1999/2000, to provide an overview of the scope of the programs and the issues that these programs address. (author)

  13. The Canadian R and D program targeted at CANDU reactors

    International Nuclear Information System (INIS)

    Moeck, E.O.

    1988-01-01

    CANDU reactors produce electricity cheaply and reliably, with miniscule risk to the population and minimal impact on the environment. About half of Ontario's electricity and a third of New Brunswick's are generated by CANDU power plants. Hydro Quebec and utilities in Argentina, India, Pakistan, and the Republic of Korea also successfully operate CANDU reactors. Romania will soon join their ranks. The proven record of excellent performance of CANDUs is due in part to the first objective of the vigorous R and D program: namely, to sustain and improve existing CANDU power-plant technology. The second objective is to develop improved nuclear power plants that will remain competitive compared with alternative energy supplies. The third objective is to continue to improve our understanding of the processes underlying reactor safety and develop improved technology to mitigate the consequences of upset conditions. These three objectives are addressed by individual R and D programs in the areas of CANDU fuel channels, reduced operating costs, reduced capital costs, reactor safety research, and IAEA safeguards. The work is carried out mainly at three centres of Atomic Energy of Canada Limited--the Chalk River Nuclear Laboratories, the Whiteshell Nuclear Research Establishment, and the Sheridan Park Engineering Laboratories--and at Ontario Hydro's Research Laboratories. Canadian universities, consultants, manufacturers, and suppliers also provide expertise in their areas of specialization

  14. Understanding Korean food culture from Korean paintings

    OpenAIRE

    Hae Kyung Chung; Kyung Rhan Chung; Hung Ju Kim

    2016-01-01

    Background: In Korea, there are many traditional foods that have developed along with the country's rich history. In addition, various food cultures have developed through agricultural traditions, ritual ceremonies, and the sharing of affection. Paintings, works of calligraphy, and music demonstrate some of these cultural characteristics of Korean foods. Further research and analysis of Korean food culture using these data sources is currently underway. Methods: This paper focuses on the c...

  15. Development of CANDU core monitoring system

    International Nuclear Information System (INIS)

    Yoon, M. Y.; Yeam, C. S.; Kwon, O. H.; Kim, K. H.

    2003-01-01

    The research was performed to develop a CANDU Core Monitoring System(CCMS) that enables operators to have efficient core management by monitoring core power distribution, burnup distribution, and the other important core variables and managing the past core history for Wolsong Nuclear Power Plant(NPP) No. 1. CCMS uses RFSP(Reactor Fueling Simulation Program) for continuous core calculation by integrating the algorithm and assumptions validated and uses the information taken from DCC(Digital Control Computer) for the purpose of producing basic input data. CCMS could be largely divided into two modules; CCMS server program and CCMS client program. CCMS server program plays the role in automatic and continuous RFSP run and management of the past output data resulting from the run using Data Base Management System(DBMS). CCMS client program enables users to monitor current and past core status with GUI(Graphic-User Interface) environment predefined. The effectiveness of CCMS was verified by comparing the data resulted from field-test of the system for about 43 hours with the data used in the field of Wolsong NPP No. 1

  16. Development of CANDU core monitoring system

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, M. Y.; Yeam, C. S.; Kwon, O. H.; Kim, K. H. [Institute for Advanced Engineering, Yongin (Korea, Republic of)

    2003-07-01

    The research was performed to develop a CANDU Core Monitoring System(CCMS) that enables operators to have efficient core management by monitoring core power distribution, burnup distribution, and the other important core variables and managing the past core history for Wolsong Nuclear Power Plant(NPP) No. 1. CCMS uses RFSP(Reactor Fueling Simulation Program) for continuous core calculation by integrating the algorithm and assumptions validated and uses the information taken from DCC(Digital Control Computer) for the purpose of producing basic input data. CCMS could be largely divided into two modules; CCMS server program and CCMS client program. CCMS server program plays the role in automatic and continuous RFSP run and management of the past output data resulting from the run using Data Base Management System(DBMS). CCMS client program enables users to monitor current and past core status with GUI(Graphic-User Interface) environment predefined. The effectiveness of CCMS was verified by comparing the data resulted from field-test of the system for about 43 hours with the data used in the field of Wolsong NPP No. 1.

  17. The CANDU alarm analysis tool (CAAT)

    Energy Technology Data Exchange (ETDEWEB)

    Davey, E C; Feher, M P; Lupton, L R [Control Centre Technology Branch, ON (Canada)

    1997-09-01

    AECL undertook the development of a software tool to assist alarm system designers and maintainers based on feedback from several utilities and design groups. The software application is called the CANDU Alarm Analysis Tool (CAAT) and is being developed to: Reduce by one half the effort required to initially implement and commission alarm system improvements; improve the operational relevance, consistency and accuracy of station alarm information; record the basis for alarm-related decisions; provide printed reports of the current alarm configuration; and, make day-to-day maintenance of the alarm database less tedious and more cost-effective. The CAAT assists users in accessing, sorting and recording relevant information, design rules, decisions, and provides reports in support of alarm system maintenance, analysis of design changes, or regulatory inquiry. The paper discusses the need for such a tool, outlines the application objectives and principles used to guide tool development, describes the how specific tool features support user design and maintenance tasks, and relates the lessons learned from early application experience. (author). 4 refs, 2 figs.

  18. Industrial process heat from CANDU reactors

    International Nuclear Information System (INIS)

    Hilborn, J.S.; Seddon, W.A.; Barnstaple, A.G.

    1980-08-01

    It has been demonstrated on a large scale that CANDU reactors can produce industrial process steam as well as electricity, reliably and economically. The advantages of cogeneration have led to the concept of an Industrial Energy Park adjacent to the Bruce Nuclear Power Development in the province of Ontario. For steam demands between 300,000 and 500,00 lb/h (38-63 kg/s) and an annual load factor of 80%, the estimated cost of nuclear steam at the Bruce site boundary is $3.21/MBtu ($3.04GJ), which is at least 30% cheaper than oil-fired steam at the same site. The most promising near term application of nuclear heat is likely to be found within the energy-intensive chemical industry. Nuclear energy can substitute for imported oil and coal in the eastern provinces if the price remains competitive, but low cost coal and gas in the western provinces may induce energy-intensive industries to locate near those sources of energy. In the long term it may be feasible to use nuclear heat for the mining and extraction of oil from the Alberta tar sands. (auth)

  19. Advanced CANDU Design With Negative Power Feedback

    International Nuclear Information System (INIS)

    Andang-Widi-Harto; Muslim

    2004-01-01

    The problem of positive power feedback in the recent PHWR-CANDU design, especially related to coolant void increase, will be overcame by the use of dual moderator concept, in which two moderator systems are used, i.e. a main moderator outside the calandria tube and an annular moderator inside the annular space. Annular moderator is allowed to boil in the case of overheating. The numerical calculations have been performed for two core design namely HWR-DM-ST and HWR-DM-XI which can reach burn up of 16,000 and 17,500 MWd/ ton U respectively. The results for the two designs is that the values of k at fully annular moderator filling condition are 1.0054 (HWR-DM-ST) and 1.0019 (HWR-DM-XI), while at completely empty annular moderator condition are 0.9634 (HWR-DM-ST) and 0.9143 (HWR-DM-XI). The decrease of coolant flow rate from 3,043 kg/s to 853 kg/s decrease k values of 0.0109 (HWR-DM-ST) and 0.0232 (HWR-DM-XI). While increasing inlet coolant enthalpy from 2,950 kJ/kg to 3,175 kJ/kg decreases of k values of 0.0074 (HWR-DM-ST) and 0.0239 (HWR-DM-XI). Thus, it can be summarized that the HWR-DM design has negative power reactivity feedback.(author)

  20. Development Directions For CANDU and Slowpoke Reactors

    International Nuclear Information System (INIS)

    Brooks, Gordon L.

    1990-01-01

    This paper provides a broader-based discussion of overall development directions foreseen for CANDU reactors, particularly those which have further evolved sine the earlier paper. The paper then discusses development directions for the Slowpokes Energy System which is a small nuclear heat source intended to meet local heating needs for building complexes and municipal heating systems. In evolving a sound development direction, it is necessary, firstly, to address the question of requirements, viz., what are the requirements which future nuclear power plants must satisfy if they are to be successful? Today, some in the nuclear industry believe that the most important of such requirements relates to the need for 'safer' reactors. Indeed, some proponents of this view would seem to suggest that if only we could develop such 'safer' reactors, suddenly all of our problem s with public acceptance would disappear and utilities would form long lines waiting to purchase such marvellous machines. I do not share such a simplistic view nor, indeed, do many of my colleagues in the international nuclear power industry

  1. Stochastic maintenance optimization at Candu power plants

    International Nuclear Information System (INIS)

    Doyle, E.K.; Duchesne, T.; Lee, C.G.; Cho, D.I.

    2004-01-01

    The use of various innovative maintenance optimization techniques at Bruce has lead to cost effective preventive maintenance applications for complex systems as previously reported at ICONE 6 in New Orleans (1996). Further refinement of the station maintenance strategy was evaluated via the applicability of statistical analysis of historical failure data. The viability of stochastic methods in Candu maintenance was illustrated at ICONE 10 in Washington DC (2002). The next phase consists of investigating the validity of using subjective elicitation techniques to obtain component lifetime distributions. This technique provides access to the elusive failure statistics, the lack of which is often referred to in the literature as the principal impediment preventing the use of stochastic methods in large industry. At the same time the technique allows very valuable information to be captured from the fast retiring 'baby boom generation'. Initial indications have been quite positive. The current reality of global competition necessitates the pursuit of all financial optimizers. The next construction phase in the power generation industry will soon begin on a worldwide basis. With the relatively high initial capital cost of new nuclear generation all possible avenues of financial optimization must be evaluated and implemented. (authors)

  2. The CANDU alarm analysis tool (CAAT)

    International Nuclear Information System (INIS)

    Davey, E.C.; Feher, M.P.; Lupton, L.R.

    1997-01-01

    AECL undertook the development of a software tool to assist alarm system designers and maintainers based on feedback from several utilities and design groups. The software application is called the CANDU Alarm Analysis Tool (CAAT) and is being developed to: Reduce by one half the effort required to initially implement and commission alarm system improvements; improve the operational relevance, consistency and accuracy of station alarm information; record the basis for alarm-related decisions; provide printed reports of the current alarm configuration; and, make day-to-day maintenance of the alarm database less tedious and more cost-effective. The CAAT assists users in accessing, sorting and recording relevant information, design rules, decisions, and provides reports in support of alarm system maintenance, analysis of design changes, or regulatory inquiry. The paper discusses the need for such a tool, outlines the application objectives and principles used to guide tool development, describes the how specific tool features support user design and maintenance tasks, and relates the lessons learned from early application experience. (author). 4 refs, 2 figs

  3. The evolution of CANDU containment design

    International Nuclear Information System (INIS)

    Pendergast, Duane R.; Meneley, Daniel A.

    1995-01-01

    This paper reviews Canadian containment design, reflects on forces which have shaped the evolving designs and contemplates future containment design subject to existing constraints of Canadian and International nuclear power regulations. The discussion mentions modifications which could play a role in easing customary restrictions on siting while meeting the intent and literal definition of Canadian licensing requirements. Many different containment concepts have been considered and deployed over the years. Professor Birkhofer points out that the earliest systems installed were 'dry containment'. These rely on large containment buildings designed to contain all the the steam from a loss of coolant. It was soon realized the large water inventory of boiling water reactors would require very large and expensive containment buildings. This led to the development of pressure suppression concepts. Various pressure suppression schemes have been devised to condense steam and thus reduce design pressure and volume requirements. Examples include the forcing of steam into condensing water pools of BWR's steam condensation by passage into arrays of ice, dousing spray systems in CANDU reactors, and the spray condenser systems of late model Soviet designed VVER - 440 reactors. The fundamental concept of energy removal by steam condensation can result in a small containment volume requiring low design pressure capability

  4. Modeling flow-accelerated corrosion in CANDU

    International Nuclear Information System (INIS)

    Burrill, K.A.

    1995-11-01

    Flow-accelerated corrosion (FAC) of large areas of carbon steel in various circuits of CANDU plants generates significant quantities of corrosion products. As well, the relatively rapid corrosion rate can lead to operating difficulties with some components. Three areas in the plant are identified and a simple model of mass-transfer controlled corrosion of the carbon steel is derived and applied to these areas. The areas and the significant finding for each are given below: A number of lines in the feedwater system generate sludge by FAC, which causes steam generator fouling. Prediction of the steady-state iron concentration at the feedtrain outlet compares well with measured values. Carbon steel outlet feeders connect the reactor core with the steam generators. The feeder surface provides the dissolved iron through FAC, which fouls the primary side of the steam generator tubes, and can lead to derating of the plant and difficulty in tube inspection. Segmented carbon steel divider plates in the steam generator primary head leak at an increasing rate with time. The leakage rate is strongly dependent on the tightness of the overlapping joints. which undergo FAC at an increasing rate with time. (author) 7 refs., 5 tabs., 6 figs

  5. Spent fuels program

    International Nuclear Information System (INIS)

    Shappert, L.B.

