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Sample records for candu fuel bundle

  1. Using Advanced Fuel Bundles in CANDU Reactors

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  3. Subchannel analysis of CANDU 37-element fuel bundles

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-08-15

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-06-15

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  17. Post-irradiation examination of CANDU fuel bundles fuelled with (Th, Pu)O2

    International Nuclear Information System (INIS)

    AECL has extensive experience with thoria-based fuel irradiations as part of an ongoing R&D program on thorium within the Advanced Fuel Cycles Program. The BDL-422 experiment was one component of the thorium program that involved the fabrication and irradiation testing of six Bruce-type bundles fuelled with (Th, Pu)O2 pellets. The fuel was manufactured in the Recycle Fuel Fabrication Laboratories (RFFL) at Chalk River allowing AECL to gain valuable experience in fabrication and handling of thoria fuel. The fuel pellets contained 86.05 wt. % Th and 1.53 wt. % Pu in (Th, Pu)O2. The objectives of the BDL-422 experiment were to demonstrate the ability of 37-element geometry (Th, Pu)O2 fuel bundles to operate to high burnups up to 1000 MWh/kgHE (42 MWd/kgHE), and to examine the (Th, Pu)O2 fuel performance. This paper describes the post-irradiation examination (PIE) results of BDL-422 fuel bundles irradiated to burnups up to 856 MWh/kgHE (36 MWd/kgHE), with power ratings ranging from 52 to 67 kW/m. PIE results for the high burnup bundles (>1000 MWh/kgHE) are being analyzed and will be reported at a later date. The (Th, Pu)O2 fuel performance characteristics were superior to UO2 fuel irradiated under similar conditions. Minimal grain growth was observed and was accompanied by benign fission gas release and sheath strain. Other fuel performance parameters, such as sheath oxidation and hydrogen distribution, are also discussed. (author)

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

    International Nuclear Information System (INIS)

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

  19. Installation of an irradiated fuel bundle discharge counter at Bruce NGS-B 3 000 MW(e) CANDU power station

    International Nuclear Information System (INIS)

    Design, manufacture and installation of an irradiated fuel bundle discharge counter for the multi-unit CANDU Bruce NGS-B Generating Station involved contributions from the International Atomic Energy Agency (Agency), designers (AECL), contractors, manufacturers, utility and the regulatory agency. The installation at Bruce NGS-B was the first made by the Agency as a retrofit to a multi-unit CANDU reactor approaching its fist critical operation, where the whole project was the responsibility of the Agency and where the original design of the reactor had not had provision for the Agency equipment. The scheduling and integration of the installation into the normal activities involved in starting up a 3 000 MW(e) multi-unit generating station were successfully achieved. The Agency has demonstrated the capability and performance of the fuel discharge counter

  20. CANDU bundle junction. Misalignment probability and pressure-drop correlation

    International Nuclear Information System (INIS)

    The pressure drop over the bundle junction is an important component of the pressure drop in a CANDU (Canada Deuterium Uranium) fuel channel. This component can represent from ∼ 15% for aligned bundles to ∼ 26% for rotationally misaligned bundles, and is dependent on the degree of misalignment. The geometry of the junction increases the mixing between subchannels, and hence improves the thermal performance of the bundle immediately downstream. It is therefore important to model the junction's performance adequately. This paper summarizes a study sponsored by COG (CANDU Owners Group) and an NSERC (National Science and Engineering Research Council) Industrial Research Grant, undertaken, at CRL (Chalk River Laboratories) to identify and develop a bundle-junction model for potential implementation in the ASSERT (Advanced Solution of Subchannel Equations in Reactor Thermalhydraulics) subchannel code. The work reported in this paper consists of two components of this project: an examination of the statistics of bundle misalignment, demonstrating that there are no preferred positions for the bundles and therefore all misalignment angles are equally possible; and, an empirical model for the single-phase pressure drop across the junction as a function of the misalignment angle. The second section of this paper includes a brief literature review covering the experimental, analytical and numerical studies concerning the single-phase pressure drop across bundle junctions. 32 refs., 9 figs

  1. Fuel for advanced CANDU reactors

    International Nuclear Information System (INIS)

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

  2. Application of Be-free Zr-based amorphous sputter coatings as a brazing filler metal in CANDU fuel bundle manufacture

    International Nuclear Information System (INIS)

    Amorphous sputter coatings of Be-free multi-component Zr-based alloys were applied as a novel brazing filler metal for Zircaloy-4 brazing. By applying the homogeneous and amorphous-structured layers coated by sputtering the crystalline targets, the highly reliable joints were obtained with the formation of predominantly grown α-Zr grains owing to a complete isothermal solidification, exhibiting high tensile and fatigue strengths as well as excellent corrosion resistance, which were comparable to those of Zircaloy-4 base metal. The present investigation showed that Be-free and Zr-based multi-component amorphous sputter coatings can offer great potential for brazing Zr alloys and manufacturing fuel rods in CANDU fuel bundle system. (author)

  3. Second International Conference on CANDU Fuel

    International Nuclear Information System (INIS)

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

  4. Candu reactors with thorium fuel cycles

    International Nuclear Information System (INIS)

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

  5. Fuel condition in Canadian CANDU 6 reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  6. Fuel condition in Canadian CANDU 6 reactors

    International Nuclear Information System (INIS)

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

  7. A prediction method of the effect of radial heat flux distribution on critical heat flux in CANDU fuel bundles

    International Nuclear Information System (INIS)

    Fuel irradiation experiments to study fuel behaviors have been performed in the experimental loops of the National Research Universal (NRU) Reactor at Atomic Energy of Canada Limited (AECL) Chalk River Laboratories (CRL) in support of the development of new fuel technologies. Before initiating a fuel irradiation experiment, the experimental proposal must be approved to ensure that the test fuel strings put into the NRU loops meet safety margin requirements in critical heat flux (CHF). The fuel strings in irradiation experiments can have varying degrees of fuel enrichment and burnup, resulting in large variations in radial heat flux distribution (RFD). CHF experiments performed in Freon flow at CRL for full-scale bundle strings with a number of RFDs showed a strong effect of RFD on CHF. A prediction method was derived based on experimental CHF data to account for the RFD effect on CHF. It provides good CHF predictions for various RFDs as compared to the data. However, the range of the tested RFDs in the CHF experiments is not as wide as that required in the fuel irradiation experiments. The applicability of the prediction method needs to be examined for the RFDs beyond the range tested by the CHF experiments. The Canadian subchannel code ASSERT-PV was employed to simulate the CHF behavior for RFDs that would be encountered in fuel irradiation experiments. The CHF predictions using the derived method were compared with the ASSERT simulations. It was observed that the CHF predictions agree well with the ASSERT simulations in terms of CHF, confirming the applicability of the prediction method in fuel irradiation experiments. (author)

  8. CANDU spent fuel dry storage interim technique

    International Nuclear Information System (INIS)

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

  9. Modelling nuclear fuel vibrations in horizontal CANDU reactors

    International Nuclear Information System (INIS)

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

  10. 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)

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

    International Nuclear Information System (INIS)

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

  12. Numerical model for thermal and mechanical behaviour of a CANDU 37-element bundle

    International Nuclear Information System (INIS)

    Prediction of transient fuel bundle deformations is important for assessing the integrity of fuel and the surrounding structural components under different operating conditions including accidents. For numerical simulation of the interactions between fuel bundle and pressure tube, a reliable numerical bundle model is required to predict thermal and mechanical behaviour of the fuel bundle assembly under different thermal loading conditions. To ensure realistic representations of the bundle behaviour, this model must include all of the important thermal and mechanical features of the fuel bundle, such as temperature-dependent material properties, thermal viscoplastic deformation in sheath, fuel-to-sheath interactions, endplate constraints and contacts between fuel elements. In this paper, we present a finite element based numerical model for predicting macroscopic transient thermal-mechanical behaviour of a complete 37-element CANDU nuclear fuel bundle under accident conditions and demonstrate its potential for being used to investigate fuel bundle to pressure tube interaction in future nuclear safety analyses. This bundle model has been validated against available experimental and numerical solutions and applied to various simulations involving steady-state and transient loading conditions. (author)

  13. Basic research and industrialization of CANDU advanced fuel

    International Nuclear Information System (INIS)

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

  14. 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

  15. 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.

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

    International Nuclear Information System (INIS)

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

  17. Assessment of ASSERT-PV for prediction of post-dryout heat transfer in CANDU bundles

    International Nuclear Information System (INIS)

    Highlights: • Assessment of the new Canadian subchannel code ASSERT-PV 3.2 for PDO sheath temperature prediction. • CANDU 28-, 37- and 43-element bundle PDO experiments. • Prediction improvement of ASSERT-PV 3.2 over previous code versions. • Sensitivity study of the effect of PDO model options. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of subchannel flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against PDO tests performed during five full-size CANDU bundle experiments conducted between 1992 and 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element bundles. A total of 10 PDO test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompassed the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for existing CANDU reactors. Code predictions of maximum PDO fuel-sheath temperature were compared against measurements from the SL PDO tests to quantify the code's prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, separate-effects sensitivity studies quantified the contribution of each PDO model change or enhancement to the improvement in PDO heat transfer prediction. Overall, the assessment demonstrated significant improvement in prediction of PDO sheath temperature in horizontal fuel channels containing CANDU bundles

  18. Assessment of ASSERT-PV for prediction of post-dryout heat transfer in CANDU bundles

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

    Highlights: • Assessment of the new Canadian subchannel code ASSERT-PV 3.2 for PDO sheath temperature prediction. • CANDU 28-, 37- and 43-element bundle PDO experiments. • Prediction improvement of ASSERT-PV 3.2 over previous code versions. • Sensitivity study of the effect of PDO model options. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of subchannel flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against PDO tests performed during five full-size CANDU bundle experiments conducted between 1992 and 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element bundles. A total of 10 PDO test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompassed the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for existing CANDU reactors. Code predictions of maximum PDO fuel-sheath temperature were compared against measurements from the SL PDO tests to quantify the code's prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, separate-effects sensitivity studies quantified the contribution of each PDO model change or enhancement to the improvement in PDO heat transfer prediction. Overall, the assessment demonstrated significant improvement in prediction of PDO sheath temperature in horizontal fuel channels containing CANDU bundles.

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-31

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

  1. Assessment of ASSERT-PV for prediction of critical heat flux in CANDU bundles

    International Nuclear Information System (INIS)

    Highlights: • Assessment of the new Canadian subchannel code ASSERT-PV 3.2 for CHF prediction. • CANDU 28-, 37- and 43-element bundle CHF experiments. • Prediction improvement of ASSERT-PV 3.2 over previous code versions. • Sensitivity study of the effect of CHF model options. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against five full-scale CANDU bundle experiments conducted in 1990s and in 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element (CANFLEX) bundles. A total of 15 CHF test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompassed the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for CANDU reactors. Code predictions of channel dryout power and axial and radial CHF locations were compared against measurements from the SL CHF tests to quantify the code prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, the sensitivity studies evaluated the contribution of each CHF model change or enhancement to the improvement in CHF prediction. Overall, the assessment demonstrated significant improvement in prediction of channel dryout power and axial and radial CHF locations in horizontal fuel channels containing CANDU bundles

  2. Assessment of ASSERT-PV for prediction of critical heat flux in CANDU bundles

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-15

    Highlights: • Assessment of the new Canadian subchannel code ASSERT-PV 3.2 for CHF prediction. • CANDU 28-, 37- and 43-element bundle CHF experiments. • Prediction improvement of ASSERT-PV 3.2 over previous code versions. • Sensitivity study of the effect of CHF model options. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against five full-scale CANDU bundle experiments conducted in 1990s and in 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element (CANFLEX) bundles. A total of 15 CHF test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompassed the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for CANDU reactors. Code predictions of channel dryout power and axial and radial CHF locations were compared against measurements from the SL CHF tests to quantify the code prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, the sensitivity studies evaluated the contribution of each CHF model change or enhancement to the improvement in CHF prediction. Overall, the assessment demonstrated significant improvement in prediction of channel dryout power and axial and radial CHF locations in horizontal fuel channels containing CANDU bundles.

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

    International Nuclear Information System (INIS)

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

  4. Enthalpy and void distributions in subchannels of PHWR fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Park, J. W.; Choi, H.; Rhee, B. W. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    Two different types of the CANDU fuel bundles have been modeled for the ASSERT-IV code subchannel analysis. From calculated values of mixture enthalpy and void fraction distribution in the fuel bundles, it is found that net buoyancy effect is pronounced in the central region of the DUPIC fuel bundle when compared with the standard CANDU fuel bundle. It is also found that the central region of the DUPIC fuel bundle can be cooled more efficiently than that of the standard fuel bundle. From the calculated mixture enthalpy distribution at the exit of the fuel channel, it is found that the mixture enthalpy and void fraction can be highest in the peripheral region of the DUPIC fuel bundle. On the other hand, the enthalpy and the void fraction were found to be highest in the central region of the standard CANDU fuel bundle at the exit of the fuel channel. This study shows that the subchannel analysis is very useful in assessing thermal behavior of the fuel bundle that could be used in CANDU reactors. 10 refs., 4 figs., 2 tabs. (Author)

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

    International Nuclear Information System (INIS)

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

  6. Future fuel cycle development for CANDU reactors

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-04-01

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

  11. Temperature effect of DUPIC fuel in CANDU reactor

    International Nuclear Information System (INIS)

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

  12. Thermochemical modelling of advanced CANDU reactor fuel

    Science.gov (United States)

    Corcoran, Emily Catherine

    2009-04-01

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

  13. Shielding calculations for spent CANDU fuel transport cask

    International Nuclear Information System (INIS)

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

  14. 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).

  15. Filler metals for containers holding irradiated fuel bundles

    International Nuclear Information System (INIS)

    One of the procedures being considered for the disposal of Canadian deuterium uranium (CANDU) irradiated fuel bundles is to place the bundles in containers, fill the containers with metal, and place them underground. This investigation deals with the selection of the filler metal with particular reference to the reaction rate with, and bonding of the filler metal to, the fuel sheathing (Zircaloy 4) and potential container materials. Lead, zinc, and aluminium alloys were examined as potential filler metals. (U.K.)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  18. Feasibility study of CANDU-9 fuel handling system

    International Nuclear Information System (INIS)

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

  19. Analysis of F/M duty cycle and O/M cost for four-bundle shift refuelling scheme in CANDU6 NPP

    International Nuclear Information System (INIS)

    A four-bundle shift refuelling method, a refuelling scheme that can reduces local flux peak compared to the current eight-bundle shift refuelling method used in CANDU6 NPP, is analyzed to see how much Fuel Handling System load and management cost increase are required due to the change. The current eight-bundle shift refuelling method requires to refuel eight fuel bundles from a single fuel channel, and to refuel 14 fuel channels in a week on average assuming that the reactor is in a steady state. The four-bundle shift refuelling method increases Fuelling Machine duty cycle and operator load. The study showed that the refuelling scheme change from the eight-to four-bundle shift increases the operation and maintenance cost about 35% from the current figure by conservative estimate and that the Fuel Handling System has enough flexibility to meet the demand of a more frequent refuelling scheme

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  5. 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).

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

    International Nuclear Information System (INIS)

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

  7. Research and development for CANDU fuel channels and fuel

    International Nuclear Information System (INIS)

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

  8. Inspection of Candu Nuclear Reactor Fuel Channels

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  10. Dry storage of irradiated CANDU fuel at Pickering NGS

    International Nuclear Information System (INIS)

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

  11. CHF Enhancement of Advanced 37-Element Fuel Bundles

    Directory of Open Access Journals (Sweden)

    Joo Hwan Park

    2015-01-01

    Full Text Available A standard 37-element fuel bundle (37S fuel bundle has been used in commercial CANDU reactors for over 40 years as a reference fuel bundle. Most CHF of a 37S fuel bundle have occurred at the elements arranged in the inner pitch circle for high flows and at the elements arranged in the outer pitch circle for low flows. It should be noted that a 37S fuel bundle has a relatively small flow area and high flow resistance at the peripheral subchannels of its center element compared to the other subchannels. The configuration of a fuel bundle is one of the important factors affecting the local CHF occurrence. Considering the CHF characteristics of a 37S fuel bundle in terms of CHF enhancement, there can be two approaches to enlarge the flow areas of the peripheral subchannels of a center element in order to enhance CHF of a 37S fuel bundle. To increase the center subchannel areas, one approach is the reduction of the diameter of a center element, and the other is an increase of the inner pitch circle. The former can increase the total flow area of a fuel bundle and redistributes the power density of all fuel elements as well as the CHF. On the other hand, the latter can reduce the gap between the elements located in the middle and inner pitch circles owing to the increasing inner pitch circle. This can also affect the enthalpy redistribution of the fuel bundle and finally enhance CHF or dry-out power. In this study, the above two approaches, which are proposed to enlarge the flow areas of the center subchannels, were considered to investigate the impact of the flow area changes of the center subchannels on the CHF enhancement as well as the thermal characteristics by applying a subchannel analysis method.

  12. General overview of CANDU advanced fuel cycles program

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-07-01

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

  14. Post-irradiation examination of the 37M fuel bundle at Chalk River Laboratories (AECL)

    Energy Technology Data Exchange (ETDEWEB)

    Armstrong, J. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Daniels, T. [Ontario Power Generation, Pickering, Ontario (Canada); Montin, J. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2014-03-15

    The modified (-element (37M) fuel bundle was designed by Ontario Power Generation (OPG) to improve Critical Heat Flux (CHF) performance in ageing pressure tubes. A modification of the conventional 37-element fuel bundle design, the 37M fuel bundle allows more coolant flow through the interior sub-channels by way of a smaller central element. A demonstration irradiation (DI) of thirty-two fuel bundles was completed in 2011 at OPG's Darlington Nuclear Generating Station to confirm the suitability of the 37M fuel bundles for full core implementation. In support of the DI, fuel elements were examined in the Chalk River Laboratories Hot Cells. Inspection activities included: Bundle and element visual examination; Bundle and element dimensional measurements; Verification of bundle and element integrity; and Internal Gas Volume Measurements. The inspection results for 37M were comparable to that of conventional 37-element CANDU fuel. Fuel performance parameters of the 37M DI fuel bundle and fuel elements were within the range observed for similarly operated conventional 37-element CANDU fuel. Based on these Post Irradiation Examination (PIE) results, 37M fuel performed satisfactorily. (author)

  15. Study on Unigraphics Drawing Modeling Method for 37-Element and CANFLEX Fuel Bundle

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Yu Mi; Park, Joo Hwan

    2010-03-15

    The CANFLEX bundle contains 43 elements of two different diameters. It has two rings of small diameter elements on the outside, and eight elements (with diameter slightly larger than those in the standard 37-Element bundle) in the center. This larger number of small diameter elements on the outside of the CANFLEX bundle enhances thermo-hydraulic capability, resulting in a higher power capability and an improvement in operating safety margins. As a Result of advanced fuel design for CANFLEX fuel bundles, components consisting of fuel bundles are more complicated. Hence, the detailed modeling of components is inevitable in order to analyze the fuel performance by computational fluid dynamics. In this report, the basic design of the advanced fuel for CANDU reactors was carried out and the methodology for the modeling of fuel bundle were described. Firstly, the components consisting of fuel bundles were separately modeled and saved with different file names. The final feature of fuel bundle was accomplished by an assembling process of components. Since this report developed the modeling methodology based on the Unigraphics program, the basic explanations for the software were given first, and the complete modeling of 37-elements and CANFLEX fuel bundles were provided. The components of CANFLEX fuel bundles were also compared with that of 37-elements fuel bundles. Although, in this report, the modeling methodology is applied only to 37-elements and CANFLEX fuel bundles, this methodology may be applicable to the newly designed fuel bundles which are to be developed in the future

  16. Proceedings of the international conference on CANDU fuel

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  20. CANDU fuel quality and how it is achieved

    International Nuclear Information System (INIS)

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

  1. Candu advanced fuel cycles: key to energy sustainability

    International Nuclear Information System (INIS)

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

  2. Economic analysis of alternative options in CANDU fuel cycle

    International Nuclear Information System (INIS)

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

  3. CANDU type fuel behavior evaluation - a probabilistic approach

    International Nuclear Information System (INIS)

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

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

    Institute of Scientific and Technical Information of China (English)

    HUO Xiao-Dong; XIE Zhong-Sheng

    2004-01-01

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

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

    International Nuclear Information System (INIS)

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

  6. Thermalhydraulics of advanced 37-element fuel bundle in crept pressure tubes

    Directory of Open Access Journals (Sweden)

    Park Joo Hwan

    2016-01-01

    Full Text Available A CANDU-6 reactor, which has 380 fuel channels of a pressure tube type, is suffering from aging or creep of the pressure tubes. Most of the aging effects for the CANDU primary heat transport system were originated from the horizontal crept pressure tubes. As the operating years of a CANDU reactor proceed, a pressure tube experiences high neutron irradiation damage under high temperature and pressure. The crept pressure tube can deteriorate the Critical Heat Flux (CHF of a fuel channel and finally worsen the reactor operating performance and thermal margin. Recently, the modification of the central subchannel area with increasing inner pitch length of a standard 37-element fuel bundle was proposed and studied in terms of the dryout power enhancement for the uncrept pressure tube since a standard 37-element fuel bundle has a relatively small flow area and high flow resistance at the central region. This study introduced a subchannel analysis for the crept pressure tubes loaded with the inner pitch length modification of a standard 37-element fuel bundle. In addition, the subchannel characteristics were investigated according to the flow area change of the center subchannels for the crept pressure tubes. Also, it was discussed how much the crept pressure tubes affected the thermalhydraulic characteristics of the fuel channel as well as the dryout power for the modification of a standard 37-element fuel bundle.

  7. Input modelling of ASSERT-PV V2R8M1 for RUFIC fuel bundle

    International Nuclear Information System (INIS)

    This report describes the input modelling for subchannel analysis of CANFLEX-RU (RUFIC) fuel bundle which has been developed for an advanced fuel bundle of CANDU-6 reactor, using ASSERT-PV V2R8M1 code. Execution file of ASSERT-PV V2R8M1 code was recently transferred from AECL under JRDC agreement between KAERI and AECL. SSERT-PV V2R8M1 which is quite different from COBRA-IV-i code has been developed for thermalhydraulic analysis of CANDU-6 fuel channel by subchannel analysis method and updated so that 43-element CANDU fuel geometry can be applied. Hence, ASSERT code can be applied to the subchannel analysis of RUFIC fuel bundle. The present report was prepared for ASSERT input modelling of RUFIC fuel bundle. Since the ASSERT results highly depend on user's input modelling, the calculation results may be quite different among the user's input models. The objective of the present report is the preparation of detail description of the background information for input data and gives credibility of the calculation results

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

  9. Validation of the ASSERT subchannel code: Prediction of critical heat flux in standard and nonstandard CANDU bundle geometries

    International Nuclear Information System (INIS)

    The ASSERT code has been developed to address the three-dimensional computation of flow and phase distribution and fuel element surface temperatures within the horizontal subchannels of Canada uranium deuterium (CANDU) pressurized heavy water reactor fuel channels and to provide a detailed prediction of critical heat flux (CHF) distribution throughout the bundle. The ASSERT subchannel code has been validated extensively against a wide repertoire of experiments; its combination of three-dimensional prediction of local flow conditions with a comprehensive method of predicting CHF at these local conditions makes it a unique tool for predicting CHF for situations outside the existing experimental database. In particular, ASSERT is an appropriate tool to systematically investigate CHF under conditions of local geometric variations, such as pressure tube creep and fuel element strain. The numerical methodology used in ASSERT, the constitutive relationships incorporated, and the CHF assessment methodology are discussed. The evolutionary validation plan is also discussed and early validation exercises are summarized. More recent validation exercises in standard and nonstandard geometries are emphasized

  10. Validation of the assert subchannel code: Prediction of CHF in standard and non-standard Candu bundle geometries

    International Nuclear Information System (INIS)

    The ASSERT code has been developed to address the three-dimensional computation of flow and phase distribution and fuel element surface temperatures within the horizontal subchannels of CANDU PHWR fuel channels, and to provide a detailed prediction of critical heat flux (CHF) distribution throughout the bundle. The ASSERT subchannel code has been validated extensively against a wide repertoire of experiments; its combination of three-dimensional prediction of local flow conditions with a comprehensive method of prediting CHF at these local conditions, makes it a unique tool for predicting CHF for situations outside the existing experimental data base. In particular, ASSERT is an appropriate tool to systematically investigate CHF under conditions of local geometric variations, such as pressure tube creep and fuel element strain. This paper discusses the numerical methodology used in ASSERT, the constitutive relationships incorporated, and the CHF assessment methodology. The evolutionary validation plan is discussed, and early validation exercises are summarized. The paper concentrates, however, on more recent validation exercises in standard and non-standard geometries

  11. Validation of the ASSERT subchannel code for prediction of CHF in standard and non-standard CANDU bundle geometries

    International Nuclear Information System (INIS)

    The ASSERT code has been developed to address the three-dimensional computation of flow and phase distribution and fuel element surface temperatures within the horizontal subchannels of CANDU PHWR fuel channels, and to provide a detailed prediction of critical heat flux distribution throughout the bundle. The ASSERT subchannel code has been validated extensively against a wide repertoire of experiments; its combination of three-dimensional prediction of local flow conditions with a comprehensive method of predicting critical heat flux (CHF) at these local conditions makes it a unique tool for predicting CHF for situations outside the existing experimental data base. In particular, ASSERT is the only tool available to systematically investigate CHF under conditions of local geometric variations, such as pressure tube creep and fuel element strain. This paper discusses the numerical methodology used in ASSERT, the constitutive relationships incorporated, and the CHF assessment methodology. The evolutionary validation plan is discussed, and early validation exercises are summarized. The paper concentrates, however, on more recent validation exercises in standard and non-standard geometries. 28 refs., 12 figs

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

    International Nuclear Information System (INIS)

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

  13. Third international conference on CANDU fuel

    International Nuclear Information System (INIS)

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

  14. Static stress analysis of CANFLEX fuel bundles

    International Nuclear Information System (INIS)

    The static stress analysis of CANFLEX bundles is performed to evaluate the fuel structural integrity during the refuelling service. The structure analysis is carried out by predicting the drag force, stress and displacements of the fuel bundle. By the comparison of strength tests and analysis results, the displacement values are well agreed within 15%. The analysis shows that the CANFLEX fuel bundle keep its structural integrity. 24 figs., 6 tabs., 12 refs. (Author) .new

  15. Hydraulic characteristics of HANARO fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Cho, S.; Chung, H. J.; Chun, S. Y.; Yang, S. K.; Chung, M. K. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    This paper presents the hydraulic characteristics measured by using LDV (Laser Doppler Velocimetry) in subchannels of HANARO, KAERI research reactor, fuel bundle. The fuel bundle consists of 18 axially finned rods with 3 spacer grids, which are arranged in cylindrical configuration. The effects of the spacer grids on the turbulent flow were investigated by the experimental results. Pressure drops for each component of the fuel bundle were measured, and the friction factors of fuel bundle and loss coefficients for the spacer grids were estimated from the measured pressure drops. Implications regarding the turbulent thermal mixing were discussed. Vibration test results measured by using laser vibrometer were presented. 9 refs., 12 figs. (Author)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  2. Safe, permanent disposal of used CANDU fuel

    International Nuclear Information System (INIS)

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

  3. Procurement and supply of CANDU fuels

    International Nuclear Information System (INIS)

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

  4. A model for fission product distribution in CANDU fuel

    International Nuclear Information System (INIS)

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

  5. Development of a CANDU fuel channel model to assess the effect of a pressure tube creep on the safety related parameters

    International Nuclear Information System (INIS)

    Recently the effect of pressure tube creep on the reactor safety in CANDUs emerges as an important issue of safety analysis due to a need for an extended operation. The accident analysis for the aged plants needs to incorporate major degradations of the plant performance in the safety analysis. In this paper, a CATHENA fuel channel model for studying the effects of the vertical offset of the fuel bundles in a crept pressure tube on the fuel and pressure tube cooling is developed. The current practice of the CANDU safety analysis assumes that the fuel bundles stay in a manner concentric to the pressure tube centerline even in the crept pressure tubes, whereas in reality the bundles sit at the bottom of the pressure tube. With this point in mind, 37-pin models with and without vertical offset of the bundle in the crept fuel channel are developed and tested for Reactor Outlet Header (ROH) 100% break LOCA accident, and results compared. As a result, it was found that the difference between the uncrept fuel channel model and the two crept fuel channel models, a concentric one and another vertically offset one, is quite significant, whereas the difference between the two crept fuel channel models is insignificant. Therefore it is concluded that the use of the concentric crept fuel channel model for the aged CANDU-6 safety analysis is justifiable for the first 200 sec into an accident. (author)

  6. Quality control of CANDU6 fuel element in fabrication process

    International Nuclear Information System (INIS)

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

  7. In-pile test of Qinshan PWR fuel bundle

    International Nuclear Information System (INIS)

    In-pile test of Qinshan Nuclear Power Plant PWR fuel bundle has been conducted in HWRR HTHP Test loop at CIAE. The test fuel bundle was irradiated to an average burnup of 25000 Mwd/tU. The authors describe the structure of (3 x 3-2) test fuel bundle, structure of irradiation rig, fuel fabrication, irradiation conditions, power and fuel burnup. Some comments on the in-pile performance for fuel bundle, fuel rod and irradiation rig were made

  8. TRIGA spent fuel bundles safe storage

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  11. Neutronic calculations regarding the new LEU 6 x 6 fuel bundle for 14 MW TRIGA - SSR, in order to increase the reactor power up to 21 MW

    Energy Technology Data Exchange (ETDEWEB)

    Iorgulis, C.; Ciocanescu, M.; Preda, M.; Mladin, M. [Institute of Nuclear Research, Pitesti (Romania)

    1998-07-01

    In order to meet the increasing demands of terminal flux for the experimental devices which will be loaded with CANDU natural uranium pins (or clusters), is necessary to rise the reactor power up to 21 MW. In this respect we consider in our evaluations a new 6x6 TRIGA fuel bundle geometry (the actual fuel bundle contains 5x5 pins). This paper will contain a comparative analysis regarding: flux and power distribution across the 29 fuel bundles standard core, and managements patters, in order to maximize the discharge fuel burnup and core lifetime. (author)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  3. 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)

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

    Holmes, Bradford

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  8. LVRF fuel bundle manufacture for Bruce

    Energy Technology Data Exchange (ETDEWEB)

    Pant, A. [Zircatec Precision Industries, Port Hope, Ontario (Canada)

    2005-12-15

    In response to the Power Uprate program at Bruce Power, Zircatec has committed to introduce, by Spring 2006 a new manufacturing line for the production of 43 element Bruce LVRF bundles containing Slightly Enriched Uranium (SEU) with a centre pin of blended dysprosia/urania (BDU). This is a new fuel design and is the first change in fuel design since the introduction of the current 37-element fuel over 20 years ago. Introduction of this new line has involved the introduction of significant changes to an environment that is not used to rapid changes with significant impact. At ZPI we have been able to build on our innovative capabilities in new fuel manufacturing, the strength and experience of our core team, and on our prevailing management philosophy of 'support the doer'. The presentation will discuss some of the novel aspects of this fuel introduction and the mix of innovative and classical project management methods that are being used to ensure that project deliveries are being met. Supporting presentations will highlight some of the issues in more detail. (author)

  9. Thorium utilization in Candu reactors

    International Nuclear Information System (INIS)

    In this study, means of thorium utilization in a CANDU reactor are considered. A once through thorium-DUPIC cycle is analyzed in detail. CANDU has the best neutron economy among the commercially available power reactors, which makes it suitable for many different fuel cycle options. A review of the available fuel cycles is also done in the scope of this study to select an economically viable cycle which does not impose profound changes in the neutronic properties of the core that require remodeling of core and related systems. To create a good model ot the CANDU core for the necessary calculations, the steady state properties of CANDU reactor are analyzed. It is assumed that approximation ot refueling as moving the bundles at a constant velocity is valid. This approximation leads to a corollary; The average cross sections of two adjacent bidirectionally refueled channels are independent of axial location. This is also veritied. A result of this corollary the CANDU core can be modeled only in radial direction in cylindirical geometry. The steady state CANDU core model is prepared using the actual power values and these values are sought in the results. The control systems which effect the neutron flux shape are introduced into the model later in the form of additional absorption cross section and lower diffusion coefficient. The results are in good agreement with the actual values. Several different thorium-DUPIC fuel bundle configurations are considered and the one with 12 Th02 elements in the third ring is found to have similar burnup dependent cross-sections and location infinite multiplication factors. Using the model created, the bundle is tested also in the tull core model and it is tound that this bundle configuration complies with the current refueling scheme. That is, no changes are necessary in the refuelind rate or the control systems. A higher conversion ratio of 0.82 is attained, while the excess reactivity of the core is found to decrease by 0.01 Ak

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

    International Nuclear Information System (INIS)

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

  11. Status of the demonstration irradiation program of the new fuel bundle CANFLEX-NU in Korea

    International Nuclear Information System (INIS)

    In the late part of 1999, the Korea Electric Power Corporation has initiated a program CANFLEX-NU (Natural Uranium) fuel in the Wolsong Generating Station (WGS) - no.1 which has been operating since 1983, because the CANFLEX could be used to recover some of a CANDU heat transport system operation margins that had decreased due to The Korea Ministry of Science and Technology (MOST) has recognized the successful demonstration irradiation of 24 CANFLEX bundles at the Pt. Lepreau Generating Station in Canada, as final verification of the CANFLEX design in preparation for full core conversion. Therefore, MOST has pushed and gave a financial support to a KEPRI/KAERI Joint Industrialization Program of CANFLEX-NU Fuel, which will be for 3 years from 2000 November, to validate CANFLEX-NU fuel bundle performance in direct conditions of relevance under the Korean licensing requirements as well as to evaluate the fuel fabrication capability, and to produce a safety analysis report for the full-core implementation. The economic benefits of CANFLEX-NU fuel are directly dependent on the thermalhydraulic performance. Switching from the existing 37-element fuel to the CANFLEX fuel will be largely driven by the economic benefits to be realized. Showing a positive result in the economic evaluation as well as successfully demonstrating the CANFLEX fuel irradiation in WGS-no. 1, the full-core implementation of the fuel at the WGS-no.1 in Korea will proceed by starting the licensing process at around 2003 April because the safety report for the full-core conversion will be ready by 2003 March. This paper describes the status of CANFLEX-NU fuel industrialization program in Korea, as well as the fuel design features. It summarizes the plan of CANFLEX-NU fuel demonstration irradiation at the WGS-no. 1 in Korea and the status of documentation for the demonstration irradiation as well as for the CANFLEX-NU full-core implementation. (author)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  16. Calculation of power coefficient in CANFLEX-NU fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Min, Byung Joo; Jun, Ji Su; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-11-01

    Changes in power level affect reactivity due to its dependence on fuel and coolant temperatures. The power coefficient of reactivity is related to the fuel temperature coefficient through the change in fuel temperature per percent change in power. In addition, power level changes are followed by changes in coolant temperature and density which contribute to the reactivity effect. In this report, the power coefficient of CANFLEX-NU was calculated and the result would be compared with that of CANDU-6 reactor which is operating. 8 refs., 43 figs., 2 tabs. (Author)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

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

    International Nuclear Information System (INIS)

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

  20. 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).

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  3. Thermal Analysis of CANDU Spent Fuel Bay Cooling System

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  11. Evolution of procurement and supply conditions for CANDU fuels

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  13. Chop-leach fuel bundle residues densification by melting

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, R.G.; Griggs, B.

    1976-11-01

    Two melting processes were investigated for the densification of fuel bundle residues: Industoslag melting and graphite crucible melting. The Industoslag process, with prior decontamination and sorting, can produce ingots of Zircaloy, stainless steel and Inconel of a quality suitable for refabrication and reuse. The process can also melt oxidized mixtures of fuel bundle residues for direct storage. Eutectic mixtures of these materials can be melted in graphite at temperatures of 1300/sup 0/C. Hydrogen absorption experiments with the zirconium-rich alloys show the alloys to be potential tritium reservoirs. 13 figures.

  14. Interconnection of bundled solid oxide fuel cells

    Science.gov (United States)

    Brown, Michael; Bessette, II, Norman F; Litka, Anthony F; Schmidt, Douglas S

    2014-01-14

    A system and method for electrically interconnecting a plurality of fuel cells to provide dense packing of the fuel cells. Each one of the plurality of fuel cells has a plurality of discrete electrical connection points along an outer surface. Electrical connections are made directly between the discrete electrical connection points of adjacent fuel cells so that the fuel cells can be packed more densely. Fuel cells have at least one outer electrode and at least one discrete interconnection to an inner electrode, wherein the outer electrode is one of a cathode and and anode and wherein the inner electrode is the other of the cathode and the anode. In tubular solid oxide fuel cells the discrete electrical connection points are spaced along the length of the fuel cell.

  15. Development of a best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B.D.; Lee, W.J.; Lim, H.S. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-03-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool , i.e. RELAP5/MOD3. This scope of project is a third step of the whole project, and expand the RELAP5/MOD3/CANDU version for implementation of LOCA Analysis. There are three main area of model development, i.e. Moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs. Newly developed version, namely RELAP5/MOD3/CANDU+ is applicable to CANDU plant analysis with keeping the function of light water reactor analysis. The limited validations of model installation were performed. Assessment of CHF model using AECL separated effect test and calculation for Wolsong 2 plant were performed also for the applicability test of the developed version. 15 refs., 37 figs., 8 tabs. (Author)

  16. Development of best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B. D.; Lee, W. J.; Lim, H. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2000-03-15

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model if existing PWR auditing tool, i.e. RELAP5/MOD3. This scope of project is a third step of the whole project, and expand the RELAP5/MOD3/CANDU version for implementation of LOCA analysis. There are three main area of model development, i.e. moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs. Newly developed version, namely RELAP5/MOD3/CANDU+ is applicable to CANDU plant analysis with keeping the function of light water reactor analysis. The limited validations of model installation were performed. Assessment of CHF model using AECL separated effect test and calculation for Wolsong 2 plant were performed also for the applicability test of the developed version.

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-07-01

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

  19. The 24 CANFLEX-NU bundle demonstration irradiation at Wolsong-1 generating station-bundle manufacture and QA, fuel handling aspects, flasking and shipping and pie for the irradiated fuel, and follow-up documentation

    International Nuclear Information System (INIS)

    Korea Ministry of Science and Technology(MOST) has pushed and given a financial support to a KEPRI/KAERI Joint Industrialization Program of CANFLEX-NU Fuel as one of Korea's National Nuclear Mid- and Long Term R and D Program. The Industrialization Program will be conducted for 3 years from 2000 November to efficiently utilize the CANFLEX fuel technology developed by KAERI and AECL jointly, where the KAERI's works have been conducted under the Korea's national program of the mid- and long-term nuclear R and D programs since 1992. This document is a report to guideline the following activities on the safety assessment for the 24 CANFLEX-NU (CANDU Flexible fuelling-Natural Uranium) fuel bundle demonstration irradiation at Wolsong-1 Generating Station: 'bundle manufacture and QA', 'Fuel handling aspects such as loading fuel, de-fuelling and segregation, and visual in-bay examinations', 'Flasking and shipping', 'Post-irradiation examination', and 'Follow-up documentation to be produced'

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  3. The effect of radial power profile of DUPIC bundle on CHF

    International Nuclear Information System (INIS)

    The axial and ring power profiles of DUPIC bundle are much different from those of reference 37-element fuel bundle since a DUPIC fuel bundle is re-fabricated using spent PWR fuel and 2-bundle shift refuelling scheme is proposed to CANDU-6 reactor. In case that the ring power profile of a fuel bundle is altered, the flow and enthalpy distribution of subchannels and the radial position of CHF occurrence will be changed. Similarly, the axial power profile of a fuel channel affects CHF and axial position of CHF occurrence as well as axial enthalpy, quality and pressure distribution. The ring power profile of the DUPIC bundle as increasing burnup is altered and flattened compared to 37-element bundle and each fuel bundle in a fuel channel has a different ring power profile from the other bundles at different axial position in the same fuel channel. Therefore, how to consider the burnup or ring power effect on CHF is very important to DUPIC thermalhydraulic analysis. At present study, thermalhydraulic analysis of the DUPIC bundle was performed in consideration of ring power profile effect on CHF. The subchannel enthalpy, mass flux and CHF distribution for 0 burnup to discharged burnup (18,000 MWD/THM) of DUPIC bundle were evaluated using ASSERT subchannel code. The results were compared to those of 37-element bundle and the compatability of DUPIC bundle with an existing CANDU-6 was presented in a CHF point of view

  4. Coupling Systems of Five CARA Fuel Bundles for Atucha I

    International Nuclear Information System (INIS)

    This paper describe the mechanical design of two options for the coupling systems of five CARA fuel bundles, to be used in the Atucha I nuclear power plant. These systems will be hydraulic tested in a low pressure loop to know their hydraulic loss of pressure

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

    International Nuclear Information System (INIS)

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

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

    Directory of Open Access Journals (Sweden)

    Bergelson Boris R.

    2008-01-01

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

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

    International Nuclear Information System (INIS)

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

  8. CAT reconstruction and potting comparison of a LMFBR fuel bundle

    International Nuclear Information System (INIS)

    A standard Liquid Metal Fast Breeder Reactor (LMFBR) subassembly used in the Experimental Breeder Reactor II (EBR-II) was investigated, by remote techniques, for fuel bundle distortion by both nondestructive and destructive methods, and the results from both methods were compared. The non-destructive method employed neutron tomography to reconstruct the locations of fuel elements through the use of a maximum entropy reconstruction algorithm known as MENT. The destructive method consisted of ''potting'' (a technique that embeds and permanently fixes the fuel elements in a solid matrix) the subassembly, and then cutting and polishing the individual sections. The comparison indicated that the tomography reconstruction provided good results in describing the bundle geometry and spacer-wire locations, with the overall resolution being on the order of a spacer-wire diameter. A dimensional consistency check indicated that the element and spacer-wire dimensions were accurately reproduced in the reconstruction

  9. Monte Carlo Few-Group Constant Generation for CANDU 6 Core Analysis

    Directory of Open Access Journals (Sweden)

    Seung Yeol Yoo

    2015-01-01

    Full Text Available The current neutronics design methodology of CANDU-PHWRs based on the two-step calculations requires determining not only homogenized two-group constants for ordinary fuel bundle lattice cells by the WIMS-AECL lattice cell code but also incremental two-group constants arising from the penetration of control devices into the fuel bundle cells by a supercell analysis code like MULTICELL or DRAGON. As an alternative way to generate the two-group constants necessary for the CANDU-PHWR core analysis, this paper proposes utilizing a B1 theory augmented Monte Carlo (MC few-group constant generation method (B1 MC method which has been devised for the PWR fuel assembly analysis method. To examine the applicability of the B1 MC method for the CANDU 6 core analysis, the fuel bundle cell and supercell calculations are performed using it to obtain the two-group constants. By showing that the two-group constants from the B1 MC method agree well with those from WIMS-AECL and that core neutronics calculations for hypothetical CANDU 6 cores by a deterministic diffusion theory code SCAN with B1 MC method generated two-group constants also agree well with whole core MC analyses, it is concluded that the B1 MC method is well qualified for both fuel bundle cell and supercell analyses.

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

    International Nuclear Information System (INIS)

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

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

    Indian Academy of Sciences (India)

    B R Bergelson; A S Gerasimov; G V Tikhomirov

    2007-02-01

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

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

    Directory of Open Access Journals (Sweden)

    Bergelson Boris R.

    2008-01-01

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

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

    International Nuclear Information System (INIS)

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

  14. Scenarios for the transmutation of actinides in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-12-15

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-06-15

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

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

    International Nuclear Information System (INIS)

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

  19. Cost and schedule reduction for next-generation Candu

    International Nuclear Information System (INIS)

    AECL has developed a suite of technologies for CanduR 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. Investigation of the CANLUB/sheath interface in CANDU fuel at extended burnup by XPS and SEM/WDX

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  2. The button effect of CANFLEX bundle on the critical heat flux and critical channel power

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Jun, Jisu; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Dimmick, G. R.; Bullock, D. E.; Inch, W. [Atomic Energy of Canada Limited, Ontario (Canada)

    1997-12-31

    A CANFLEX (CANdu FLEXible fuelling) 43-element bundle has developed for a CANDU-6 reactor as an alternative of 37-element fuel bundle. The design has two diameter elements (11.5 and 13.5 mm) to reduce maximum element power rating and buttons to enhance the critical heat flux (CHF), compared with the standard 37-element bundle. The freon CHF experiments have performed for two series of CANFLEX bundles with and without buttons with a modelling fluid as refrigerant R-134a and axial uniform heat flux condition. Evaluating the effects of buttons of CANFLEX bundle on CHF and Critical Channel Power (CCP) with the experimental results, it is shown that the buttons enhance CCP as well as CHF. All the CHF`s for both the CANFLEX bundles are occurred at the end of fuel channel with the high dryout quality conditions. The CHF enhancement ratio are increased with increase of dryout quality for all flow conditions and also with increase of mass flux only for high pressure conditions. It indicates that the button is a useful design for CANDU operating condition because most CHF flow conditions for CANDU fuel bundle are ranged to high dryout quality conditions. 5 refs., 11 figs. (Author)

  3. CANDU 9 design

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  5. Incentives for improvement of CANDU

    International Nuclear Information System (INIS)

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

  6. Joint studies on large CANDU

    International Nuclear Information System (INIS)

    CANDU PHWRs have demonstrated generic benefits which will be continued in future designs. These include economic benefits due to low operating costs, business potential, strategic benefits due to fuel cycle flexibility and operational benefits. These benefits have been realized in Korea through the operation of Wolsong 1, resulting in further construction of PHWRs at the same site. The principal benefit, low electricity cost, is due to the high capacity factor and the low fuel cost for CANDU. The CANDU plant at Wolsong has proven to be a safe, reliable and economical electricity producer. The ability of PHWR to burn natural uranium ensures security of fuel supply. Following successful Technology Transfer via the Wolsong 2,3 and 4 project, future opportunity exists between Korea and Canada for continuing co-operation in research and development to improve the technology base, for product development partnerships, and business opportunities in marketing and building PHWR plants in third countries. High reliability, through excellent design, well-controlled operation, efficient maintenance and low operating costs is critical to the economic viability of nuclear plants. CANDU plants have an excellent performance record. The four operating CANDU 6 plants, operated by four utilities in three countries, are world performance leaders. The CANDU 9 design, with higher output capacity, will help to achieve better site utilization and lower electricity costs. Being an evolutionary design, CANDU 9 assures high performance by utilizing proven systems, and component designs adapted from operating CANDU plants (Bruce B, Darlington and CANDU 6). All system and operating parameters are within the operating proven range of current plants. KAERI and AECL have an agreement to perform joint studies on future PHWR development. The objective of the joint studies is to establish the requirements for the design of future advanced CANDU PHWR including the utility need for design improvements

  7. Studies of a larger fuel bundle for the ABWR improved evolutionary reactor

    International Nuclear Information System (INIS)

    Studies for an Improved Evolutionary Reactor (IER) based on the Advanced Boiling Water Reactor (ABWR) were initiated in 1990. The author summarizes the current status of the core and fuel design. A core and fuel design based on a BWR K-lattice fuel bundle with a pitch larger than the conventional BWR fuel bundle pitch is under investigation. The core and fuel design has potential for improved core design flexibility and improved reactor transient response. Furthermore, the large fuel bundle, coupled with a functional control rod layout, can achieve improvement of operation and maintenance, as well as improvement of overall plant economy

  8. LVRF fuel bundle manufacture for Bruce - project update

    Energy Technology Data Exchange (ETDEWEB)

    Pant, A. [Zircatec Precision Industries, Port Hope, Ontario (Canada)

    2005-07-01

    In response to the Power Uprate program at Bruce Power, Zircatec has committed to introduce, by Spring 2006 a new manufacturing line for the production of 43 element Bruce LVRF bundles containing Slightly Enriched Uranium (SEU) with a centre pin of blended dysprosia/urania (BDU). This is a new fuel design and is the first change in fuel design since the introduction of the current 37 element fuel over 20 years ago. Introduction of this new line has involved the introduction of significant changes to an environment that is not used to rapid changes with significant impact. At ZPI we have been able to build on our innovative capabilities in new fuel manufacturing, the strength and experience of our core team, and on our prevailing management philosophy of 'support the doer'. The presentation will discuss some of the novel aspects of this fuel introduction and the mix of innovative and classical project management methods that are being used to ensure that project deliverables are being met. Supporting presentations will highlight some of the issues in more detail. (author)

  9. LVRF fuel bundle manufacture for Bruce - project update

    International Nuclear Information System (INIS)

    In response to the Power Uprate program at Bruce Power, Zircatec has committed to introduce, by Spring 2006 a new manufacturing line for the production of 43 element Bruce LVRF bundles containing Slightly Enriched Uranium (SEU) with a centre pin of blended dysprosia/urania (BDU). This is a new fuel design and is the first change in fuel design since the introduction of the current 37 element fuel over 20 years ago. Introduction of this new line has involved the introduction of significant changes to an environment that is not used to rapid changes with significant impact. At ZPI we have been able to build on our innovative capabilities in new fuel manufacturing, the strength and experience of our core team, and on our prevailing management philosophy of 'support the doer'. The presentation will discuss some of the novel aspects of this fuel introduction and the mix of innovative and classical project management methods that are being used to ensure that project deliverables are being met. Supporting presentations will highlight some of the issues in more detail. (author)

  10. CFD simulations of the single-phase and two-phase coolant flow of water inside the original and modified CANDU 37-element bundles

    International Nuclear Information System (INIS)

    Single-phase (inlet temperature of 180° C, outlet pressure of 9 MPa, total power of 2 MW and flow rate of 13.5 Kg/s), and two-phase (inlet temperature of 265° C, outlet pressure of 10 MPa, total power of 7.126 MW and flow rate of 19 Kg/s) water flows inside a CANDU thirty seven element fuel string are simulated using a Computational Fluid Dynamics (CFD) code with parallel processing and results are presented in this paper. The analyses have been performed for the original and modified (11.5 mm center element diameter) designs with skewed cosine axial heat flux distribution and 5.1% diametral creep of the pressure tube. The CFD results are in good agreement with the expected temperature and velocity cross-sectional distributions. The effect of the pressure tube creep on the flow bypass is detected, and the CHF improvement in the core region of the modified design is confirmed. The two-phase flow model reasonably predicted the void distribution and the role of interfacial drag on increasing the pressure drop. In all CFD models, the appendages were shown to enhance the production of cross flows and their corresponding flow mixing and asymmetry. (author)

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

    International Nuclear Information System (INIS)

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

  12. An integrated CANDU system

    International Nuclear Information System (INIS)

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

  13. Laser dismantling of PHWR spent fuel bundles and decladding of fuel pins in the highly radioactive hot cells

    International Nuclear Information System (INIS)

    Full text: For reprocessing of PHWR fuel, fuel bundles are at present chopped mechanically into small pieces of pins using high tonnage mechanical press before dissolution. The existing method of bundle dismantling is purely mechanical using very high force for chopping. A laser based automated bundle dismantling system is developed. In the system, end-plates of bundle, which holds the fuel pins together, are cut using Nd-YAG laser to separate the bundles into pins. In addition to pin separation, the pins are to be chopped into small pieces using a small mechanical chopper. Since the spent fuel is highly radioactive, all these operations are performed remotely in hot cells. Post irradiation examination also requires dismantling of bundles into pins so that they can select the pins for the further examinations. In both these applications laser dismantling remains the most. important step and this system has been developed and tested. This paper describes the experience gained during the development efforts

  14. Enhancing the seismic capability of the on-power refueling system of the CANDU reactor

    International Nuclear Information System (INIS)

    The CANDU reactor assembly includes several hundred horizontal fuel channels, each containing twelve fuel bundles, arranged in a square lattice, and supported by the reactor structures. CANDU operates on natural uranium or other low fissile content fuel, and is refueled on-power, with either four or eight fuel bundles in a channel being replaced during each refueling operation. The fueling machines clamp onto the opposite ends of the fuel channel to be refueled. The seismic capacity of this refueling system is evaluated in terms of its dynamic response during an earthquake. This paper describes the approach adopted to enhance the seismic capability of the fueling machine and calandria assembly for earthquakes of O.3g ground acceleration covering a broad range of soil conditions ranging from soft to hard. A detailed, 3-D finite element seismic model of the fueling machine and calandria assembly system is developed to calculate the seismic responses of the structure. Some relatively simple hardware design changes have been considered to increase the seismic capacity of the CANDU 6 reactor. These changes in the fueling machine and calandria assembly of the CANDU 6 reactor are briefly described. They have been incorporated into the finite element seismic model of the system. Most of these design changes have already been considered and implemented in other CANDU reactor projects. The current CANDU 6 reactor design fully meets the requirements of seismic qualification for sites with potential for O.2g ground acceleration where the seismic loads need to be combined with the other design loads for the support and pressure boundary components to demonstrate compliance with the applicable Code requirements. In the present study it is demonstrated that, with relatively simple hardware changes, the fueling machine and calandria assembly of the CANDU 6 reactor can withstand earthquakes of O.3g ground acceleration. Based on the current study and some preliminary analysis of the

  15. Cap assembly for a bundled tube fuel injector

    Energy Technology Data Exchange (ETDEWEB)

    LeBegue, Jeffrey Scott; Melton, Patrick Benedict; Westmoreland, III, James Harold; Flanagan, James Scott

    2016-04-26

    A cap assembly for a bundled tube fuel injector includes an impingement plate and an aft plate that is disposed downstream from the impingement plate. The aft plate includes a forward side that is axially separated from an aft side. A tube passage extends through the impingement plate and the aft plate. A tube sleeve extends through the impingement plate within the tube passage towards the aft plate. The tube sleeve includes a flange at a forward end and an aft end that is axially separated from the forward end. A retention plate is positioned upstream from the impingement plate. A spring is disposed between the retention plate and the flange. The spring provides a force so as to maintain contact between at least a portion of the aft end of the tube sleeve and the forward side of the aft plate.

  16. MENT reconstruction and potting comparison of a LMFBR fuel bundle

    International Nuclear Information System (INIS)

    Since the advent of computer-assisted-tomography (CAT), the CAT techniques have been rapidly expanded to the nuclear industry. A number of investigators have applied these techniques to reconstruct the fuel bundle configuration inside a subassembly with various degrees of resolution; however, there has been little data available on the accuracy of these reconstructions, and no comparisons have been made with the internal structure of actual irradiated subassemblies. Some efforts have utilized pretest mock-ups to calibrate the CAT algorithms, but the resulting mock-up configurations do not necessarily represent an actual subassembly, so an exact comparison has been lacking. The purpose of this paper is to present the results of a comparison between a CAT reconstruction of an irradiated subassembly and the destructive examination of the same subassembly

  17. Ultrasonic imaging methods for quality evaluation of the CANDU fuel bundle welds

    International Nuclear Information System (INIS)

    For more than 20 years, the quality control of the end-cap/end-plates welds and of the brazed appendage joints is made by destructive methods (metallographic examinations or mechanical tests) on specimens sampled from production. Having a very limited statistics, these destructive methods are useful only to indicate 'trends' of the production quality, not for detecting infrequent single defect events. It is recognized that nondestructive examination techniques are required to achieve sufficient 'visibility' of the production quality, at a statistically significant sampling rate. For this reason, the INR-Ultra-acoustics R and D has developed the MICROSCAN equipment family for high resolution ultrasonic imaging, with performances close to the Acoustic Microscopy domain. The paper makes a presentation of the performances of the MICROSCAN 02/03 equipment for B and C-scan high-frequency ultrasonic imaging. Experimental results are presented and comparisons are make with metallographic examinations. (Author) 8 Figs., 4 Refs

  18. Specifications for reactor physics experiments on CANFLEX-RU fuel

    International Nuclear Information System (INIS)

    This is to describe reactor physics experiments to be performed in the ZED-2 reactor to study CANFLEX-RU fuel bundles in CANDU-type fuel channels. The experiments are to provide benchmark quality validation data for the computer codes and associated nuclear databases used for physics calculations, in particular WIMS-AECL. Such validation data is likely to be a requirement by the regulator as condition for licensing a CANDU reactor based on an enriched fuel cycle

  19. Moving towards sustainable thorium fuel cycles

    International Nuclear Information System (INIS)

    The CANDU 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 design. These features facilitate the introduction and full exploitation of thorium fuel cycles in CANDU reactors in an evolutionary fashion. Thoria (ThO2) based fuel offers both fuel performance and safety advantages over urania (UO2) based fuel, due its higher thermal conductivity which results in lower fuel-operating temperatures at similar linear element powers. Thoria fuel has demonstrated lower fission gas release than UO2 under similar operating powers during test irradiations. In addition, thoria has a higher melting point than urania and is far less reactive in hypothetical accident scenarios owing to the fact that it has only one oxidation state. This paper examines one possible strategy for the introduction of thorium fuel cycles into CANDU reactors. In the short term, the initial fissile material would be provided in a heterogeneous bundle of low-enriched uranium and thorium. The medium term scenario uses homogeneous Pu/Th bundles in the CANDU reactor, further increasing the energy derived from the thorium. In the long term, the full energy potential from thorium would be realized through the recycle of the U-233 in the used fuel. With U-233 recycle in CANDU reactors, plutonium would then only be required to top up the fissile content to achieve the desired burnup. (author)

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

    International Nuclear Information System (INIS)

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

  1. Automation in inspection of PHWR fuel elements & bundles at Nuclear Fuel Complex

    International Nuclear Information System (INIS)

    Nuclear Fuel Complex (NFC), Hyderabad, a constituent of Department of Atomic Energy, India manufactures fuel for all Indian nuclear power reactors. Currently NFC manufactures both 19 element & 37 element bundles for catering to the requirement of 220 MWe & 540 MWe PHWRs. In order to meet the growing needs for the Nuclear Fuel, NFC engaged in expansion of the production facilities. This calls for enhanced throughput at various inspection stages keeping in tandem with the production & for achieving this objective, NFC has chosen automation. This paper deals with automation of the inspection line at NFC. (author)

  2. Detector response in a CANDU low void reactivity core

    International Nuclear Information System (INIS)

    The response of the in-core flux detectors to the CANFLEX Low-Void-Reactivity Fuel (LVRF) [1] bundles for use in the CANDU reactor at Bruce nuclear generation station has been studied. The study was based on 2 detector types - platinum (Pt)-clad Inconel and pure Inconel detectors, and 2 fuel types - LVRF bundles and natural-uranium (NU) bundles. Both detectors show a decrease of thermal-neutron-flux to total-photon-flux ratio when NU fuel bundles are replaced by LVRF bundles in the reactor core (7% for Inconel and 9% for Pt-clad detectors). The ratio of the prompt component of the net electron current to the total net electron current (PFe) of the detectors however shows a different response. The use of LVRF bundles in place of NU fuel bundles in the reactor core did not change the PFe of the Pt-clad Inconel detector but increased the PFe of the pure Inconel detector by less than 2%. The study shows that the Inconel detector has a larger prompt-detector response than that of the platinum-clad detector; it reacts to the change of fluxes in the reactor core more readily. On the other hand, the Pt-clad detector is less sensitive to perturbations of the neutron-to-gamma ratio. Nevertheless the changes in an absolute sense are minimal; one does not anticipate a change of the flux-monitoring system if the NU fuel bundles are replaced with the CANFLEX LVRF bundles in the core of the Bruce nuclear generating station. (authors)

  3. RU fuel development program for an advanced fuel cycle in Korea

    International Nuclear Information System (INIS)

    Korea is a unique country, having both PWR and CANDU reactors. Korea can therefore exploit the natural synergism between the two reactor types to minimize overall waste production, and maximize energy derived from the fuel, by ultimately burning the spent fuel from its PWR reactors in CANDU reactors. As one of the possible fuel cycles, Recovered Uranium (RU) fuel offers a very attractive alternative to the use of Natural Uranium (NU) and slightly enriched uranium (SEU) in CANDU reactors. Potential benefits can be derived from a number of stages in the fuel cycle: no enrichment required, therefore no enrichment tails, direct conversion to UO2, lower sensitivity to 234U and 236U absorption in the CANDU reactor, and expected lower cost relative to NU and SEU. These benefits all fit well with the PWR-CANDU fuel cycle synergy. RU arising from the conventional reprocessing of European and Japanese oxide spent fuel by 2000 is projected to be approaching 25,000 te. The use of RU fuel in a CANDU 6 reactor should result in no serious radiological difficulties and no requirements for special precautions and should not require any new technologies for the fuel fabrication and handling. The use of the CANDU Flexible Fueling (CANFLEX) bundle as the carrier for RU will be fully compatible with the reactor design, current safety and operational requirements, and there will be improved fuel performance compared with the CANDU 37-element NU fuel bundle. Compared with the 37-element NU bundle, the RU fuel has significantly improved fuel cycle economics derived from increased burnups, a large reduction in both fuel requirements and spent fuel, arisings, and the potential lower cost for RU material. There is the potential for annual fuel cost savings in the range of one-third to two-thirds, with enhanced operating margins using RU in the CANFLEX bundle design. These benefits provide the rationale for justifying R and D efforts on the use of RU fuel for advanced fuel cycles in CANDU

  4. CANDU market prospects

    International Nuclear Information System (INIS)

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

  5. Direct reuse of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  6. A Model to Reproduce the Response of the Gaseous Fission Product Monitor (GFPM) in a CANDU{sup R} 6 Reactor (An Estimate of Tramp Uranium Mass in a Candu Core)

    Energy Technology Data Exchange (ETDEWEB)

    Mostofian, Sara; Boss, Charles [AECL Atomic Energy of Canada Limited, 2251 Speakman Drive, Mississauga Ontario L5K 1B2 (Canada)

    2008-07-01

    In a Canada Deuterium Uranium (Candu) reactor, the fuel bundles produce gaseous and volatile fission products that are contained within the fuel matrix and the welded zircaloy sheath. Sometimes a fuel sheath can develop a defect and release the fission products into the circulating coolant. To detect fuel defects, a Gaseous Fission Product Monitoring (GFPM) system is provided in Candu reactors. The (GFPM) is a gamma ray spectrometer that measures fission products in the coolant and alerts the operator to the presence of defected fuel through an increase in measured fission product concentration. A background fission product concentration in the coolant also arises from tramp uranium. The sources of the tramp uranium are small quantities of uranium contamination on the surfaces of fuel bundles and traces of uranium on the pressure tubes, arising from the rare defected fuel element that released uranium into the core. This paper presents a dynamic model that reproduces the behaviour of a GFPM in a Candu 6 plant. The model predicts the fission product concentrations in the coolant from the chronic concentration of tramp uranium on the inner surface of the pressure tubes (PT) and the surface of the fuel bundles (FB) taking into account the on-power refuelling system. (authors)

  7. Testing and implementation program for the modified Darlington 37-element fuel bundle

    International Nuclear Information System (INIS)

    To mitigate the effects of reactor ageing, a design modification to the 37-element fuel is proposed in which the diameter of the centre element will be reduced to 11.5 mm from 13.1 mm. The testing and implementation phase for the 37-element fuel bundle modification is discussed in this paper. The initial plan for testing is to perform a set of out-reactor tests to assess the endurance, acoustic response and cross-flow behaviour of the revised fuel bundle design. The initial schedule outlines activities that will enable OPG to implement full core fuelling of the modified bundle within the next three to four years. (author)

  8. Manufacturing of 37-element fuel bundles for PHWR 540 - new approach

    Energy Technology Data Exchange (ETDEWEB)

    Arora, U.K.; Sastry, V.S.; Banerjee, P.K.; Rao, G.V.S.H.; Jayaraj, R.N. [Nuclear Fuel Complex, Dept. Atomic Energy, Government of India, Hyderabad (India)