    1983-01-01

    The goal of this task is to support the Domestic Spent Fuel Storage Program through studies involving the transport of spent fuel. A catalog was developed to provide authoritative, timely, and accessible transportation information for persons involved in the transport of irradiated reactor fuel. The catalog, drafted and submitted to the Transportation Technology Center, Sandia National Laboratories, for their review and approval, covers such topics as federal, state, and local regulations, spent fuel characteristics, cask characteristics, transportation costs, and emergency response information

  6. Carbon-14 behavior in a cement-dominated environment: Implications for spent CANDU resin waste disposal

    International Nuclear Information System (INIS)

    Dayal, R.; Reardon, E.J.

    1994-01-01

    Cement based waste forms and concrete engineered barriers are expected to play a key role in providing 14 C waste containment and control 14 C migration for time periods commensurate with its hazardous life of about 50,000 years. The main thrust of this study was, therefore, to evaluate the performance of cement based waste forms with regard to 14 C containment. Of particular importance are the geochemical processes controlling 14 C solubility and release under anticipated cement dominated low and intermediate level waste repository conditions. Immobilization of carbonate-form exchange resin in grout involves transfer of sorbed 14 CO 3 2- ions, through exchange for hydroxyl ions from the grout slurry, followed by localized precipitation of solid calcium carbonate at the cement/resin interface in the grout matrix. Carbon-14 release behavior can be attributed to the dissolution characteristics and solubility of calcite present in the cement based waste form. The groundwater flow regime can exert a pronounced effect both on the near-field chemistry and the leaching behavior of 14 C. For a cement dominated repository, at relatively low-flow or stagnant groundwater conditions, the alkaline near-field chemical environments inhibits the release of 14 C from the cemented waste form. Under high flow conditions, the near-field environment is characterized by relatively neutral pH conditions which promote calcite dissolution, thus resulting in 14 C release from the waste form

  7. Radioactive effluents from CANDU 6 reactors during normal operation

    International Nuclear Information System (INIS)

    Boss, C.R.; Allsop, P.J.

    1995-12-01

    During routine operation of a CANDU 6 reactor, various gaseous, liquid, and solid radioactive wastes are generated. The layout of the CANDU 6 reactor and the design of its systems ensure that these are minimized, but small quantities of gaseous and liquid wastes are continually discharged at very low concentrations. This report discusses the make-up of these chronically generated gaseous and liquid effluents. From a safety perspective, the doses to individual members of the public resulting from radioactive wastes chronically discharged from CANDU 6 reactors have been negligible. Similarly, doses to the regional and global populations have been negligible, generally less than 0.001% of background. While far below regulatory limits, releases of tritium, noble gases and gross β - -γ have been the most radiologically significant emissions, while radioiodine and particulates have had the greatest potential to deliver public dose. (author). 8 refs., 16 tabs., 3 figs

  8. Evolution of the CANDU control centre retrofit and new stations

    International Nuclear Information System (INIS)

    Olmstead, R.A.; Mitchell, W.

    1991-01-01

    Significant event data from operating nuclear plants in many countries consistently indicates human errors are the root cause for 40-60% of operating station significant events. Because so much information is already in digital form, opportunities exist to improve the CANDU control centre with retrofits that exploit this information. These opportunities are enhanced because of rapid technological development in computers and electronics, coupled with significant progress in the behavioural sciences that greatly increases our knowledge of the cognitive strengths and weaknesses of human beings. CANDU control rooms are undergoing retrofits and for future CANDU stations, a new concept of the control centre is emerging. The objective is to significantly reduce the incidence of human error, reduce operations and maintenance costs and improve both reliability and safety

  9. Evolution of the CANDU control centre retrofit and new stations

    Energy Technology Data Exchange (ETDEWEB)

    Olmstead, R A [AECL-CANDU, ON (Canada); Mitchell, W [Ontario Hydro, Darlington Nuclear Generating Station, Bowmanville, ON (Canada)

    1991-04-01

    Significant event data from operating nuclear plants in many countries consistently indicates human errors are the root cause for 40-60% of operating station significant events. Because so much information is already in digital form, opportunities exist to improve the CANDU control centre with retrofits that exploit this information. These opportunities are enhanced because of rapid technological development in computers and electronics, coupled with significant progress in the behavioural sciences that greatly increases our knowledge of the cognitive strengths and weaknesses of human beings. CANDU control rooms are undergoing retrofits and for future CANDU stations, a new concept of the control centre is emerging. The objective is to significantly reduce the incidence of human error, reduce operations and maintenance costs and improve both reliability and safety.

  10. Distinctive safety aspects of the CANDU-PHW reactor design

    International Nuclear Information System (INIS)

    Kugler, G.

    1980-01-01

    Two lectures are presented in this report. They were prepared in response to a request from IAEA to provide information on the 'Special characteristics of the safety analysis of heavy water reactors' to delegates from member states attending the Interregional Training Course on Safety Analysis Review, held at Karlsruhe, November 19 to December 20, 1979. The CANDU-PHW reactor is used as a model for discussion. The first lecture describes the distinctive features of the CANDU reactor and how they impact on reactor safety. In the second lecture the Canadian safety philosophy, the safety design objective, and other selected topics on reactor safety analysis are discussed. The material in this report was selected with a view to assisting those not familiar with the CANDU heavy water reactor design in evaluating the distinctive safety aspects of these reactors. (auth)

  11. Simulation of transient heat transfer in MACSTOR/KN-400 module storing irradiated CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sabourin, G. [Atomic Energy of Canada Limited, Montreal, Quebec (Canada); Lee, K.-H.; Yoon, J.-H.; Choi, B.-I.; Lee, H.-Y.; Song, M.-J. [KHNP, Nuclear Environment Technology Inst., Taejon (Korea, Republic of)

    2004-07-01

    Korea Hydro and Nuclear Power (KHNP), in collaboration with Atomic Energy of Canada Limited (AECL), are developing a new module for the dry storage of spent fuel from the four CANDU 6 nuclear reactors at the Wolsong site in South Korea, the MACSTOR/KN-400. The simulation of transient conditions for AECL's spent fuel dry storage systems, presented in this paper, has not been performed before and is considered a major achievement of the present work. In a fist step, CATHENA was compared to MACSTOR-200 temperature measurements and the accuracy of the results were very good. In a second step, CATHENA was applied to the MACSTOR/KN-400. Four cases were performed for the MACSTOR/KN-400: Off-normal cases in summer and winter and reduced air flow cases in summer and winter. The maximum local concrete temperatures were predicted to be 63{sup o}C for the off-normal case and 65{sup o}C in the reduced air flow case. The maximum temperature gradients in the concrete are predicted to be 28{sup o}C for the off-normal case and 30{sup o}C in the reduced air flow case, incorporating a 3{sup o}C uncertainty. This paper shows that the maximum temperature for the module is expected to meet the temperature limitations of appropriate standards. (author)

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

    International Nuclear Information System (INIS)

    1995-01-01

    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

  13. Analysis of ASTEC code adaptability to severe accident simulation for CANDU type reactors

    International Nuclear Information System (INIS)

    Constantin, Marin; Rizoiu, Andrei

    2008-01-01

    In order to prepare the adaptation of the ASTEC code to CANDU NPP severe accident analysis two kinds of activities were performed: - analyses of the ASTEC modules from the point of view of models and options, followed by CANDU exploratory calculation for the appropriate modules/models; - preparing the specifications for ASTEC adaptation for CANDU NPP. The paper is structured in three parts: - a comparison of PWR and CANDU concepts (from the point of view of severe accident phenomena); - exploratory calculations with some ASTEC modules- SOPHAEROS, CPA, IODE, CESAR, DIVA - for CANDU type reactors specific problems; - development needs analysis - algorithms, methods, modules. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-12-31

    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.

  15. CANDU steam generator life management: laboratory data and plant experience

    International Nuclear Information System (INIS)

    Tapping, R.L.; Nickerson, J.H.; Subash, N.; Wright, M.D.

    2001-10-01

    As CANDU reactors enter middle age, and the potential value of the plants in a deregulated market is realized, life management and life extension issues become increasingly important. An accurate assessment of critical components, such as the CANDU 6 steam generators (SGs), is crucial for successful life extension, and in this context, material issues are a key factor. For example, service experience with Alloy 900 tubing indicates very low levels of degradation within CANDU SGs; the same is also noted worldwide. With little field data for extrapolation, life management and life extension decisions for the tube bundles rely heavily on laboratory data. Similarly, other components of the SGs, in particular the secondary side internals, have only limited inspection data upon which to base a condition assessment. However, in this case there are also relatively little laboratory data. Decisions on life management and life extension are further complicated--not only is inspection access often restricted, but repair or replacement options for internal components are, by definition, also limited. The application of CANDU SG life management and life extension requires a judicious blend of in-service data, laboratory research and development (R and D) and materials and engineering judgment. For instance, the available laboratory corrosion and fretting wear data for Alloy 800 SG tubing have been compared with plant experience (with all types of tubing), and with crevice chemistry simulations, in order to provide an appropriate inspection guide for a 50-year SG life. A similar approach has been taken with other SG components, where the emphasis has been on known degradation mechanisms worldwide. This paper provides an outline of the CANDU SG life management program, including the results to date, a summary of the supporting R and D program showing the integration with condition assessment and life management activities, and the approach taken to life extension for a typical

  16. Future generations of CANDU: advantages and development with passive safety

    International Nuclear Information System (INIS)

    Duffey, R. B.

    2006-01-01

    Atomic Energy of Canada Limited (AECL) advances water reactor and CANDLT technology using an evolutionary development strategy. This strategy ensures that innovations are based firmly on current experience and keeps our development programs focused on one reactor concept, reducing risks, development costs, and product development cycle times. It also assures our customers that our products will never become obsolete or unsupported, and the continuous line of water reactor development is secure and supported into the future. Using the channel reactor advantage of modularity, the subdivided core has the advantage of passive safety by heat removal to the low- pressure moderator. With continuous improvements, the Advanced CANDU Reactor TM (ACR-1000TM) concept will likely remain highly competitive for a number of years and leads naturally to the next phase of CANDU development, namely the Generation IV CANDU -SCWR concept. This is conventional water technology, since supercritical boilers and turbines have been operating for some time in coal-fired power plants. Significant cost, safety, and performance advantages would result from the CANDU-SCWR concept, plus the flexibility of a range of plant sizes suitable for both small and large electric grids, and the ability for co-generation of electric power, process heat, and hydrogen. In CANDU-SCWR, novel developments are included in the primary circuit layout and channel design. The R and D in Canada is integrated with the Generation IV international Forum (GIF) plans, and has started on examining replaceable insulating liners that would ensure channel life, and on providing completely passive reactor decay heat removal directly to the moderator heat sink without forced cooling. In the interests of sustainability, hydrogen production by a CANDU- SCWR is also be included as part of the system requirements, where the methods for hydrogen production will depend on the outlet temperature of the reactor

  17. CANDU fuel quality and how it is achieved

    International Nuclear Information System (INIS)

    Gacesa, M.; Quarrington, G.R.; Tarasuk, W.R.; Carrick, I.R.; Pawliw, J.; McGregor, G.; Debnam, H.R.; Proos, L.

    1980-07-01

    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)

  18. Nuclear safety risk control in the outage of CANDU unit

    International Nuclear Information System (INIS)

    Wu Mingliang; Zheng Jianhua

    2014-01-01

    Nuclear fuel remains in the core during the outage of CANDU unit, but there are still nuclear safety risks such as reactor accidental criticality, fuel element failure due to inability to properly remove residual heat. Furthermore, these risks are aggravated by the weakening plant system configuration and multiple cross operations during the outage. This paper analyzes the phases where there are potential nuclear safety risks on the basis of the typical critical path arrangement of the outage of Qinshan NPP 3 and introduces a series of CANDU-specific risk control measures taken during the past plant outages to ensure nuclear safety during the unit outage. (authors)

  19. The status of safeguarding 600 MW(e) CANDU reactors

    International Nuclear Information System (INIS)

    Von Baeckmann, A.; Rundquist, D.E.; Pushkarjov, V.; Smith, R.M.; Zarecki, C.W.

    1982-09-01

    There has been extensive work in the development of CANDU safeguards since the last International Conference on Nuclear Power, and this has resulted in the development of improved equipment for the safeguards system now being installed in the 600 MW(e) CANDU generating stations. The overall system is designed to improve on the existing IAEA safeguards and to provide adequate coverage for each plausible nuclear material diversion route. There is sufficient sensitivity and redundancy to enable the timely detection of the possible diversion of significant quantities of nuclear material

  20. A JAVA applet to simulate a CANDU reactor

    International Nuclear Information System (INIS)

    Varin, E.; Desarmenien, J.

    2004-01-01

    Here we present a CANDU nuclear power plant simulator, directly available on a web page. The developed applet has two mains objectives: to expose the CANDU technology to a large public on the internet; and to construct a realistic simulator to be used as a pedagogical tool for nuclear introduction to high school or under-graduate students. The neutronic behavior and control algorithms of the reactor are simulated. Java programming language enables a very flexible environment for public information and user interaction with the plant. Examples of shutdown and power maneuver are explained. (author)

  1. Effects contributing to positive coolant void reactivity in CANDU

    International Nuclear Information System (INIS)

    Whitlock, J.J.; Garland, W.J.; Milgram, M.S.

    1995-01-01

    The lattice cell code WIMS-AECL (Ref. 3) is used to model a typical CANDU lattice cell, using nominal geometric bucklings, the PIJ option, and 69-group Winfrith library. The effect of cell voiding is modeled as a 100% instantaneous removal of coolant from the lattice. This is conservative because of the neglect of time dependence and partial core voiding, considered more plausible in CANDU. Results are grouped into three spectral groups: fast neutrons (0.821 to 10 MeV), epithermal neutrons (0.625 eV to 0.821 MeV), and thermal neutrons (<0.625 eV)

  2. 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)

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

    International Nuclear Information System (INIS)

    Causey, A.R.; Cheadle, B.A.; Coleman, C.E.; Price, E.G.

    1996-02-01

    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 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 with those for current channels. The approaches to these issues are discussed briefly in this report. (author). 14 refs., 1 tab., 8 figs

  4. Results of the CANDU 3 probabilistic safety assessment

    Energy Technology Data Exchange (ETDEWEB)

    Jaitly, R K [Atomic Energy of Canada Ltd., Saskatoon, SK (Canada)

    1996-12-31

    The purpose of the Conceptual Probabilistic Safety Assessment (PSA) of the CANDU 3 reactor was to provide safety assistance in the early stages of design to ensure that the design included adequate redundancy and functional separation of the mitigating systems; the final design should therefore give better results, particularly after modifications involving control, electrical power, instrument air, and service water. The initial PSA gave a total CANDU 3 core damage frequency of 7.8 x 10{sup -6}/year. 4 refs., 1 fig.

  5. 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)

    1998-12-31

    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)

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

    International Nuclear Information System (INIS)

    Coleman, C.E.; Cheadle, B.A.; Causey, A.R.; Doubt, G.L.; Fong, R.W.L.; Venkatapathi, S.

    1996-03-01

    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)

  7. Risk Importance Determination Process of CANDU Maintenance Rule Function

    International Nuclear Information System (INIS)

    Seo, Mi Ro; Jo, Ha Yan; Hwang, Mi Jeong

    2009-01-01

    In Korea, Maintenance Rule (MR) programs development for all PWR were completed. However, in case of PHWR (CANDU type, Wolsong Unit 1,2,3,4), the study of MR program was delayed, since the design concepts and operating experiences are different from those of PWR. This paper describes the process and results for the risk importance determination process of functions in scope. The risk importance was determined by PSA Basic Event Mapping and Delphi method. For Delphi evaluation, Delphi evaluation item for CANDU need to be developed because the design and normal operation functions are different from PWR

  8. Application of fuel management calculation codes for CANDU reactor

    International Nuclear Information System (INIS)

    Ju Haitao; Wu Hongchun

    2003-01-01

    Qinshan Phase III Nuclear Power Plant adopts CANDU-6 reactors. It is the first time for China to introduce this heavy water pressure tube reactor. In order to meet the demands of the fuel management calculation, DRAGON/DONJON code is developed in this paper. Some initial fuel management calculations about CANDU-6 reactor of Qinshan Phase III are carried out using DRAGON/DONJON code. The results indicate that DRAGON/DONJON can be used for the fuel management calculation for Qinshan Phase III

  9. Determination of representative CANDU feeder dimensions for engineering simulator

    International Nuclear Information System (INIS)

    Cho, S.; Muzumdar, A.

    1996-01-01

    This paper describes a logic for selection of representative channel groups and a methodology for determination of representative CANDU feeder dimensions and the pressure drops between inlet/outlet header and fuel channel in the primary loop. A code, MEDOC, was developed based on this logic and methodology and helps perform a calculation of representative feeder dimensions for a selected channel group on the basis of feeder geometry data (fluid volume, mass flow rate, loss factor) and given property data (pressure, quality, density) at inlet/outlet header. The representative feeder dimensions calculated based on this methodology will be useful for the engineering simulator for the CANDU type reactor. (author)

  10. Assessment studies on plutonium recycle in CANDU reactors

    International Nuclear Information System (INIS)

    1978-11-01

    This paper describes the CANDU reactor system in detail and goes on to explore the potential for using the system with plutonium recycle fuelling to improve fuel utilisation and to meet the long-term challenge of economic supplies of nuclear fuel. The paper includes comments on costs and non-proliferation aspects. It concludes that: recycle fuelling is feasible with little modification to the reactor design and no degradation of safety, and could offer over 50% savings in uranium requirements. However, recycle fuelling costs do not appear competitive with natural uranium in the CANDU system under current economic conditions

  11. Design of a transportation cask for irradiated CANDU fuel

    International Nuclear Information System (INIS)

    Nash, K.E.; Gavin, M.E.

    1983-01-01

    A major step in the development of a large-scale transportation system for irradiated CANDU fuel is being made by Ontario Hydro in the design and construction of a demonstration cask by 1988/89. The system being designed is based on dry transportation with the eventual fully developed system providing for dry fuel loading and unloading. Research carried out to date has demonstrated that it is possible to transport irradiated CANDU fuel in a operationally efficient and simple manner without any damage which would prejudice subsequent automated fuel handling

  12. Load-following performance and assessment of CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Tayal, M.; Floyd, M.; Rattan, D.; Xu, Z.; Manzer, A.; Lau, J. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Kohn, E. [Ontario Power Generation, Fuel and Fuel Channel Analysis Dept., Toronto, Ontario (Canada)

    1999-09-01

    Load following of nuclear reactors is now becoming an economic necessity in some countries. When nuclear power stations are operated in a load-following mode, the reactor and the fuel may be subjected to step changes in power on a weekly, daily, or even hourly basis, depending on the grid's needs. This paper updates the previous surveys of load-following capability of CANDU fuel, focusing mainly on the successful experience at the Bruce B station. As well, initial analytical assessments are provided that illustrate the capability of CANDU fuel to survive conditions other than those for which direct in-reactor evidence is available. (author)

  13. CANDU - a versatile reactor for plutonium disposition or actinide burning

    International Nuclear Information System (INIS)

    Chan, P.S.W.; Gagnon, M.J.N.; Boczar, P.G.; Ellis, R.J.; Verrall, R.A.

    1997-10-01

    High neutron economy, on-line refuelling, and a simple fuel-bundle design result in a high degree of versatility in the use of the CANDU reactor for the disposition of weapons-derived plutonium and for the annihilation of long-lived radioactive actinides, such as plutonium, neptunium, and americium isotopes, created in civilian nuclear power reactors. Inherent safety features are incorporated into the design of the bundles carrying the plutonium and actinide fuels. This approach enables existing CANDU reactors to operate with various plutonium-based fuel cycles without requiring major changes to the current reactor design. (author)

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

    International Nuclear Information System (INIS)

    Vieru, G.

    2001-01-01

    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)

  15. Improving the calandria tubes for CANDU reactors

    International Nuclear Information System (INIS)

    Coleman, C.E.; Fong, R.W.L.; Doubt, G.L.

    1997-01-01

    CANDU calandria tubes are made from annealed Zircaloy-2 sheet formed into a cylinder and welded along its length to make the tube. The current calandria tubes have given exemplary service for many years. With more stringent regulations and the need to accommodate warm cooling water in tropical countries, we started a development program to increase the margins for failure during postulated accidents. These improvements involve increasing the tube strength and optimising the heat-transfer from an excessively hot fuel channel to the cool moderator. If the postulated accident involves a pressure tube break, it would be desirable if the calandria tube withstood the full pressure of the heat-transport system. The weakest link in current calandria tubes is the weld. Thickening the weld can increase the strength by 20% while seamless tubes can be 45% stronger than current tubes. The latter tubes can hold full system pressure for many hours without failure. If during the postulated accident the fuel and pressure tube become excessively hot but do not touch the calandria tube, the radiant heat loss must be maximised. Current calandria tubes have an absorptivity (emissivity) of about 0.2. To protect the fuel and the fuel channel we have devised a finish to the inside surface of the calandria tube that increases the emissivity to 0.7. If during the postulated accident the hot pressure tube touches the cool calandria tube, the contact conductance and the critical heat flux must be optimised to ensure nucleate boiling of the moderator at the outside surface of the calandria tube and therefore efficient exploitation of the moderator as a heat sink. In laboratory tests small ridges on the inside surface and roughening of the outside surface have been shown to increase the margins against failure and increase the possible moderator temperatures thus providing the opportunity to decrease the cost of the moderator heat-exchange system and remove restrictions on reactor operation in

  16. A new CANDU-600 containment structure

    International Nuclear Information System (INIS)

    Serban, V.; Bobei, M.; Gheorghiu, M.; Popescu, M.; Stanciu, M.; Dinica, D.; Alexandru, C.

    1994-01-01

    This paper is presenting a structure made of reinforced concrete with rectangular cross-section, box-divided, prefabricated and modulled on a bay 6.5 m wide and 4.5 m high, and provided with a steel liner. The building has an overall basement in which the steel liner is embedded and which is supporting the building walls. The inner structure is common to the containment as well and it is carried out for each room (generally 6.5 m by 6.5 m) having intermediar floors at the necessary elevations. The containment dimensions, on horizontal plane are 6 x 6.5 m by 5 x 6.5 m and the total height of the side walls is 30.5 m. The containment is closed in A-C direction by a prefabricated semi-cylinder which is supported by the side walls and 5 intermediate arches. The fuel transfer deck structure is common to the inner structure and the containment structure. The Calandria vault is a separate individual structure located above E1. 100. For CANDU-600 main equipment the same arrangement was maintained, some unsignificant modifications being made, for example, the access areas located in the four corners of the building as well as the location of some auxiliary systems. The paper is also including a set of 1:200 scale drawings, comments on the construction manner and the results of the building structural analysis. The suggested solution is evidencing economical benefits facilities in the operation and construction of the plant and it is specially recommended for areas with high seismic events. (author)

  17. Conceptual designs for very high-temperature CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bushby, S.J.; Dimmick, G.R.; Duffey, R.B. [Atomic Energy of Canada Ltd., Chalk River, Ontario (Canada)

    2000-07-01

    Although its environmental benefits are demonstrable, nuclear power must be economically competitive with other energy sources to ensure it retains, or increases, its share of the changing and emerging energy markets of the next decades. In recognition of this, AECL is studying advanced reactor concepts with the goal of significant reductions in capital cost through increased thermodynamic efficiency and plant simplification. The program, generically called CANDU-X, examines concepts for the future, but builds on the success of the current CANDU designs by keeping the same fundamental design characteristics: excellent neutron economy for maximum flexibility in fuel cycle; an efficient heavy-water moderator that provides a passive heat sink under upset conditions; and, horizontal fuel channels that enable on-line refueling for optimum fuel utilization and power profiles. Retaining the same design fundamentals takes maximum advantage of the existing experience base, and allows technological and design improvements developed for CANDU-X to be incorporated into more evolutionary CANDU plants in the short to medium term. Three conceptual designs have been developed that use supercritical water (SCW) as a coolant. The increased coolant temperature results in the thermodynamic efficiency of each CANDU-X concept being significantly higher than conventional nuclear plants. The first concept, CANDU-X Mark 1, is a logical extension of the current CANDU design to higher operating temperatures. To take maximum advantage of the high heat capacity of water at the pseudo-critical temperature, water at nominally 25 MPa enters the core at 310{sup o}C, and exits at {approx}410{sup o}C. The high specific heat also leads to high heat transfer coefficients between the fuel cladding and the coolant. As a result, Zr-alloys can be used as cladding, thereby retaining relatively high neutron economy. The second concept, CANDU-X NC, is aimed at markets that require smaller simpler distributed

  18. Conceptual designs for very high-temperature CANDU reactors

    International Nuclear Information System (INIS)

    Bushby, S.J.; Dimmick, G.R.; Duffey, R.B.

    2000-01-01

    Although its environmental benefits are demonstrable, nuclear power must be economically competitive with other energy sources to ensure it retains, or increases, its share of the changing and emerging energy markets of the next decades. In recognition of this, AECL is studying advanced reactor concepts with the goal of significant reductions in capital cost through increased thermodynamic efficiency and plant simplification. The program, generically called CANDU-X, examines concepts for the future, but builds on the success of the current CANDU designs by keeping the same fundamental design characteristics: excellent neutron economy for maximum flexibility in fuel cycle; an efficient heavy-water moderator that provides a passive heat sink under upset conditions; and, horizontal fuel channels that enable on-line refueling for optimum fuel utilization and power profiles. Retaining the same design fundamentals takes maximum advantage of the existing experience base, and allows technological and design improvements developed for CANDU-X to be incorporated into more evolutionary CANDU plants in the short to medium term. Three conceptual designs have been developed that use supercritical water (SCW) as a coolant. The increased coolant temperature results in the thermodynamic efficiency of each CANDU-X concept being significantly higher than conventional nuclear plants. The first concept, CANDU-X Mark 1, is a logical extension of the current CANDU design to higher operating temperatures. To take maximum advantage of the high heat capacity of water at the pseudo-critical temperature, water at nominally 25 MPa enters the core at 310 o C, and exits at ∼410 o C. The high specific heat also leads to high heat transfer coefficients between the fuel cladding and the coolant. As a result, Zr-alloys can be used as cladding, thereby retaining relatively high neutron economy. The second concept, CANDU-X NC, is aimed at markets that require smaller simpler distributed power plants

  19. Design of a Multi-Spectrum CANDU-based Reactor, MSCR, with 37-element fuel bundles using SERPENT code

    International Nuclear Information System (INIS)

    Hussein, M.S.; Bonin, H.W.; Lewis, B.J.; Chan, P.