    2003-07-01

    Nuclear Fuel Complex (NFC), established in early seventies, is a major industrial unit of Department of Atomic Energy. NFC is responsible for the supply of fuel bundles to all the 220 MWe PHWRs presently in operation. For supplying fuel bundles for the forthcoming 540 MWe PHWRs, NEC is dovetailing 37-element fuel bundle manufacturing facilities in the existing plants. In tune with the philosophy of self-reliance, emphasis is given to technology upgradation, higher customer satisfaction and application of modern quality control techniques. With the experience gained over the years in manufacturing 19-element fuel bundles, NEC has introduced resistance welding of appendages on fuel tubes prior to loading of UO{sub 2} pellets, use of bio-degradable cleaning agents, simple diagnostic tools for checking the equipment condition, on line monitoring of variables, built-in process control methods and total productive maintenance concepts in the new manufacturing facility. Simple material handling systems have been contemplated for handling of the fuel bundles. This paper highlights the flow-sheet adopted for the process, design features of critical equipment and the methodology for fabricating the 37-element fuel bundles, 'RIGHT FIRST TIME'. (author)

  9. Manufacturing of 37-element fuel bundles for PHWR 540 - new approach

    International Nuclear Information System (INIS)

    Nuclear Fuel Complex (NFC), established in early seventies, is a major industrial unit of Department of Atomic Energy. NFC is responsible for the supply of fuel bundles to all the 220 MWe PHWRs presently in operation. For supplying fuel bundles for the forthcoming 540 MWe PHWRs, NEC is dovetailing 37-element fuel bundle manufacturing facilities in the existing plants. In tune with the philosophy of self-reliance, emphasis is given to technology upgradation, higher customer satisfaction and application of modern quality control techniques. With the experience gained over the years in manufacturing 19-element fuel bundles, NEC has introduced resistance welding of appendages on fuel tubes prior to loading of UO2 pellets, use of bio-degradable cleaning agents, simple diagnostic tools for checking the equipment condition, on line monitoring of variables, built-in process control methods and total productive maintenance concepts in the new manufacturing facility. Simple material handling systems have been contemplated for handling of the fuel bundles. This paper highlights the flow-sheet adopted for the process, design features of critical equipment and the methodology for fabricating the 37-element fuel bundles, 'RIGHT FIRST TIME'. (author)

  10. The behaviour of Phenix fuel pin bundle under irradiation

    International Nuclear Information System (INIS)

    An entire Phenix sub-assembly has been mounted and sectioned after irradiation. The examination of cross-sections revealed the effects of mechanical interaction in the bundle (ovalisations and contacts between clads). According to analysis of the sodium channels, cooling of the pin bundle remained uniform. (author)

  11. Design and fabrication of remote welding system for the fuel bundle assembly

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S.S.; Lee, J.W.; Park, G.I. [Korea Atomic Energy Research Inst., Daejeon (Korea, Republic of)

    2011-07-01

    Remote fuel bundle welding equipment in a hot-cell was designed and fabricated. To achieve this, a preliminary investigation of hands-on fuel fabrication outside a hot-cell was conducted with a consideration of the constraints caused by welding in a hot-cell. Some basic experiments were also carried out to improve the end-plate welding process for fuel bundle fabrication. The resistance welding equipment using end-plate welding was also improved. It was found that resistance welding was more suitable for joining an end-plate to end caps in a hot-cell. The optimum conditions for end-plate welding for remote operation were also obtained. Preliminary performances to improve the resistance welding process were also examined, and the resistance welding process was determined to be the best in the hot-cell environment for fuel bundle fabrication. The greatest advantage of fuel bundle welding equipment would be a commercialized welding process in which there is extensive production experience. This paper presents an outline of the developed welding equipment for a fuel bundle fabrication and reviews the conceptual design of remote welding equipment using a master-slave manipulator. The design of the remote welding equipment using the Pro-Engineer method was also reviewed. Furthermore the mechanical considerations and a mock-up simulation test were described. Finally, its performance test results were presented for a mock-up of remote fuel bundle welding equipment. (author)

  12. Assessment of CANDU reactor physics effects using a simplified whole-core MCNP model

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S

    2002-07-01

    A whole-core Monte Carlo n-particle (MCNP) model of a simplified CANDU reactor was developed and used to study core configurations and reactor physics phenomena of interest in CANDU safety analysis. The resulting reactivity data were compared with values derived from corresponding WIMS-AECL/RFSP, two-neutron-energy-group diffusion theory core simulations, thereby extending the range of CANDU-related code-to-code benchmark comparisons to include whole-core representations. These comparisons show a systematic discrepancy of about 6 mk between the respective absolute k{sub eff} values, but very good agreement to within about -0.15 {+-} 0.06 mk for the reactivity perturbation induced by G-core checkerboard coolant voiding. These findings are generally consistent with the results of much simpler uniform-lattice comparisons involving only WIMS-AECL and MCNP. In addition, MCNP fission-energy tallies were used to evaluate other core-wide properties, such as fuel bundle and total-channel power distributions, as well as intra-bundle details, such as outer-fuel-ring relative power densities and outer-ring fuel element azimuthal power variations, which cannot be determined directly from WIMS-AECL/RFSP core calculations. The average MCNP values for the ratio of outer fuel element to average fuel element power density agreed well with corresponding values derived from WIMS-AECL lattice-cell cases, showing a small systematic discrepancy of about 0.5 %, independent of fuel bum-up. For fuel bundles containing the highest-power fuel elements, the maximum peak-to-average outer-element azimuthal power variation was about 2.5% for cases where a statistically significant trend was observed, while much larger peak-to-average outer-element azimuthal power variations of up to around 42% were observed in low-power fuel bundles at the core/radial-neutron-reflector interface. (author)

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

    International Nuclear Information System (INIS)

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

  14. Development of a best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, M. K.; Lee, W. J. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-03-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool , i.e. RELAP5/MOD3. This scope of project is a fourth step of the whole project, applying the RELAP5/MOD3/CANDU+ version for the real CANDU plant LOCA Analysis and D2O leakage incident. There are three main models under investigation, i.e. Moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs, especially when CANDU LOCA is tested. Also, for Wolsung unit 1 D2O leakage incident analysis, the plant behavior is predicted with the newly developed version for the first 1000 seconds after onset of the incident, with the main interest aiming for system pressure, level control system, and thermal hydraulic transient behavior of the secondary system. The model applied for this particular application includes heat transfer model of nuclear fuel assembly, decay heat model, and MOV (Motor Operated Valve) model. Finally, the code maintenance work, mainly correcting the known errors, is presented. 12 refs., 26 figs., 3 tabs. (Author)

  15. Development of best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Hwnag, M

    2001-04-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool , i.e. RELAP5/MOD3. This scope of project is a fourth step of the whole project, applying the RELAP5/MOD3/CANDU+ version for the real CANDU plant LOCA Analysis and D2O leakage incident. There are three main models under investigation, i.e. Moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs, especially when CANDU LOCA is tested. Also, for Wolsung unit 1 D2O leakage incident analysis, the plant behavior is predicited with the newly developed version for the first 1000 seconds after onset of the incident, with the main interest aiming for system pressure, level control system, and thermal hydraulic transient behavior of the secondary system. The model applided for this particular application includes heat transfer model of nuclear fuel assembly, decay heat model, and MOV (Motor Operated Valve) model. Finally, the code maintenance work, mainly correcting the known errors, is presented.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  18. Parametric study of thermo-mechanical behaviour of 19-element PHWR fuel bundle having AHWR fuel material

    International Nuclear Information System (INIS)

    AHWR Th-LEU of 4.3 weight % 235U enrichment is a fuel design option for its trial irradiation in Indian PHWRs. The important component of this option is the large enhancement in the average discharge burn-up from the core. A parametric study of the 19-element fuel bundle, with natural uranium currently is being used in all operating 220 MWe PHWRs, has been carried out for AHWR Th-LEU fuel material by computer code FUDA MOD2. The important fuel parameters such as fuel temperature, fission gas release, fuel swelling and sheath strain have been analyzed for required fuel performance. With Th-LEU, average discharge burnups of about 25,000 MW-d/TeHE can be achieved. The FUDA code (Fuel Design Analysis code) MOD2 version has been used in the fuel element analysis. The code takes into account the inter-dependence of different parameters like fuel pellet temperatures, pellet expansions, fuel-sheath gap heat transfer, sheath strain and stresses, fission gas release and gas pressures, fuel densification etc. Thermo-mechanical analysis of fuel element having AHWR material is carried out for the bundle power histories reaching up to design burn-up 40000 MWd/TeHE. The resultant parameters such as fuel temperature, sheath plastic strain and fission gas pressure for AHWR fuel element were compared with respective thermo-mechanical parameters for similar fuel bundle element with natural uranium as fuel material. (author)

  19. Behavior of mixed-oxide fuel elements in a tight bundle under duty-cycle conditions

    International Nuclear Information System (INIS)

    The irradiation behavior of the TOB-10 fuel pins was comparable with that obtained in the single pin tests. There was no significant effect that could be directly attributed to tight bundle configuration. The postirradiation examination data provided information on the axial migration of cesium and its effect on cladding strain. Severe fuel/cladding chemical interaction (FCCI), which resulted in substantial cladding thinning and probably restricted venting of fission gas from the fuel column into the pin plena, apparently caused the earlier-than-expected cladding breaches in the D9-clad pins. No such severe FCCI was noted in the 316SS-clad pins. At the time of test termination, the overall cladding strain from creep and swelling was insufficient to cause bundle closure. Consequently, there would have been minimal pin bundle-duct interaction in the subassembly. Neither of the breaches appeared to be induced by pin bundle-duct interaction. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-07-01

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

  1. System for supporting a bundled tube fuel injector within a combustor

    Energy Technology Data Exchange (ETDEWEB)

    LeBegue, Jeffrey Scott; Melton, Patrick Benedict; Westmoreland, III, James Harold; Flanagan, James Scott

    2016-06-21

    A combustor includes an end cover having an outer side and an inner side, an outer barrel having a forward end that is adjacent to the inner side of the end cover and an aft end that is axially spaced from the forward end. An inner barrel is at least partially disposed concentrically within the outer barrel and is fixedly connected to the outer barrel. A fluid conduit extends downstream from the end cover. A first bundled tube fuel injector segment is disposed concentrically within the inner barrel. The bundled tube fuel injector segment includes a fuel plenum that is in fluid communication with the fluid conduit and a plurality of parallel tubes that extend axially through the fuel plenum. The bundled tube fuel injector segment is fixedly connected to the inner barrel.

  2. The small (or large) modular CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Meneley, D.; Harvel, G. [Univ. of Ontario Inst. of Tech., Oshawa, Ontario (Canada)

    2013-07-01

    This presentation outlines the design for small (or large) modular CANDU. The origins of this work go back many years to a comment by John Foster, then President of AECL CANDU. Foster noted that the CANDU reactor, with its many small fuel channels, was like a wood campfire. To make a bigger fire, just throw on some more logs (channels). If you want a smaller fire, just use fewer logs. The design process is greatly simplified.

  3. Feasibility of Accident-Tolerant FCM Replacement Fuel for CANDUs

    International Nuclear Information System (INIS)

    For enhanced accident tolerance, an innovative fuel concept, the fully ceramic microencapsulated (FCM) fuel based on the particle fuel concept of a gas-cooled reactor, is proposed to replace the conventional UO2 fuel bundle of existing and advanced CANDU reactors. In this study, the feasibility of replacing conventional UO2 fuel bundle with the accident-tolerant FCM fuel bundle has been assessed in view of core neutronics compatibility, accident-tolerance, and fuel cycle management. From the study, it was demonstrated that the FCM replacement fuel can provide resolution to CANDU generic issues by ensuring not only enhanced accident tolerance, but also an improved fuel cycle management. The accident-tolerant FCM fuel concept is proposed for replacing the conventional UO2 fuel bundle in CANDUs. The FCM fuel is shown to be neutronically compatible with existing core and the core residence time can be increased by more than 100 days. Accident-tolerance is remarkably enhanced by key features of the FCM fuel: it is refractory, thermo-mechanically and chemically stable, and fission product retentive. Less fuel feed and discharge obtained with the FCM fuel provide large savings in the spent fuel management burden charge and reduces the burden to the spent fuel storage facility in the long run. The smaller amount of minor actinides in the discharge bundles, together with the fission product retention and corrosion resistant features of the FCM fuel, should facilitate the long-term dry disposals of the spent fuel. From this study, it has been demonstrated that the CANDU FCM fuel is a feasible and viable option for CANDU reactors. The technology readiness level of the FCM fuel design and manufacturing is close to a lead test bundle loading for near-term deployment

  4. Investigation of neutronic behavior in a CANDU reactor with different (Am, Th, 235U)O2 fuel matrixes

    International Nuclear Information System (INIS)

    Recently thorium-based fuel matrixes are taken into consideration for nuclear waste incineration because of thorium proliferation resistance feature moreover its breeding or convertor ability in both thermal and fast reactors. In this work, neutronic influences of adding Am to (Th-235U)O2 on effective delayed neutron fraction, reactivity coefficients and burn up of a fed CANDU core has been studied using MCNPX 2.6.0 computational code. Different atom fractions of Am have been introduced in the fuel matrix to evaluate its effects on neutronic parameters of the modeled core. The computational data show that adding 2% atom fraction of Am to thorium-based fuel matrix won't noticeably change reactivity coefficients in comparison with the fuel matrix containing 1% atom fraction of Am. The use of 2% atom fraction of Am resulted in a higher delayed neutron fraction. According to the obtained data, 32.85 GWd burn up of the higher Americium-containing fuel matrix resulted in 55.2%, 26.5%, 41.9% and 2.14% depletion of 241Am, 243Am, 235U and 232Th respectively. 132.8 kg of 233U fissile element is produced after the burn up time and the nuclear core multiplication factor increases in rate of 2390 pcm. The less americium-containing fuel matrix resulted in higher depletion of 241/243Am, 235U and 232Th while the nuclear core effective multiplication factor increases in rate of 5630 pcm after the burn up time with 9.8 kg additional 233U production.

  5. Research reactor fuel bundle design review by means of hydrodynamic testing

    International Nuclear Information System (INIS)

    During the design steps of a fuel bundle for a nuclear reactor, some vibration tests are usually necessary to verify the prototype dynamical response characteristics and the structural integrity. To perform these tests, the known hydrodynamic loop facilities are used to evaluate the vibrational response of the bundle under the different flow conditions that may appear in the reactor. This paper describes the tests performed on a 19 plate fuel bundle prototype designed for a low power research reactor. The tests were done in order to know the dynamical characteristics of the plates and also of the whole bundle under different flow rate conditions. The paper includes a description of the test facilities and the results obtained during the dynamical characterization tests and some preliminary comments about the tests under flowing water are also presented. (author)

  6. A Model to Reproduce the Response of the Gaseous Fission Product Monitor (GFPM) in a CANDUR 6 Reactor (An Estimate of Tramp Uranium Mass in a Candu Core)

    International Nuclear Information System (INIS)

    In a Canada Deuterium Uranium (Candu) reactor, the fuel bundles produce gaseous and volatile fission products that are contained within the fuel matrix and the welded zircaloy sheath. Sometimes a fuel sheath can develop a defect and release the fission products into the circulating coolant. To detect fuel defects, a Gaseous Fission Product Monitoring (GFPM) system is provided in Candu reactors. The (GFPM) is a gamma ray spectrometer that measures fission products in the coolant and alerts the operator to the presence of defected fuel through an increase in measured fission product concentration. A background fission product concentration in the coolant also arises from tramp uranium. The sources of the tramp uranium are small quantities of uranium contamination on the surfaces of fuel bundles and traces of uranium on the pressure tubes, arising from the rare defected fuel element that released uranium into the core. This paper presents a dynamic model that reproduces the behaviour of a GFPM in a Candu 6 plant. The model predicts the fission product concentrations in the coolant from the chronic concentration of tramp uranium on the inner surface of the pressure tubes (PT) and the surface of the fuel bundles (FB) taking into account the on-power refuelling system. (authors)

  7. Total evaluation of in bundle void fraction measurement test of PWR fuel assembly

    International Nuclear Information System (INIS)

    Nuclear Power Engineering Corporation is performing the various proof or verification tests on safety and reliability of nuclear power plants under the sponsorship of the Ministry of International Trade and Industry. As one program of these Japanese national projects, an in bundle void fraction measurement test of a pressurized water reactor (PWR) fuel assembly was started in 1987 and finished at the end of 1994. The experiments were performed using the 5 x 5 square array rod bundle test sections. The rod bundle test section simulates the partial section and full length of a 17 x 17 type Japanese PWR fuel assembly. A distribution of subchannel averaged void fraction in a rod bundle test section was measured by the gamma-ray attenuation method using the stationary multi beam systems. The additional single channel test was performed to obtain the required information for the calibration of the rod bundle test data and the assessment of the void prediction method. Three test rod bundles were prepared to analyze an axial power distribution effect, an unheated rod effect, and a grid spacer effect. And, the obtained data were used for the assessment of the void prediction method relevant to the subchannel averaged void fraction of PWR fuel assemblies. This paper describes the outline of the experiments, the evaluation of the experimental data and the assessment of void prediction method

  8. Temperature Distributions in LMR Fuel Pin Bundles as Modeled by COBRA-IV-I

    Science.gov (United States)

    Wright, Steven A.; Stout, Sherry

    2005-02-01

    Most pin type reactor designs for space power or terrestrial applications group the fuel pins into a number of relatively large fuel pin bundles or subassemblies. Fuel bundles for terrestrial liquid metal fast breeders reactors typically use 217 - 271 pins per sub-assembly, while some SP100 designs use up to 331 pins in a central subassembly that was surrounded by partial assemblies. Because thermal creep is exponentially related to temperature, small changes in fuel pin cladding temperature can make large differences in the lifetime in a high temperature liquid metal reactor (LMR). This paper uses the COBRA-IV-I computer code to determine the temperature distribution within LMR fuel bundles. COBRA-IV-I uses the sub-channel analysis approach to determine the enthalpy (or temperature) and flow distribution in rod bundles for both steady-state and transient conditions. The COBRA code runs in only a few seconds and has been benchmarked and tested extensively over a wide range of flow conditions. In this report the flow and temperature distributions for two types of lithium cooled space reactor core designs were calculated. One design uses a very tight fuel pin packing that has a pitch to diameter ratio of 1.05 (small wire wrap with a diameter of 392 μm) as proposed in SP100. The other design uses a larger pitch to diameter ratio of 1.09 with a larger more conventional sized wire wrap diameter of 1 mm. The results of the COBRA pin bundle calculations show that the larger pitch-to-diameter fuel bundle designs are more tolerant to local flow blockages, and in addition they are less sensitive to mal-flow distributions that occur near the edges of the subassembly.

  9. Structural integrity evaluations of CANDU pressure tubes

    International Nuclear Information System (INIS)

    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 (Kr and Lr). 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 Kr and Lr, 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)

  10. A theoretical model of the mean grain size evolution during the liquid phase sintering in CANDU advanced fuels

    International Nuclear Information System (INIS)

    In order to enhance the fuel burnup in CANDU reactors, the microstructure of the fuel pellet must by modified, in order to accommodate the fission gas released during irradiation. A large grain size structure is able to retain a large fraction of the released gases. The addition of some oxides to the UO2 powder leads to the liquid phase ocarina during sintering. Starting from the basic theory of the grain growth in solid and liquid phase sintering and from the experimental microstructure characterization, a model of the mean size evolution in the presence of very small liquid fraction is proposed. As a function of dopant concentration, solubility limits and liquid phase composition, this model explicitly takes account of the continuous distribution of the additive into the matrix during grain growth process of the wetting properties of the liquid phase and of all transfer processes governing the microstructure evolution: grain boundary diffusion, surface reaction and diffusion through the liquid phase. The model validation was done for UO2-Nb2O3 and UO2-TiO2 systems, in isothermal treatment at 1700 deg. C. The model predictions are confirmed by the qualitative and quantitative experimental results, given by SEM, X-ray diffraction investigations. (authors)

  11. Radiation effects in glass and glass-ceramic waste forms for the immobilization of CANDU UO2 fuel reprocessing waste

    International Nuclear Information System (INIS)

    AECL has investigated three waste forms for the immobilization of high-level liquid wastes that would arise if used CANDU fuels were reprocessed at some time in the future to remove fissile materials for the fabrication of new power reactor fuel. These waste forms are borosilicate glasses, aluminosilicate glasses and titanosilicate glass-ceramics. This report discusses the potential effects of alpha, beta and gamma radiation on the releases of radionuclides from these waste forms as a result of aqueous corrosion by groundwaters that would be present in an underground waste disposal vault. The report discusses solid-state damage caused by radiation-induced atomic displacements in the waste forms as well as irradiation of groundwater solutions (radiolysis), and their potential effects on waste-form corrosion and radionuclide release. The current literature on radiation effects on borosilicate glasses and in ceramics is briefly reviewed, as are potential radiation effects on specialized waste forms for the immobilization of 129I, 85Kr and 14C. (author). 104 refs., 9 tabs., 5 figs

  12. Flow-induced vibration and acoustic behaviour of CANFLEX-LVRF bundles in a Bruce B NGS fuel channel

    International Nuclear Information System (INIS)

    Frequency/temperature sweep tests were performed in a high-temperature/high-pressure test channel to determine the acoustic and flow-induced vibration characteristics of the CANFLEX-LVRF bundle. The vibratory response of CANFLEX-LVRF bundles was compared with that of 37-element fuel bundles under Bruce B NGS fuel channel normal operating conditions. The tests were performed with a 12-bundle string of CANFLEX-LVRF bundles as well as a mixed string for the transition core. The tests showed that the LVRF bundles performed as required without failure or gross geometry changes. The mixed fuel strings behaved in a manner similar to that of a string of CANFLEX-LVRF bundles. (author)

  13. Measurement and CFD calculation of spacer loss coefficient for a tight-lattice fuel bundle

    International Nuclear Information System (INIS)

    Highlights: • Experiment and CFD analysis evaluated the pressure drop in a spacer grid. • The measurement and CFD errors for the spacer loss coefficient were estimated. • The spacer loss coefficient for the dual-cooled annular fuel bundle was determined. • The CFD prediction agrees with the measured spacer loss coefficient within 8%. - Abstract: An experiment and computational fluid dynamics (CFD) analysis were performed to evaluate the pressure drop in a spacer grid for a dual-cooled annular fuel (DCAF) bundle. The DCAF bundle for the Korean optimum power reactor (OPR1000) is a 12 × 12 tight-lattice rod array with a pitch-to-diameter ratio of 1.08 owing to a larger outer diameter of the annular fuel rod. An experiment was conducted to measure the pressure drop in spacer grid for the DCAF bundle. The test bundle is a full-size 12 × 12 rod bundle with 11 spacer grid. The test condition covers a Reynolds number range of 2 × 104–2 × 105 by changing the temperature and flow rate of water. A CFD analysis was also performed to predict the pressure drop through a spacer grid using the full-size and partial bundle models. The pressure drop and loss coefficient of a spacer grid were predicted and compared with the experimental results. The CFD predictions of spacer pressure drop and loss coefficient agree with the measured values within 8%. The spacer loss coefficient for the DCAF bundle is estimated to be approximately 1.50 at a nominal operating condition of OPR1000, i.e., Re = 4 × 105

  14. Progress of the DUPIC fuel compatibility analysis (II) - thermal-hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Choi, Hang Bok

    2005-03-01

    Thermal-hydraulic compatibility of the DUPIC fuel bundle with a 713 MWe Canada deuterium uranium (CANDU-6) reactor was studied by using both the single channel and sub-channel analysis methods. The single channel analysis provides the fuel channel flow rate, pressure drop, critical channel power, and the channel exit quality, which are assessed against the thermal-hydraulic design requirements of the CANDU-6 reactor. The single channel analysis by the NUCIRC code showed that the thermal-hydraulic performance of the DUPIC fuel is not different from that of the standard CANDU fuel. Regarding the local flow characteristics, the sub-channel analysis also showed that the uncertainty of the critical channel power calculation for the DUPIC fuel channel is very small. As a result, both the single and sub-channel analyses showed that the key thermal-hydraulic parameters of the DUPIC fuel channel do not deteriorate compared to the standard CANDU fuel channel.

  15. CANDU safety under severe accidents

    International Nuclear Information System (INIS)

    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

  16. Physics and Control Assessment of AN 850 Mw(e) - Leu-Candu Reactor.

    Science.gov (United States)

    Barbone, Michelangelo

    The physics and control assessment of an 850 MW(e) Low Enriched Uranium CANDU Pressurized Heavy Water (LEU -CANDU-PHW) reactor constitute the major objective of this thesis. The use of Low Enriched Uranium fuel in the present CANDU nuclear power generating stations is recognized as economically beneficial due to reduced fuelling costs. The LEU fuel cycle is also recognized as a stepping stone to transit from the present CANDU-PHW once-through natural Uranium cycle to advanced cycles such as those based on Plutonium recycle, once-through Th + U-235 cycle, Thorium with Uranium recycle and net U-235 feed, Thorium with Uranium recycle and Plutonium feed. However, although the use of Low Enriched Uranium in the present CANDU-PHW reactor has economic advantages, and it would act as a technical bridge between the present cycle and advanced cycles, technical problems in different areas of reactor physics and fuel management were anticipated. The present thesis research work adresses the areas of reactor physics, fuel management, and control (in particular, the spatial control of large CANDU-PHW reactors). The main conclusions that have been drawn following these studies are as follows: (1) The Low Enriched Uranium Cycle is feasible in a CANDU-PHW reactor of present design and provided that: (a) The enrichment is kept relatively low (that is, about 1% instead of 0.711%); (b) the number of bundles to be replaced at every refuelling operation is about one-half that of the natural Uranium fuel case; (c) The channels are refuelled in the same direction as the coolant. (2) The response of an LEU-CANDU-PHW reactor to reactivity perturbation such as single- and two-channel refuelling operation, shim transient, shutdown-start-up transient with enrichment levels of 0.9% and 1.2% is essentially very similar {provided that certain conditions in (1) are respected} to that of the natural uranium reactor core case without any reactor reoptimization. The general behaviour of the reactor

  17. CANDU development

    International Nuclear Information System (INIS)

    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 320C. 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

  18. ELESTRES 2.1 computer code for high burnup CANDU fuel performance analysis

    International Nuclear Information System (INIS)

    The ELESTRES (ELEment Simulation and sTRESses) computer code models the thermal, mechanical and micro structural behaviours of CANDU® fuel element under normal operating conditions. The main purpose of the code is to calculate fuel temperatures, fission gas release, internal gas pressure, fuel pellet deformation, and fuel sheath strains in fuel element design analysis and assessments. It is also used to provide initial conditions for evaluating fuel behaviour during high temperature transients. ELESTRES 2.1 was developed for high burnup fuel application, based on an industry standard tool version of the code, through the implementation or modification to code models such as fission gas release, fuel pellet densification, flux depression (radial power distribution in the fuel pellet), fuel pellet thermal conductivity, fuel sheath creep, fuel sheath yield strength, fuel sheath oxidation, two dimensional heat transfer between the fuel pellet and the fuel sheath; and an automatic finite element meshing capability to handle various fuel pellet shapes. The ELESTRES 2.1 code design and development was planned, implemented, verified, validated, and documented in accordance with the AECL software quality assurance program, which meets the requirements of the Canadian Standards Association standard for software quality assurance CSA N286.7-99. This paper presents an overview of the ELESTRES 2.1 code with descriptions of the code's theoretical background, solution methodologies, application range, input data, and interface with other analytical tools. Code verification and validation results, which are also discussed in the paper, have confirmed that ELESTRES 2.1 is capable of modelling important fuel phenomena and the code can be used in the design assessment and the verification of high burnup fuels. (author)

  19. Posttest examination of the VVER-1000 fuel rod bundle CORA-W2

    International Nuclear Information System (INIS)

    The bundle meltdown experiment CORA-W2, representing the behavior of a Russian type VVER-1000 fuel element, with one B4C/stainless steel absorber rod was selected by the OECD/CSNI as International Standard Problem (ISP-36). The experimental results of CORA-W2 serve as data base for comparison with analytical predictions of the high-temperature material behavior by various code systems. The first part of the experimental results is described in KfK 5363 (1994), the second part is documented in this report which contains the destructive post-test examination results. The metallographical and analytical (SEM/EDX) post-test examinations were performed in Germany and Russia and are summarized in five individual contributions. The upper half of the bundle is completely oxidized, the lower half has kept the fuel rods relatively intact. The post-test examination results show the strong impact of the B4C absorber rod and the stainless steel grid spacers on the ''low-temperature'' bundle damage initiation and progression. The B4C absorber rod completely disappeared in the upper half of the bundle. The multicomponent melts relocated and formed coolant channel blockages on solidification with a maximum extent of about 30% in the lower part of the bundle. At temperatures above the melting point of the ZrNb1 cladding extensive fuel dissolution occurred. (orig.)

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

    International Nuclear Information System (INIS)

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

  1. Measurements of instant-release source terms for 137Cs, 90Sr, 99Tc, 129I and 14C in used CANDU fuels

    International Nuclear Information System (INIS)

    Combined gap and grain-boundary inventories of 137Cs, 129I, 90Sr, 99Tc and 14C were measured in 15 used CANDU fuel elements by leaching crushed fuel samples. A good correlation between the combined gap and grain-boundary inventories of 137Cs and 129I was found, suggesting that these fission products exhibit similar behavior in CANDU fuel. The expected correlation between combined gap and grain-boundary inventories of 137Cs and 129I with calculated fission-gas release to the gap and grain boundaries could only be confirmed for lower power fuels (90Sr were higher than expected and showed no correlation with calculated fission-gas release. No values for the combined gap and grain-boundary inventories of 99Tc were obtained because 99Tc in used fuel samples is very insoluble and appears to require oxidation prior to dissolution. Combined gap and grain-boundary inventories of 14C appeared to be independent of fuel power or burnup. (orig.)

  2. Enhanced candu 6 reactor: status

    International Nuclear Information System (INIS)

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

  3. Computation of two phase flow distribution and CHF in horizontal bundles using the assert subchannel code

    International Nuclear Information System (INIS)

    The ASSERT subchannel code has been developed specifically to model flow and phase distributions within CANDU fuel bundles. ASSERT uses a drift-flux model which permits the phases to have unequal velocities, and can thus model phase separation tendencies which may occur in horizontal flow. The basic principles of ASSERT are outlined, and computed results are compared against data from various experiments for validation purposes. The paper concludes with an example of the use of the code to predict critical heat flux in CANDU geometries

  4. Results of international standard problem No. 36 severe fuel damage experiment of a VVER fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Firnhaber, M. [Gesellschaft fuer Anlagen-und Reaktorsicherheit, Koeln (Germany); Yegorova, L. [Nuclear Safety Institute of Russian Research Center, Moscow (Russian Federation); Brockmeier, U. [Ruhr-Univ. of Bochum (Germany)] [and others

    1995-09-01

    International Standard Problems (ISP) organized by the OECD are defined as comparative exercises in which predictions with different computer codes for a given physical problem are compared with each other and with a carefully controlled experimental study. The main goal of ISP is to increase confidence in the validity and accuracy of analytical tools used in assessing the safety of nuclear installations. In addition, it enables the code user to gain experience and to improve his competence. This paper presents the results and assessment of ISP No. 36, which deals with the early core degradation phase during an unmitigated severe LWR accident in a Russian type VVER. Representatives of 17 organizations participated in the ISP using the codes ATHLET-CD, ICARE2, KESS-III, MELCOR, SCDAP/RELAP5 and RAPTA. Some participants performed several calculations with different codes. As experimental basis the severe fuel damage experiment CORA-W2 was selected. The main phenomena investigated are thermal behavior of fuel rods, onset of temperature escalation, material behavior and hydrogen generation. In general, the calculations give the right tendency of the experimental results for the thermal behavior, the hydrogen generation and, partly, for the material behavior. However, some calculations deviate in important quantities - e.g. some material behavior data - showing remarkable discrepancies between each other and from the experiments. The temperature history of the bundle up to the beginning of significant oxidation was calculated quite well. Deviations seem to be related to the overall heat balance. Since the material behavior of the bundle is to a great extent influenced by the cladding failure criteria a more realistic cladding failure model should be developed at least for the detailed, mechanistic codes. Regarding the material behavior and flow blockage some models for the material interaction as well as for relocation and refreezing requires further improvement.

  5. Prediction of temperature distribution in a fast reactor spent fuel bundle

    International Nuclear Information System (INIS)

    A simple mathematical model is described for predicting temperature distribution in a spent fuel bundle. The model takes into account γ-ray leakage, radiant and conductive heat transports between the various fuel pins arranged in a triangular array and enclosed in a hexagonal shaped tube containing gaseous medium. With the geometry of the fuel bundle the configuration factors between various fuel pins can be calculated. The configuration factors along with the heat generation rates, net γ-ray leakage, surface emissivity, conductivity of the enclosed medium and the temperature of the hexagonal tube can be used to estimate the temperature distribution with the help of the computer code TICOFUSA developed on the basis of this model. (author)

  6. Development of analysis system and analysis on reactor physics for CANDU advanced fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Bong Gi; Bae, Chang Joon; Kwon, Oh Sun [Korea Power Electric Corporation, Taejon (Korea, Republic of)

    1997-07-01

    characteristics of reactor physics for CANFLEX-NU fuel core were calculated using final fuel design data. The results of analysis showed that there was no impact on reactor operations and safety. The above results of calculations and analysis were described in the physics design for CANFLEX-NU= fuel core. Various fuel models were evaluated for selecting high burnup fuel using recovered uranium. It is judged to be worse effects for reactor safety Hence, the use of graphite within fuel was proposed and its results showed to be better. The analysis system of reactor physics for design and analysis of high burnup fuel was evaluated. Lattice codes and core code were reviewed. From the results, the probability of WIMS-AECL and HELIOS is known to be high for analysis of high burnup fuel. For the core code, RFSP, it was evaluated that the simplified 2 group equation should be replaced by explicit 2 group equation. (Author) 32 refs., 25 tabs., 79 figs.

  7. Measurement of the composition of noble-metal particles in high-burnup CANDU fuel by wavelength dispersive X-ray microanalysis

    Energy Technology Data Exchange (ETDEWEB)

    Hocking, W.H.; Szostak, F.J

    1999-09-01

    An investigation of the composition of the metallic inclusions in CANDU fuel, which contain Mo, Tc, Ru, Rh and Pd, has been conducted as a function of burnup by wavelength dispersive X-ray (WDX) microanalysis. Quantitative measurements were performed on micrometer sized particles embedded in thin sections of fuel using elemental standards and the ZAF method. Because the fission yields of the noble metals change with burnup, as a consequence of a shift from almost entirely {sup 235}U fission to mainly {sup 239}Pu fission, their inventories were calculated from the fuel power histories using the WIMS-Origin code for comparison with experiment. Contrary to expectations that the oxygen potential would be buffered by progressive Mo oxidation, little evidence was obtained for reduced incorporation of Mo in the noble-metal particles at high burnup. These surprising results are discussed with respect to the oxygen balance in irradiated CANDU fuels and the likely intrinsic and extrinsic sinks for excess oxygen. (author)

  8. Build your own Candu reactor

    International Nuclear Information System (INIS)

    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.)

  9. Extrapolating power-ramp performance criteria for current and advanced CANDU fuels

    Energy Technology Data Exchange (ETDEWEB)

    Tayal, M.; Chassie, G.G

    2000-06-01

    To improve the precision and accuracy of power-ramp performance criteria for high-burnup fuel, we have examined in-reactor fuel performance data as well as out-reactor test data. The data are consistent with some of the concepts used in the current formulations for defining fuel failure thresholds, such as size of power-ramp and extent of burnup. Our review indicates that there is a need to modify some other aspects of the current formulations; therefore, a modified formulation is presented in this paper. The improvements mainly concern corrodent concentration and its relationships with threshold stress for failure. The new formulation is consistent with known and expected trends such as strength of Zircaloy in corrosive environment, timing of the release of fission products to the pellet-to-sheath gap, CANLUB coating, and fuel burnup. Because of the increased precision and accuracy, the new formulation is better able to identify operational regimes that are at risk of power-ramp failures; this predictive ability provides enhanced protection to fuel against power-ramp defects. At die same time, by removing unnecessary conservatisms in other areas, the new formulation permits a greater range of defect-free operational envelope as well as larger operating margins in regions that are, in fact, not prone to power-ramp failures. (author)

  10. Modelling of iodine-induced stress corrosion cracking in CANDU fuel

    Science.gov (United States)

    Lewis, B. J.; Thompson, W. T.; Kleczek, M. R.; Shaheen, K.; Juhas, M.; Iglesias, F. C.

    2011-01-01

    Iodine-induced stress corrosion cracking (I-SCC) is a recognized factor for fuel-element failure in the operation of nuclear reactors requiring the implementation of mitigation measures. I-SCC is believed to depend on certain factors such as iodine concentration, oxide layer type and thickness on the fuel sheath, irradiation history, metallurgical parameters related to sheath like texture and microstructure, and the mechanical properties of zirconium alloys. This work details the development of a thermodynamics and mechanistic treatment accounting for the iodine chemistry and kinetics in the fuel-to-sheath gap and its influence on I-SCC phenomena. The governing transport equations for the model are solved with a finite-element technique using the COMSOL Multiphysics® commercial software platform. Based on this analysis, this study also proposes potential remedies for I-SCC.

  11. CANDU steam generator life management

    International Nuclear Information System (INIS)

    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)

  12. The demonstration irradiation of the CANFLEX-NU fuel bundle in Wolsong NGS 1

    International Nuclear Information System (INIS)

    A demonstration irradiation (DI) of 24 CANFLEX-NU fuel bundles in the high power Q07 channel and low power L21 channel of Wolsong Power Generation Station-1 had been successfully conducted jointly by KEPRI/KHNP/KAERI in the period of 2002 July to 2004 January. The tracking of the reactor operation data showed that the reactor has been stably operated during the DI. One CANFLEX bundle irradiated in the Q07 channel had a typical history of high power and high burnup, where the maximum element linear power rating was ∼ 42 kW/m at the burnup of ∼ 50 MWh/kgU and ∼ 35 kW/m at the discharge element burnup of ∼ 210 MWh/kgU. While, another CANFLEX bundles also irradiated in the Q07 channel had a typical history of power ramping, where the maximum element power ramping-up or -down rate was 28 kW/m. The unusual performance and integrity of the CANFLEX elements could not be found in the ELESTRES predictions and also the in-bay visual examinations showed that all the bundles were intact, free of defects and appeared to be in good condition as expected. Therefore, it is concluded that the demonstration irradiation shows the validation of the CANFLEX bundle performance with direct conditions of relevance under the Korean licensing requirements and the KNFC fuel fabrication capability, and provides the rationale for the decision to perform the full-conversion of CANFLEX fuel in WPGS-1. (author)

  13. Development of neural network for analysis of local power distributions in BWR fuel bundles

    International Nuclear Information System (INIS)

    A neural network model has been developed to learn the local power distributions in a BWR fuel bundle. A two layers neural network with total 128 elements is used for this model. The neural network learns 33 cases of local power peaking factors of fuel rods with given enrichment distribution as the teacher signals, which were calculated by a fuel bundle nuclear analysis code based on precise physical models. This neural network model studied well the teacher signals within 1 % error. It is also able to calculate the local power distributions within several % error for the different enrichment distributions from the teacher signals when the average enrichment is close to 2 %. This neural network is simple and the computing speed of this model is 300 times faster than that of the precise nuclear analysis code. This model was applied to survey the enrichment distribution to meet a target local power distribution in a fuel bundle, and the enrichment distribution with flat power shape are obtained within short computing time. (author)

  14. Development of neural network simulating power distribution of a BWR fuel bundle

    International Nuclear Information System (INIS)

    A neural network model is developed to simulate the precise nuclear physics analysis program code for quick scoping survey calculations. The relation between enrichment and local power distribution of BWR fuel bundles was learned using two layers neural network (ENET). A new model is to introduce burnable neutron absorber (Gadolinia), added to several fuel rods to decrease initial reactivity of fresh bundle. The 2nd stages three layers neural network (GNET) is added on the 1st stage network ENET. GNET studies the local distribution difference caused by Gadolinia. Using this method, it becomes possible to survey of the gradients of sigmoid functions and back propagation constants with reasonable time. Using 99 learning patterns of zero burnup, good error convergence curve is obtained after many trials. This neural network model is able to simulate no learned cases fairly as well as the learned cases. Computer time of this neural network model is about 100 times faster than a precise analysis model. (author)

  15. Resistance factors, two phase multipliers and void fractions for best estimate flow calculations in Dodewaard fuel bundles

    International Nuclear Information System (INIS)

    Values are given for resistance factors, two phase multipliers and core and chimney void fractions in the fuel and chimney to be used in best estimate calculations of the flow in Dodewaard fuel bundles. The resistance factors are based on single phase experimental data for a mockup of the Dodewaard fuel bundle. The two phase multipliers are determined from two phase measurements of mockups of other fuel bundles for nuclear reactors. This is also true for the in bundle void fractions. The void fractions in the chimney have been validated by measured void fractions in large diameter pipes. The recommended changes to the existing input for calculations are somewhat larger than the uncertainties in the measurements. (author). 37 refs.; 48 figs.; 4 tabs

  16. Comparison of MCNP4B and WIMS-AECL calculations of coolant-void-reactivity effects for uniform lattices of CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    1999-07-01

    This paper compares the results of coolant-void reactivity (CVR) reactor-physics calculations performed using the Monte Carlo N-particle transport code, MCNP version 4B, with those obtained using Atomic Energy of Canada Limited's (AECL's) latest version of the Winfrith improved multigroup scheme (WIMS) code, WIMS-AECL version 2-5c. Cross sections derived from the evaluated nuclear data file version B-VT (ENDF/B-VI) are used for both the WIMS-AECL and MCNP4B calculations. The comparison is made for uniform lattices at room temperature containing either fresh natural uranium or mixed oxide (MOX) 37-element CANDU fuel. The MOX fuel composition corresponds roughly to that of irradiated CANDU fuel at a burnup of about 4500 MWd/tU. The level of agreement between the CVR predictions of WIMS-AECL and MCNP4B is studied as a function of lattice buckling (a measure of the curvature of the neutron-flux distribution) over the range from 0.0 to 4.1 m{sup -2}. For the cases studied, it is found that the absolute keff values calculated by WIMS-AECL are higher than those of MCNP4B by several mk (1 mk is a change of 0.001 in keff), amounts that depend on the fuel type being modelled and the particular cross-section data used. However, the agreement between WIMS-AECL and MCNP4B is much better for the CVR (i.e., the {delta}keff on coolant voiding), and is relatively insensitive to the fuel type. (author)

  17. Comparison of MCNP4B and WIMS-AECL calculations of coolant-void-reactivity effects for uniform lattices of CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S

    1999-05-01

    This paper compares the results of coolant-void reactivity (CVR) reactor-physics calculations performed using the Monte Carlo N-particle transport code, MCNP version 4B, with those obtained using Atomic Energy of Canada Limited's (AECL's) latest version of the Winfrith improved multigroup scheme (WIMS) code, WIMS-AECL version 2-5c. Cross sections derived from the evaluated nuclear data file version B-VI (ENDF/B-VI) are used for both the WIMS-AECL and MCNP4B calculations. The comparison is made for uniform lattices at room temperature containing either fresh natural uranium or mixed oxide (MOX) 37-element CANDU fuel. The MOX fuel composition corresponds roughly to that of irradiated CANDU fuel at a burnup of about 4500 MWd/tU. The level of agreement between the CVR predictions of WIMS-AECL and MCNP4B is studied as a function of lattice buckling (a measure of the curvature of the neutron-flux distribution) over the range from 0.0 to 4.1 m{sup -2} . For the cases studied, it is found that the absolute k values calculated by WIMS-AECL are higher than those of MCNP4B by several mk (1 mk is a change of 0.001 in k), amounts that depend on the fuel type being modelled and the particular cross-section data used. However, the agreement between WIMS-AECL and MCNP4B is much better for the CVR (i.e., the {delta}k on coolant voiding), and is relatively insensitive to the fuel type. (author)

  18. A Mechanistic Approach for the Prediction of Critical Power in BWR Fuel Bundles

    Science.gov (United States)

    Chandraker, Dinesh Kumar; Vijayan, Pallipattu Krishnan; Sinha, Ratan Kumar; Aritomi, Masanori

    The critical power corresponding to the Critical Heat Flux (CHF) or dryout condition is an important design parameter for the evaluation of safety margins in a nuclear fuel bundle. The empirical approaches for the prediction of CHF in a rod bundle are highly geometric specific and proprietary in nature. The critical power experiments are very expensive and technically challenging owing to the stringent simulation requirements for the rod bundle tests involving radial and axial power profiles. In view of this, the mechanistic approach has gained momentum in the thermal hydraulic community. The Liquid Film Dryout (LFD) in an annular flow is the mechanism of CHF under BWR conditions and the dryout modeling has been found to predict the CHF quite accurately for a tubular geometry. The successful extension of the mechanistic model of dryout to the rod bundle application is vital for the evaluation of critical power in the rod bundle. The present work proposes the uniform film flow approach around the rod by analyzing individual film of the subchannel bounded by rods with different heat fluxes resulting in different film flow rates around a rod and subsequently distributing the varying film flow rates of a rod to arrive at the uniform film flow rate as it has been found that the liquid film has a strong tendency to be uniform around the rod. The FIDOM-Rod code developed for the dryout prediction in BWR assemblies provides detailed solution of the multiple liquid films in a subchannel. The approach of uniform film flow rate around the rod simplifies the liquid film cross flow modeling and was found to provide dryout prediction with a good accuracy when compared with the experimental data of 16, 19 and 37 rod bundles under BWR conditions. The critical power has been predicted for a newly designed 54 rod bundle of the Advanced Heavy Water Reactor (AHWR). The selected constitutive models for the droplet entrainment and deposition rates validated for the dryout in tube were

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

    International Nuclear Information System (INIS)

    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

  20. Single and two-phase flow pressure drop for CANFLEX bundle

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Jun, Ji Su; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Dimmick, G. R.; Bullock, D. E. [Atomic Energy of Canada Limited, Ontario (Canada)

    1998-12-31

    Friction factor and two-phase flow frictional multiplier for a CANFLEX bundle are newly developed and presented in this paper. CANFLEX as a 43-element fuel bundle has been developed jointly by AECL/KAERI to provide greater operational flexibility for CANDU reactor operators and designers. Friction factor and two-phase flow frictional multiplier have been developed by using the experimental data of pressure drops obtained from two series of Freon-134a (R-134a) CHF tests with a string of simulated CANFLEX bundles in a single phase and a two-phase flow conditions. The friction factor for a CANFLEX bundle is found to be about 20% higher than that of Blasius for a smooth circular pipe. The pressure drop predicted by using the new correlations of friction factor and two-phase frictional multiplier are well agreed with the experimental pressure drop data of CANFLEX bundle within {+-} 5% error. 11 refs., 5 figs. (Author)

  1. CANDU nuclear power system

    International Nuclear Information System (INIS)

    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

  2. CANDU operating experience

    International Nuclear Information System (INIS)

    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

  3. Study of thermal hydraulic behavior of supercritical water flowing through fuel rod bundles

    International Nuclear Information System (INIS)

    Investigations on thermal-hydraulic behavior in Supercritical Water Reactor (SCWR) fuel assembly have obtained a significant attention in the international SCWR community because of its potential to obtain high thermal efficiency and compact design. Present work deals with CFD analysis to study the flow and heat transfer behavior of supercritical water in 4 metre long 7-pin fuel bundle using commercial CFD package ANSYS CFX for single phase steady state conditions. Considering the symmetric conditions, 1/12th part of the fuel rod bundle is taken as a domain of analysis. RNG K-epsilon model with scalable wall functions is used for modeling the turbulence behavior. Constant heat flux boundary condition is applied at the fuel rod surface. IAPWS equations of state are used to compute thermo-physical properties of supercritical water. Sharp variations in its thermo-physical properties (specific heat, density) are observed near the pseudo-critical temperature causing sharp change in heat transfer coefficient. The pseudo-critical point initially appears in the gaps among heated fuel rods, and then spreads radially outward reaching the adiabatic wall as the flow goes downstream. The enthalpy gain in the centre of the channel is much higher than that in the wall region. Non-uniformity in the circumferential distribution of surface temperature and heat transfer coefficient is observed which is in agreement with published literature. Heat transfer coefficient is high on the rod surface near the tight region and decreases as the distance between rod surfaces increases. (author)

  4. DUPIC fuel compatibility assessment

    International Nuclear Information System (INIS)

    The purpose of this study is to assess the compatibility of DUPIC (Direct Use of Spent PWR Fuel in CANDU Reactors) fuel with the current CANDU 6 reactor, which is one of the technology being developed to utilize the spent PWR fuel in CANDU reactors. The Phase II study of this project includes the analysis of impact on the reactor safety, the development of core design technology, the development of fuel supply technology of optimal composition, and feasibility analysis on localization and license of DUPIC fuel. From the reactor safety analysis results, it is known that DUPIC fuel satisfies the safety limit of reactor containment and public dose for single failure. But, the safety limit may be exceeded for dual failure. Therefore, more analysis is needed for the removal of excessive conservatism in accident analysis methodology and modification of transient fuel behavior analysis methodology. The results of the validation calculations of core design methodology have confirmed that the current core analysis system is acceptable for the feasibility study of the DUPIC fuel compatibility analysis. The results of compatibility and fuel fabrication have shown that DUPIC fuel is technically feasible. For practical use and licensing, however, more research items required in the practical use, fuel rod and bundle design and fuel loading are should be performed. When these items are performed and resolved, the compatibility of the DUPIC fuel is achieved, and, eventually, the possibility of DUPIC fuel licensing can be confirmed

  5. PHWR advanced fuel R and D for the 21st Century in Korea

    International Nuclear Information System (INIS)

    cycles, SEU, LWR/PHWR synergism (RU, MOX, TANDEM) and Th could be also considered for CANDU fuel cycles in the 21st century, because CANDU fuel cycle flexibility arises naturally from excellent neutron economy, on-power fuelling, and simple fuel design. The advanced fuel cycles can be utilized in the CANFLEX 43-element fuel bundle. However, the CANDU-PHWR advanced fuel R and D for the 21st century in Korea shall take into account 'nuclear safety', 'nuclear waste', 'non-proliferation' and 'economics' in the favor of the arguments from the public, international environment, and utilities. (author)

  6. Characteristics of DUPIC fuel fabrication technology

    International Nuclear Information System (INIS)

    The concept of DUPIC (Direct use of spent PWR fuel in CANDU reactor) fuel is recently being taken a growing interest as an innovative fuel in terms of proliferation resistant fuel cycle for reusing spent PWR fuel in CANDU reactor without wet reprocessing process. The fabrication processes are composed of decladding of spent PWR fuel rods, OREOX processing to produce the powder feedstock from the spent PWR fuel meat. Once the resinterable powder is prepared, the fuel fabrication processes such as pelletizing, rod and bundle manufacturing processes are similar to the conventional powder/pellet routes for the fuel fabrication except that all the processes should be performed in hot cell by remote manner. The characteristics of the DUPIC fuel fabrication processes and the status of the equipment development at KAERI is described. (author)

  7. Reactor physics and economic aspects of the CANDU reactor system

    International Nuclear Information System (INIS)

    A history of the development of the CANDU system is given along with a fairly detailed description of the 600 MW(e) CANDU reactor. Reactor physics calculation methods are described, as well as comparisons between calculated reactor physics parameters and those measured in research and power reactors. An examination of the economics of CANDU in the Ontario Hydro system and a comparison between fossil fuelled and light water reactors is presented. Some physics, economics and resources aspects are given for both low enriched uranium and thorium-fuelled CANDU reactors. Finally the RβD program in Advanced Fuel Cycles is briefly described

  8. Assessment of RELAP5/CANDU+ code for regulatory auditing analysis of CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Young Seok; Kim, Hho Jung; Yang, Chae Yong

    2001-12-15

    The objectives of this study are to undertake the verification and validation of RELAP5/CANDU+ code, which is developed in this project, by simulating the B8711 test of RD-14 facility, and to examine the properties of this code by doing the sensitivity analysis for experimental prediction modes about thermal-hydraulics phenomena in CANDU reactor systems added to this code. The B8711 test was an experiment of a 45% ROH break for simulating large LOCA. Also, in this study, the methods for making input cards related to CANDU options are described, so that some users can use the RELAP5/CANDU+ code with easy. RELAP/CANDU+ code can choose the options of Henry-Fauske mode, Ransom-Trapp model, and Moody model for prediction of the critical mass flow. It is examined that Henry-Fauske model and Ransom-Trapp model are considered properly, but Moody model is still required to be improved. Heat transfer correlations available in RELAP5/CANDU+ code for CANDU-type reactors are a horizontal stratified model, a fuel heat-up model and D2O/H2O CHF correlations, and these models take an important role to improve the predictability of the experimental procedures. It is concluded that RELAP5/CANDU+ code is useful for the auditing of the accident analysis of CANDU reactors, and the results of the sensitivity analysis for thermal-hydraulic models examined in this study are valuable for the actual auditing of real CANDU-type power plants.

  9. An analytic study on LBLOCA for CANDU type reactor using MARS-KS/CANDU

    International Nuclear Information System (INIS)

    This study provides the simulation results using MARS-KS/CANDU code for the Large Break LOCA of CANDU type reactor. The purpose of the study is to evaluate the capability of MARS-KS/CANDU for simulating the actual plants (Wolsong 2/3/4). The steady state and the transient analysis results were provided. After the sensitivity study depend on break size, the case that 35% of the inlet header known as the accident that has the most limiting effect on the temperature of the fuel sheath was calculated. In order to evaluate the results, the results were compared with those of CATHENA simulation. (author)

  10. Compatibility analysis of DUPIC fuel (part 3) - radiation physics analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chun Soo; Bae, Dae Seok; Kim, Kyung Su; Park, Byung Yun; Koh, Young Kown

    2000-04-01

    As a part of the compatibility analysis of DUPIC fuel in CANDU reactors, the radiation physics calculations have been performed for the CANDU primary shielding system, thermal shield, radiation damage, transportation cask and storage. At first, the primary shield system was assessed for the DUPIC fuel core, which has shown that the dose rates and heat deposition rates through the primary shield of the DUPIC fuel core are not much different from those of natural uranium core because the power levels on the core periphery are similar for both cores. Secondly, the radiation effects on the critical components and the themal shields were assessed when the DUPIC fuel is loaded in CANDU reactors. Compared with the displacement per atom (DPA) of the critical component for natural uranium core, that for the DUPIC fuel core was increased by -30% for the innermost groove and the weld points and by -10% for the corner of the calandria subshells and annular plates in the calandria, respectivdely. Finally, the feasibility study of the DUPIC fuel handling was performed, which has shown that all handling and inspection of the DUPIC fuel bundles be done remotely and behind a shielding wall. For the transportation of the DUPIC fuel, the preliminary study has shown that there shold be no technical problem th design a transportation cask for the fresh and spent DUPIC fuel bundles. For the storage of the fresh and spent DUPIC fuels, there is no the criticality safety problem unless the fuel bundle geometry is destroyed.

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

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Won Young; Min, Byung Joo; Park, Joo Hwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2006-07-01

    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{sup 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.

  12. Consideration of subchannel area of a 37-element fuel to enhance CHF

    International Nuclear Information System (INIS)

    CANDU-6 reactor has 380 fuel channels of a pressure tube type, which provides an independent flow passage, and each pressure tube contains 12 fuel bundles horizontally. The CHF of a CANDU fuel bundle in a horizontal fuel channel is one of the important parameters determining the thermalhydraulic safety margin as well as the trip set point of the Regional Overpower Protection (ROP) system. Hence, the CHF enhancement of a CANDU fuel bundle has been an issue for a long time and can be affected by the geometric configuration of the fuel elements as well as several appendages such as the end-plates, bearing pads, and spacers attached to the fuel elements. This paper considers the modification of the inner ring radius of a standard 37-element fuel bundle to enhance the CHF, since the CHFs of a standard 37-element fuel bundle preferably occur at the peripheral subchannels of the center rod, owing to the relative small flow area or high flow resistance under high flow conditions or the normal operating conditions of a CANDU reactor. Subchannel analysis techniques using the ASSERT-PV code were applied to investigate the local CHF characteristics according to the inner ring radius variation for the original diameter of the pressure tube. It was found that the modification of the inner ring radius is very effective in enhancing the dryout power of the fuel bundle under the reactor operating conditions through an enthalpy re-distribution of the subchannels and change in the local locations of the first CHF occurrences. (author)

  13. Economics of CANDU-PHW

    International Nuclear Information System (INIS)

    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

  14. The transition from natural uranium to 1.2% SEU in a CANDU with repositioned reactivity devices

    International Nuclear Information System (INIS)

    The transition from natural uranium to 1.2% slightly enriched uranium (SEU) has been modelled in an 1100-day time-dependent fuel-management simulation for a CANDU 6 in which the reactivity devices have been relocated. The fuelling strategy was simply to replace natural uranium by SEU, two bundles at a time. Two main challenges were encountered during the transition: 1) maintaining criticality when the transition first began, when the SEU was added to the low-worth locations at the channel ends; and 2) keeping maximum bundle powers to within acceptable limits when there were two or four SEU bundles in the channel. The ability to relocate the adjuster rods was found to have little impact on the ability to shape the axial power distribution during the transition. Alternate strategies are suggested for facilitating the transition

  15. Economics of CANDU

    International Nuclear Information System (INIS)

    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

  16. OpenFOAM Analysis of CANDU-6 Moderator Flow

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyoung Tae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Chang, Se-Myong [Kunsan National University, Gunsan (Korea, Republic of)

    2015-10-15

    In this study OpenFOAM (Open Field Operation and Manipulation), an open source CFD solver, is used to simulate the three-dimensional moderator flow in calandria tank of CANDU-6 reactor improving the computational efficiency by parallel computing which does not need any proprietary license. A prototype of CANDU-6 reactor is numerically analyzed about three-dimensional moderator flow in calandrian tank with OpenFOAM, an open source CFD code. The horizontal fuel channels in a CANDU-6 reactor (a pressurized heavy water reactor) are submerged in the heavy water (D{sub 2}O) pool which is contained by a cylindrical tank, calandria. Each fuel channel consists of concentric tubes: a Pressure Tube (PT) and a Calandria Tube (CT). And the CO{sub 2} gas is filled between these tubes. Consequently, a heat flux is rapidly transferred to the outer CT so that a film boiling may occur in CT. As a result, it is important to keep the subcooling in the moderator. It is one of the major concerns in the CANDU safety analyses to estimate the local subcooling margin of the moderator inside the calandria tank. Previous experimental studies showed that the film boiling would be unlikely to occur if the local moderator subcooling is sufficient. Therefore, an accurate prediction of the moderator temperature distribution in the calandria tank is needed to confirm the channel integrity. There have been numerous computational efforts to estimate the thermal hydraulics in the calandria tank using CFD codes. Hadaller et al. obtained a tube bank pressure drop model for tube bundle region of the calandria tank and implemented it into the MODTURC{sub C}LAS code. Yoon et al. used the CFX code to develop a CFD model with a porous media approach for the core region. However, it is known that porous media modeling provide only average values of flow velocities and temperatures and do not give any information about local flow variables near tube solid walls, which are necessary to implement accurate heat

  17. Combustor having mixing tube bundle with baffle arrangement for directing fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hughes, Michael John; McConnaughhay, Johnie Franklin

    2016-08-23

    A combustor includes a tube bundle that extends radially across at least a portion of the combustor. The tube bundle includes an upstream surface axially separated from a downstream surface, and a plurality of tubes extend from the upstream surface through the downstream surface to provide fluid communication through the tube bundle. A barrier extends radially inside the tube bundle between the upstream and downstream surfaces, and a baffle extends axially inside the tube bundle between the upstream surface and the barrier.