    2015-01-01

    The burning of highly-enriched uranium and plutonium from dismantled nuclear warhead material in the new design nuclear power plants represents an important step towards nonproliferation. The blending of these highly enriched uranium and plutonium with with uranium dioxide from the spent fuel of CANDU reactors, or mixing it with depleted uranium would need a very long time to dispose of this material. Consequently, considering that more efficient transmutation of actinides occurs in fast neutron reactors, a novel Multi-Spectrum CANDU Reactor, has been designed on the basis of the CANDU6 reactor with two concentric regions. The simulations of the MSCR were carried out using the SERPENT code. The inner or fast neutron spectrum core is fuelled by different levels of enriched uranium oxides. The helium is used as a coolant in the fast neutron core. The outer or the thermal neutron spectrum core is fuelled with natural uranium with heavy water as both moderator and coolant. Both cores use 37- element fuel bundles. The size of the two cores and the percentage level of enrichment of the fresh fuel in the fast core were optimized according to the criticality safety of the whole reactor. The excess reactivity, the regeneration factor, radial and axial flux shapes of the MSCR reactor were calculated at different of the concentration of fissile isotope 235 U of uranium fuel at the fast neutron spectrum core. The effect of variation of the concentration of the fissile isotope on the fluxes in both cores at each energy bin has been studied. (author)

  20. Design of a Multi-Spectrum CANDU-based Reactor, MSCR, with 37-element fuel bundles using SERPENT code

    Energy Technology Data Exchange (ETDEWEB)

    Hussein, M.S.; Bonin, H.W.; Lewis, B.J.; Chan, P., E-mail: mohamed.hussein@rmc.ca, E-mail: bonin-h@rmc.ca, E-mail: lewis-b@rmc.ca, E-mail: Paul.Chan@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, ON (Canada)

    2015-07-01

    The burning of highly-enriched uranium and plutonium from dismantled nuclear warhead material in the new design nuclear power plants represents an important step towards nonproliferation. The blending of these highly enriched uranium and plutonium with with uranium dioxide from the spent fuel of CANDU reactors, or mixing it with depleted uranium would need a very long time to dispose of this material. Consequently, considering that more efficient transmutation of actinides occurs in fast neutron reactors, a novel Multi-Spectrum CANDU Reactor, has been designed on the basis of the CANDU6 reactor with two concentric regions. The simulations of the MSCR were carried out using the SERPENT code. The inner or fast neutron spectrum core is fuelled by different levels of enriched uranium oxides. The helium is used as a coolant in the fast neutron core. The outer or the thermal neutron spectrum core is fuelled with natural uranium with heavy water as both moderator and coolant. Both cores use 37- element fuel bundles. The size of the two cores and the percentage level of enrichment of the fresh fuel in the fast core were optimized according to the criticality safety of the whole reactor. The excess reactivity, the regeneration factor, radial and axial flux shapes of the MSCR reactor were calculated at different of the concentration of fissile isotope {sup 235}U of uranium fuel at the fast neutron spectrum core. The effect of variation of the concentration of the fissile isotope on the fluxes in both cores at each energy bin has been studied. (author)

  1. Assessment of CANDU physics codes using experimental data - II: CANDU core physics measurements

    International Nuclear Information System (INIS)

    Roh, Gyu Hong; Jeong, Chang Joon; Choi, Hang Bok

    2001-11-01

    Benchmark calculations of the advanced CANDU reactor analysis tools (WIMS-AECL, SHETAN and RFSP) and the Monte Carlo code MCNP-4B have been performed using Wolsong Units 2 and 3 Phase-B measurement data. In this study, the benchmark calculations have been done for the criticality, boron worth, reactivity device worth, reactivity coefficient, and flux scan. For the validation of the WIMS-AECL/SHETANRFSP code system, the lattice parameters of the fuel channel were generated by the WIMS-AECL code, and incremental cross sections of reactivity devices and structural material were generated by the SHETAN code. The results have shown that the criticality is under-predicted by -4 mk. The reactivity device worths are generally consistent with the measured data except for the strong absorbers such as shutoff rod and mechanical control absorber. The heat transport system temperature coefficient and flux distributions are in good agreement with the measured data. However, the moderator temperature coefficient has shown a relatively large error, which could be caused by the incremental cross-section generation methodology for the reactivity device. For the MCNP-4B benchmark calculation, cross section libraries were newly generated from ENDF/B-VI release 3 through the NJOY97.114 data processing system and a three-dimensional full core model was developed. The simulation results have shown that the criticality is estimated within 4 mk and the estimated reactivity worth of the control devices are generally consistent with the measurement data, which implies that the MCNP code is valid for CANDU core analysis. In the future, therefore, the MCNP code could be used as a reference tool to benchmark design and analysis codes for the advanced fuels for which experimental data are not available

  2. Korean nuclear reactor strategy for the early 21st century

    International Nuclear Information System (INIS)

    Lee, Byong Whi; Shin, Young Kyun

    1991-01-01

    The system analysis for Korean nuclear power reactor option is made on the basis of reliability, cost minimization, finite uranium resource availability and nuclear engineering manpower supply constraints. The reference reactor scenarios are developed considering the future electricity demand, nuclear share, current nuclear power plant standardization program and manufacturing capacity. The levelized power generation cost, uranium requirement and nuclear engineering professionals demand are estimated for each reference reactor scenarios and nuclear fuel cycle options from the year 1990 up to the year 2030. Based on the outcomes of the analysis, uranium resource utilization, reliability and nuclear engineering manpower requirements are sensitive to the nuclear reactor strategy and associated fuel cycle whereas the system cost is not. APWR, CANDU: FBR strategy is to be the best option for Korea. However, APWR, CANDU: Passive Safe Reactor (PSR) vFBR strategy should be also considered as a contingency for growing national concerns on nuclear safety and public acceptance deterioration in the future. FBR development and establishment of related fuel cycle should be started as soon as possible considering the uranium shortage anticipated between 2007 and 2032. It should be noted that the increasing use of nuclear energy to minimize the greenhouse effects in the early 21st century would accelerate the uranium resource depletion. The study also concludes that the current level of nuclear engineering professionals employment is not sufficient until 2010 for the establishment of nuclear infrastructure. (Author)

  3. Spent nuclear fuel storage

    International Nuclear Information System (INIS)

    Romanato, Luiz Sergio

    2005-01-01

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

  4. Modelling nuclear fuel vibrations in horizontal CANDU reactors

    International Nuclear Information System (INIS)

    Jagannath, D.V.; Oldaker, I.E.

    1976-01-01

    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)

  5. Addressing severe accidents in the CANDU 9 design

    International Nuclear Information System (INIS)

    Nijhawan, S.M.; Wight, A.L.; Snell, V.G.

    1998-01-01

    CANDU 9 is a single-unit evolutionary heavy-water reactor based on the Bruce/Darlington plants. Severe accident issues are being systematically addressed in CANDU 9, which includes a number of unique features for prevention and mitigation of severe accidents. A comprehensive severe accident program has been formulated with feedback from potential clients and the Canadian regulatory agency. Preliminary Probabilistic Safety Analyses have identified the sequences and frequency of system and human failures that may potentially lead to initial conditions indicating onset of severe core damage. Severe accident consequence analyses have used these sequences as a guide to assess passive heat sinks for the core, and containment performance. Estimates of the containment response to mass and energy injections typical of postulated severe accidents have been made and the results are presented. We find that inherent CANDU severe accident mitigation features, such as the presence of large water volumes near the fuel (moderator and shield tank), permit a relatively slow severe accident progression under most plant damage states, facilitate debris coolability and allow ample time for the operator to arrest the progression within, progressively, the fuel channels, calandria vessel or shield tank. The large-volume CANDU 9 containment design complements these features because of the long times to reach failure

  6. Proceedings of the fourth international conference on CANDU maintenance

    International Nuclear Information System (INIS)

    1997-01-01

    These proceedings record the information presented at the 4th International Conference on CANDU Maintenance held November 16-18,1997 in Toronto, Canada. The papers for these proceedings were prepared on component maintenance, human performance, steam generator leak detection, fuel channel inspections, rotating equipment maintenance, surveillance programs, inspection techniques, valve maintenance, steam generator repairs and performance, reactor aging management and preventative maintenance

  7. Proceedings of the fourth international conference on CANDU maintenance

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-07-01

    These proceedings record the information presented at the 4th International Conference on CANDU Maintenance held November 16-18,1997 in Toronto, Canada. The papers for these proceedings were prepared on component maintenance, human performance, steam generator leak detection, fuel channel inspections, rotating equipment maintenance, surveillance programs, inspection techniques, valve maintenance, steam generator repairs and performance, reactor aging management and preventative maintenance.

  8. Feasibility Study for Cobalt Bundle Loading to CANDU Reactor Core

    International Nuclear Information System (INIS)

    Park, Donghwan; Kim, Youngae; Kim, Sungmin

    2016-01-01

    CANDU units are generally used to produce cobalt-60 at Bruce and Point Lepreau in Canada and Embalse in Argentina. China has started production of cobalt-60 using its CANDU 6 Qinshan Phase III nuclear power plant in 2009. For cobalt-60 production, the reactor’s full complement of stainless steel adjusters is replaced with neutronically equivalent cobalt-59 adjusters, which are essentially invisible to reactor operation. With its very high neutron flux and optimized fuel burn-up, the CANDU has a very high cobalt-60 production rate in a relatively short time. This makes CANDU an excellent vehicle for bulk cobalt-60 production. Several studies have been performed to produce cobalt-60 using adjuster rod at Wolsong nuclear power plant. This study proposed new concept for producing cobalt-60 and performed the feasibility study. Bundle typed cobalt loading concept is proposed and evaluated the feasibility to fuel management without physics and system design change. The requirement to load cobalt bundle to the core was considered and several channels are nominated. The production of cobalt-60 source is very depend on the flux level and burnup directly. But the neutron absorption characteristic of cobalt bundle is too high, so optimizing design study is needed in the future

  9. Emergency core cooling systems in CANDU nuclear power plants

    International Nuclear Information System (INIS)

    1981-12-01

    This report contains the responses by the Advisory Committee on Nuclear Safety to three questions posed by the Atomic Energy Control Board concerning the need for Emergency Core Cooling Systems (ECCS) in CANDU nuclear power plants, the effectiveness requirement for such systems, and the extent to which experimental evidence should be available to demonstrate compliance with effectiveness standards

  10. The Candu system - The way for nuclear autonomy

    International Nuclear Information System (INIS)

    Martinez, H.

    1987-01-01

    The experience acquired by Canada during the development of Candu System is presented. Some basic foundations of technology transfer are defined and, the conditions of canadian nuclear industry to provide developing countries, technical assistence for acquisition of nuclear energy autonomy, are analysed. (M.C.K.) [pt

  11. CANDU steam generator tubing material service experience and allied development

    International Nuclear Information System (INIS)

    Hart, A.E.; Lesurf, J.E.

    1976-01-01

    This paper covers the following aspects for the tube materials in CANDU-PHW steam generators: inservice performance with respect to tube leaks and coolant activity attributable to boiler tube corrosion, selection of tube materials for use with non-boiling and boiling primary coolants, supporting development on corrosion, vibration, fretting wear, tube inspection, leak detection and plugging of defective tubes. (author)

  12. Feasibility Study for Cobalt Bundle Loading to CANDU Reactor Core

    Energy Technology Data Exchange (ETDEWEB)

    Park, Donghwan; Kim, Youngae; Kim, Sungmin [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    CANDU units are generally used to produce cobalt-60 at Bruce and Point Lepreau in Canada and Embalse in Argentina. China has started production of cobalt-60 using its CANDU 6 Qinshan Phase III nuclear power plant in 2009. For cobalt-60 production, the reactor’s full complement of stainless steel adjusters is replaced with neutronically equivalent cobalt-59 adjusters, which are essentially invisible to reactor operation. With its very high neutron flux and optimized fuel burn-up, the CANDU has a very high cobalt-60 production rate in a relatively short time. This makes CANDU an excellent vehicle for bulk cobalt-60 production. Several studies have been performed to produce cobalt-60 using adjuster rod at Wolsong nuclear power plant. This study proposed new concept for producing cobalt-60 and performed the feasibility study. Bundle typed cobalt loading concept is proposed and evaluated the feasibility to fuel management without physics and system design change. The requirement to load cobalt bundle to the core was considered and several channels are nominated. The production of cobalt-60 source is very depend on the flux level and burnup directly. But the neutron absorption characteristic of cobalt bundle is too high, so optimizing design study is needed in the future.

  13. A short history of the CANDU nuclear power system

    Energy Technology Data Exchange (ETDEWEB)

    Brooks, G L

    1993-04-01

    This paper provides a short historical summary of the evolution of the CANDU nuclear power system with emphasis on the roles played by Ontario Hydro and private sector companies in Ontario in collaboration with Atomic Energy of Canada Limited (AECL). (author). 1 fig., 61 refs.