  18. CONSIDERATIONS FOR THE DEVELOPMENT OF A DEVICE FOR THE DECOMMISSIONING OF THE HORIZONTAL FUEL CHANNELS IN THE CANDU 6 NUCLEAR REACTOR. PART 6 - PRESENTATION OF THE DECOMMISSIONING DEVICE

    Directory of Open Access Journals (Sweden)

    Gabi ROSCA FARTAT

    2015-05-01

    Full Text Available The objective of this paper is to present a possible solution for the designing of a device for the decommissioning of the horizontal fuel channels in the CANDU 6 nuclear reactor. The decommissioning activities are dismantling, demolition, controlled removal of equipment, components, conventional or hazardous waste (radioactive, toxic in compliance with the international basic safety standards on radiation protection. One as the most important operation in the final phase of the nuclear reactor dismantling is the decommissioning of fuel channels. For the fuel channels decommissioning should be taken into account the detailed description of the fuel channel and its components, the installation documents history, adequate radiological criteria for decommissioning guidance, safety and environmental impact assessment, including radiological and non-radiological analysis of the risks that can occur for workers, public and environment, the description of the proposed program for decommissioning the fuel channel and its components, the description of the quality assurance program and of the monitoring program, the equipments and methods used to verify the compliance with the decommissioning criteria, the planning of performing the final radiological assessment at the end of the fuel channel decommissioning. These will include also, a description of the proposed radiation protection procedures to be used during decommissioning. The dismantling of the fuel channel is performed by one device which shall provide radiation protection during the stages of decommissioning, ensuring radiation protection of the workers. The device shall be designed according to the radiation protection procedures. The decommissioning device assembly of the fuel channel components is composed of the device itself and moving platform support for coupling of the selected channel to be dismantled. The fuel channel decommissioning device is an autonomous device designed for

  19. Application of Genetic Algorithm methodologies in fuel bundle burnup optimization of Pressurized Heavy Water Reactor

    International Nuclear Information System (INIS)

    Highlights: • We study and compare Genetic Algorithms (GA) in the fuel bundle burnup optimization of an Indian Pressurized Heavy Water Reactor (PHWR) of 220 MWe. • Two Genetic Algorithm methodologies namely, Penalty Functions based GA and Multi Objective GA are considered. • For the selected problem, Multi Objective GA performs better than Penalty Functions based GA. • In the present study, Multi Objective GA outperforms Penalty Functions based GA in convergence speed and better diversity in solutions. - Abstract: The work carried out as a part of application and comparison of GA techniques in nuclear reactor environment is presented in the study. The nuclear fuel management optimization problem selected for the study aims at arriving appropriate reference discharge burnup values for the two burnup zones of 220 MWe Pressurized Heavy Water Reactor (PHWR) core. Two Genetic Algorithm methodologies namely, Penalty Functions based GA and Multi Objective GA are applied in this study. The study reveals, for the selected problem of PHWR fuel bundle burnup optimization, Multi Objective GA is more suitable than Penalty Functions based GA in the two aspects considered: by way of producing diverse feasible solutions and the convergence speed being better, i.e. it is capable of generating more number of feasible solutions, from earlier generations. It is observed that for the selected problem, the Multi Objective GA is 25.0% faster than Penalty Functions based GA with respect to CPU time, for generating 80% of the population with feasible solutions. When average computational time of fixed generations are considered, Penalty Functions based GA is 44.5% faster than Multi Objective GA. In the overall performance, the convergence speed of Multi Objective GA surpasses the computational time advantage of Penalty Functions based GA. The ability of Multi Objective GA in producing more diverse feasible solutions is a desired feature of the problem selected, that helps the

  20. FLASHPOINT - a tool to routinely calculate the heat load in the irradiated fuel bays

    Energy Technology Data Exchange (ETDEWEB)

    Vyskocil, E.; Morrison, C.; Gifford, E.; Inglot, A.; Kozlowski, K.; Gocmanac, M. [AMEC NSS, Reactor and Radiation Physics, Toronto, Ontario (Canada); Parlatan, Y. [Ontario Power Generation, Safety Analysis Improvement Project Dept., Pickering, Ontario (Canada); Alabasha, H. [Bruce Power, Nuclear Safety Analysis and Support, Toronto, Ontario (Canada)

    2013-07-01

    At the recommendation of the World Association of Nuclear Operators (WANO), a tool was developed as an enhancement of NuFLASH (Nuclear Fuel Location and Storage History) in order to routinely calculate the Irradiated Fuel Bay (IFB) heat load. It uses information stored in NuFLASH regarding the location and details of spent fuel bundle properties to calculate the decay power on a bundle by bundle basis and then sum the decay powers of all bundles in a particular IFB. FLASHPOINT employs a two-step approximation of the bundle irradiation history based on the record of the life cycle for each individual fuel bundle. The primary parameter affecting the decay power of any individual irradiated CANDU fuel bundle following its discharge from core is the period of time elapsed since the bundle last operated at power within the reactor. The remaining factors influencing the decay power of an individual fuel bundle concern the irradiation history of that bundle while in core. The accuracy of the FLASHPOINT methodology has been assessed primarily through comparison of results obtained using the two step history representation implemented in FLASHPOINT against results from a more detailed ORIGEN-S calculation of the decay heat based on the SORO power history for a randomly selected sample of bundles. The results for individual bundles and the aggregate group are presented and the accuracy of the two-step approximation is demonstrated to be acceptable. (author)

  1. Advances in the manufacture of clad tubes and components for PHWR fuel bundle

    International Nuclear Information System (INIS)

    Fuel bundles for Pressurized Heavy Water Reactors (PHWRs) consists of Uranium di-oxide pellets encapsulated into thin wall Zircaloy clad tubes. Other components such as end caps, bearing pads and spacer pads are the integral elements of the fuel bundle. As the fuel assembly is subjected to severe operating conditions of high temperature and pressure in addition to continual irradiation exposure, all the components are manufactured conforming to stringent specifications with respect to chemical composition, mechanical & metallurgical properties and dimensional tolerances. The integrity of each component is ensured by NDE at different stages of manufacture. The manufacturing route for fuel tubes and components comprise of a combination of thermomechanical processing and each process step has marked effect on the final properties. The fuel tubes are manufactured by processing the extruded blanks in four stage cold pilgering with intermediate annealing and final stress relieving operation. The bar material is produced by hot extrusion followed by multi-pass swaging and intermediate annealing. Spacer pads and bearing pads are manufactured by blanking and coining of Zircaloy sheet which is made by a combination of hot and cold rolling operations. Due to the small size and stringent dimensional requirements of these appendages, selection of production route and optimization of process parameters are important. This paper discusses about various measures taken for improving the recoveries and mechanical and corrosion properties of the tube, sheet and bar materials being manufactured at Nuclear Fuel Complex, Hyderabad For the production of clad tubes, modifications at extrusion stage to reduce the wall thickness variation, introduction of ultrasonic testing of extruded blanks, optimization of cold working and heat treatment parameters at various stages of production etc. were done. The finished bar material is subjected to 100% Ultrasonic and eddy current testing to ensure

  2. CANDU safety under severe accidents

    International Nuclear Information System (INIS)

    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

  3. CONSIDERATIONS FOR THE DEVELOPMENT OF A DEVICE FOR THE DECOMMISSIONING OF THE HORIZONTAL FUEL CHANNELS IN THE CANDU 6 NUCLEAR REACTOR. PART 9 - CUTTING AND EXTRACTING DEVICE FUNCTIONING

    Directory of Open Access Journals (Sweden)

    Constantin POPESCU

    2015-05-01

    Full Text Available This paper presents a constructive solution proposed by the authors in order to achieve of a cutting and extracting device for the decommissioning of the horizontal fuel channels in the CANDU 6 nuclear reactor. The Cutting and Extraction Device (CED performs the dismantling, cutting and extraction of the fuel channel components. It's a flexible and modular device, which is designed to work inside the fuel channel and has the following functions: moving with variable speed, temperature monitoring and video surveillance inside the pipe, unblock and extract the channel closure plug (from End Fitting - EF, unblock and extract the channel shield plug (from Lattice Tube - LT, block and cut the middle of the pressure tube, block and cut the end of the pressure tube, block and extract the half of pressure tube. The Cutting and Extraction Device (CED consists of following modules: guiding-fixing module, traction modules, cutting module, guiding-extracting module and flexible elements for modules connecting. The guiding-fixing module is equipped with elastic guiding rollers and fixing claws in working position, the traction modules are provided with variable pitch rollers for allowing variable travel speed through the fuel channel. The cutting module is positioned in the middle of the device and it is equipped with three knife rolls for pressure tube cutting, using a system for cutting place video surveillance and pyrometers for monitoring cutting place temperature. The Cutting and Extraction Device (CED is fully automated, connected by wires to a Programmable Logic Controller (PLC and controlled from a Human Machine Interface (HMI. The design of the Cutting and Extraction Device (CED shall be achieved according to the particular features of the fuel channel components to be dismantled and to ensure radiation protection of workers.

  4. Fuel composition optimization in a 78-element fuel bundle for use in a pressure tube type supercritical water-cooled reactor

    International Nuclear Information System (INIS)

    A 78-element fuel bundle containing a plutonium-thorium fuel mixture has been proposed for a Generation IV pressure tube type supercritical water-cooled reactor. In this work, using a lattice cell model created with the code DRAGON,the lattice pitch, fuel composition (fraction of PuO2 in ThO2) and radial enrichment profile of the 78-element bundle is optimized using a merit function and a metaheuristic search algorithm.The merit function is designed such that the optimal fuel maximizes fuel utilization while minimizing peak element ratings and coolant void reactivity. A radial enrichment profile of 10 wt%, 11 wt% and 20 wt% PuO2 (inner to outer ring) with a lattice pitch of 25.0 cm was found to provide the optimal merit score based on the aforementioned criteria. (author)

  5. FEED 1.6: modelling of hydrogen diffusion and precipitation in fuel bundle zircaloy components

    International Nuclear Information System (INIS)

    An as-fabricated Zircaloy component in a CANDU® fuel bundle has certain amount of hydrogen. In addition, the Zircaloy component pickups hydrogen during operation, where sheath oxidation occurs on the water side. Hydrogen content in the Zircaloy component will change due to the diffusion under gradients of concentration and temperature. A hydrostatic stress gradient may also have some effect on hydrogen diffusion. When the local concentration of hydrogen exceeds the terminal solid solubility (TSS), hydrides will start to form (i.e., hydride precipitation). Because hydrides have a negative effect on material properties (e.g., lower ductility), the hydrogen content in Zircaloy sheath needs to be limited to ensure that the sheath strength is not affected. The FEED (Finite Element Estimate for Diffusion) code was developed to predict the local hydrogen concentration and formation of hydride. The FEED 1.6 code has the following capabilities: Model transient Hydrogen/Deuterium (H/D) diffusion in Zircaloy components (e.g., fuel sheath, endcap and endcap weld); Model H/D pickup in Zircaloy sheath; Account for the effect of gradients of concentration, temperature and stress; and, Model transient hydride precipitation and re-dissolutions. This paper describes the FEED 1.6 code, including theory, models, and some validation examples. (author)

  6. FEED 1.6: modelling of hydrogen diffusion and precipitation in fuel bundle zircaloy components

    Energy Technology Data Exchange (ETDEWEB)

    Lai, L.; Xu, Z.; Jiang, Q.; Cheng, G. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)

    2010-07-01

    An as-fabricated Zircaloy component in a CANDU® fuel bundle has certain amount of hydrogen. In addition, the Zircaloy component pickups hydrogen during operation, where sheath oxidation occurs on the water side. Hydrogen content in the Zircaloy component will change due to the diffusion under gradients of concentration and temperature. A hydrostatic stress gradient may also have some effect on hydrogen diffusion. When the local concentration of hydrogen exceeds the terminal solid solubility (TSS), hydrides will start to form (i.e., hydride precipitation). Because hydrides have a negative effect on material properties (e.g., lower ductility), the hydrogen content in Zircaloy sheath needs to be limited to ensure that the sheath strength is not affected. The FEED (Finite Element Estimate for Diffusion) code was developed to predict the local hydrogen concentration and formation of hydride. The FEED 1.6 code has the following capabilities: Model transient Hydrogen/Deuterium (H/D) diffusion in Zircaloy components (e.g., fuel sheath, endcap and endcap weld); Model H/D pickup in Zircaloy sheath; Account for the effect of gradients of concentration, temperature and stress; and, Model transient hydride precipitation and re-dissolutions. This paper describes the FEED 1.6 code, including theory, models, and some validation examples. (author)

  7. Thermal-hydraulic design calculations for the annular fuel element with replaceable test bundles (TOAST) on the test zone position 205 of KNK II/3

    International Nuclear Information System (INIS)

    Annular fuel elements are foreseen in KNK II as carrier elements for irradiation inserts and test bundles. For the third core a reloadable annular element on position 205 is foreseen, in which replaceable 19-pin test bundles (TOAST) shall be irradiated. The present report deals with the thermal-hydraulic design of the annular carrier element and the test bundle, whereby the test bundle required additional optimization. The code CIA has been used for the calculations. Start of irradiation of the subassembly is planned at the beginning of the third core operation. After optimization of the pin-spacer geometry in the test bundle, design calculations for both bundles were performed, whereby thermal coupling between both was taken into account. The calculated mass-flows and temperature distributions are given for the nominal and the eccentric element configuration. The calculated bundle pressure losses have been corrected according to experimental results

  8. Key thrusts in next generation CANDU. Annex 10

    International Nuclear Information System (INIS)

    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. PARTICULARITIES REGARDING THE OPERATING PROCESS OF THE CUTTING AND EXTRACTION DEVICE IN THE CANDU HORIZONTAL FUEL CHANNELS PRESSURE TUBE DECOMMISSIONING PART II: CUTTING AND EXTRACTING PRESSURE TUBE PROCESS

    Directory of Open Access Journals (Sweden)

    Constantin POPESCU

    2016-05-01

    Full Text Available This paper presents some details of operation process for a Cutting and Extraction Device (CED in order to achieve the decommissioning of the horizontal fuel channels pressure tube in the CANDU 6 nuclear reactor. The most important characteristic of the Cutting and Extraction Device (CED is his capability of totally operator’s protection against the nuclear radiation during pressure tube decommissioning. The cutting and extracting pressure tube processes present few particularities due to special adopted technical solutions: a special module with three cutting rollers (system driven by an actuator, a guiding-extracting and connecting module (three fixing claws which are piloted by an actuator and block the device in the connecting position with extracting plugs. The Cutting and Extraction Device (CED is a train of modules equipped with special systems to be fully automated, connected with a Programmable Logic Controller (PLC and controlled by an operator panel type Human Machine Interface (HMI. All processes are monitored by video cameras. In case of error, the process is automatically stopped, the operator receiving an error message and the last sequence could be reinitialized or aborted due to safety reasons.

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

    International Nuclear Information System (INIS)

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

  11. Heat Transfer Enhancement By Three-Dimensional Surface Roughness Technique In Nuclear Fuel Rod Bundles

    Science.gov (United States)

    Najeeb, Umair

    This thesis experimentally investigates the enhancement of single-phase heat transfer, frictional loss and pressure drop characteristics in a Single Heater Element Loop Tester (SHELT). The heater element simulates a single fuel rod for Pressurized Nuclear reactor. In this experimental investigation, the effect of the outer surface roughness of a simulated nuclear rod bundle was studied. The outer surface of a simulated fuel rod was created with a three-dimensional (Diamond-shaped blocks) surface roughness. The angle of corrugation for each diamond was 45 degrees. The length of each side of a diamond block is 1 mm. The depth of each diamond block was 0.3 mm. The pitch of the pattern was 1.614 mm. The simulated fuel rod had an outside diameter of 9.5 mm and wall thickness of 1.5 mm and was placed in a test-section made of 38.1 mm inner diameter, wall thickness 6.35 mm aluminum pipe. The Simulated fuel rod was made of Nickel 200 and Inconel 625 materials. The fuel rod was connected to 10 KW DC power supply. The Inconel 625 material of the rod with an electrical resistance of 32.3 kO was used to generate heat inside the test-section. The heat energy dissipated from the Inconel tube due to the flow of electrical current flows into the working fluid across the rod at constant heat flux conditions. The DI water was employed as working fluid for this experimental investigation. The temperature and pressure readings for both smooth and rough regions of the fuel rod were recorded and compared later to find enhancement in heat transfer coefficient and increment in the pressure drops. Tests were conducted for Reynold's Numbers ranging from 10e4 to 10e5. Enhancement in heat transfer coefficient at all Re was recorded. The maximum heat transfer co-efficient enhancement recorded was 86% at Re = 4.18e5. It was also observed that the pressure drop and friction factor increased by 14.7% due to the increased surface roughness.

  12. Marketing CANDU internationally

    International Nuclear Information System (INIS)

    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.)

  13. CANDU in-reactor quantitative visual-based inspection techniques

    Science.gov (United States)

    Rochefort, P. A.

    2009-02-01

    This paper describes two separate visual-based inspection procedures used at CANDU nuclear power generating stations. The techniques are quantitative in nature and are delivered and operated in highly radioactive environments with access that is restrictive, and in one case is submerged. Visual-based inspections at stations are typically qualitative in nature. For example a video system will be used to search for a missing component, inspect for a broken fixture, or locate areas of excessive corrosion in a pipe. In contrast, the methods described here are used to measure characteristic component dimensions that in one case ensure ongoing safe operation of the reactor and in the other support reactor refurbishment. CANDU reactors are Pressurized Heavy Water Reactors (PHWR). The reactor vessel is a horizontal cylindrical low-pressure calandria tank approximately 6 m in diameter and length, containing heavy water as a neutron moderator. Inside the calandria, 380 horizontal fuel channels (FC) are supported at each end by integral end-shields. Each FC holds 12 fuel bundles. The heavy water primary heat transport water flows through the FC pressure tube, removing the heat from the fuel bundles and delivering it to the steam generator. The general design of the reactor governs both the type of measurements that are required and the methods to perform the measurements. The first inspection procedure is a method to remotely measure the gap between FC and other in-core horizontal components. The technique involves delivering vertically a module with a high-radiation-resistant camera and lighting into the core of a shutdown but fuelled reactor. The measurement is done using a line-of-sight technique between the components. Compensation for image perspective and viewing elevation to the measurement is required. The second inspection procedure measures flaws within the reactor's end shield FC calandria tube rolled joint area. The FC calandria tube (the outer shell of the FC) is

  14. Thermal-hydraulic analysis of flow blockage in a supercritical water-cooled fuel bundle with sub-channel code

    International Nuclear Information System (INIS)

    Highlights: • COBTA-SC code shows good suitability for the blockage analysis of SCWR fuel bundle. • Several thermal-hydraulic models are incorporated and evaluated for the flow blockage of SCWR-FQT bundle. • The axial/circumferential heat conduction of fuel and heat transfer correlation are identified as the important models. • The peak cladding temperature can be reduced effectively by the safety measures of SCWR-FQT. - Abstract: Sub-channel code is nowadays the most applied method for safety analysis and thermal-hydraulic simulation of fuel assembly. It plays an indispensable role to predict the detail thermal-hydraulic behavior of the supercritical water-cooled reactor (SCWR) fuel assembly because of the strong non-uniformity within the fuel bundle. Since the coolant shows a strong variation of physical thermal property near the pseudo critical line, the local blockage in an assembly of a SCWR is of importance to safety analysis. Due to the low specific heat of supercritical water with high temperature, the blockage and the subsequent flow reduction at the downstream of the blockage will yield particular high cladding temperature. To analyze the local thermal-hydraulic parameters in the supercritical water reactor-fuel qualification test (SCWR-FQT) fuel bundle with a flow blockage caused by detachment of the wire wrap, the sub-channel code COBRA-SC is unitized. The code is validated by some blockage experiments, and it reveals a good feasibility and accuracy for the SCWR and blockage flow analysis. Some new models, e.g. the axial and circumferential heat conduction model, turbulent mixing models, pressure friction models and heat transfer correlations, are incorporated in COBRA-SC code. And their influence on the cladding temperature and mass flow distribution are evaluated and discussed. Based on the results, the appropriate models for description of the flow blockage phenomenon in SCWR assembly is identified and recommended. A transient analysis of the

  15. Validation study of thermal-hydraulic analysis program spiral for fuel pin bundle of sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Full text of publication follows: Japan Nuclear Cycle Development Institute (JNC) has been developing a numerical simulation system in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel subassemblies of sodium-cooled fast reactors under various operating conditions such as normal operation, transient condition or deformed geometry condition from the viewpoint of the assessment of fuel pin structure integrity. This paper describes the validation study of SPIRAL that is one component code of the numerical simulation system and contributes to detailed simulations of local flow and temperature fields in a wire-wrapped fuel pin bundle. SPIRAL is a multi-dimensional finite element method code that can treat complicated geometries like a fuel pin bundle. For numerical stabilization, one can choose Streamline Upwind Petrov Galerkin method and Balancing Tensor Diffusivity method. Semi-implicit solution scheme (fractional step method) developed by Ramaswamy is used for time integration. As the pressure equation matrix solver, ICCG or Gaussian elimination is applied. Energy conservation equations of coolant and structure are also solved and therefore temperature distributions of both coolant and fuel pins can be calculated. Several turbulence models, high/low Reynolds number isotropic/anisotropic models, were incorporated to the code. The code was parallelized using MPI for enhancing simulation efficiency. Pre-processor is also available for numerical grid generation for wire-wrapped fuel pin bundles by curvilinear coordinate system. Fundamental validity related to solving mass, momentum and energy conservation equations and applicability of turbulence models were confirmed by simulating several basic problems. As typical examples, two kinds of simulations using high Re number models, backward facing step flow and 4- fuel-pin bundle in rectangular duct, are introduced in this paper. The simulation results indicate that RNG k-ε model shows relatively

  16. Integrated Planar Solid Oxide Fuel Cell: Steady-State Model of a Bundle and Validation through Single Tube Experimental Data

    Directory of Open Access Journals (Sweden)

    Paola Costamagna

    2015-11-01

    Full Text Available This work focuses on a steady-state model developed for an integrated planar solid oxide fuel cell (IP-SOFC bundle. In this geometry, several single IP-SOFCs are deposited on a tube and electrically connected in series through interconnections. Then, several tubes are coupled to one another to form a full-sized bundle. A previously-developed and validated electrochemical model is the basis for the development of the tube model, taking into account in detail the presence of active cells, interconnections and dead areas. Mass and energy balance equations are written for the IP-SOFC tube, in the classical form adopted for chemical reactors. Based on the single tube model, a bundle model is developed. Model validation is presented based on single tube current-voltage (I-V experimental data obtained in a wide range of experimental conditions, i.e., at different temperatures and for different H2/CO/CO2/CH4/H2O/N2 mixtures as the fuel feedstock. The error of the simulation results versus I-V experimental data is less than 1% in most cases, and it grows to a value of 8% only in one case, which is discussed in detail. Finally, we report model predictions of the current density distribution and temperature distribution in a bundle, the latter being a key aspect in view of the mechanical integrity of the IP-SOFC structure.

  17. The CANDU contribution to environmentally friendly energy production

    International Nuclear Information System (INIS)

    National prosperity is based on the availability of affordable, energy supply. However, this need is tempered by a complementary desire that the energy production and utilization will not have a major impact on the environment. The CANDU energy system, including a next generation of CANDU designs, is a major primary energy supply option that can be an important part of an energy mix to meet Canadian needs. CANDU nuclear power plants produce energy in the form of medium pressure steam. The advanced version of the CANDU design can be delivered in unit modules ranging from 400 to 1200 MWe. This Next Generation of CANDU designs features lower cost, coupled with robust safety margins. Normally this steam is used to drive a turbine and produce electricity. However, a fraction of this steam (large or small) may alternatively be used as process steam for industrial consumption. Options for such steam utilization include seawater desalination, oil sands extraction and heating. The electricity may be delivered to an electrical grid or alternatively used to produce quantities of hydrogen. Hydrogen is an ideal clean transportation fuel because its use only produces water. Thus, a combination of CANDU generated electricity and hydrogen distribution for vehicles is an available, cost-effective route to dramatically reduce emissions from the transportation sector. The CANDU energy system contributes to environmental protection and the prevention of climate change because of its very low emission. The CANDU energy system does not produce any NOx, SOx or greenhouse gas (notably CO2) emissions during operation. In addition, the CANDU system operates on a fully closed cycle with all wastes and emissions fully monitored, controlled and managed throughout the entire life cycle of the plant. The CANDU energy system is an environmentally friendly and flexible energy source. It can be an effective component of a total energy supply package, consistent with Canadian and global climate

  18. Fuel cell integral bundle assembly including ceramic open end seal and vertical and horizontal thermal expansion control

    Science.gov (United States)

    Zafred, Paolo R.; Gillett, James E.

    2012-04-24

    A plurality of integral bundle assemblies contain a top portion with an inlet fuel plenum and a bottom portion containing a base support, the base supports a dense, ceramic air exhaust manifold having four supporting legs, the manifold is below and connects to air feed tubes located in a recuperator zone, the air feed tubes passing into the center of inverted, tubular, elongated, hollow electrically connected solid oxide fuel cells having an open end above a combustion zone into which the air feed tubes pass and a closed end near the inlet fuel plenum, where the open end of the fuel cells rest upon and within a separate combination ceramic seal and bundle support contained in a ceramic support casting, where at least one flexible cushion ceramic band seal located between the recuperator and fuel cells protects and controls horizontal thermal expansion, and where the fuel cells operate in the fuel cell mode and where the base support and bottom ceramic air exhaust manifolds carry from 85% to all of the weight of the generator.

  19. Influence of the flux axial form on the conversion rate and duration of cycle between recharging for ThPu and U{sub nat} fuels in CANDU reactors; Influence de la forme axiale du flux sur le taux de conversion et la duree du cycle entre rechargements pour du combustible ThPu et U{sub nat} dans les reacteurs CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Chambon, Richard [Laboratoire de Physique Subatomique et de Cosmologie, Universite Joseph Fourier / CNRS-IN2P3, 53 Avenue des Martyrs, F-38026 Grenoble (France)

    2007-01-15

    To face the increasing world power demand the world nuclear sector must be continuously updated and developed as well. Thus reactors of new types are introduced and advanced fuel cycles are proposed. The technological and economic feasibility and the transition of the present power park to a renewed park require thorough studies and scenarios, which are highly dependent on the reactor performances. The conversion rate and cycle span between recharging are important parameters in the scenarios studies. In this frame, we have studied the utilisation of thorium in the CANDU type reactors and particularly the influence of axial form of the flux, i.e. of the recharging mode, on the conversion rate and duration of the cycle between recharging. The results show that up to a first approximation the axial form of the flux resulting from the neutron transport calculations for assessing the conversion rate is not necessary to be taken into account. However the time span between recharging differs up to several percents if the axial form of the flux is taken into consideration in transport calculations. Thus if the burnup or the recharging frequency are parameters which influence significantly the deployment scenarios of a nuclear park an approach more refined than a simple transport evolution in a typical cell/assembly is recommended. Finally, the results of this study are not more general than for the assumed conditions but they give a thorough calculation method valid for any recharging/fuel combination in a CANDU type reactor.

  20. Simulation of the fuel rod bundle test QUENCH-03 using the integral code ASTEC V2

    Energy Technology Data Exchange (ETDEWEB)

    Kruse, Philipp; Koch, Marco K. [Bochum Univ. (Germany). Chair of Energy Systems and Energy Economics

    2010-05-15

    Failure of the main and emergency cooling-systems can lead to an accident with severe core degradation even with core meltdown. To prevent total meltdown of the uncovered and overheated core reflooding with water is an unavoidable accident management measure. The fast supply of water and the resulting increased available amount of steam can lead to crack formations and break up of the oxide layer of the fuel rods. The additionally exposed surface could result in an increased release of hydrogen due to a supplementary exothermal zirconium- steam-oxidation reaction. Within the frame of the QUENCH test-program - realised by FZK - loss of coolant accidents in LWR (Light Water Reactor) are analysed using an experimental reactor core to determine the produced amount of hydrogen, the so-called hydrogen source term. Additionally, the behaviour of the bundle with different absorber rod and cladding materials is being analysed. Based on the post-test calculations of the QUENCH tests with the severe accident code system ASTEC the capability of the code can be established and evaluated. In the following the post-test calculations of the QUENCH-03 test with ASTEC V2 are discussed. (orig.)

  1. Ultrasonic systems for high-accuracy thickness measurement of fuel bundle bearing pads and shield plug crimps

    International Nuclear Information System (INIS)

    The performance of two ultrasonic systems, remotely operated in high radiation environment, are presented. The first system is used to measure the bearing pad height of radioactive fuel bundles located in the irradiated fuel bays, at Darlington NGS. The system was designed and commissioned to achieve an accuracy of ± 20 μm. The repeatability of results is within ± 10 μm uniformity band. The measurements are independent of testing speed, water temperature, bundle temperature, pencil geometry. Possibilities and limitations of the UT system are also presented and some improved alternatives are proposed. The second system was developed for measuring the crimp height of shield plugs (special iron casting) at Bruce B - Mark Ill development. The accuracy of measurements is ± 50 μm, with a repeatability of ± 25 μm. The results are independent of shield plug thickness variation and ovality, crimp off-set and heavy-water temperature. (author)

  2. Methodology Improvement of Reactor Physics Codes for CANDU Channels Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Myung Hyun; Choi, Geun Suk; Win, Naing; Aung, Tharndaing; Baek, Min Ho; Lim, Jae Yong [Kyunghee University, Seoul (Korea, Republic of)

    2010-04-15

    As the operational time increase, pressure tubes and calandria tubes in CANDU core encounter inevitably a geometrical deformation along the tube length. A pressure tube may be sagged downward within a calandria tube by creep from irradiation. This event can bring about a problem that is serious in integrity of pressure tube. A measurement of deflection state of in-service pressure tube is, therefore, very important for the safety of CANDU reactor. In this paper, evaluation of impacts on nuclear characteristic due to fuel channel deformation were aimed in order to improve nuclear design tools for concerning the local effects from abnormal deformations. It was known that sagged pressure tube can cause the eccentric configuration of fuel bundles in pressure tube by O.6cm maximum. In this case, adverse pin power distribution and reactivity balance can affect reactor safety under normal and accidental condition. Thermal and radiation-induced creep in pressure tube would expand a tube size. It was known that maximum expansion may be 5% in volume. In this case, more coolant make more moderation in the deformed channel resulting in the increase of reactivity. Sagging of pressure tube did not cause considerable change in K-inf values. However, expansion of the pressure tube made relatively large change in K-inf. Modeling of eccentric and enlarged configuration is not easy in preparation of input geometry at both HELlOS and MCNP. On the other hand, there is no way to consider this deformation in one-dimensional homogenization tool such as WIMS code. The way of handling this deformation was suggested as the correction method of expansion effect by adjusting the number density of coolant. The number density of heavy water coolant was set to be increased as the rate of expansion increase. This correction was done in the intact channel without changing geometry. It was found that this correction was very effective in the prediction of K-inf values. In this study, further

  3. Determination of transient radial-azimuthal temperature distributions in fuel bundles under loss-of-coolant-accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Saltos, N.T.; Christensen, R.N.; Aldemir, T.

    1988-10-01

    A methodology is presented to determine the transient temperature distributions in fuel bundles under loss-of-coolant accident (LOCA) conditions using a recently developed variational technique for the solution of radial-azimuthal heat conduction in the fuel rods and the modified view factor concept proposed by Uchida and Nakamure to model the radiative heat transfer between the rods. The variational technique is based on the Lebron-Labermont restricted variational principle and represents the temperature distribution in the rods at a given time during the LOCA via parabolic and circular trial functions in the radial and azimuthal directions, respectively. The methodology is implemented to a 4 x 4 boiling water reactor fuel bundle under typical LOCA conditions to investigate the effects of changes in rod heat transfer characteristics and simplifying modeling assumptions on predicted rod temperature distributions. The results show that these effects depend on the rod location in the assembly and LOCA phase under consideration and indicate that same degree of modelling detail may not be necessary for all the rods in the bundle at all times during the LOCA.

  4. 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.

  5. Ontario Hydro CANDU operating experience

    International Nuclear Information System (INIS)

    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

  6. Ontario Hydro CANDU operating experience

    International Nuclear Information System (INIS)

    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

  7. Conceptual Study on Dismantling of CANDU Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Woo-Tae; Lee, Sang-Guk [KHNP-CRI, Daejeon (Korea, Republic of)

    2014-10-15

    In this paper, we reviewed 3D design model of the CANDU type reactor and suggested feasible cutting scheme. The structure of CANDU nuclear reactor, the calandria assembly was reviewed using 3-D CAD model for future decommissioning. Through the schematic diagram of CANDU nuclear power plant, we identified the differences between PWR and CANDU reactor assembly. Method of dismantling the fuel channels from the calandria assembly was suggested. Custom made cutter is recommended to cut all the fuel channels. The calandria vessel is recommended to be cut by band saw or plasma torch. After removal of the fuel channels, it was assumed that radiation level near the calandria vessel is not very high. For cutting of the end shields, various methods such as band saw, plasma torch, CAMC could be used. The choice of a specific method is largely dependent on radiological environment. Finally, method of cutting the embedment rings is considered. As we assume that operators could cut the rings without much radiation exposure, various industrial cutting methods are suggested to be applied. From the above reviews, we could conclude that decommissioning of CANDU reactor is relatively easy compared to that of PWR reactor. Technologies developed from PWR reactor decommissioning could be applied to CANDU reactor dismantling.

  8. Development of computational technology on heat transfer and fluid flow in a nuclear fuel bundle of advanced reactor

    International Nuclear Information System (INIS)

    The assessment of the RANS(Reynolds-Averaged Navier-Stokes) based turbulence model was conducted to establish the optimal CFD system for turbulent flow and heat transfer in reactor during the first year of the project. The RANS models used in this project are the two-equation models based on the eddy viscosity assumption and the Second-Moment Closure(SMC) models. Since the nuclear fuel assembly loaded in the nuclear reactor is a rod bundle which is square or triangular array, the predictions using the various turbulence models were compared for turbulent flow in bare square and/or triangular rod bundle and the rod bundle with the flow mixing vane. The study for the second year of the project examined the CFD model and the applicability of the CFD code for the turbulent two-phase flow. The numerical predictions of lateral distributions of void fraction, phasic velocities and turbulent kinetic energy were compared against the experimental results for upward and downward bubbly flow in a vertical tube. The boiling flows in vertical tube and rod bundle were also simulated to verify the CFD results

  9. OECD/NRC PSBT Benchmark: Investigating the CATHARE2 Capability to Predict Void Fraction in PWR Fuel Bundle

    Directory of Open Access Journals (Sweden)

    A. Del Nevo

    2012-01-01

    Full Text Available Accurate prediction of steam volume fraction and of the boiling crisis (either DNB or dryout occurrence is a key safety-relevant issue. Decades of experience have been built up both in experimental investigation and code development and qualification; however, there is still a large margin to improve and refine the modelling approaches. The qualification of the traditional methods (system codes can be further enhanced by validation against high-quality experimental data (e.g., including measurement of local parameters. One of these databases, related to the void fraction measurements, is the pressurized water reactor subchannel and bundle tests (PSBT conducted by the Nuclear Power Engineering Corporation (NUPEC in Japan. Selected experiments belonging to this database are used for the OECD/NRC PSBT benchmark. The activity presented in the paper is connected with the improvement of current approaches by comparing system code predictions with measured data on void production in PWR-type fuel bundles. It is aimed at contributing to the validation of the numerical models of CATHARE 2 code, particularly for the prediction of void fraction distribution both at subchannel and bundle scale, for different test bundle configurations and thermal-hydraulic conditions, both in steady-state and transient conditions.

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

    International Nuclear Information System (INIS)

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

  11. Candu technology: the next generation now

    International Nuclear Information System (INIS)

    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

  12. Neutronics-thermalhydraulics coupling in a CANDU SCWR

    Science.gov (United States)

    Adouki, Pierre

    In order to implement new nuclear technologies as a solution to the growing demand for energy, 10 countries agreed on a framework for international cooperation in 2002, to form the Generation IV International Forum (GIF). The goal of the GIF is to design the next generation of nuclear reactors that would be cost effective and would enhance safety. This forum has proposed several types of Generation IV reactors including the Supercritical Water-Cooled Reactor (SCWR). The SCWR comes in two main configurations: pressure vessel SCWR and pressure tube SCWR (PT-SCWR). In this study, the CANDU SCWR (a PT-SCWR) is considered. This reactor is oriented vertically and contains 336 channels with a length of 5 m. The target coolant inlet and outlet temperatures are 350 Celsius and 625 Celsius, respectively. The coolant flows downwards, and the reactor power is 2540 MWth. Various fuel designs have been considered in order not to exceed the linear element rating. However, the dependency between the core power and thermalhydraulics parameters results in the necessity to use a neutronics/thermalhydaulics coupling scheme to determine the core power and the thermalhydraulics parameters. The core power obtained has a power peaking factor of 1.4. The bundle power distribution for all channels has a peak at the third bundle from the inlet, but the value of this peak increases with the channel power. The heat-transfer coefficient and the specific-heat capacity have a peak at the same location in a channel, and this location shifts toward the inlet as the channel power increases. The exit coolant temperature increases with the channel power, while the exit coolant density and pressure decrease with the channel power. Also, higher channel powers lead to higher fuel and cladding temperatures. Moreover, as the coupling method is applied, the effective multiplication factor and the values of thermalhydaulics parameters oscillate as they converge.

  13. Studying the vibration and random hydrodynamic loads on the fuel rods bundles in the fuel assemblies of the reactor installations used at nuclear power stations equipped with VVER reactors

    Science.gov (United States)

    Solonin, V. I.; Perevezentsev, V. V.

    2012-05-01

    Random hydrodynamic loads causing vibration of fuel rod bundles in a turbulent flow of coolant are obtained from the results of pressure pulsation measurements carried out over the perimeter of the external row of fuel rods in the bundle of a full-scale mockup of a fuel assembly used in a second-generation VVER-440 reactor. It is shown that the turbulent flow structure is a factor determining the parameters of random hydrodynamic loads and the vibration of fuel rod bundles excited by these loads. The results from a calculation of random hydrodynamic loads are used for estimating the vibration levels of fuel rod bundles used in prospective designs of fuel assemblies for VVER reactors.

  14. Predictions of Critical Heat Flux Using the ASSERT-PV Subchannel Code for a CANFLEX Variant Bundle

    International Nuclear Information System (INIS)

    The ASSERT-PV subchannel code developed by AECL has been applied as a design-assist tool to the advanced CANDU1 reactor fuel bundle. Based primarily on the CANFLEX2 fuel bundle, several geometry changes (such as element sizes and pitchcircle diameters of various element rings) were examined to optimize the dryout power and pressure-drop performances of the new fuel bundle. An experiment was performed to obtain dryout power measurements for verification of the ASSERT-PV code predictions. It was carried out using an electrically heated, Refrigerant-134a cooled, fuel bundle string simulator. The axial power profile of the simulator was uniform, while the radial power profile of the element rings was varied simulating profiles in bundles with various fuel compositions and burn-ups. Dryout power measurements are predicted closely using the ASSERT-PV code, particularly at low flows and low pressures, but are overpredicted at high flows and high pressures. The majority of data shows that dryout powers are underpredicted at low inlet-fluid temperatures but overpredicted at high inlet-fluid temperatures

  15. Investigations of flow and temperature field development in bare and wire-wrapped reactor fuel pin bundles cooled by sodium

    International Nuclear Information System (INIS)

    Highlights: ► We study sodium flow and temperature development in fuel pin bundles. ► Pin diameter, number of pins, wire wrap and ligament gap are varied as parameters. ► Flow development is achieved within ∼30–40 hydraulic diameters. ► Thermal development is attained only for small pin diameter and less number of pins. ► Wire wrap and ligament gap strongly influence Nusselt number. - Abstract: Simultaneous development of liquid sodium flow and temperature fields in the heat generating pin bundles of reactor has been investigated. Development characteristics are seen to be strongly influenced by pin diameter, number of pins, helical wire-wrap, ligament gap between the last row of pins and hexcan wall and Reynolds number. Flow development is achieved within an axial length of ∼125 hydraulic diameters, for all the pin bundle configurations considered. But temperature development is attained only if the pin diameter is small or the number of pins is less. In the case of large pin diameter with more pins, temperature development could not be achieved even after a length of ∼1000 hydraulic diameters. The reason for this behavior is traced to be the weak communication among sub-channels in tightly packed bundles. It is seen that the pin Nusselt number decreases from center to periphery in a bundle. Also, if the ligament gap is narrow, the Nusselt number is large and more uniform. Flow development length is short if the Reynolds number is large and the converse is true for thermal development length. Helical wire-wrap shortens the thermal entry length and significantly enhances the global Nusselt number. But, its influence on hydrodynamic entry length is not significant

  16. Feasibility evaluation of x-ray imaging for measurement of fuel rod bowing in CFTL test bundles

    International Nuclear Information System (INIS)

    The Core Flow Test Loop (CFTL) is a high temperature, high pressure, out-of-reactor helium-circulating system. It is designed for detailed study of the thermomechanical performance, at prototypic steady-state and transient operating conditions, of electrically heated rods that simulate segments of core assemblies in the Gas-Cooled Fast Breeder reactor demonstration plant. Results are presented of a feasibility evaluation of x-ray imaging for making measurements of the displacement (bowing) of fuel rods in CFTL test bundles containing electrically heated rods. A mock-up of a representative CFTL test section consisting of a test bundle and associated piping was fabricated to assist in this evaluation

  17. SEU blending project, concept to commercial operation, Part 3: production of powder for demonstration irradiation fuel bundles

    International Nuclear Information System (INIS)

    The processes for production of Slightly Enriched Uranium (SEU) dioxide powder and Blended Dysprosium and Uranium (BDU) oxide powder that were developed at laboratory scale at Cameco Technology Development (CTD), were implemented and further optimized to supply to Zircatec Precision Industries (ZPI) the quantities required for manufacturing twenty six Low Void Reactivity (LVRF) CANFLEX fuel bundles. The production of this new fuel was a challenge for CTD and involved significant amount of work to prepare and review documentation, develop and approve new analytical procedures, and go through numerous internal reviews and audits by Bruce Power, CNSC and third parties independent consultants that verified the process and product quality. The audits were conducted by Quality Assurance specialists as well as by Human Factor Engineering experts with the objective to systematically address the role of human errors in the manufacturing of New Fuel and confirm whether or not a credible basis had been established for preventing human errors. The project team successfully passed through these audits. The project management structure that was established during the SEU and BDU blending process development, which included a cross-functional project team from several departments within Cameco, maintained its functionality when Cameco Technology Development was producing the powder for manufacturing Demonstration Irradiation fuel bundles. Special emphasis was placed on the consistency of operating steps and product quality certification, independent quality surveillance, materials segregation protocol, enhanced safety requirements, and accurate uranium accountability. (author)

  18. 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.

  19. Application of a Zircaloy/Steam Oxidation Model to a CFX-10 Code for its Validation against a CANDU Fuel Channel Experiment: CS28-2

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyoung Tae; Park, Joo Hwan; Rhee, Bo Wook

    2007-11-15

    Oxidation of a Zircaloy cladding exposed to high-temperature steam is an important phenomenon in the safety analysis of CANDU reactors during a postulated loss-of-coolant accident (LOCA), since a Zircaloy/steam reaction is highly exothermic and results in hydrogen production.

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

    International Nuclear Information System (INIS)

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

  1. 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.

  2. Plutonium Consumption Program, CANDU Reactor Project final report

    International Nuclear Information System (INIS)

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

  3. Coolability of severely degraded CANDU cores. Revised

    Energy Technology Data Exchange (ETDEWEB)

    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)

  4. Survey of considerations involved in introducing CANDU reactors into the United States

    Energy Technology Data Exchange (ETDEWEB)

    Till, C E; Bohn, E M; Chang, Y I; van Erp, J B

    1977-01-01

    The important issues that must be considered in a decision to utilize CANDU reactors in the U.S. are identified in this report. Economic considerations, including both power costs and fuel utilization, are discussed for the near and longer term. Safety and licensing considerations are reviewed for CANDU-PHW reactors in general. The important issues, now and in the future, associated with power generation costs are the capital costs of CANDUs and the factors that impact capital cost comparisons. Fuel utilization advantages for the CANDU depend upon assumptions regarding fuel recycle at present, but the primary issue in the longer term is the utilization of the thorium cycle in the CANDU. Certain safety features of the CANDU are identified as intrinsic to the concept and these features must be examined more fully regarding licensability in the U.S.

  5. Absorber materials in CANDU PHWRs

    International Nuclear Information System (INIS)

    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 the relatively benign environment of low pressure, low temperature heavy water between neighbouring rows or 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 redesigned back-fit resolved the problem. (author). 3 refs, 8

  6. Beyond designed functional margins in CANDU type NPP. Radioactive nuclei assessment in an LOCA type accident

    Directory of Open Access Journals (Sweden)

    Budu Andrei Razvan

    2015-01-01

    Full Text Available European Union's energy roadmap up to year 2050 states that in order to have an efficient and sustainable economy, with minimum or decreasing greenhouse gas emissions, along with use of renewable resources, each constituent state has the option for nuclear energy production as one desirable option. Every scenario considered for tackling climate change issues, along with security of supply positions the nuclear energy as a recommended option, an option that is highly competitive with respect to others. Nuclear energy, along with other renewable power sources are considered to be the main pillars in the energy sector for greenhouse gas emission mitigation at European level. European Union considers that nuclear energy must be treated as a highly recommended option since it can contribute to security of energy supply. Romania showed excellent track-records in operating in a safe and economically sound manner of Cernavoda NPP Units 1&2. Both Units are in top 10 worldwide in terms of capacity factor. Due to Romania's need to ensure the security of electricity supply, to meet the environmental targets and to move to low carbon generation technologies, Cernavoda Units 3&4 Project appears as a must. This Project was started in 2010 and it is expected to have the Units running by 2025. Cost effective and safety operation of a Nuclear Power Plant is made taking into consideration functional limits of its equipment. As common practice, every nuclear reactor type (technology used is tested according to the worse credible accident or equipment failure that can occur. For CANDU type reactor, this is a Loss of Cooling Accident (LOCA. In a LOCA type accident in a CANDU NPP, using RELAP/SCDAP code for fuel bundle damage assessment the radioactive nuclei are to be quantified. Recently, CANDU type NPP accidents are studied using the RELAP/SCDAP code only. The code formerly developed for PWR type reactors was adapted for the CANDU geometry and can assess the

  7. Heavy water cycle in the CANDU reactor

    International Nuclear Information System (INIS)

    Hydrogen atom has two isotopes: deuterium 1H2 and tritium 1H3. The deuterium oxide D2O is called heavy water due to its density of 1105.2 Kg/m3. 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 D2O 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, D2O management is required to preserve it. (author)

  8. Evaluation of radiation source term for the DUPIC fuel core

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok; Ryu, Ho Jin; Park, Chang Je

    2004-12-01

    The radiation source term of the DUPIC fuel CANDU reactor was estimated for the total and gap inventories of fission products. The calculation was performed by the ELESTRES code using fuel burnup and linear power distribution obtained from the reactor physics calculations. The radiation source term represented by the fission products gap inventory was 46912 TBq for the 1/4 DUPIC fuel core, while it was 75448 TBq for the natural uranium core. Such a reduction of the radiation source term for the DUPIC fuel core can be attributed to the lower linear power of the DUPIC fuel bundle caused by the flattened power distribution of the DUPIC fuel core which adopts a 2-bundle shift refueling scheme. It is therefore expected that the consequence of the loss of coolant accident for the DUPIC fuel core could be weak when compared to the natural uranium core from the viewpoint of radiation doses to the public.

  9. CANDU at the crossroads

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1990-11-01

    ''Ready for the challenge of the 90s'' was the theme of this year's gathering of the Canadian Nuclear Association held in Toronto, 3-6 June. What that challenge really entails is whether the CANDU system will survive as the last remaining alternative to the light water reactor in the world reactor market, or whether it will decline into oblivion along with the Advanced Gas Cooled reactor and so many other technically excellent systems which have fallen along the way. The fate of the CANDU system will not be determined by its technical merits, nor by its impeccable safety record. It will be determined by public perceptions and by the deliberations of an Environmental Assessment Panel established by the Government of Ontario. The debate at the Association meeting is reported. (author).

  10. Behavior of AgInCd absorber material in Zry/UO{sub 2} fuel rod simulator bundles tested at high temperatures in the CORA facility

    Energy Technology Data Exchange (ETDEWEB)

    Sepold, L.; Hagen, S.; Hofmann, P.; Schanz, G.

    2009-01-15

    The CORA experiments carried out in an out-of-pile facility at the Kernforschungszentrum Karlsruhe (KfK), Federal Republic of Germany, are part of the ''Severe Fuel Damage'' (SFD) program. The experimental program is to provide information on the failure mechanisms of Light Water Reactor (LWR) fuel elements in a temperature range from 1200 C to 2000 C and in a few cases up to 2400 C. In the CORA experiments two different bundle configurations are tested: PWR (Pressurized Water Reactor) and BWR (Boiling Water Reactor) bundles. The PWR-type assemblies usually consist of 25 rods with 16 electrically heated fuel rod simulators and nine unheated rods (full-pellet and absorber rods). Bundle CORA-5 contained one Ag/In/Cd - steel absorber rod whereas two absorber rods were used in CORA-12, CORA-15, and CORA-9. The larger bundle CORA-7 contained 5 absorber rods. CORA-12 was terminated by quenching with water from the bottom. In CORA-15 the heated and unheated rods were pressurized to achieve pronounced clad ballooning. Bundle CORA-9 was tested with a system pressure of 1.0 MPa instead of 0.22 MPa. The test bundles were subjected to temperature transients of a slow heatup rate in a steam environment. Thus, an accident sequence is simulated, which may develop from a small-break loss-of-coolant accident of a LWR. The transient phases of the tests were initiated with a temperature ramp rate of 1 K/s. The temperature escalation due to the exothermal zircaloy (Zry)-steam reaction started at about 1100 C, leading the bundles to maximum temperatures of approximately 2000 C. Rod destruction started with the failure of the absorber rod cladding at about 1200 C, i.e. about 250 K below the melting regime of steel. Penetration of the steel cladding was presumably caused by a eutectic interaction between steel and the zircaloy guide tube. The test bundles resulted in severe oxidation and partial melting of the cladding, fuel dissolution by Zry/UO{sub 2} interaction

  11. 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.

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

    International Nuclear Information System (INIS)

    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

  13. A neutronic study of the cycle PWR-CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Alberto da; Pereira, Claubia; Veloso, Maria Auxiliadora Fortini; Fortini, Angela; Pinheiro, Ricardo Brant [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear]. E-mail: albertomoc@terra.com.br; claubia@nuclear.ufmg.br; dora@nuclear.ufmg.br; fortini@nuclear.ufmg.br; rbp@nuclear.ufmg.br

    2007-07-01

    The cycle PWR-CANDU was simulated using the WIMSD-5B and ORIGEN2.1 codes. It was simulated a fuel burnup of 33,000 MWd/t for UO{sub 2} with enrichment of 3.2% and a fuel extended burnup of 45,000 MWd/t for UO{sub 2} with enrichments of 3.5%, 4.0% and 5.0% in a PWR reactor. The PWR discharged fuel was submitted to the simulation of deposition for five years. After that, it was submitted to AYROX reprocessing and used to produce a fuel to CANDU reactor. Then, it was simulated the burnup in the CANDU. Parameters such as infinite medium multiplication factor, k{sub inf}, fuel temperature coefficient of reactivity, {alpha}{sub TF}, moderator temperature coefficient of reactivity, {alpha}{sub TM}, the ratio rapid flux/total flux and the isotopic composition in the begin and the end of life were evaluated. The results showed that the fuels analyzed could be used on PWR and CANDU reactors without the need of change on the design of these reactors. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-07-01

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

  15. A comparison of the void reactivity effect between the CANDU standard and CANDU-6 SEU-43 cells

    International Nuclear Information System (INIS)

    In a CANDU type reactor the void coefficient of the reactivity is positive. The experimental data are available only for fresh fuel in cold conditions. On the other hand, taking into account the reactivity effects induced by changes of the coolant properties is often difficult. The safety analyses require an estimation of the calculation error. A comparison between models is an usual approach to obtain detailed information. In our paper a heterogeneous multi-stratified coolant model is used both for the CANDU standard fuel assembly cell and CANDU SEU-43 cell concept. The coolant is treated as a two phase (liquid and vapors) medium gravitationally separated. The results are inter-compared for different burnups in the partial or total void cases. (authors)

  16. Averaging methods of the gap heat transfer coefficients and the loss form coefficients of nuclear reactor cores loaded with different fuel bundles

    International Nuclear Information System (INIS)

    When performing transient analysis in heterogeneous nuclear reactors loaded with different types of fuel bundles is necessary to model the reactor core by a few representative fuel elements with average properties of a region containing a large number of fuel elements. The properties of these representative fuel bundles are obtained by averaging the thermal-hydraulic properties of the fuel elements contained in each region. In this paper we study the different ways to perform the averaging of the thermal-hydraulic properties that can have an influence on the transient results for licence purposes. Also we study the influence of the different averaging methods on the peak clad temperature (PCT) evolution for a LOCA, and on the critical power ratio (CPR) in the hot channels for a turbine trip transient without bypass credit.

  17. The evolution of the CANDU energy system - ready for Europe's energy future

    International Nuclear Information System (INIS)

    As air quality and climate change issues receive increasing attention, the opportunity for nuclear to play a larger role in the coming decades also increases. The good performance of the current fleet of nuclear plants is crucial evidence of nuclear's potential. The excellent record of Cernavoda-1 is an important part of this, and demonstrates the maturity of the Romanian program and of the CANDU design approach. However, the emerging energy market also presents a stringent economic challenge. Current NPP designs, while established as reliable electricity producers, are seen as limited by high capital costs. In some cases, the response to the economic challenge is to consider radical changes to new design concepts, with attendant development risks from lack of provenness. Because of the flexibility of the CANDU system, it is possible to significantly extend the mid-size CANDU design, creating a Next Generation product, without sacrificing the extensive design, delivery and operations information base for CANDU. This enables a design with superior safety characteristics while at the same time meeting the economic challenge of emerging markets. The Romanian nuclear program has progressed successfully forward, leading to the successful operation of Cernavoda-1, and the project to bring Cernavoda-2 to commercial operation. The Romanian nuclear industry has become a full-fledged member of the CANDU community, with all areas of nuclear technology well established and benefiting from international cooperation with other CANDU organizations. AECL is an active partner with Romanian nuclear organizations, both through cooperative development programs, commercial contracts, and also through the activities of the CANDU owners' Group (COG). The Cernavoda project is part of the CANDU 6 family of nuclear power plants developed by AECL. The modular fuel channel reactor concept can be modified extensively, through a series of incremental changes, to improve economics, safety

  18. Life Assurance Strategy for CANDU NPP

    International Nuclear Information System (INIS)

    Nuclear Plants have a nominal 'design life' that forms the basis of equipment specification, economic evaluations and licensing for some jurisdictions. Some component in the plant may require replacement, refurbishing and or rehabilitation during the plant 'design life'. Components which are extremely difficult or economically impossible to replace will place a limit on plant life. Rehabilitation programs completed to date on older CANDU plants to improve reliability of plant components, coupled with R and D programs, experimental data and advanced analytical methods form the basis for CANDU plant component life assurance. Life assurance is verified during plant operation by comprehensive in-service inspection programs and laboratory examinations. The paper provides an overview of the experiences to date on Refurbishment and Rehabilitation programs and some Canadian approaches on the main activities involved in scoping and managing nuclear plant life assurance. A number of proactive programs are underway to anticipate, detect and mitigate potential aging degradation at an early stage to ensure plant safety and reliability. Some of these programs include; systematic plant condition assessment, refurbishment and upgrading programs, environmental qualification programs and a program of examination of components from decommissioned reactors. These programs are part of an overall nuclear power plant maintenance strategy. Beyond life assurance, a longer term approach would be geared towards life extension as a viable option for the future. Recent CANDU designs have benefited from the early CANDU experience and are expected to require less rehabilitation. Examples of changes in CANDU 6 include fuel channel design and adopting a closed component cooling water system. New designs are based on 'design life' longer than that used for economic evaluations. The approach is to design for easy replace ability for components that can be economically replaced. Specific examples

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

    International Nuclear Information System (INIS)

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

  20. Study on the effect of the CANFLEX-NU fuel element bowing on the critical heat flux

    Energy Technology Data Exchange (ETDEWEB)

    Suk, Ho Chun; Cho, Moon Sung; Jeon, Ji Su

    2001-01-01

    The effect of the CANFLEX-NU fuel element bowing on the critical heat flux is reviewed and analyzed, which is requested by KINS as the Government design licensing condition for the use of the fuel bundles in CANDU power reactors. The effect of the gap between two adjacent fuel elements on the critical heat flux and onset-of-dryout power is studied. The reduction of the width of a single inter-rod gap from its nominal size to the minimum manufacture allowance of 1 mm has a negligible effects on the thermal-hydraulic performance of the bundle for the given set of boundary conditions applied to the CANFLEX-43 element bundle in an uncrept channel. As expected, the in-reactor irradiation test results show that there are no evidence of the element bow problems on the bundle performance.

  1. CANDU operating experience

    International Nuclear Information System (INIS)

    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)

  2. Simultaneous development of flow and temperature fields in wire-wrapped fuel pin bundles of sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Simultaneous development of flow and temperature fields in the entrance region of fast breeder reactor (FBR) fuel pin bundles with helical spacer wires has been investigated by three-dimensional computational simulations. The Reynolds number, pitch of helical spacer wire and number of pins in the bundle are systematically varied. It is found that the magnitude of mean cross-stream velocity in the fully developed region is inversely proportional to the helical pitch length and it is nearly independent of the number of pins. But, there is a strong correlation between the locations of spacer wire and the peak cross-stream velocity. Flow attains full development at an axial length of 70 times hydraulic diameter in all the cases and this length is found to be unaffected by the helical pitch length. Friction factor is seen to fluctuate periodically over a mean value and the fluctuation over each helical pitch corresponds to a specific position of helical wire. The mean value of the friction factor in the entrance region reduces below the mean value in the fully developed region contrary to that seen in ducted flows. The mean fully developed friction factor is inversely proportional to the helical pitch. But, it is independent of the number of pins in the bundle. The Nusselt number passes through multiple minima before attaining fully developed periodic fluctuations and its development is slower than that of friction factor. For larger number of pins thermal development length is larger. Traditionally, the correlations reported for fully developed flow are considered for core design. But, the present study indicates that this approach is not conservative. Further, the entrance region effects and the oscillations in the fully developed region have to be properly accounted in the core design. Nusselt number exhibits a strong dependence on helical pitch similar to that of friction factor. A correlation for Nusselt number is proposed as a function of helical pitch and other

  3. An emergency water injection system (EWIS) for future CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Marques, Andre L.F. [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil). E-mail: momarques@uol.com.br; Todreas, Neil E.; Driscoll, Michael J. [Massachusetts Inst.of Tech., Cambridge, MA (United States). Nuclear Engineering Dept.

    2000-07-01

    This paper deals with the investigation of the feasibility and effectiveness of water injection into the annulus between the calandria tubes and the pressure tubes of CANDU reactors. The purpose is to provide an efficient decay heat removal process that avoids permanent deformation of pressure tubes severe accident conditions, such as loss of coolant accident (LOCA). The water injection may present the benefit of cost reduction and better actuation of other related safety systems. The experimental work was conducted at the Massachusetts Institute of Technology (MIT), in a setup that simulated, as close as possible, a CANDU bundle annular configuration, with heat fluxes on the order of 90 kW/m{sup 2}: the inner cylinder simulates the pressure tube and the outer tube represents the calandria tube. The experimental matrix had three dimensions: power level, annulus water level and boundary conditions. The results achieved overall heat transfer coefficients (U), which are comparable to those required (for nominal accident progression) to avoid pressure tube permanent deformation, considering current CANDU reactor data. Nonetheless, future work should be carried out to investigate the fluid dynamics such as blowdown behavior, in the peak bundle, and the system lay-out inside the containment to provide fast water injection. (author)

  4. 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.

  5. Proceedings of the fourth international conference on CANDU maintenance

    International Nuclear Information System (INIS)

    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

  6. Analyse du transfert de chaleur et de la perte de pression pour des ecoulements supercritiques dans le reacteur CANDU-SCWR

    Science.gov (United States)

    Zoghlami, Sarra

    channel, some geometrical simplifications are made. In fact, we assumed that the coolant flow in the fuel channel is represented by the flow around the fuel rod, bounded by the first crown. This simplified model was used for neutronic/thermal-hydraulic coupled calculation performed with neutronic codes DRAGON/DONJON (Varin et al., 2005), to analyse the thermal-hydraulic behavior of the fuel channel in CANDU-SCWR. We observed that the coolant density at the fuel rod external surface, at the sixth fuel bundle is 3.5 times lower than the average coolant density in the fuel channel. This puts into question the assumption of considering the supercritical water flow as an homogeneous flow and the ability to build a supercritical water nuclear reactor in CANDU type, i.e., with horizontal fuel channels. In order to validate ARTHUR &barbelow;SCWR code, we compared our results to SUBCHAN code (Jiang et al., 2009), which is a thermal-hydraulic code developed to analyze CANDU-SCWR. Both codes give the same shapes and orders of magnitude for the coolant average temperature and the cladding-surface temperature. The axial distribution of the centerline temperature in the fuel rod is different. This is due to the fact that the calculations performed by the SUBCHAN code are not coupled to a neutronic code. For this reason, the thermal power distributions differs in the two codes. The variation of the mass flow influences the forced convective heat transfer, so, the distribution of thermal-physical properties in the channel. In fact, if the mass flow is reduced by 50% compared to the nominal mass flow rate, following a pump failure, the external fuel rod surface temperature exceeds the melting point, which is between 1400°C and 1455°C. This phenomenon may results in the radioactive contamination of the environment.

  7. Reactor Physics Analysis Models for a CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok

    2007-10-15

    Canada deuterium uranium (CANDU) reactor physics analysis is typically performed in three steps. At first, macroscopic cross-sections of the reference lattice is produced by modeling the reference fuel channel. Secondly macroscopic cross-sections of reactivity devices in the reactor are generated. The macroscopic cross-sections of a reactivity device are calculated as incremental cross-sections by subtracting macroscopic cross-sections of a three-dimensional lattice without reactivity device from those of a three-dimensional lattice with a reactivity device. Using the macroscopic cross-sections of the reference lattice and incremental cross-sections of the reactivity devices, reactor physics calculations are performed. This report summarizes input data of typical CANDU reactor physics codes, which can be utilized for the future CANDU reactor physics analysis.

  8. Development and commercial implementation of CANFLEX fuel in Korea

    International Nuclear Information System (INIS)

    Since the early 1990's, Korea Atomic Energy Research Institute (KAERI) and Atomic Energy of Canada Limited (AECL) have cooperated to develop, verify, and demonstrate an advanced CANDU fuel, called a CANFLEX-NU (CANdu FLEXible-Natural Uranium) fuel, which has been specially designed so as to compensate the deterioration of the heat transfer rate in a crept pressure tube and also to enhance the safety margin of it by a reduction of the maximum linear element power of a fuel bundle. Now, the CANFLEX-NU fuel is ready to be commercialized in CANDU-6 because its design and demonstration irradiation have been completed in both Korea and Canada. Currently, some CANDU plants are being refurbished as they are approaching the end of their life span. One of the refurbished CANDUs is Wolsong unit 1 in Korea. It has been operating for 26 years and the operating power before its shutdown for refurbishment was less than 90% of its full power because of a reduction of the margin for the ROP trip set point. One of the most appropriate ways, up-to now, to overcome such a power de-rating due to a crept pressure tube is the introduction of the CANFLEX-NU fuel into a CANDU. Korea is now seriously considering the commercialization of the CANFLEX-NU fuel in the Wolsong units, especially units 2, 3 and 4 which have been in operation for over 13 years. Before commercializing the CANFLEX-NU fuel in the Wolsong units, there has been requested for an economic benefits analysis for the CANFLEX-NU fuel in the Wolsong units because the Uranium price at present is very different from that during the development phase of the CANFLEX-NU fuel. The present paper shows the important characteristics of the CANFLEX NU fuel, such as the physics, thermal-hydraulics fuel performance and safety and the enhancement of reactor safety and operating performance are compared to those of the 37-element fuel bundle. Also, the results of the economic benefits analysis for the full core loading of the CANFLEX

  9. CFD activities in support of thermal-hydraulic modeling of SFR fuel bundles

    International Nuclear Information System (INIS)

    Extensive testing and validation work is being performed to assess and validate Computational Fluid Dynamics (CFD) applicability to the simulation of SFR fuel assemblies. The demonstrated robustness of the method allows extending the CFD analysis to distorted fuel configurations, which will inevitably occur during extended fuel operation. The subchannel code COBRA-IV-I-MIT is adopted to evaluate the range of applicability of lumped parameter methods. Comparisons of mixing simulations show some intrinsic limitation in the subchannel methods, but allow confirming its overall applicability to nominal and mildly deformed assembly configurations. For significantly deformed geometries CFD is the recommend approach and is applied in this work. Deformed geometries considered include duct swelling, rod swelling, rod bowing, rod twisting, and various combinations of the simple deformations. While not derived from the realistic analysis of the in-core fuel behavior, the distorted geometries have been designed to embrace all conceptual worst case scenarios. The work focuses on the evaluation of the influence of the deformation on the fuel behavior, rather than on the actual fuel performance. Such approach is driven by the objective of deriving general understanding, and evaluating the applicability of subchannel analysis codes to long life fuel design, possibly in combination with distorted-channel factors derived from the CFD analyses. (author)

  10. CONSIDERATIONS FOR THE DEVELOPMENT OF A DEVICE FOR THE DECOMMISSIONING OF THE HORIZONTAL FUEL CHANNELS IN THE CANDU 6 NUCLEAR REACTOR. PART 10 - PRESENTATION OF THE DECOMMISSIONING DEVICE OPERATING

    Directory of Open Access Journals (Sweden)

    Constantin D. STANESCU,

    2015-05-01

    Full Text Available This paper presents a solution proposed by the authors in order to achieve of a cutting and extracting device operating panel for the decommissioning of the horizontal fuel channels in the CANDU 6 nuclear reactor. The Cutting and Extraction Device (CED is fully automated, connected by wires to a Programmable Logic Controller (PLC and controlled from a Human Machine Interface (HMI. The Cutting and Extraction Device (CED performs the dismantling, cutting and extraction of the fuel channel components, moving with variable speed, temperature monitoring and video surveillance inside the pipe, unblock and extract the channel closure plug (from End Fitting - EF, unblock and extract the channel shield plug (from Lattice Tube - LT, block and cut the middle of the pressure tube, block and cut the end of the pressure tube, block and extract the half of pressure tube. All operations can be monitored and controlled from a operating panel. The PLC fully command the device in automatic or manually mode, to control the internal sensors, transducers, electrical motors, video surveillance and pyrometers for monitoring cutting place temperature. The device controller has direct access to the measured values with these sensors, interprets and processes them, preparing the next actionafter confirming the action in progress. The design of the Cutting and Extraction Device (CED shall be achieved according to the particular features of the fuel channel components to be dismantled and to ensure radiation protection of workers.