  14. Computer-aided engineering for Qinshan CANDU projects

    International Nuclear Information System (INIS)

    Huang Zhizhang; Goland, D.

    1999-01-01

    The author briefly describes AECL's work in applying computer-aided engineering tools to the Qinshan CANDU Project. The main emphases will be to introduce the major CADD software tools and their use in civil design, process design and EI and C design. Other special software tools and non-CADD tools and their applications are also briefly introduced

  15. CANDU 9 - reducing capital costs through advanced construction

    International Nuclear Information System (INIS)

    Hedges, K.; Yu, S.; Hopwood, J.

    1996-01-01

    Reducing costs is a clear priority in nuclear markets where capital reaches billions and financing is hard-won. To address that priority, AECL introduced the use of advanced construction techniques. This has been one of the key thrusts behind its development of CANDU 9. (author)

  16. A short history of the CANDU nuclear power system

    International Nuclear Information System (INIS)

    Brooks, G.L.

    1993-04-01

    This paper provides a short historical summary of the evolution of the CANDU nuclear power system with emphasis on the roles played by Ontario Hydro and private sector companies in Ontario in collaboration with Atomic Energy of Canada Limited (AECL). (author). 1 fig., 61 refs

  17. Development of CANDU ECCS performance evaluation methodology and guides

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Kwang Hyun; Park, Kyung Soo; Chu, Won Ho [Korea Maritime Univ., Jinhae (Korea, Republic of)

    2003-03-15

    The objectives of the present work are to carry out technical evaluation and review of CANDU safety analysis methods in order to assist development of performance evaluation methods and review guides for CANDU ECCS. The applicability of PWR ECCS analysis models are examined and it suggests that unique data or models for CANDU are required for the following phenomena: break characteristics and flow, frictional pressure drop, post-CHF heat transfer correlations, core flow distribution during blowdown, containment pressure, and reflux rate. For safety analysis of CANDU, conservative analysis or best estimate analysis can be used. The main advantage of BE analysis is a more realistic prediction of margins to acceptance criteria. The expectation is that margins demonstrated with BE methods would be larger that when a conservative approach is applied. Some outstanding safety analysis issues can be resolved by demonstration that accident consequences are more benign than previously predicted. Success criteria for analysis and review of Large LOCA can be developed by top-down approach. The highest-level success criteria can be extracted from C-6 and from them, the lower level criteria can be developed step-by-step, in a logical fashion. The overall objectives for analysis and review are to verify radiological consequences and frequency are met.

  18. Conceptual designs for advanced, high-temperature CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bushby, S.J. [Atomic Energy of Canada Ltd., Corrosion and Surface Science Branch, Chalk River Laboratories, Chalk River, ON (Canada); Dimmick, G.R. [Atomic Energy of Canada Ltd., Fuel Channel Thermmalhydraulics Branch, Chalk River, ON (Canada); Duffey, R.B. [Atomic Energy of Canada Ltd., Principal Scientist, Chalk River Laboratories, Chalk River, On (Canada); Spinks, N.J. [Atomic Energy of Canada Ltd., Researcher Emeritus, Chalk River Laboratories, Chalk River, ON (Canada); Burrill, K.A. [Atomic Energy of Canada Ltd., Chalk River Laboratories, Chalk River, ON (Canada); Chan, P.S.W. [Atomic Energy of Canada Ltd., Reactor Core Physics Branch, Mississauga, ON (Canada)

    2000-07-01

    AECL is studying advanced reactor concepts with the aim of significant cost reduction through improved thermodynamic efficiency and plant simplification. The program, generically called CANDU-X, also incorporates enhanced safety features, and flexible, proliferation-resistant fuel cycles, whilst retaining the fundamental design characteristics of CANDU: neutron economy, horizontal fuel channels, and a separate D{sub 2}O moderator that provides a passive heat sink. Where possible, proven, existing components and materials would be adopted, so that 'first-of-a-kind' costs and uncertainties are avoided. Three reactor concepts ranging in output from {approx}375 MW(e) to 1150 MW(e) are described. The modular design of a pressure tube reactor allows the plant size for each concept to be tailored to a given market through the addition or removal of fuel channels. Each concept uses supercritical water as the coolant at a nominal pressure of 25 MPa. Core outlet temperatures range from {approx}400degC to 625degC, resulting in substantial improvements in thermodynamic efficiencies compared to current nuclear stations. The CANDU-X Mark 1 concept is an extension of the present CANDU design. An indirect cycle is employed, but efficiency is increased due to higher coolant temperature, and changes to the secondary side; as well, the size and number of pumps and steam generators are reduced. Safety is enhanced through facilitation of thermo-siphoning of decay heat by increasing the temperature of the moderator. The CANDU-X NC concept is also based on an indirect cycle, but natural convection is used to circulate the primary coolant. This approach enhances cycle efficiency and safety, and is viable for reactors operating near the pseudo-critical temperature of water because of large changes in heat capacity and thermal expansion in that region. In the third concept (CANDUal-X), a dual cycle is employed. Supercritical water exits the core and feeds directly into a very high

  19. Development of the CANDU 66-group SN transport library

    International Nuclear Information System (INIS)

    Tsang, K.T.

    2001-01-01

    The design of the shield configuration around a nuclear reactor is strongly dependent on the neutron and photon spatial and energy distributions. The nuclear heat deposition and material damage in and surrounding the reactor core are also a function of the neutron and photon distributions. Therefore, to ensure a suitable configuration of materials for shielding or heat transfer, an accurate calculation of the particle fluxes in the reactor systems is essential. The CANDU 66-group library was developed to update the cross sections that are needed to assess the performance of CANDU bulk shields. Since about 1980, shielding analysts at Atomic Energy of Canada Limited (AECL) and Ontario Power Generation Inc. (OPGI) have been using a 38-group CANDU-specific library to perform S N transport calculations. In 1994, a new CANDU 67-group cross-section library was developed. The 67-group cross-section library was developed to provide radiation-physics analysts with up-to-date nuclear data to correct deficiencies with documentation of the old library. Although there were improvements over the 38-group library, initial use showed there were some deficiencies in the 67-group library. To correct these deficiencies, the CANDU 66-group S N transport cross-section library was developed. The 66-group library is based on the 241-group cross-section library VITAMIN-B6. Collapsing and weighting of the 241-group cross sections into 66 groups were performed using the modular code system SCALE 4.4. This paper describes how the modules in the SCALE system were applied to generate the 66-group library. The CANDU 66-group library includes both core-weighted and lattice-weighted cross sections of 235 U, 238 U, and 239 Pu with, and without, delayed fission-product photons. In addition, the 66-group library contains more response functions than did the 67-group library. Finally, the CANDU 66-group library has been validated against one-dimensional benchmark problems. The results generated with

  20. CANDU combined cycles featuring gas-turbine engines

    International Nuclear Information System (INIS)

    Vecchiarelli, J.; Choy, E.; Peryoga, Y.; Aryono, N.A.

    1998-01-01

    In the present study, a power-plant analysis is conducted to evaluate the thermodynamic merit of various CANDU combined cycles in which continuously operating gas-turbine engines are employed as a source of class IV power restoration. It is proposed to utilize gas turbines in future CANDU power plants, for sites (such as Indonesia) where natural gas or other combustible fuels are abundant. The primary objective is to eliminate the standby diesel-generators (which serve as a backup supply of class III power) since they are nonproductive and expensive. In the proposed concept, the gas turbines would: (1) normally operate on a continuous basis and (2) serve as a reliable backup supply of class IV power (the Gentilly-2 nuclear power plant uses standby gas turbines for this purpose). The backup class IV power enables the plant to operate in poison-prevent mode until normal class IV power is restored. This feature is particularly beneficial to countries with relatively small and less stable grids. Thermodynamically, the advantage of the proposed concept is twofold. Firstly, the operation of the gas-turbine engines would directly increase the net (electrical) power output and the overall thermal efficiency of a CANDU power plant. Secondly, the hot exhaust gases from the gas turbines could be employed to heat water in the CANDU Balance Of Plant (BOP) and therefore improve the thermodynamic performance of the BOP. This may be accomplished via several different combined-cycle configurations, with no impact on the current CANDU Nuclear Steam Supply System (NSSS) full-power operating conditions when each gas turbine is at maximum power. For instance, the hot exhaust gases may be employed for feedwater preheating and steam reheating and/or superheating; heat exchange could be accomplished in a heat recovery steam generator, as in conventional gas-turbine combined-cycle plants. The commercially available GateCycle power plant analysis program was applied to conduct a

  1. North Korean nuclear negotiation drama

    International Nuclear Information System (INIS)

    Kim, Jai Bok

    1995-06-01

    This book reports negotiation on North Korean nuclear among South Korea, North Korea and U. S. It includes an account about international issues on North Korean nuclear activities, a new aspect on the problems of North Korean nuclear, pressure on North Korea and startup for dialogue trying to solve problems by communication, investigation by IAEA, IAEA resolution and high tension on Korean peninsula with North Korean nuclear.

  2. North Korean nuclear negotiation drama

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jai Bok

    1995-06-15

    This book reports negotiation on North Korean nuclear among South Korea, North Korea and U. S. It includes an account about international issues on North Korean nuclear activities, a new aspect on the problems of North Korean nuclear, pressure on North Korea and startup for dialogue trying to solve problems by communication, investigation by IAEA, IAEA resolution and high tension on Korean peninsula with North Korean nuclear.

  3. Impact of aging and material structure on CANDU plant performance

    International Nuclear Information System (INIS)

    Nadeau, E.; Ballyk, J.; Ghalavand, N.

    2011-01-01

    In-service behaviour of pressure tubes is a key factor in the assessment of safety margins during plant operation. Pressure tube deformation (diametral expansion) affects fuel bundle dry out characteristics resulting in reduced margin to trip for some events. Pressure tube aging mechanisms also erode design margins on fuel channels or interfacing reactor components. The degradation mechanisms of interest are primarily deformation, loss of fracture resistance and hydrogen ingress. CANDU (CANada Deuterium Uranium, a registered trademark of the Atomic Energy of Canada Limited used under exclusive licence by Candu Energy Inc.) owners and operators need to maximize plant capacity factor and meet or exceed the reactor design life targets while maintaining safety margins. The degradation of pressure tube material and geometry are characterized through a program of inspection, material surveillance and assessment and need to be managed to optimize plant performance. Candu is improving pressure tubes installed in new build and life extension projects. Improvements include changes designed to reduce or mitigate the impact of pressure tube elongation and diametral expansion rates, improvement of pressure tube fracture properties, and reduction of the implications of hydrogen ingress. In addition, Candu provides an extensive array of engineering services designed to assess the condition of pressure tubes and address the impact of pressure tube degradation on safety margins and plant performance. These services include periodic and in-service inspection and material surveillance of pressure tubes and deterministic and probabilistic assessment of pressure tube fitness for service to applicable standards. Activities designed to mitigate the impact of pressure tube deformation on safety margins include steam generator cleaning, which improves trip margins, and trip design assessment to optimize reactor trip set points restoring safety and operating margins. This paper provides an

  4. Impact of aging and material structure on CANDU plant performance

    Energy Technology Data Exchange (ETDEWEB)

    Nadeau, E.; Ballyk, J.; Ghalavand, N. [Candu Energy Inc., Mississauga, Ontario (Canada)

    2011-07-01

    In-service behaviour of pressure tubes is a key factor in the assessment of safety margins during plant operation. Pressure tube deformation (diametral expansion) affects fuel bundle dry out characteristics resulting in reduced margin to trip for some events. Pressure tube aging mechanisms also erode design margins on fuel channels or interfacing reactor components. The degradation mechanisms of interest are primarily deformation, loss of fracture resistance and hydrogen ingress. CANDU (CANada Deuterium Uranium, a registered trademark of the Atomic Energy of Canada Limited used under exclusive licence by Candu Energy Inc.) owners and operators need to maximize plant capacity factor and meet or exceed the reactor design life targets while maintaining safety margins. The degradation of pressure tube material and geometry are characterized through a program of inspection, material surveillance and assessment and need to be managed to optimize plant performance. Candu is improving pressure tubes installed in new build and life extension projects. Improvements include changes designed to reduce or mitigate the impact of pressure tube elongation and diametral expansion rates, improvement of pressure tube fracture properties, and reduction of the implications of hydrogen ingress. In addition, Candu provides an extensive array of engineering services designed to assess the condition of pressure tubes and address the impact of pressure tube degradation on safety margins and plant performance. These services include periodic and in-service inspection and material surveillance of pressure tubes and deterministic and probabilistic assessment of pressure tube fitness for service to applicable standards. Activities designed to mitigate the impact of pressure tube deformation on safety margins include steam generator cleaning, which improves trip margins, and trip design assessment to optimize reactor trip set points restoring safety and operating margins. This paper provides an

  5. CANDU fuel - fifteen years of power reactor experience

    International Nuclear Information System (INIS)

    Fanjoy, G.R.; Bain, A.S.