  11. Technology transfer in CANDU marketing

    International Nuclear Information System (INIS)

    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

  12. Development of neural network for predicting local power distributions in BWR fuel bundles considering burnable neutron absorber

    International Nuclear Information System (INIS)

    A neural network model is under development to predict the local power distribution in a BWR fuel bundle as a high speed simulator of precise nuclear physical analysis model. The relation between 235U enrichment of fuel rods and local peaking factor (LPF) has been learned using a two-layered neural network model ENET. The training signals used were 33 patterns having considered a line symmetry of a 8x8 assembly lattice including 4 water rods. The ENET model is used in the first stage and a new model GNET which learns the change of LPFs caused by burnable neutron absorber Gadolinia, is added to the ENET in the second stage. Using this two-staged model EGNET, total number of training signals can be decreased to 99. These training signals are for zero-burnup cases. The effect of Gadolinia on LPF has a large nonlinearity and the GNET should have three layers. This combined model of EGNET can predict the training signals within 0.02 of LPF error, and the LPF of a high power rod is predictable within 0.03 error for Gadolinia rod distributions different from the training signals when the number of Gadolinia rods is less than 10. The computing speed of EGNET is more than 100 times faster than that of a precise nuclear analysis model, and EGNET is suitable for scoping survey analysis. (author)

  13. Transient non-boiling heat transfer in a fuel rod bundle during accidental power excursions

    International Nuclear Information System (INIS)

    The physical problem studied is the transient non-boiling heat transfer of a cylindrical fuel rod consisting of fuel, gap, and cladding to a steady, fully developed turbulent flow. The fuel pin is assumed to be located in the interior region of a subassembly with regular triangular or square arrangements. The turbulent velocity field as well as turbulent transport properties are specified as functions of the coordinates normal to the axial flow direction. The heat generation within the fuel may be specified as an arbitrary function of the three spatial coordinates and time. A digital computer program has been developed. On the basis of finite-difference techniques, to solve the governing partial differential equations with their associated subsidiary conditions. Results have been obtained for a series of exponential power transients of interest to safety of liquid-metal and water cooled nuclear reactors. The general physical features of transient convective heat transfer as explored by previous investigators have qualitatively been substantiated by the present analysis. Emphasis has been devoted to investigate the differences of heat-transfer (coefficient) results from multi-region analysis including a realistic fuel rod model and single-region analysis for the coolant region only. A comparison with the engineering relationships for turbulent liquid-metal cooling by Stein, which are an extension of the heat transfer coefficient concept to account for transient heat fluxes, clearly demonstrates that, at the parameters studied, Stein's approach tends to largely overestimate the convective heat transfer at early times

  14. The CANDU experience in Romania

    International Nuclear Information System (INIS)

    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

  15. Fuel cycles

    International Nuclear Information System (INIS)

    AECL publications, from the open literature, on fuels and fuel cycles used in CANDU reactors are listed in this bibliography. The accompanying index is by subject. The bibliography will be brought up to date periodically

  16. SLARette operations at CANDU 6 stations

    International Nuclear Information System (INIS)

    The Objective of the SLARette system as an ongoing maintenance device was to design a system which could be installed in reactor, commissioned and operational in less than 48 hours and removed in less than 24 hours, thus allowing a selected number of fuel channels to be SLARed during the normal yearly maintenance outage. This required a relatively simple design that does not require any changes to the Station Fuel Handling Systems and is portable, reliable and easily maintained. Over the last 10 years virtually every part of the SLARette system has been enhanced to further reduce man-hour and man-rem expenditure and it has now been used successfully on four CANDU 6 stations and many fuel channels. (author)

  17. Post-irradiation examination of DUPIC fuel

    International Nuclear Information System (INIS)

    The DUPIC (Direct Use of spent PWR fuel In CANDU reactors) program is designed to demonstrate the feasibility of fabricating DUPIC fuel and irradiating it under CANDU® operating conditions. The scope of the Canadian part in the program included the fabrication of three DUPIC fuel elements, irradiation testing in the NRU research reactor and assessment of DUPIC fuel performance relative to CANDU UO2 fuel. AECL successfully fabricated three 37-element geometry DUPIC elements. The irradiation testing of the three DUPIC elements in the NRU reactor in a demountable element bundle commenced in 1999 and was completed in 2003. The three DUPIC elements were irradiated to burnups ranging from 251 MWh/kgHE (10 MWd/kgHE) to 517 MWh/kgHE (22 MWd/kgHE), with maximum powers ranging from 49 kW/m to 51 kW/m. Post-irradiation examination (PIE) of the elements was conducted in the hot cells at Chalk River Laboratories (CRL) after their discharge from the NRU loops. DUPIC fuel exhibited comparable sheath strain and fission-gas release (FGR) to that observed in similarly operated UO2 fuel. The pellet microstructural behavior of DUPIC fuel (e.g., grain growth) is consistent with UO2 fuel irradiated to higher burnup than that of the DUPIC fuel. PIE results for the first DUPIC element were previously reported at the 7th CANDU Fuel Conference. This paper describes sheath strain and microstructure results of the second and third DUPIC elements. The FGR and other post-irradiation examination (PIE) results of the DUPIC fuel will be reported at a later date. (author)

  18. Basic research and industrialization of CANDU advanced fuel - Effect of transverse convex curvature on boiling heat transfer and ONB point of nucleate fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyung Chun; Lee, Young; Lee, Sung Hong [Pusan National University, Pusan (Korea)

    2000-04-01

    Recently, the effect of convex curvature on heat transfer should not be ignored when the radius of curvature tends to be small and/or associated with high heat transfer rate cases. Both analytical and experimental studies were performed to prove the effect of transverse convex curvature on the boiling heat transfer in concentric annuli flows. The effect of the transverse convex surface curvature on ONB are studied analytically in the case of reactor and evaporator. It is seen that the inner wall heat flux depends on R/sub i/, Rc, Re, Pr, {alpha}, and the {theta} of working fluid. An experimental study on the incipience of nucleate boiling is performed as a verification ad extension of previous analyses. Through flow visualization, the results show that the most dominant parameter to affect the heat flux at ONB is found to be the surface curvature. The heat flux data at ONB increases with the Re and the subcooling, and the effect of subcooling on ONB becomes smaller with decreasing Re. The heat flux at ONB increases rapidly as increase in {alpha} due to higher convective motion of bulk flow. Comparison between both results are accomplished with respect to the relative enhancement due to the convex curvature. The relative heat transfer enhancement ratio shows a good agreement between theory and experiment qualitatively and quantitatively. In conclusion, the obtained results suggest that the effect transverse convex curvature appears significantly in the boiling heat transfer. Therefore, it can be clearly expected that the effect should be more strong at the case of critical heat flux condition which is the most important design goal of the advanced nuclear fuel rods. 30 refs., 78 figs. (Author)

  19. Proceedings of the third international conference on CANDU maintenance

    International Nuclear Information System (INIS)

    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. Advanced CANDU reactor: an optimized energy source of oil sands application

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL) is developing the ACR-700TM (Advanced CANDU Reactor-700TM) to meet customer needs for reduced capital cost, shorter construction schedule, high capacity factor while retaining the benefits of the CANDU experience base. The ACR-700 is based on the concept of CANDU horizontal fuel channels surrounded by heavy water moderator. The major innovation of this design is the use of slightly enriched uranium fuel in a CANFLEX bundle that is cooled by light water. This ensures: higher main steam pressures and temperatures providing higher thermal efficiency; a compact and simpler reactor design with reduced capital costs and shorter construction schedules; and reduced heavy water inventory compared to existing CANDU reactors. ACR-700 is not only a technically advanced and cost effective solution for electricity generating utilities, but also a low-cost, long-life and sustainable steam source for increasing Alberta's Oil Sand production rates. Currently practiced commercial surface mining and extraction of Oil Sand resources has been well established over the last three decades. But a majority of the available resources are somewhat deeper underground require in-situ extraction. Economic removal of such underground resources is now possible through the Steam Assisted Gravity Drainage (SAGD) process developed and proto-type tested in-site. SAGD requires the injection of large quantities of high-pressure steam into horizontal wells to form reduced viscosity bitumen and condensate mixture that is then collected at the surface. This paper describes joint AECL studies with CERI (Canadian Energy Research Institute) for the ACR, supplying both electricity and medium-pressure steam to an oil sands facility. The extensive oil sands deposits in northern Alberta are a very large energy resource. Currently, 30% of Canda's oil production is from the oil sands and this is expected to expand greatly over the coming decade. The bitumen deposits in the

  1. Three-component U-Pu-Th fuel for plutonium irradiation in heavy water reactors

    Directory of Open Access Journals (Sweden)

    Peel Ross

    2016-01-01

    Full Text Available This paper discusses concepts for three-component fuel bundles containing plutonium, uranium and thorium for use in pressurised heavy water reactors, and cases for and against implementation of such a nuclear energy system in the United Kingdom. Heavy water reactors are used extensively in Canada, and are deploying within India and China, whilst the UK is considering the use of heavy water reactors to manage its plutonium inventory of 140 tonnes. The UK heavy water reactor proposal uses a mixed oxide (MOX fuel of plutonium in depleted uranium, within the enhanced CANDU-6 (EC-6 reactor. This work proposes an alternative heterogeneous fuel concept based on the same reactor and CANFLEX fuel bundle, with eight large-diameter fuel elements loaded with natural thorium oxide and 35 small-diameter fuel elements loaded with a MOX of plutonium and reprocessed uranium stocks from UK MAGNOX and AGR reactors. Indicative neutronic calculations suggest that such a fuel would be neutronically feasible. A similar MOX may alternatively be fabricated from reprocessed <5% enriched light water reactor fuel, such as the fuel of the AREVA EPR reactor, to consume newly produced plutonium from reprocessing, similar to the DUPIC (direct use of PWR fuel in CANDU process.

  2. Quality Products - The CANDU Approach

    International Nuclear Information System (INIS)

    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

  3. Nuclear fuel supply view in Argentina

    Energy Technology Data Exchange (ETDEWEB)

    Cirimello, R.O. [Comision Nacional de Energia Atomica, Conuar SA (Argentina)

    1997-07-01

    The Argentine Atomic Energy Commission promoted and participated in a unique achievement in the R and D system in Argentina: the integration of science technology and production based on a central core of knowledge for the control and management of the nuclear fuel cycle technology. CONUAR SA, as a fuel manufacturer, FAE SA, the manufacturer of Zircaloy tubes, CNEA and now DIOXITEC SA producer of Uranium Dioxide, have been supply, in the last ten years, the amount of products required for about 1300 Tn of equivalent U content in fuels. The most promising changes for the fuel cycle economy is the Slight Enriched Uranium project which begun in Atucha I reactor. In 1997 seventy five fuel assemblies, equivalent to 900 Candu fuel bundles, will complete its irradiation. (author)

  4. CANDU refurbishment - managing the life cycle

    International Nuclear Information System (INIS)

    All utilities that operate a nuclear power plant have an integrated plan for managing the condition of the plant systems, structures and components. With a sound plant life management program, after about 25 years of operation, replacement of certain reactor core components can give an additional 25 to 30 years of operation. This demonstrates the long-term economic strength of CANDU technology and justifies a long-term commitment to nuclear power. Indeed, replacement of pressure tubes and feeders with the most recent technology will also lead to increased capacity factors - due to reduced requirements for feeder inspections and repair, and eliminating the need for fuel channel spacer relocation which have caused additional and longer maintenance outages. Continuing the operation of CANDU units parallels the successful life extensions of reactors in other countries and provides the benefits of ongoing reliable operation, at an existing plant location, with the continued support of the host community. The key factors for successful, optimum management of the life cycle are: ongoing, effective plant life management programs; careful development of refurbishment scope, taking into account system condition assessments and a systematic safety review; and, a well-planned and well-executed retubing and refurbishment outage, where safety and risk management is paramount to ensure a successful project The paper will describe: the benefits of extended plant life; the outlook for refurbishment; the life management and refurbishment program; preparations for retubing of the reactor core; and, enhanced performance post-retubing. Given the potential magnitude of the program over the next 10 years, AECL will maintain a lead role providing overall support for retubing and plant Life Cycle Management programs and the CANDU Owners Group will provide a framework for collaboration among its Members. (author)

  5. Detail design of test loop for FIV in fuel bundle and preliminary test

    Energy Technology Data Exchange (ETDEWEB)

    Sim, Woo Gunl; Lee, Wan Young; Kim, Sung Won [Hannam University, Taejeon (Korea)

    2002-04-01

    It is urgent to develop the analytical model for the structural/mechanical integrity of fuel rod. In general, it is not easy to develop a pure analytical model. Occasionally, experimental results have been utilized for the model.Because of this reason, it is required to design proper test loop. Using the optimized test loop, With the optimized test loop, the dynamic behaviour of the rod will be evaluated and the critical flow velocity, which the rod loses the stability in, will be measured for the design of the rod. To verify the integrity of the fuel rod, it is required to evaluate the dynamic behaviour and the critical flow velocity with the test loop. The test results will be utilized to the design of the rod. Generally, the rod has a ground vibration due to turbulence in wide range of flow velocity and the amplitude of vibration becomes larger by the resonance, in a range of the velocity where occurs vortex. The rod loses stability in critical flow velocity caused by fluid-elastic instability. For the purpose of the present work to perform the conceptional design of the test loop, it is necessary (1) to understand the mechanism of the flow-induced vibration and the related experimental coefficients, (2) to evaluate the existing test loops for improving the loop with design parameters and (3) to decide the design specifications of the major equipments of the loop. 35 refs., 14 figs., 4 tabs. (Author)

  6. ACR technology for CANDU enhancements

    International Nuclear Information System (INIS)

    The ACR-1000 design retains many essential features of the original CANDU plant design. As well as further-enhanced safety, the design also focuses on operability and maintainability, drawing on valuable customer input and OPEX. The engineering development of the ACR-1000 design has been accompanied by a research and confirmatory testing program. This program has extended the database of knowledge on the CANDU design. The ACR-1000 design has been reviewed by the Canadian regulator, the Canadian Nuclear Safety Commission (CNSC) which concluded that there are no fundamental barriers to licensing the ACR-1000 design in Canada after completing three phases of the pre-project design review. The generic PSAR for the ACR-1000 design was completed in September 2009. The PSAR contains the ACR-1000 design details, the safety and design methodology, and the safety analysis that demonstrate the ACR-1000 safety case and compliance with Canadian and international regulatory requirements and expectations. The ACR technology developed during the ACR-1000 Engineering Program and the supporting development testing has had a major impact beyond the ACR program itself: Improved CANDU components and systems; Enhanced engineering processes and engineering tools, which lead to better product quality, and better project efficiency; and Improved operational performance. This paper provides a summary of technology arising from the ACR program that has been incorporated into new CANDU designs such as the EC6, or can be applied for servicing operating CANDU reactors. (author)

  7. PARTICULARITIES REGARDING THE OPERATING PROCESS OF THE CUTTING AND EXTRACTION DEVICE IN THE CANDU HORIZONTAL FUEL CHANNELS PRESSURE TUBE DECOMMISSIONING PART I: MOVEMENT AND FIXING DEVICE INSIDE THE PRESSURE TUBE

    Directory of Open Access Journals (Sweden)

    Constantin POPESCU

    2016-05-01

    Full Text Available This paper presents some details of operation process for a Cutting and Extraction Device (CED in order to achieve the decommissioning of the horizontal fuel channels pressure tube in the CANDU 6 nuclear reactor. The most important characteristic of the Cutting and Extraction Device (CED is his capability of totally operator’s protection against the nuclear radiation during pressure tube decommissioning. The movement and fixing processes present few particularities due to special adopted technical solutions: train guiding-fixing modules equipped with elastic guiding rollers and fixing claws, traction modules with elastic rollers and variable pitch, also with propriety to adapt the system according to various dimensions of the tube. The Cutting and Extraction Device (CED is a train of modules equipped with special systems to be fully automated, connected with a Programmable Logic Controller (PLC and controlled by an operator panel type Human Machine Interface (HMI. All processes are monitored by video cameras. In case of error, the process is automatically stopped, the operator receiving an error message and the last sequence could be reinitialized or aborted due to safety reasons

  8. Analysis of a homogenous and heterogeneous stylized half core of a CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    EL-Khawlani, Afrah [Physics Department, Sana' a (Yemen); Aziz, Moustafa [Nuclear and radiological regulatory authority, Cairo (Egypt); Ismail, Mahmud Yehia; Ellithi, Ali Yehia [Cairo Univ. (Egypt). Faculty of Science

    2015-03-15

    The MCNPX (Monte Carlo N-Particle Transport Code System) code has been used for modeling and simulation of a half core of CANDU (CANada Deuterium-Uranium) reactor, both homogenous and heterogeneous model for the reactor core are designed. The fuel is burnt in normal operation conditions of CANDU reactors. Natural uranium fuel is used in the model. The multiplication factor for homogeneous and heterogeneous reactor core is calculated and compared during fuel burnup. The concentration of both uranium and plutonium isotopes are analysed in the model. The flux and power distributions through channels are calculated.

  9. Compatibility analysis of DUPIC fuel(4) - thermal hydraulic analysis

    International Nuclear Information System (INIS)

    Thermal-hydraulic compatibility of the DUPIC fuel bundle in the CANDU reactor has been studied. The critical channel power, the critical power ratio, the channel exit quality and the channel flow are calculated for the DUPIC and the standard fuels by using the NUCIRC code. The physical models and associated parametric values for the NUCIRC analysis of the fuels are also presented. Based upon the slave channel analysis, the critical channel power and the critical power ratios have been found to be very similar for the two fuel types. The same dryout model is used in this study for the standard and the DUPIC fuel bundles. To assess the dryout characteristics of the DUPIC fuel bundle, the ASSERT-PV code has been used for the subchannel analysis. Based upon the results of the subchannel analysis, it is found that the dryout location and the power for the two fuel types are indeed very similar. This study shows that thermal performance of the DUPIC fuel is not significantly different from that of the standard fuel

  10. Three-dimensional porous media based numerical investigation of spatial power distribution effect on advanced nuclear fuel rod bundles critical power

    Energy Technology Data Exchange (ETDEWEB)

    Stosic, Zoran V. [Framatome ANP GmbH . NBTT, Erlangen (Germany)], e-mail: Zoran.Stosic@Framatome-ANP.de; Stevanovic, Vladimir D. [Framatome ANP GmbH, Erlangen (Germany); Iguchi, Tadashi [Japan Atomic Energy Research Institute (JAERI), Ibaraki (Japan)

    2001-07-01

    The influence of spatial power generation shape on thermal-hydraulics behaviour of the fuel rod bundle has been investigated. Particularly, the occurrence of the local Boiling Transition has been analysed, indicating that conditions for the Critical Heat Flux (CHF) are reached somewhere within the boiling water channels in the assembly. The two-phase coolant flow within the bundle is represented with the two-fluid model in 3D space. The porous medium concept is applied in the simulation of the two-phase flow through the rod bundle implying nonequilibrium thermal and flow conditions. The governing equations in three-dimensions are discretized with the control volume method. The 3D numerical simulation and analyses of thermal-hydraulics in a complex geometry of an advanced nuclear fuel assembly are performed for conditions of a partial and/or complete rods uncovering indicating occurrence of high quality CHF - Dryout. The obtained results from numerical simulations are compared with experimental Critical Power data obtained from full scale tests. Employed is an electrically heated test rod bundle with real 1:1 geometry. Different radial and axial power distributions are used with wide range of inlet mass flow rates (2 - 19 kg/s) and coolant inlet subcooling (25 - 185 kJ/kg). The coolant pressure, equal to 6.9 MPa, is typical for BWRs conditions. Comparison of the predicted Critical Power values with measured data shows encouraging agreements for all analysed power distributions and the results completely reflect measured two-phase mixture cross flows, steam void distribution and spatial positions of Dryout onsets. Based on performed numerical investigation, an improvement of Dryout criteria is proposed. Dynamic effects of power shape change on spatial thermal hydraulics and hence on CHF occurrence as well as the influence of transfer function on thermal hydraulics under cyclic power and/or flow rate changes are also being analysed. Experiments for such verifications

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

    International Nuclear Information System (INIS)

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

  12. CONSIDERATIONS FOR THE DEVELOPMENT OF A DEVICE FOR THE DECOMMISSIONING OF THE HORIZONTAL FUEL CHANNELS IN THE CANDU 6 NUCLEAR REACTOR. PART 7 - FUNCTIONING OF THE DECOMMISSIONING DEVICE

    Directory of Open Access Journals (Sweden)

    Gabi ROSCA FARTAT

    2015-05-01

    Full Text Available The scope of this paper is to achieve the device functioning steps for the commissioning of the horizontal fuel channels of calandria vessel. The dismantling of the fuel channel is performed by one device which shall provide radiation protection during the stages of decommissioning, ensuring radiation protection of the workers. For the decommissioning operation design shall be taken to ensure all aspects of security, environmental protection during decommissioning operation steps and creating and implementing work procedures resulting from developed decommissioning plan. The fuel channel decommissioning device is designed for dismantling and extraction of the fuel channel and its components. The decommissioning operation consists of following major steps: platform with device positioning to the fuel channel to be dismantled; coupling and locking the device at the fuel channel; unblock, extract and store the channel closure plug; unblock, extract and store the channel shield plug; block and cut the middle and the end of the pressure tube; block, extract and store the end fitting; block, extract and store the half of pressure tube; mounting of the extended closing plug. The operations steps are performed by the Cutting and Extraction Device and by the extraction actuator from the device handling elements assembly. After each step of dismantling is necessary the confirmation its finalization in order to perform the next operation step. The dismantling operation steps of the fuel channel components are repeated for all the 380 channels of the reactor, from the front of calandria side (plane R as well as the rear side (plane R'.

  13. Data base for a CANDU-PHW operating on a once-through natural uranium cycle

    International Nuclear Information System (INIS)

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

  14. The influence of lattice structure and composition on the coolant void reactivity in CANDU

    International Nuclear Information System (INIS)

    The neutronic mechanisms responsible for the production of positive void reactivity in CANDU lattices are presented. Based on an understanding of these, a method of void reactivity reduction is developed. Calculations show that a zero void reactivity lattice, which requires neutron poison in the fuel and fuel enrichment, is possible. (Author) 1 ref., 8 tabs

  15. Evaluation the coolant void reactivity of assemblies of CANDU-6 lattices

    International Nuclear Information System (INIS)

    It is known that the coolant void reactivity (CVR) is positive in an isolated CANDU-6 fuel cell. Here we will attempt to evaluate the CVR for a heterogeneous assembly of CANDU-6 cells using the code DRAGON. Because the existing version of DRAGON has inherent geometry limitations, CANDU-6 assembly calculations cannot be performed in the exact geometry. Accordingly, an approximate model for the geometry must be considered. Here, the model we propose consists in replacing the annular fuel pins by equivalent square fuel pins. This model can be used to calculate the effect of homogeneous or heterogeneous draining of the coolant in a single cell. It will also be used to evaluate the effect of fine mesh discretization at the corner of the lattices on cell eigenvalue as well as the CVR for the checkerboard voiding of a 2 x 2 assembly of cells. (author)

  16. CFD simulating the transient thermal–hydraulic characteristics in a 17 × 17 bundle for a spent fuel pool under the loss of external cooling system accident

    International Nuclear Information System (INIS)

    Highlights: • A 3-D CFD is adopted to simulate transient behaviors in an SFP under the accident. • This model realistically simulates a 17 × 17 bundle, rid of porous media approach. • The loss of external cooling system accident for an SFP is assumed in this paper. • Thermal–hydraulic characteristics in a bundle are strongly influenced by grids. • The results confirm temperature rising rate used in Maanshan NPP is conservative. - Abstract: This paper develops a three-dimensional (3-D) transient computational fluid dynamics (CFD) model to simulate the thermal–hydraulic characteristics in a fuel bundle located in a spent fuel pool (SFP) under the loss of external cooling system accident. The SFP located in the Maanshan nuclear power plant (NPP) is selected herein. Without adopting the porous media approach usually used in the previous CFD works, this model uses a real-geometry simulation of a 17 × 17 fuel bundle, which can obtain the localized distributions of the flow and heat transfer during the accident. These distribution characteristics include several peaks in the axial distributions of flow, pressure, temperature, and Nusselt number (Nu) near the support grids, the non-uniform distribution of secondary flow, and the non-uniform temperature distribution due to flow mixing between rods, etc. According to the conditions adopted in the Procedure 597.1 (MNPP Plant Procedure 597.1, 2010) for the management of the loss-of-cooling event of the spent fuel pool in the Maanshan NPP, the temperature rising rate predicted by the present model can be equivalent to 1.26 K/h, which is the same order as that of 3.5 K/h in the this procedure. This result also confirms that the temperature rising rate used in the Procedure 597.1 for the Maanshan NPP is conservative. In addition, after the loss of external cooling system, there are about 44 h for the operator to repair the malfunctioning system or provide the alternative water source for the pool inventory to

  17. Systems analysis of the CANDU 3 Reactor

    International Nuclear Information System (INIS)

    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. Interactions in Zircaloy/UO2 fuel rod bundles with Inconel spacers at temperatures above 1200deg C (posttest results of severe fuel damage experiments CORA-2 and CORA-3)

    International Nuclear Information System (INIS)

    In the CORA experiments test bundles of usually 16 electrically heated fuel rod simulators and nine unheated rods are subjected to temperature transients of a slow heatup rate in a steam environment. Thus, an accident sequence is simulated, which may develop from a small-break loss-of-coolant accident of an LWR. An aim of CORA-2, as a first test of its kind, was also to gain experience in the test conduct and posttest handling of UO2 specimens. CORA-3 was performed as a high-temperature test. The transient phases of CORA-2 and CORA-3 were initiated with a temperature ramp rate of 1 K/s. The temperature escalation due to the exothermal zircaloy(Zry)-steam reaction started at about 1000deg C, leading the bundles to maximum temperatures of 2000deg C and 2400deg C for tests CORA-2 and CORA-3, respectively. The test bundles resulted in severe oxidation and partial melting of the cladding, fuel dissolution by Zry/UO2 interaction, complete Inconel spacer destruction, and relocation of melts and fragments to lower elevations in the bundle, where extended blockages have formed. In both tests the fuel rod destruction set in together with the formation of initial melts from the Inconel/Zry interaction. The lower Zry spacer acted as a catcher for relocated material. In test CORA-2 the UO2 pellets partially disintegrated into fine particles. This powdering occurred during cooldown. There was no physical disintegration of fuel in test CORA-3. (orig./MM)

  19. Materials performance in CANDU reactors: The first 30 years and the prognosis for life extension and new designs

    International Nuclear Information System (INIS)

    A number of CANDU reactors have now been in-service for more than 30 years, and several are planning life extensions. This paper summarizes the major corrosion degradation operating experience of various out-of-core (i.e., excluding fuel channels and fuel) materials in-service in currently operating CANDU reactors. Also discussed are the decisions that need to be made for life extension of replaceable and non-replaceable components such as feeders and steam generators, and materials choices for new designs, such as the advanced CANDU reactor (ACR) and enhanced CANDU-6. The basis for these choices, including a brief summary of the R and D necessary to support such decisions is provided. Finally we briefly discuss the materials and R and D needs beyond the immediate future, including new concepts to improve plant operability and component reliability

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

    International Nuclear Information System (INIS)

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

  1. Data base for a CANDU-PHW operating on a once-through, slightly enriched uranium cycle (AECL-6594)

    International Nuclear Information System (INIS)

    This report, prepared for INFCE, gives data for an extrapolated 1000 MW(e) CANDU-PHW design operating on a once-through fuel cycle with a feed fuel of slightly enriched uranium - 1.2 weight % U-235 in uranium. The effects of varying fuel enrichment, maximum channel power, and economic parameters are also discussed

  2. Data base for a CANDU-PHW operating on a once-through, slightly enriched uranium cycle

    International Nuclear Information System (INIS)

    This report, prepared for INFCE, gives data for an extrapolated 1000 MW(e) CANDU-PHW design operating on a once-through fuel cycle with a feed fuel of slightly enriched uranium: 1.2 weight percent U-235 in uranium. The effects of varying fuel enrichment, maximum channel power, and economic parameters are also discussed. (author)

  3. A model for coolant void reactivity evaluation in assemblies of CANDU cells

    International Nuclear Information System (INIS)

    Coolant void reactivity is a very important safety parameter in CANDU reactor analysis. Here we evaluate the coolant void reactivity in a 2 x 2 heterogeneous assembly of CANDU cells using the code DRAGON. Since the current version of DRAGON can only treat the coolant void reactivity for a single CANDU cell, an approximate model for the geometry must be considered to perform assembly calculations in a 2 x 2 pattern. The model we propose consists of replacing the annular fuel pins by equivalent square fuel pins. The equivalence between annular fuel pins and square fuel pins is brought about by homogenizing the fuel plus its sheath and subsequently conserving the reaction rates between the two geometries using a SPH equivalence procedure. The approximate CANDU cells constructed using square pins were used to perform the transport calculations in 2 x 2 assembly patterns. In addition, the model was used to evaluate coolant void reactivity in 2 x 2 checkerboard voiding patterns. These calculations reflect more accurately the actual voiding situation being studied. This helps in assessing the effects due to the coupling of neutrons born in one cell to those born in the neighbouring cells

  4. Development of multi-dimensional thermal hydraulic modeling using mixing factors for wire wrapped fuel pin bundles with inter-subassembly heat transfer in fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Nishimura, M.; Kamide, H.; Ohshima, H. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1996-10-01

    Temperature distributions in fuel subassemblies of fast reactors interactively affect heat transfer from center to outer region of the core (inter-subassembly heat transfer) and cooling capability of an inter-wrapper flow, as well as maximum cladding temperature. The prediction of temperature distribution in the sub-assembly is, therefore one of the important issues for the reactor safety assessment. To treat the complex phenomena in the core, a multi-dimensional thermal hydraulic analysis is the most promising method. From the studies on the multi-dimensional thermal hydraulic modeling for the fuel sub-assemblies, the modeling have been recommended through the analysis of sodium experiments using driver subassembly test rig PLANDTL-DHX and blanket subassembly test rig CCTL-CFR. Computations of steady states experiments in the test rigs using the above modeling showed quite good agreement to the experimental data. In the present study, the use of this modeling was extended to transient analyses, and its applicability was examined. Firstly, non-dimensional parameters used to determine the mixing factors were modified from the ones based on bundle-averaged values to the ones by local values. Secondly, a new threshold function was derived and introduced to cut off the mixing factor of thermal plumes under inertia force dominant conditions. In the results of this validation, the accuracy was comparable between the modeling and the experimental instrumentation. Thus the present modeling is capable of predicting the thermal hydraulic fields of the wire wrapped fuel pin bundles with inter-subassembly heat transfer under the conditions from rated steady operations to transitions toward natural circulation decay heat removal modes. (J.P.N.)

  5. CANDU reactors. Experience and innovation

    International Nuclear Information System (INIS)

    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

  6. Input modelling of PHT system stability analysis for CANFLEX-RU bundle by SOPHT code

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Jun, Ji Su; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-01-01

    The overall objective of this report is to undertake the stability analysis of primary heat transport (PHT) system for a CANDU reactor to be loaded with CANFLEX-RU bundle which is to provide a vehicle for the economic use of recycled uranium and for the economic provision of additional operating margins in aging CANDU reactors. The modelling report for the input data of SOPHT code is required in order to give the specific and accurate information for flow stability calculation. This report is consisted of the several sections which are described the usage of control cards, calculation methods for input data generation, the comparison of specific input data set for 37-element, CANFLEX-NU and CANFLEX-RU bundles and calculation results for the steady state condition. Those input data set prepared will be used for the flow stability analysis of PHT system of CANDU reactor by SOPHT code. (author). 1 ref., 2 figs., 16 tabs.

  7. LBB in Candu plants

    Energy Technology Data Exchange (ETDEWEB)

    Kozluk, M.J.; Vijay, D.K. [Ontario Hydro Nuclear, Toronto, Ontario (Canada)

    1997-04-01

    Postulated catastrophic rupture of high-energy piping systems is the fundamental criterion used for the safety design basis of both light and heavy water nuclear generating stations. Historically, the criterion has been applied by assuming a nonmechanistic instantaneous double-ended guillotine rupture of the largest diameter pipes inside of containment. Nonmechanistic, meaning that the assumption of an instantaneous guillotine rupture has not been based on stresses in the pipe, failure mechanisms, toughness of the piping material, nor the dynamics of the ruptured pipe ends as they separate. This postulated instantaneous double-ended guillotine rupture of a pipe was a convenient simplifying assumption that resulted in a conservative accident scenario. This conservative accident scenario has now become entrenched as the design basis accident for: containment design, shutdown system design, emergency fuel cooling systems design, and to establish environmental qualification temperature and pressure conditions. The requirement to address dynamic effects associated with the postulated pipe rupture subsequently evolved. The dynamic effects include: potential missiles, pipe whipping, blowdown jets, and thermal-hydraulic transients. Recent advances in fracture mechanics research have demonstrated that certain pipes under specific conditions cannot crack in ways that result in an instantaneous guillotine rupture. Canadian utilities are now using mechanistic fracture mechanics and leak-before-break assessments on a case-by-case basis, in limited applications, to support licensing cases which seek exemption from the need to consider the various dynamic effects associated with postulated instantaneous catastrophic rupture of high-energy piping systems inside and outside of containment.