    1977-01-01

    CANDU (Canada Deuterium Uranium) fuel has operated in power reactors since 1962. Analyses of performance statistics, supplemented by examinations of fuel from power reactors and experimental loops have yielded: (a) A thorough understanding of the fundamental behaviour of CANDU fuel. (b) Data showing that the predicted high utilization of uranium has been achieved. Actual fuelling costs in 1976 at the Pickering Generating Station are 1.2 m$/kWh (1976 Canadian dollars) with the simple oncethrough natural-UO 2 fuel cycle. (c) Criteria for operation, which have led to the current very low defect rate of 0.03% of all assemblies and to ''CANLUB'' fuel, which has a graphite interlayer between the fuel and sheath to reduce defects on power increases. (d) Proof that the short length (500 mm), collapsible cladding features of the CANDU bundle are successful and that the fuel can operate at high-power output (current peak outer-element linear power is 58 +- 15% kW/m). Involvement by the utility in all stages of fuel development has resulted in efficient application of this fundamental knowledge to ensure proper fuel specifications, procurement, scheduling into the reactor and feedback to developers, designers and manufacturers. As of mid-1976 over 3 x 10 6 individual elements have been built in a well-estabilished commercially competitive fuel fabrication industry and over 2 x 10 6 elements have been irradiated. Only six defects have been attributed to faulty materials or fabrication, and the use of high-density UO 2 with low-moisture content precluded defects from hydrogen contamination and densification. Development work on UO 2 and other fuel cycles (plutonium and thorium) is continuing, and, because CANDU reactors use on-power fuelling, bundles can be inserted into power reactors for testing. Thus new fuel designs can be quickly adopted to ensure that the CANDU system continues to provide low-cost energy with high reliability

  6. Development of Dynamic Spent Nuclear Fuel Environmental Effect Analysis Model

    International Nuclear Information System (INIS)

    Jeong, Chang Joon; Ko, Won Il; Lee, Ho Hee; Cho, Dong Keun; Park, Chang Je

    2010-07-01

    The dynamic environmental effect evaluation model for spent nuclear fuel has been developed and incorporated into the system dynamic DANESS code. First, the spent nuclear fuel isotope decay model was modeled. Then, the environmental effects were modeled through short-term decay heat model, short-term radioactivity model, and long-term heat load model. By using the developed model, the Korean once-through nuclear fuel cycles was analyzed. The once-through fuel cycle analysis was modeled based on the Korean 'National Energy Basic Plan' up to 2030 and a postulated nuclear demand growth rate until 2150. From the once-through results, it is shown that the nuclear power demand would be ∼70 GWe and the total amount of the spent fuel accumulated by 2150 would be ∼168000 t. If the disposal starts from 2060, the short-term decay heat of Cs-137 and Sr-90 isotopes are W and 1.8x10 6 W in 2100. Also, the total long-term heat load in 2100 will be 4415 MW-y. From the calculation results, it was found that the developed model is very convenient and simple for evaluation of the environmental effect of the spent nuclear fuel

  7. Spent sulfite liquor developments

    Energy Technology Data Exchange (ETDEWEB)

    Black, H H

    1958-01-01

    A review of methods of utilizing spent sulfite liquor, including evaporation and burning, fermentation to produce yeast or alcohol, production of vanillin and lignosulfonates, and use as a roadbinder.

  8. Spent fuel management

    International Nuclear Information System (INIS)

    2005-01-01

    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

  9. Korean Screen Cultures

    DEFF Research Database (Denmark)

    Screen Cultures sets out to redress this imbalance with a broad selection of essays spanning both North and South as well as different methodological approaches, from ethnographic and audience studies to cultural materialist readings. The first section of the book, «The South», highlights popular media...... new approaches to Korean popular culture beyond national borders and includes work on K-pop and Korean television drama. This book is a vital addition to existing scholarship on Korean popular culture, offering a unique view by providing an imaginary unification of the two Koreas negotiated through...

  10. Understanding Korean food culture from Korean paintings

    Directory of Open Access Journals (Sweden)

    Hae Kyung Chung

    2016-03-01

    Conclusion: Beyond the basic value of food in maximizing nutrients and energy, Korean food culture has developed distinctive cultural characteristics through more than 5,000 years of agricultural history. Although the genre paintings analyzed in this paper are limited to a certain era, this paper will serve as a milestone in providing direction for future studies.

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

    International Nuclear Information System (INIS)

    Johnson, A.B.

    1978-01-01

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

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

    International Nuclear Information System (INIS)

    Basso, R.A.

    1990-07-01

    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

  13. Preliminary evaluation of licensing issues associated with U. S. -sited CANDU-PHW nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    van Erp, J B

    1977-12-01

    The principal safety-related characteristics of current CANDU-PHW power plants are described, and a distinction between those characteristics which are intrinsic to the CANDU-PHW system and those that are not is presented. An outline is given of the main features of the Canadian safety and licensing approach. Differences between the U.S. and Canadian approach to safety and licensing are discussed. Some of the main results of the safety analyses, routinely performed for CANDU-PHW reactors, are presented. U.S.-NRC General Design Criteria are evaluated as regards their applicability to CANDU-PHW reactors; vice-versa the CANDU-PHW reactor is evaluated with respect to its conformance to the U.S.-NRC General Design Criteria. A number of design modifications are proposed to be incorporated into the CANDU-PHW reactor in order to facilitate its introduction into the U.S.

  14. Improved operability of the CANDU 9 control centre

    International Nuclear Information System (INIS)

    Macbeth, M. J.; Webster, A.

    1996-01-01

    The next generation CANDU nuclear power plant being designed by AECL is the 900 MWe class CANDU 9 station. It is based upon the Darlington CANDU station design which is among the world leaders in capacity factor with low Operation, Maintenance and Administration (OM and A) costs. This Control Centre design includes the proven functionality of existing CANDU control centres (including the Wolsong 2,3, and 4 control centre improvements, such as the Emergency Core Cooling panels), the characteristics identified by systematic design with human factors analysis of operations requirements and the advanced features needed to improve station operability which is made possible by the application of new technology. The CANDU 9 Control Centre provides plant staff with an improved operability capability due to the combination of proveness, systematic design with human factors engineering and enhanced operating features. Significant features which contribute to this improved operability include: · Standard NSP, BOP and F/H panels with controls and indicators integrated by a standard display/presentation philosophy. · Common plant parameter signal database for extensive monitoring, checking, display and annunciation. · Powerful annunciation system allowing alarm filtering, prioritizing and interrogation to enhance staff recognition of events, plant state and required corrective procedural actions. · The use of an overview display to present immediate and uncomplicated plant status information to facilitate operator awareness of unit status in a highly readable and recognizable format. · Extensive cross checking of similar process parameters amongst themselves, with the counterpart safety system parameters and as well as with 'signature' values obtained from known steady state conditions. · Powerful calculation capabilities, using the plant wide database, providing immediate recognizable and readable and readable output data on plant state information and plant state change

  15. Fuel channel design improvements for large CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Villamagna, A; Price, E G; Field, G J [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

    1996-12-31

    From the initial designs used in NPD and Douglas point reactors, the CANDU fuel channel and its components have undergone considerable development. Two major designs have evolved: the Pickering/CANDU 6 design which has 12 fuel bundles in the core and where the new fuel is inserted into the inlet end, and the Bruce/Darlington design which has 13 bundles in the channel and where new fuel is inserted into the outlet end. In the development of a single unit CANDU reactor of the size of a Bruce or Darlington unit which would use a Darlington design calandria, the decision has been made to use the CANDU 6 fuel channel rather than the Darlington design. The CANDU 6 channel has provided excellent performance and will not encounter the degree of maintenance required for the Bruce/Darlington design. The channel design in turn influences the fuelling machine/fuel handling concepts required. The changes to the CANDU 6 fuel channel design to incorporate it in the large unit are small. In fact, the changes that are proposed relate to the desire to increase margins between pressure tube properties and design conditions or ameliorate the consequences of postulated accident conditions, rather than necessary adaptation to the larger unit. Better properties have been achieved in the pressure tube material resulting from alloy development program over the past 10 years. Pressure tubes can now he made with very low hydrogen concentrations so that the hydrogen picked up as deuterium will not exceed the terminal solid solubility for the in-core region in 30 years. The improvements in metal chemistry allow the production of high toughness tubes that retain a high level of toughness during service. A small increase in wall thickness will reduce the dimensional changes without significantly affecting burnup. Changes to increase safety margins from postulated accidents are concentrated on containing the consequences of pressure tube damage. The changes are concentrated on the calandria tube

  16. Risk assessment in spent fuel storage and transportation

    International Nuclear Information System (INIS)

    Pandimani, S.

    1989-01-01

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

  17. Performance of pressure tubes in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rodgers, D.; Griffiths, M.; Bickel, G.; Buyers, A.; Coleman, C.; Nordin, H.; St Lawrence, S. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2016-06-15

    The pressure tubes in CANDU reactors typically operate for times up to about 30 years prior to refurbishment. The in-reactor performance of Zr-2.5Nb pressure tubes has been evaluated by sampling and periodic inspection. This paper describes the behavior and discusses the factors controlling the behaviour of these components. The Zr–2.5Nb pressure tubes are nominally extruded at 815{sup o}C, cold worked nominally 27%, and stress relieved at 400 {sup o}C for 24 hours, resulting in a structure consisting of elongated grains of hexagonal close-packed alpha-Zr, partially surrounded by a thin network of filaments of body-centred-cubic beta-Zr. These beta-Zr filaments are meta-stable and contain about 20% Nb after extrusion. The stress-relief treatment results in partial decomposition of the beta-Zr filaments with the formation of hexagonal close-packed alpha-phase particles that are low in Nb, surrounded by a Nb-enriched beta-Zr matrix. The material properties of pressure tubes are determined by variations in alpha-phase texture, alpha-phase grain structure, network dislocation density, beta-phase decomposition, and impurity concentration that are a function of manufacturing variables. The pressure tubes operate at temperatures between 250 {sup o}C and 310 {sup o}C with coolant pressures up to about 11 MPa in fast neutron fluxes up to 4 x 10{sup 17} n·m{sup -2}·s{sup -1} (E > 1 MeV) and the properties are modified by these conditions. The properties of the pressure tubes in an operating reactor are therefore a function of both manufacturing and operating condition variables. The ultimate tensile strength, fracture toughness, and delayed hydride-cracking properties (velocity (V) and threshold stress intensity factor (K{sub IH})) change with irradiation, but all reach a nearly limiting value at a fluence of less than 10{sup 25} n·m{sup -2} (E > 1 MeV). At this point the ultimate tensile strength is raised about 200 MPa, toughness is reduced by about 50%, V increases

  18. Korean Affairs Report

    Science.gov (United States)

    1985-09-20

    Pyongyang station, the South Korean delegation boarded 16 Mercedes Benz sedans and seven buses, then headed for the Koryo Hotel. The South Korean...money-making, loudly advertising "relief of hungry children, "relief of the destitute" and "programme for the rebirth of narcotic addicts. It was with... market are twin pains these days. When they convert what they earned on the bearish stock exchange into dol- lars, their returns show minus gains

  19. CANDU physics considerations for the disposition of weapons-grade plutonium

    International Nuclear Information System (INIS)

    Pitre, J.; Chan, P.; Dastur, A.

    1995-01-01

    At the request of the US Department of Energy AECL has examined the feasibility of using CANDU for the disposition of weapons grade plutonium. Utilizing existing CANDU technology, the feasibility of using MOX (mixed oxide) fuel in an existing CANDU reactor was studied. The results of this study indicate that the target disposition for disposal of weapons grade plutonium can be met without the requirement of any major modifications to existing plant design. (author). 3 refs., 4 tabs., 5 figs

  20. CANDU physics considerations for the disposition of weapons-grade plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Pitre, J; Chan, P; Dastur, A [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

    1996-12-31

    At the request of the US Department of Energy AECL has examined the feasibility of using CANDU for the disposition of weapons grade plutonium. Utilizing existing CANDU technology, the feasibility of using MOX (mixed oxide) fuel in an existing CANDU reactor was studied. The results of this study indicate that the target disposition for disposal of weapons grade plutonium can be met without the requirement of any major modifications to existing plant design. (author). 3 refs., 4 tabs., 5 figs.

  1. Fuel-management simulations for once-through thorium fuel cycle in CANDU reactors

    International Nuclear Information System (INIS)

    Chan, P.S.W.; Boczar, P.G.; Ellis, R.J.; Ardeshiri, F.

    1999-01-01

    High neutron economy, on-power refuelling and a simple fuel bundle design result in unsurpassed fuel cycle flexibility for CANDU reactors. These features facilitate the introduction and exploitation of thorium fuel cycles in existing CANDU reactors in an evolutionary fashion. Detailed full-core fuel-management simulations concluded that a once-through thorium fuel cycle can be successfully implemented in an existing CANDU reactor without requiring major modifications. (author)

  2. On Being There: Korean in Japan.

    Science.gov (United States)

    Maher, John C.; Kawanishi, Yumiko

    1995-01-01

    Examines the experiences of ethnic Koreans and the Korean language in Japan since 1945, focusing on Japanese attitudes toward Koreans and the Korean language, provisions for Korean and bilingual education, and the increasing influence of Koreans in Japanese society. (18 references) (MDM)

  3. Spent fuel storage facility, Kalpakkam

    International Nuclear Information System (INIS)

    Shreekumar, B.; Anthony, S.

    2017-01-01

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

  4. Spent fuel storage and isolation

    International Nuclear Information System (INIS)

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

    1979-02-01

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

  5. Disposal of spent fuel

    International Nuclear Information System (INIS)

    Blomeke, J.O.; Ferguson, D.E.; Croff, A.G.

    1978-01-01

    Based on preliminary analyses, spent fuel assemblies are an acceptable form for waste disposal. The following studies appear necessary to bring our knowledge of spent fuel as a final disposal form to a level comparable with that of the solidified wastes from reprocessing: 1. A complete systems analysis is needed of spent fuel disposition from reactor discharge to final isolation in a repository. 2. Since it appears desirable to encase the spent fuel assembly in a metal canister, candidate materials for this container need to be studied. 3. It is highly likely that some ''filler'' material will be needed between the fuel elements and the can. 4. Leachability, stability, and waste-rock interaction studies should be carried out on the fuels. The major disadvantages of spent fuel as a disposal form are the lower maximum heat loading, 60 kW/acre versus 150 kW/acre for high-level waste from a reprocessing plant; the greater long-term potential hazard due to the larger quantities of plutonium and uranium introduced into a repository; and the possibility of criticality in case the repository is breached. The major advantages are the lower cost and increased near-term safety resulting from eliminating reprocessing and the treatment and handling of the wastes therefrom

  6. A design concept of underground facilities for the deep geologic disposal of spent fuel

    International Nuclear Information System (INIS)

    Lee, Jong Youl; Choi, Heui Joo; Choi, Jong Won; Hahn, Pil Soo

    2005-01-01

    Spent nuclear fuel from nuclear power plants can be disposed in the underground repository. In this paper, a concept of Korean Reference HLW disposal System (KRS-1) design is presented. Though no site for the underground repository has been specified in Korea, but a generic site with granitic rock is considered for reference spent fuel repository design. To implement the concept, design requirements such as spent fuel characteristics and capacity of the repository and design principles were established. Then, based on these requirements and principles, a concept of the disposal process, the facilities and the layout of the repository was developed

  7. Natural uranium equivalent fuel an innovative design for proven CANDU technology

    Energy Technology Data Exchange (ETDEWEB)

    Pineiro, F.; Ho, K.; Khaial, A.; Boubcher, M.; Cottrell, C.; Kuran, S., E-mail: fabricia.pineiro@candu.com [Candu Energy Inc., Mississauga, ON (Canada); Zhenhua, Z.; Zhiliang, M. [Third Qinshan Nuclear Power Company, Haiyan, Zhejiang (China)

    2015-07-01

    The high neutron economy, on-power refuelling capability and fuel bundle design simplicity in CANDU reactors allow for the efficient utilization of alternative fuels. Candu Energy Inc. (Candu), in collaboration with the Third Qinshan Nuclear Power Company (TQNPC), the China North Nuclear Fuel Corporation (CNNFC), and the Nuclear Power Institute of China (NPIC), has successfully developed an advanced fuel called Natural Uranium Equivalent (NUE). This innovative design consists of a mixture of recycled and depleted uranium, which can be implemented in existing CANDU stations thereby bringing waste products back into the energy stream, increasing fuel resources diversity and reducing fuel costs. (author)

  8. A study to develop the domestic functional requirements of the specific safety systems of CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Man Woong [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Lee, Jae Young; Park, Kun Chul [Handong Global Univ., Pohang (Korea, Republic of)] (and others)

    2003-03-15

    The present research has been made to develop and review critically the functional requirements of the specific safety systems of CANDU such as SDS-1, SDS2, ECCS, and containment. Based on R documents for this, a systematic study was made to develop the domestic regulation statements. Also, the conventional laws are carefully reviewed to see the compatibility to CANDU. Also, the safety assessment method for CANDU was studied by reviewing C documents and recommendation of IAEA. Through the present works, the vague policy in the CANDU safety regulation is cleaning up in a systematic form and a new frame to measure the objective risk of nuclear power plants was developed.

  9. A study to develop the domestic functional requirements of the specific safety systems of CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Man Woong; Lee, Jae Young; Bang, Kwang Hyun [Handong Global Univ., Pohang (Korea, Republic of)] (and others)

    2001-03-15

    The present research has been made to develop and review critically the functional requirements of the specific safety systems of CANDU such as SOS-1, SOS-2, ECCS and containment. Based on R documents for this, a systematic study was made to develop the domestic regulation statements. Also, the conventional laws are carefully reviewed to see the compatibility to CANDU. Also, the safety assessment method for CANDU was studied by reviewing C documents and recommendation of IAEA. Through the present works, the vague policy in the CANDU safety regulation is cleaning up in a systematic form and a new frame to measure the objective risk of nuclear power plants was developed.

  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. Guidebook on spent fuel storage

    International Nuclear Information System (INIS)

    1984-01-01

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

  12. Spent fuel pyroprocessing demonstration

    International Nuclear Information System (INIS)

    McFarlane, L.F.; Lineberry, M.J.

    1995-01-01

    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

  13. Spent Fuel in Chile

    International Nuclear Information System (INIS)

    López Lizana, F.

    2015-01-01

    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)

  14. Designing for human performance - CANDU electrical panels - then and now

    International Nuclear Information System (INIS)

    Patterson, B.K.

    2003-01-01

    Experience has shown that few changes are made to CANDU control center panels once plants are in service. There are several reasons for this, for example, operational conservatism, outage minimization, avoidance of retraining effort, and perhaps lack of appreciation of improvement opportunities. Since the initial CANDU 6 plants were designed, the industry has moved forward with adoption of new approaches to control room interface design that have introduced changes in design principles and emphasis. At Point Lepreau, the electrical panels have remained virtually unchanged since initial design. The author, the original electrical system commissioning engineer, reviews the design considerations that led to the initial panel design, reviews some of the operational limitations discovered during use, and offers some suggestions on improvement directions, if these panels were being designed today, that would offer an implementation to better support operator tasks. (author)

  15. The integrity of CANDU fuel during load following

    International Nuclear Information System (INIS)

    Tayal, M.; Manzer, A.M.; Sejnoha, R.; Hains, A.J.

    1989-08-01

    This paper summarizes data and analyses of integrity and of physics of CANDU fuel during load following. Measurements of irradiated fuel show that power cycles do not enhance release of fission gas. Data from research reactors show that the power cycles cause cyclic strains in the sheath. Finite element analyses show that the cyclic strains give highly multiaxial stresses in the sheath. The stresses and the strains are well into the plastic range. The cyclic loads 'use up' some fraction of the sheath's resistance to environmentally-assisted cracking (EAC), depending on the details of the fuel design and of then power cycles. The balance of the sheath's resistance to EAC continues to be available to counteract static loads. Thousands of fuel bundles have experienced many power cycles in research and in commercial reactors. Overall integrity of fuel bundles is well over 99%. Thus, CANDU fuel continues to show good performance in both base-load and load-following reactors

  16. Damping considerations in CANDU feeder pipe design and analysis

    International Nuclear Information System (INIS)

    Usmani, S.A.; Saleem, M.A.; So, G.

    1990-01-01

    Recent developments in pipe damping indicate a trend towards more realistic and less conservative values, which result in less rigid and safer pipe designs. The CANDU-PHW (Canada deuterium uranium, pressurized heavy water) reactor feeder pipe designs have applied similar approaches which permit seismic qualifications without overly restraining these compact arrays of pipes to cater for the large creep and thermal anchor movement. This paper reviews the feeder design aspects, especially pertaining to the design provisions, experimental verification and analytical modelling for seismic qualification in the light of recent pipe dynamic developments. Using illustrative examples, comparison of seismic analysis results is provided for the ASME Code Case N-411 dampings, and those traditionally used in the feeder seismic qualification. The results confirm acceptability of the traditional approach which permit simplified analysis to demonstrate seismic qualificationqualification of CANDU feeder pipes

  17. CANDU with supercritical water coolant: conceptual design features

    International Nuclear Information System (INIS)

    Spinks, N.

    1997-01-01

    An advanced CANDU reactor, with supercritical water as coolant, has many attractive design features. The pressure exceeds 22 MPa but coolant temperatures in excess of 370 degrees C can be reached without encountering the two-phase region with its associated fuel-dry-out and flow-instability problems. Increased coolant temperature leads to increased plant thermodynamic efficiency reducing unit energy cost through reduced specific capital cost and reduced fueling cost. Increased coolant temperature leads to reduced void reactivity via reduced coolant in-core density. Light water becomes a coolant option. To preserve neutron economy, an advanced fuel channel is needed and is described below. A supercritical-water-cooled CANDU can evolve as fuel capabilities evolve to withstand increasing coolant temperatures. (author)

  18. CANDU type fuel behavior evaluation - a probabilistic approach

    International Nuclear Information System (INIS)

    Moscalu, D.R.; Horhoianu, G.; Popescu, I.A.; Olteanu, G.

    1995-01-01

    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)

  19. Thermal hydraulic simulation of the CANDU nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Carvalho, Athos M.S.S. de; Ramos, Mario C.; Costa, Antonella L.; Fernandes, Gustavo H.N., E-mail: athos1495@yahoo.com.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciência e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq), Rio de janeiro, RJ (Brazil)

    2017-07-01

    The CANDU (Canada Deuterium Uranium) is a Canadian-designed power reactor of PHWR type (Pressurized Heavy Water Reactor) that uses heavy water (deuterium oxide) for moderator and coolant, and natural uranium for fuel. There are about 47 reactors of this type in operation around the world generating more than 23 GWe, highlighting the importance of this kind of device. In this way, the main purpose of this study is to develop a thermal hydraulic model for a CANDU reactor to aggregate knowledge in this line of research. In this way, a core modeling was performed using RELAP5-3D code. Results were compared with reference data to verify the model behavior in steady state operation. Thermal hydraulic parameters as temperature, pressure and mass flow rate were verified and the results are in good agreement with reference data, as it is being presented in this work. (author)

  20. The use of digital computers in CANDU shutdown systems

    International Nuclear Information System (INIS)

    Gilbert, R.S.; Komorowski, C.W.

    1986-01-01

    This paper summarizes the application of computers in CANDU shutdown systems. A general description of systems that are already in service is presented along with a description of a fully computerized shutdown system which is scheduled to enter service in 1987. In reviewing the use of computers in the shutdown systems there are three functional areas where computers have been or are being applied. These are (i) shutdown system monitoring, (ii) parameter display and testing and (iii) shutdown initiation. In recent years various factors (References 1 and 2) have influenced the development and deployment of systems which have addressed two of these functions. At the present time a system is also being designed which addresses all of these areas in a comprehensive manner. This fully computerized shutdown system reflects the previous design, and licensing experience which was gained in earlier applications. Prior to describing the specific systems which have been designed a short summary of CANDU shutdown system characteristics is presented

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

    International Nuclear Information System (INIS)

    Suk, Ho Chun; Hwang, Woan; Jeong, Young Hwan; Jung, Sung Hoon

    1991-07-01

    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)

  2. The CANDU man-machine interface and simulator training

    International Nuclear Information System (INIS)

    Hinchley, E.M.; Yanofsky, N.

    1982-09-01

    The most significant features of the man-machine interface for CANDU power stations are the extensive use of computer-driven colour graphics displays and the small number of manual controls. The man-machine interface in CANDU stations is designed to present the operator with concise, easy-to-understand information. Future developments in the use of computers in safety shutdown systems, and the use of data highway technologies in plant regulating systems will present special requirements and new opportunities in the application of human factors engineering to the control room. Good man-machine interaction depends on operator training as much as on control room design. In Canada computerized training simulators, which indicate plant response to operator action, are being introducted for operator training. Such simulators support training in normal operation of all plant systems and also in the fault management tasks following malfunctions

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

    International Nuclear Information System (INIS)

    Jeong, Chang Joon; Choi, Hang Bok

    1998-01-01

    Reactivity device characteristics for a CANDU-6 reactor loaded with DUPIC fuel have been assessed. A transport code WIMS-AECL and a three-dimensional diffusion code RFSP were used for the lattice parameter generation and the core calculation, respectively. Three major reactivity devices have been assessed for their inherent functions. For the zone controller system, damping capability for spatial oscillation was investigated. The restart capability of the adjuster system was investigated. The shim operation and power stepback calculation were also performed to confirm the compatibility of the current adjuster rod system. The mechanical control absorber was assessed for the capability to compensate the temperature reactivity feedback following a power reduction. This study has shown that the current reactivity device systems retain their functions when used in a DUPIC fuel CANDU reactor

  4. Reactor physics assessment of modified 37-element CANDU fuel bundles

    International Nuclear Information System (INIS)

    Pristavu, R.; Rizoiu, A.

    2016-01-01

    Reducing the central element diameter in order to improve the total flow area of CANDU fuel bundle and redistribute the power density of all remaining elements was studied in Canada and Korea when considering the effect of aging pressure tube diametral creep. The aim of this paper is to study the modified bundle behavior using the transport codes WIMS and DRAGON. In calculations, a WIMS nuclear data library on 172 energy groups was used. 2-D transport calculations were performed with WIMS and DRAGON, leading to similar results in estimated cell parameters. Additionally, 3-D DRAGON calculations were carried on in order to evaluate the local flux distribution shift, as well as the incremental cross sections for supercells containing modified CANDU bundles and reactivity devices. The overall effect of using modified fuel bundles was meaningless for both cell and supercell parameters, thus ensuring this possibility of fuel improvement for thermal-hydraulic purposes only. (authors)

  5. Integrated aerosol and thermalhydraulics modelling for CANDU safety analysis

    International Nuclear Information System (INIS)

    McDonald, B.H.; Hanna, B.N.

    1990-08-01

    Analysis of postulated accidents in CANDU reactors that could result in severe fuel damage requires the ability to model the formation of aerosols containing fission product materials and the transport of these aerosols from the fuel, through containment, to any leak to the atmosphere. Best-estimate calculations require intimate coupling and simultaneous solution of all the equations describing the entire range of physical and chemical phenomena involved. The prototype CATHENA/PACE-3D has been developed for integrated calculation of thermalhydraulic and aerosol events in a CANDU reactor during postulated accidents. Examples demonstrate the ability of CATHENA/PACE-3D to produce realistic flow and circulation patterns and reasonable accuracy in solution of two simple fluid-flow test cases for which analytical solutions exist

  6. Fabrication of PWR fuel assembly and CANDU fuel bundle

    International Nuclear Information System (INIS)

    Lee, G.S.; Suh, K.S.; Chang, H.I.; Chung, S.H.

    1980-01-01

    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

  7. Electrical, control and information systems in the Enhanced CANDU 6

    International Nuclear Information System (INIS)

    De Grosbois, J.; Raiskums, G.; Soulard, M.

    2011-01-01

    This paper describes the electrical, control, and information system (EC and I) design feature improvements of the Enhanced CANDU 6 (EC6). These additional features are carefully integrated into the EC6 design platform, and are engineered with consideration of operational feedback, human factors, and leveraging the advantages of digital instrumentation and control (I and C) technology to create a coherent I and C architecture in support of safe and high performance operation. The design drivers for the selection of advanced features are also discussed. The EC6 nuclear power plant is a mid-sized Pressurized Heavy Water Reactor design, based on the highly successful CANDU 6 family of power plants, and upgraded to meet today's Canadian and international safety requirements and to satisfy Generation 3 design expectations. (author)

  8. Canadian CANDU plant data systems for technical surveillance and analysis

    International Nuclear Information System (INIS)

    Deverno, M.; Pothier, H.; Xian, C.; Grosbois, J. De; Bosnich, M.

    1996-01-01

    Plant data systems are emerging as a critical plant support system technology. In particular, plant-wide Historical Data Systems (HDS) are pivotal to the successful implementation of technical surveillance and analysis programs supporting plant operations, maintenance, safety, and licensing activities. In partnership with Canadian CANDU utility and design organizations, AECL has conducted a review of current Canadian CANDU HDS approaches with emphasis on understanding the existing functionality and uses of plant historical data systems, their future needs and benefits. The results is a vision of a plant-wide HDS providing seamless access to both near real-time and historical data, user tool-kits for data visualization and analysis, and data management of the large volume of data acquired during the life of a plant. The successful implementation of the HDS vision will lead to higher capability and capacity factors while minimizing Operations, Maintenance, and Administration (OM and A) costs. (author). 5 refs, 3 figs

  9. The burnable poisons utilization for fissile enriched CANDU fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Serghiuta, D; Nainer, O [Team 3 Solutions, Don Mills, ON (Canada)

    1996-12-31

    Utilization of burnable poison for the fissile enriched fueled CANDU 6 Mk1 core is investigated. The main incentives for this analysis are the reduction of void reactivity effects, the maximization of the fissile content of fresh fuel bundles, and the achievement of better power shape control, in order to preserve the power envelope of the standard 37 rod fuel bundle. The latter allows also the preservation of construction parameters of the standard core (for example: number and location of reactivity devices). It also permits the use of regular shift fueling schemes. The paper makes analyses of MOX weapons-grade plutonium and 1.2% SEU fueled CANDU 6 Mk 1 cores. (author). 6 refs., 4 tabs., 10 figs.

  10. Once-through uranium thorium fuel cycle in CANDU reactors

    International Nuclear Information System (INIS)

    Ozdemir, S.; Cubukcu, E.

    2000-01-01

    In this study, the performance of the once-through uranium-thorium fuel cycle in CANDU reactors is investigated. (Th-U)O 2 is used as fuel in all fuel rod clusters where Th and U are mixed homogeneously. CANDU reactors have the advantage of being capable of employing various fuel cycle options because of its good neutron economy, continuous on line refueling ability and axial fuel replacement possibility. For lattice cell calculations transport code WIMS is used. WIMS cross-section library is modified to achieve precise lattice cell calculations. For various enrichments and Th-U mixtures, criticality, heavy element composition changes, diffusion coefficients and cross-sections are calculate. Reactor core is modeled by using the diffusion code CITATION. We conclude that an overall saving of 22% in natural uranium demand can be achieved with the use of Th cycle. However, slightly enriched U cycle still consumes less natural Uranium and is a lot less complicated. (author)

  11. Canadian CANDU plant data systems for technical surveillance and analysis

    Energy Technology Data Exchange (ETDEWEB)

    Deverno, M; Pothier, H; Xian, C [Atomic Energy of Canada Ltd., Control Centre Technology Branch, Fredericton, NB (Canada); Grosbois, J De; Bosnich, M [Atomic Energy of Canada Ltd., Control Centre Technology Branch, Chalk River, ON (Canada). Chalk River Labs.

    1997-12-31

    Plant data systems are emerging as a critical plant support system technology. In particular, plant-wide Historical Data Systems (HDS) are pivotal to the successful implementation of technical surveillance and analysis programs supporting plant operations, maintenance, safety, and licensing activities. In partnership with Canadian CANDU utility and design organizations, AECL has conducted a review of current Canadian CANDU HDS approaches with emphasis on understanding the existing functionality and uses of plant historical data systems, their future needs and benefits. The results is a vision of a plant-wide HDS providing seamless access to both near real-time and historical data, user tool-kits for data visualization and analysis, and data management of the large volume of data acquired during the life of a plant. The successful implementation of the HDS vision will lead to higher capability and capacity factors while minimizing Operations, Maintenance, and Administration (OM and A) costs. (author). 5 refs, 3 figs.

  12. Present and future of Korean nuclear power

    International Nuclear Information System (INIS)

    Min, K-H

    2014-01-01

    'Full text:' The Korean nuclear power industry has devoted itself to technological development and self-reliance over the last 30 years since Kori unit 1, the first nuclear power plant commenced its commercial operation in 1978. As a result of such efforts and accumulated experiences, the Korean nuclear power industry has developed the OPR 1000 and APR 1400 units and is almost completing the development of the APR+ as a 1,500MW class reactor with its own technologies of design and manufacturing. Also, the Korean nuclear power industry has been able to build a strong supply chain from engineering, manufacturing, construction, and fuel supply, to operation and maintenance. At present, Korea is operating 23 commercial power reactors with a total installed capacity of 20,716 MW, accounting for 25 percent of the installed capacity and one third of the nation's total electricity generation. Also, the share of nuclear power generation capacity will be 29 percent by 2035 in the Long Term Energy Development Plan and 43 GW of nuclear energy capacity will be needed. Thanks to nuclear power generation as an essential driving force, Korea has been able to supply cheap and stable electricity. However, amid the growing public concerns about nuclear safety after the Fukushima accident, the Korean government and related organizations are exerting its utmost effort in all areas, for example, enhancing nuclear safety and safety culture, carrying out management innovation, and communicating with the public in order to enhance transparency. Also, the Korean government launched the Public Engagement Commission on spent nuclear fuel (SNF) management in 2013, which is tasked to initiate public consultation & discussion and submit recommendation to government after in-depth review and analysis on SNF management options by the end of 2014. Nuclear power has become very essential part of national economy in Korea because Korea has virtually no indigenous energy resources and

  13. Present and future of Korean nuclear power

    Energy Technology Data Exchange (ETDEWEB)

    Min, K-H [Korea Atomic Industrial Forum, Inc., Seoul (Korea, Republic of)

    2014-07-01

    'Full text:' The Korean nuclear power industry has devoted itself to technological development and self-reliance over the last 30 years since Kori unit 1, the first nuclear power plant commenced its commercial operation in 1978. As a result of such efforts and accumulated experiences, the Korean nuclear power industry has developed the OPR 1000 and APR 1400 units and is almost completing the development of the APR+ as a 1,500MW class reactor with its own technologies of design and manufacturing. Also, the Korean nuclear power industry has been able to build a strong supply chain from engineering, manufacturing, construction, and fuel supply, to operation and maintenance. At present, Korea is operating 23 commercial power reactors with a total installed capacity of 20,716 MW, accounting for 25 percent of the installed capacity and one third of the nation's total electricity generation. Also, the share of nuclear power generation capacity will be 29 percent by 2035 in the Long Term Energy Development Plan and 43 GW of nuclear energy capacity will be needed. Thanks to nuclear power generation as an essential driving force, Korea has been able to supply cheap and stable electricity. However, amid the growing public concerns about nuclear safety after the Fukushima accident, the Korean government and related organizations are exerting its utmost effort in all areas, for example, enhancing nuclear safety and safety culture, carrying out management innovation, and communicating with the public in order to enhance transparency. Also, the Korean government launched the Public Engagement Commission on spent nuclear fuel (SNF) management in 2013, which is tasked to initiate public consultation & discussion and submit recommendation to government after in-depth review and analysis on SNF management options by the end of 2014. Nuclear power has become very essential part of national economy in Korea because Korea has virtually no indigenous energy resources and

  14. Heat sink management during CANDU low level operation

    International Nuclear Information System (INIS)

    Wang Liansheng

    2008-01-01

    This paper introduces the practice of low-level operation with opening on the main heat transport system during an outage for a Candu-6 nuclear power plant, analyses the risks of losing heat sink during this condition, and points out the safety measures and management requirement for controlling such risks. This paper can be used as a reference for improving and optimizing the heat sink management for the coming outages. (author)

  15. Role of water lubricated bearings in Candu reactors

    International Nuclear Information System (INIS)

    Kumar, Ashok N.

    1999-01-01

    During the twentieth century a great emphasis was placed in understanding and defining the operating regime of oil and grease lubricated components. Major advances have been made through elastohydrodynamic lubrication theory in the quantifying the design life of heavily loaded components such as rolling element bearings and gears. Detailed guidelines for the design of oil and grease lubricated components are widely available and are being applied to the successful design of these components. However similar guidelines for water lubricated components are either not available or not well documented. It is often forgotten that the water was used as a lubricant in several components as far back as 1884 B.C. During the twentieth century the water lubricated components continued to play a major role in some high technology industries such as in the power generation plants. In CANDU nuclear reactors water lubrication of several critical components always occupied a pride place and in most cases the only practical mode of lubrication of several critical components always occupied a pride place and in most cases the only practical mode of lubrication. This paper presents some examples of the major water lubricated components in a CANDU reactors. Major part of the paper is focused on presenting an example of successful operating history of water lubricated bearings used in the HT pumps are presented. Both types of bearings have been qualified by tests for operation under normal as well as under more severe postulated condition of loss-of-coolant-accident (LOCA). These bearings have been designed to operate for the 30 years in the existing CANDU 6 (600 MW) reactors. However for the next generation of CANDU 6 reactors which go into service in the year 2003, the HT pump bearing life has been extended to 40 years. (author)

  16. Use of enriched uranium as a fuel in CANDU reactors

    International Nuclear Information System (INIS)

    Zech, H.J.

    1976-08-01

    The use of slightly enriched uranium as a fuel in CANDU-reactors is studied in a simple parametric way. The results show the possibility of 1) about 30% savings in natural uranium consumption 2) about 35% increase in the utilization of the natural uranium 3) a decrease in fuelling costs to about 70 - 80% of the normal case of natural uranium fuelling. (orig.) [de

  17. Proceedings of the third international conference on CANDU maintenance

    International Nuclear Information System (INIS)

    1995-01-01

    The third international conference on Candu maintenance included sessions on the following topics: predictive maintenance, reliability improvements, steam generator monitoring, tools and instrumentation, valve performance, fuel channel inspection and maintenance, steam generator maintenance, environmental qualification, predictive maintenance, instrumentation and control, steam generator cleaning, decontamination and radiation protection, inspection techniques, maintenance program strategies and valve packing experience, remote tooling/ robotics and fuel handling. The individual papers have been abstracted separately

  18. Advances in fuel channel technology for CANDU reactors

    International Nuclear Information System (INIS)

    Cheadle, B.A.; Coleman, C.E.

    1994-05-01

    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

  19. Proceedings of the third international conference on CANDU maintenance

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-12-31

    The third international conference on Candu maintenance included sessions on the following topics: predictive maintenance, reliability improvements, steam generator monitoring, tools and instrumentation, valve performance, fuel channel inspection and maintenance, steam generator maintenance, environmental qualification, predictive maintenance, instrumentation and control, steam generator cleaning, decontamination and radiation protection, inspection techniques, maintenance program strategies and valve packing experience, remote tooling/ robotics and fuel handling. The individual papers have been abstracted separately.

  20. Spent fuel reprocessing options

    International Nuclear Information System (INIS)

    2008-08-01

    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