  8. Bundling biodiversity

    OpenAIRE

    Heal, Geoffrey

    2002-01-01

    Biodiversity provides essential services to human societies. Many of these services are provided as public goods, so that they will typically be underprovided both by market mechanisms (because of the impossibility of excluding non-payers from using the services) and by government-run systems (because of the free rider problem). I suggest here that in some cases the public goods provided by biodiversity conservation can be bundled with private goods and their value to consumers captured in th...

  9. CANDU safety analysis system establishment

    Energy Technology Data Exchange (ETDEWEB)

    Min, Byung Joo; Rhee, B. W.; Park, J. H.; Kim, H. T.; Choi, H. B.; Shim, J. I.; Yoon, C.; Yang, M. K

    2002-03-01

    To develop CANDU safety analysis system, methodology, and assessment technology, GAIs from CNSC and GSIs drived by IAEA are summarized. Furthermore, the following safety items are investigated in the present study. - It is intended to secure credibility of the void reactivity in the stage of nuclear design and analysis. The measurement data concerned with the void reactivity were reviewed and used to assess the physics code such as POWDERPUFS-V/RFSP, and the lattice code such as WIMS-AECL and MCNP-4B. - Reviewing the Final Safety Analysis Report for Wolsong-2/3/4 Units, the safety analysis methodology, classification for accident scenarios, safety analysis codes, their interface, etc. were examined. - The development of 3D CFD transient analysis model has been performed to predict local subcooling of the moderator in the vicinity of Calandria tubes in a CANDU-6 reactor in the case of Large LOCA transient. - The trip coverage analysis methodology based on CATHENA code is developed. The simulation of real plant transient showed good agreement. The trip coverage map was generated successfully for two typical depressurization and pressurization event. - The multi-dimensional analysis methodology for hydrogen distribution and hydrogen burning phenomena in PHWR containment is developed using GOTHIC code. The multi-dimensional analysis predicts the local hydrogen behaviour compared to the lumped parameter model.

  10. CANDU plant life management - An integrated approach

    International Nuclear Information System (INIS)

    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

  11. Cobalt-60 production in CANDU power reactors

    International Nuclear Information System (INIS)

    The technology for cobalt-60 production in CANDU reactors, designed and developed by MDS Nordion and AECL, has been safely,economically and successfully employed in CANDU reactors with over 195 reactor years of production. Today over forty percent of the world's disposable medical supplies are made safer through sterilization using cobalt-60 sources from MDS Nordion. Over the past 40 years, MDS Nordion with its CANDU reactor owner partners, has safely and reliably shipped more than 500 million curies of cobalt-60 sources to customers around the world

  12. REBEKA bundle experiments

    International Nuclear Information System (INIS)

    This report is a summary of experimental investigations describing the fuel rod behavior in the refilling and reflooding phase of a loss-of-coolant accident of a PWR. The experiments were performed with 5x5 and 7x7 rod bundles, using indirectly electrically heated fuel rod simulators of full length with original PWR-KWU-geometry, original grid spacers and Zircaloy-4-claddings (Type Biblis B). The fuel rod simulators showed a cosine shaped axial power profile in 7 steps and continuous, respectively. The results describe the influence of the different parameters such as bundle size on the maximum coolant channel blockage, that of the cooling on the size of the circumferential strain of the cladding (azimuthal temperature distribution) a cold control rod guide thimble and the flow direction (axial temperature distribution) on the resulting coolant channel blockage. The rewetting behavior of different fuel rod simulators including ballooned and burst Zircaloy claddings is discussed as well as the influence of thermocouples on the cladding temperature history and the rewetting behavior. All results prove the coolability of a PWR in the case of a LOCA. Therefore, it can be concluded that the ECC-criteria established by licensing authorities can be fulfilled. (orig./HP)

  13. Hydrogen analysis using MELCOR at CANDU plant

    Energy Technology Data Exchange (ETDEWEB)

    Song, Sung Chu; Park, Jae Hong; Kim, Han Chul [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2011-10-15

    As a part of a submitted Wolsong Severe Accident Management Plan(SAMP) at the end of 2009, KINS needs to the review of it. In this the focus is on the hydrogen behavior during the station blackout(SBO) because the hydrogen explosion during severe accident in CANDU plants such as Wolsong units has been safety issue. A SBO occurs with failure of the emergency power system when loss of class IV electric power happens because the loss of class IV power is the most dominant internal event causing core damage for Wolsong units. And following a SBO event, most of the engineered safety features(ESFs), including hydrogen igniters, are inoperable except the passive systems such as Dousing systems. Hydrogen is generated due to Zr-steam reaction in fuel channels and core debris oxidation in the suspended debris beds, jet breakup of molten debris in the water pool of the reactor vault and molten core-concrete interaction (MCCI). A combustible gas control system consisting of Passive Autocatalytic Recombiner(PAR) is currently installed at Wolsong unit 1. The results of MELCOR analysis are presented

  14. The third generation CANDU control room

    International Nuclear Information System (INIS)

    In CANDU stations, as in most complex industrial plants, the man/machine interface design has progressed through three generations. First Generation control rooms consisted entirely on fixed, discrete components (handswitches, indicator lights, strip chart, recorder, annunciator windows, etc.). Human factors input was based on intuitive common sense factors which varied considerably from one designer to another. Second Generation control rooms incorporated video display units and keyboards in the control panels. Computer information processing and display are utilized. There is systematic application of human factors through ergonomic and anthropometric standards and cookbooks. The human factors are applied mainly to the physical layout of the control panels and the physical manipulation performed by the operators. Third Generation control rooms exploit the dramatic performance/cost improvements in computer, electronic display and communication technologies of the 1980's. Further applications of human factors address the cognitive aspects of operator performance. At AECL, second generation control rooms were installed on CANDU stations designed in the mid 70s and early 80s. Third generation features will be incorporated in the CANDU 3 station design and future CANDU stations. There have been significant improvements in the man/machine interface in CANDU stations over the past three decades. The continuing rapid technological developments in computers and electronics coupled with an increasing understanding and application of human factors principles is leading to further enhancements. This paper outlines progress achieved in earlier stations and highlights the features of the CANDU 3rd generation control room. (author). 13 refs, 5 figs

  15. Drift effects in CANDU reactors

    International Nuclear Information System (INIS)

    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

  16. Mitigating aging in CANDU plants

    International Nuclear Information System (INIS)

    Aging degradation is a phenomenon we all experience throughout life, both on a personal basis and in business. Many industries have been successful in postponing the inevitable impact on their related systems and components through programs to maintain long-term reliability, maintainability and safety. However, this has not always been the case for nuclear power. While all power plants are experiencing the world trend of increasing operating costs with age, few (if any) have been able to fully define the parameters that solve the aging equation, particularly in relation to major components. Inspection and preventive maintenance have not been effective in predicting life-limiting degradation and failure. In CANDU nuclear plants, utilities are taking a comprehensive approach in dealing with the aging problem. Programs have been established to identify the current condition and degradation mechanisms of critical components, the failure of which would impact negatively on station competitiveness and safety. These include subcomponents under the general headings of reactor components, civil structures, piping (nuclear and conventional), steam generators, turbines and cables. In support of these efforts, R and D projects have been defined under the CANDU Owners Group to deal with generic issues on aging common to its members (e.g., investigation of degradation mechanisms, development of tools and techniques to mitigate the effects of aging, etc.). This paper describes recent developments of this cost-shared program with specific reference to concrete aging and crack repairs, flow-assisted corrosion in piping, elastomer service life, cable aging, degradation mechanisms in steam generators and lubricant breakdown. (author)

  17. CANDU reactors, their regulation in Canada, and the identification of relevant NRC safety issues

    International Nuclear Information System (INIS)

    Atomic Energy of Canada, Limited (AECL) and its subsidiary in the US, are considering submitting the CANDU 3 design for standard design certification under 10 CFR Part 52. CANDU reactors are pressurized heavy water power reactors. They have some substantially different safety responses and safety systems than the LWRs that the commercial power reactor licensing regulations of the US Nuclear Regulatory Commission (NRC) have been developed to deal with. In this report, the authors discuss the basic design characteristics of CANDU reactors, specifically of the CANDU 3 where possible, and some safety-related consequences of these characteristics. The authors also discuss the Canadian regulatory provisions, and the CANDU safety systems that have evolved to satisfy the Canadian regulatory requirements as of December 1992. Finally, the authors identify NRC regulations, mainly in 10 CFR Parts 50 and 100, with issues for CANDU 3 reactor designs. In all, eleven such regulatory issues are identified. They are: (1) the ATWS rule (section 50.62); (2) station blackout (section 50.63); (3) conformance with Standard Review Plan (SRP); (4) appropriateness of the source term (section 50.34(f) and section 100.11); (5) applicability of reactor coolant pressure boundary (RCPB) requirements (section 50.55a, etc); (6) ECCS acceptance criteria (section 50.46)(b); (7) combustible gas control (section 50.44, etc); (8) power coefficient of reactivity (GDC 11); (9) seismic design (Part 100); (10) environmental impacts of the fuel cycle (section 51.51); and (11) (standards section 50.55a)

  18. CANDU 9 safety enhancements and licensability

    International Nuclear Information System (INIS)

    The CANDU 9 design has followed the evolutionary product development approach that has characterized the CANDU family of nuclear power plants. In addition to utilizing proven equipment and systems from operating stations, the CANDU 9 design has looked ahead to incorporate design and safety enhancements necessary to meet evolving utility and regulatory requirements both in Canada and overseas. To demonstrate licensability in Canada, and to assure overseas customers that the design had independent regulatory review in the country of origin, the pre-project Basic Engineering Program included an extensive two year formal review by the Canadian regulatory authority, the Atomic Energy Control Board (AECB). Documentation submitted for this licensing review included the licensing basis, safety requirements and safety analysis necessary to demonstrate compliance with regulations as well as to assess system design and performance. The licensing review was successfully completed in 1997 January. In addition, to facilitate licensability in Korea, CANDU 9 incorporates feedback from the application of Korean licensing requirements to the CANDU 6 reactors at Wolsong site. (author)

  19. CANDU safety and licensing framework and process

    International Nuclear Information System (INIS)

    Nuclear Safety is a shared responsibility of the Industry, public and the Government. The International Atomic Energy Agency's (IAEA) safety fundamentals, basic objectives and safety guides lay down the principles from which requirements, recommendations and methodologies for safety design of Nuclear Power Plants (NPP) are derived. Within the framework of the international regulations and those of the Canadian Nuclear Safety Commission (CNSC), this paper will discuss the overall safety objectives, the defence in depth philosophy guiding CANDU safety, as well as the licensing process defined to meet all applicable CNSC regulations. The application of such philosophy to the ACR design and safety approach will also be discussed along with aspects of its implementation. The role of deterministic analysis, and Probabilistic Safety Analysis (PSA) in the design and licensing process of the Advanced CANDU Reactor will be discussed. Postulated initiating events and their combinations, acceptance criteria, CANDU margins and limits, supporting methodologies and computer codes used in safety analysis will be reviewed. The paper will also note intrinsic safety characteristics of CANDU, some of the ACR passive safety features built-in by design, CANDU distinctive features with respect to severe core damage, mechanisms of heat rejection in those extreme conditions, emergency coolant injection system features and other post accident mitigating systems. Update on the ACR Canadian and US licensing progress will also be provided. (authors)

  20. Candu plant life management - safeguarding the investment

    International Nuclear Information System (INIS)

    With a large number of the CANDU NPPs getting to the mid-point of their design life, a number of programs have been initiated by the Utilities and AECL to proactively manage plant aging, hence ensuring a safe and reliable operation for the remaining design life. An integrated aging management program is being formulated in response to the regulatory authority the Atomic Energy Control Board (AECB). This integrated program will take into account safety, performance and economic requirements. Elements of this program cover a variety of areas; Phase 1 covers life assessment studies of the critical systems, structures and components (SSCs) including a methodology for defining the critical SSCs. Other programs within Phase 2 address the preparation of a unified set of industry guidelines covering maintenance and inspection requirements during early, middle and late years of plant operation. A 'technology watch' program has also been initiated. The objective of this program is to identify far in advance potential aging phenomena and failure modes that could affect plant performance along with the inspection and maintenance required to monitor and mitigate such aging effects. The importance of maintaining the plant within a well-established 'licensing basis' envelope is also discussed. Plant licensing is carried out for an initial set of conditions and equipment status which may vary during plant life. Canadian utilities are required by the licensing authority, the AECB, to assess the impact of aging on the licensing case for each plant in operation. So far, the main focus of these assessments has been the impact of aging on key parameters such as fuel thermal margin and containment leak rate. Ongoing activities covering characterization of aging, maintenance, monitoring, and appropriate refurbishments are examined. Emphasis of R and D performed to characterize degradation mechanisms, support inspection and fitness for service guidelines are also briefly outlined. (author)

  1. COBRA-IV PC: A personal computer version of COBRA-IV-I for thermal-hydraulic analysis of rod bundle nuclear fuel elements and cores

    International Nuclear Information System (INIS)

    COBRA-IV PC is a modified version of COBRA-IV-I, adapted for use with most IBM PC and PC-compatible desktop computers. Like COBRA-IV-I, COBRA-IV PC uses the subchannel analysis approach to determine the enthalpy and flow distribution in rod bundles for both steady-state and transient conditions. The steady-state and transient solution schemes used in COBRA-IIIC are still available in COBRA-IV PC as the implicit solution scheme option. An explicit solution scheme is also available, allowing the calculation of severe transients involving flow reversals, recirculations, expulsions, and reentry flows, with a pressure or flow boundary condition specified. In addition, several modifications have been incorporated into COBRA-IV PC to allow the code to run on the PC. These include a reduction in the array dimensions, the removal of the dump and restart options, and the inclusion of several code modifications by Oregon State University, most notably, a critical heat flux correlation for boiling water reactor fuel and a new solution scheme for cross-flow distribution calculations. 7 refs., 8 figs., 1 tab

  2. Desktop Severe Accident Graphic Simulator Module for CANDU6 : PSAIS

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. Y.; Song, Y. M. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The ISAAC ((Integrated Severe Accident Analysis Code for CANDU Plant) code is a system level computer code capable of performing integral analyses of potential severe accident progressions in nuclear power plants, whose main purpose is to support a Level 2 probabilistic safety assessment or severe accident management strategy developments. The code has the capability to predict a severe accident progression by modeling the CANDU6- specific systems and the expected physical phenomena based on the current understanding of the unique accident progressions. The code models the sequence of accident progressions from a core heatup, pressure tube/calandria tube rupture after an uncovery from inside and outside, a relocation of the damaged fuel to the bottom of the calandria, debris behavior in the calandria, corium quenching after a debris relocation from the calandria to the calandria vault and an erosion of the calandria vault concrete floor, a hydrogen burn, and a reactor building failure. Along with the thermal hydraulics, the fission product behavior is also considered in the primary system as well as in the reactor building.

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

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young [Korea Atomic Energy Research Institute, Daeduk (Korea, Republic of); Usmani, S.A. [Atomic Energy of Canada Ltd., Toronto (Canada)

    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.

  4. ACR-1000TM - advanced Candu reactor design

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL) has developed the Advanced CANDU ReactorTM- 1000 (ACR-1000TM) as an evolutionary advancement of the current CANDU 6TM reactor. This evolutionary advancement is based on AECL's in-depth knowledge of CANDU structures, systems, components and materials, gained during 50 years of continuous construction, engineering and commissioning, as well as on the experience and feedback received from operators of CANDU plants. The ACR design retains the proven strengths and features of CANDU reactors, while incorporating innovations and state-of-the-art technology. These innovations improve economics, inherent safety characteristics, and performance, while retaining the proven benefits of the CANDU family of nuclear power plants. The Canadian nuclear reactor design evolution that has reached today's stage represented by the ACR-1000, has a long history dating back to the early 1950's. In this regard, Canada is in a unique situation, shared only by a very few other countries, where original nuclear power technology has been invented and further developed. The ACR design has been reviewed by domestic and international regulatory bodies, and has been given a positive regulatory opinion about its licensability. The Canadian regulator, the Canadian Nuclear Safety Commission (CNSC) completed the Phase 1 and Phase 2 pre-project design reviews in December 2008 and August 2009, respectively, and concluded that there are no fundamental barriers to licensing the ACR-1000 design in Canada. The final stage of the ACR-1000 design is currently underway and will be completed by fall of 2011, along with the final elements of the safety analyses and probabilistic safety analyses supporting the finalized design. The generic Preliminary Safety Analysis Report (PSAR) for the ACR-1000 was completed in September 2009. The PSAR demonstrates ACR-1000 safety case and compliance with Canadian and international regulatory requirements and expectations. (authors)

  5. Neutronic parameters calculations of a CANDU reactor

    International Nuclear Information System (INIS)

    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)

  6. The CANTEACH project: preserving CANDU technical knowledge

    International Nuclear Information System (INIS)

    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

  7. Recent experience related to neutronic transients in Ontario Hydro CANDU nuclear generating stations

    International Nuclear Information System (INIS)

    Ontario Hydro presently operates 18 CANDU reactors in the province of Ontario, Canada. All of these reactors are of the CANDU Pressurized Heavy Water design, although their design features differ somewhat reflecting the evolution that has taken place from 1971 when the first Pickering unit started operation to the present as the Darlington units are being placed in service. Over the last three years, two significant neutronic transients took place at the Pickering Nuclear Generating Station 'A' (NGS A) one of which resulted in a number of fuel failures. Both events provided valuable lessons in the areas of operational safety, fuel performance And accident analysis. The events and the lessons learned are discussed in this paper

  8. CFX analysis of the CANDU moderator thermal-hydraulics in the Stern Lab. Test Facility

    International Nuclear Information System (INIS)

    A numerical calculation with the commercial CFD code CFX is conducted for a test facility simulating the CANDU moderator thermal-hydraulics. Two kinds of moderator thermal-hydraulic tests at Stern Laboratories Inc. were performed in the full geometric configuration of the CANDU moderator circulating vessel, which is called a Calandria, housing a matrix of horizontal rod bundles simulating the Calandria tubes. The first of these tests is the pressure drop measurement of a cross flow in the horizontal rod bundles. The other is the local temperature measurement on the cross section of the horizontal cylinder vessel simulating the Calandria. In the present study the full geometric details of the Calandria are incorporated in the grid generation of the computational domain to which the boundary conditions for each experiment are applied. The numerical solutions are reviewed and compared with the available test data. It is shown that the present CFD prediction without the empirical correlation based on the pressure drop test is in good agreement with the test results. The prediction becomes more accurate, as the flow conditions become more turbulent with a higher Reynolds number. However, the temperature fluctuation is observed during iteration steps for a steady-state simulation of the thermal-hydraulic test. This result shows that the flow and temperature distribution inside the moderator tank may not be stable in the actual test

  9. Exporting apocalypse: CANDU reactors and nuclear proliferation

    International Nuclear Information System (INIS)

    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

  10. Strategic Aspects of Bundling

    International Nuclear Information System (INIS)

    The increase of bundle supply has become widespread in several sectors (for instance in telecommunications and energy fields). This paper review relates strategic aspects of bundling. The main purpose of this paper is to analyze profitability of bundling strategies according to the degree of competition and the characteristics of goods. Moreover, bundling can be used as price discrimination tool, screening device or entry barriers. In monopoly case bundling strategy is efficient to sort consumers in different categories in order to capture a maximum of surplus. However, when competition increases, the profitability on bundling strategies depends on correlation of consumers' reservations values. (author)

  11. Optimization and implementation study of plutonium disposition using existing CANDU Reactors. Final report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

    Since early 1994, the Department of Energy has been sponsoring studies aimed at evaluating the merits of disposing of surplus US weapons plutonium as Mixed Oxide (MOX) fuel in existing commercial Canadian Pressurized Heavy Water reactors, known as CANDU`s. The first report, submitted to DOE in July, 1994 (the 1994 Executive Summary is attached), identified practical and safe options for the consumption of 50 to 100 tons of plutonium in 25 years in some of the existing CANDU reactors operating the Bruce A generating station, on Lake Huron, about 300 km north east of Detroit. By designing the fuel and nuclear performance to operate within existing experience and operating/performance envelope, and by utilizing existing fuel fabrication and transportation facilities and methods, a low cost, low risk method for long term plutonium disposition was developed. In December, 1995, in response to evolving Mission Requirements, the DOE requested a further study of the CANDU option with emphasis on more rapid disposition of the plutonium, and retaining the early start and low risk features of the earlier work. This report is the result of that additional work.

  12. Mechanistic modeling of thermal-mechanical deformation of CANDU pressure tube under localized high temperature condition

    International Nuclear Information System (INIS)

    Thermal strain deformation is a pressure tube failure mechanism. The main objective of this paper is to develop mechanistic models to evaluate local thermal-mechanical deformation of a pressure tube in CANDU reactor and to investigate fuel channel integrity under localized contact between fuel elements and pressure tube. The consequence of concern is potential creep strain failure of a pressure tube and calandria tube. The initial focus will be on the case where a fuel rod contacts the pressure tube at full power with highly cooling condition

  13. Ultrasonic measurement method of calandria tube sagging in CANDU reactor

    International Nuclear Information System (INIS)

    Calandria tube wrapping each pressure tube is one of the key structural components of CANDU reactor (calandria) which is consisted of many pressure tubes containing nuclear fuel assemblies. As the calandria tube (made of zirconium alloy) is sagging due to its thermal and irradiation creep during the plant operation, it possibly contacts with liquid injection nozzle crossing beneath the calandria tube, which subsequently results in difficulties on the safe operation. It is therefore necessary to check the gap for the confirmation of no contacts between the two tubes, calandria tube and liquid inject ion tube, with a proper measure during the life of plant. In this study, an ultrasonic measurement method was selected among several methods investigated. The ultrasonic device being developed for the measurement of the gap was introduced and its preliminary performance test results were presented here

  14. Bundling in Telecommunications

    OpenAIRE

    Begoña García-Mariñoso; Xavier Martinez-Giralt; Pau Olivella

    2008-01-01

    The paper offers an overview of the literature on bundling in the telecommunications sector and its application in the Spanish market. We argue that the use of bundling in the provision of services is associated to technological reasons. Therefore, there appears no need to regulate bundling activities. However, this is not to say that other related telecom markets should not be scrutinized and regulated, or that the regulator should not pay attention to other bundling-related anticompetitive ...

  15. Technology spin-offs from a CANDU development program

    International Nuclear Information System (INIS)

    Both Enhanced CANDU 6 (EC6) and ACR-1000 design retain many essential features of the operating CANDU 6 plant design. As well as further-enhanced safety, the design also focuses on operability and maintainability, drawing on valuable customer input and OPEX. The engineering development of the ACR-1000 design has been accompanied by a research and confirmatory testing program. The ACR technology developed during the ACR-1000 Basic Engineering Program and the supporting development testing has extended the database of knowledge on the CANDU design. This paper provides a summary of technology arising from the ACR program that has been incorporated into new CANDU designs such as the Enhanced CANDU 6 (EC6), or can be applied for servicing operating CANDU reactors. (author)

  16. The disposal of Canada's nuclear fuel waste: a study of postclosure safety of in-room emplacement of used CANDU fuel in copper containers in permeable plutonic rock. Volume 4: biosphere model

    International Nuclear Information System (INIS)

    AECL (Atomic Energy of Canada Limited) has developed a disposal concept for Canada's nuclear fuel waste, which calls for a vault deep in plutonic rock of the Canadian Shield. The concept has been fully, documented in an environmental impact statement (EIS) for review by a panel under the Canadian Environmental Assessment Agency. The EIS includes the results of the EIS postclosure assessment case study to address the long term safety of the disposal concept. To more fully demonstrate the flexibility of the disposal concept and our assessment methodology, we are now carrying out another postclosure assessment study, which involves different assumptions and engineering options than those used in the EIS. In response to these changes, we have updated the BIOTRAC (BIOsphere Transport and Assessment Code) model developed for the EIS postclosure assessment case study. The main changes made to the BIOTRAC model are the inclusion of 36Cl, 137Cs, 239Np and 243Am; animals inhalation pathway; International Commission on Radiological Protection 60/61 human internal dose conversion factors; all the postclosure assessment nuclides in the dose calculations for non-human biota; and groundwater dose limits for 14C, 16C1 and 129I for non-human biota to parallel these limits for humans. We have also reviewed and changed several parameter values, including evasion rates of gaseous nuclides from soil and release fractions of various nuclides from domestic water, and indicated changes that affect the geosphere/biosphere interface model. These changes make the BIOTRAC model more flexible. As a result of all of these changes, the BIOTRAC model has been significantly expanded and improved, although the changes do not greatly affect model predictions. The modified model for the present study is called BIOTRAC2 (BIOTRAC - Version 2). The full documentation of the BIOTRAC2 model includes the report by Davis et al. (1993a) and this report. (author). 105 refs., 13 tabs., 8 figs

  17. Efficiency and accuracy of the perturbation response coefficient generation method for whole core comet calculations in BWR and CANDU configurations

    International Nuclear Information System (INIS)

    The coarse mesh transport method (COMET) is a highly accurate and efficient computational tool which predicts whole-core neutronics behaviors for heterogeneous reactor cores via a pre-computed eigenvalue-dependent response coefficient (function) library. Recently, a high order perturbation method was developed to significantly improve the efficiency of the library generation method. In that work, the method's accuracy and efficiency was tested in a small PWR benchmark problem. This paper extends the application of the perturbation method to include problems typical of the other water reactor cores such as BWR and CANDU bundles. It is found that the response coefficients predicted by the perturbation method for typical BWR bundles agree very well with those directly computed by the Monte Carlo method. The average and maximum relative errors in the surface-to-surface response coefficients are 0.02%-0.05% and 0.06%-0.25%, respectively. For CANDU bundles, the corresponding quantities are 0.01%-0.05% and 0.04% -0.15%. It is concluded that the perturbation method is highly accurate and efficient with a wide range of applicability. (authors)

  18. Backup and Ultimate Heat Sinks in CANDU Reactors For Prolonged SBO Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Nitheanandan, T.; Brown, M. J. [Atomic Energy of Canada Limited, Ontario (Canada)

    2013-10-15

    In a pressurized heavy water reactor, following loss of the primary coolant, severe core damage would begin with the depletion of the liquid moderator, exposing the top row of internally-voided fuel channels to steam cooling conditions on the inside and outside. The uncovered fuel channels would heat up, deform and disassemble into core debris. Large inventories of water passively reduce the rate of progression of the accident, prolonging the time for complete loss of engineered heat sinks. The efficacy of available backup and ultimate heat sinks, available in a CANDU 6 reactor, in mitigating the consequences of a prolonged station blackout scenario was analysed using the MAAP4-CANDU code. The analysis indicated that the steam generator secondary side water inventory is the most effective heat sink during the accident. Additional heat sinks such as the primary coolant, moderator, calandria vault water and end shield water are also able to remove decay heat; however, a gradually increasing mismatch between heat generation and heat removal occurs over the course of the postulated event. This mismatch is equivalent to an additional water inventory estimated to be 350,000 kg at the time of calandria vessel failure. In the Enhanced CANDU 6 reactor ∼2,040,000 kg of water in the reserve water tank is available for prolonged emergencies requiring heat sinks.

  19. Kinetic parameter calculation as function of burn-up of candu reactor

    International Nuclear Information System (INIS)

    Kinetic parameter calculation as function of burn-up of candu reactor. Kinetic marameter calculation as function of burp-up of CANDU reactor with Canflex fuel type-CANDU has been done. This type of fuel is currently being develop, so kinetic parameter such as effective delay neutron fraction (.......), delay neutron decay constant ( .... ) and prompt neutron generation time ( ...... ) are very important for analysis of reactor operation safety. WIMS-CRNL code was used to generate macroscopic cross section and reaction rate based on transport theory. Fast and thermal neutron velocity and macroscopic cross section fission product of the unit cell were determined by KINETIC Code. The result of calculation showed that the value of effective delay neutron fraction was 7,785616 x 10-3 at the beginning of operation at burn-up of 0 MWD/T and after the reactor operated at burn-up of 7,2231 x 10-3 MWD/T was 4,962766 x 10-3, or reduced by 36%. The value of prompt generation time was 9,982703 x 10-4 s at the beginning of operation at burn-up of 0 MWD/T and 8,965416 x 10-4 s after the reactor operated at burn-up of 7,2231 x 103 MWD/T, or reduced by 10%. The result of calculation showed that the values of effective delay neutron fraction and prompt neutron generation time are still great enough

  20. Modelling and simulation of dynamic characteristics of CANDU-SCWR

    International Nuclear Information System (INIS)

    Owing to the thermal properties of supercritical water and features of heat transfer correlation under supercritical pressure, a detailed thermal-hydraulic model with movable boundary of is developed for CANDU-SCWR (Supercritical Water-Cooled Reactor). Steady-state results of the model agree well with the design data. The dynamic responses of CANDU-SCWR to different disturbances are simulated and characteristics are analyzed. A dynamic model for ACR is also developed using CATHENA. Differences between dynamic characteristics of CANDU-SCWR and those of ACR are highlighted and investigated. It is concluded that CANDU-SCWR has a larger time constant, but with a higher response amplitude. (author)

  1. Advanced CANDU reactor development: a customer-driven program

    International Nuclear Information System (INIS)

    The Advanced CANDU Reactor (ACR) product development program is well under way. The development approach for the ACR is to ensure that all activities supporting readiness for the first ACR project are carded out in parallel, as parts of an integrated whole. In this way design engineering, licensing, development and testing, supply chain planning, construct ability and module strategy, and planning for commissioning and operations, all work in synergy with one another. Careful schedule management :ensures that program focus stays on critical path priorities.'This paper provides an overview of the program, with an emphasis on integration to ensure maximum project readiness, This program management approach is important now that AECL is participating as the reactor vendor in Dominion Energy's DOE-sponsored Combined Construction/Operating License (COL) program. Dominion Energy selected the ACR-700 as their reference reactor technology for purposes of demonstrating the COL process. AECL's development of the ACR is unique in that pre-licensing activities are being carded out parallel in the USA and Canada, via independent, but well-communicated programs. In the short term, these programs are major drivers of ACR development. The ACR design approach has been to optimize to achieve major design objectives: capital cost reduction, robust design with ample margins, proveness by using evolutionary change from existing :reference plants, design for ease :of operability. The ACR development program maintains these design objectives for each of the program elements: Design: .Carefully selected design innovations based on the SEU fuel/light water coolant:/heavy water moderator approach. Emphasis on lessons-learned review from operating experience and customer feedback Licensing: .Safety case based on strengths of existing CANDU plus benefits of optimised design Development and Test: Choice of materials, conditions to enable incremental testing building on existing CANDU and LWR

  2. A review of CANDU plant lifetime management

    International Nuclear Information System (INIS)

    In recent years, plant lifetime management(PLIM) including life extension has become the focus of the nuclear industry worldwide due to a number of factors which have arisen over the past decade : new siting difficulties, imbalance of power supply and demand, and high construction costs. In order to solve the problems, the PLIM program is being developed for the purpose of life extension and improvement of plant availability and safety. This paper describes the current activities and prospects of AECL and CANDU utilities, the conceptional evaluation results for the degradation mechanisms, and PLIM regulatory aspects. In addition, this paper provides the applicability of CANDU PLIM to Wolsong Unit 1 which has been operated for 17 years

  3. CANDU 9 design for hydrogen in containment

    International Nuclear Information System (INIS)

    CANDU 9 is a single unit plant whose design is based on proven CANDU technology with some design improvements in plant performance and safety. One improvement is in the area of post-accident hydrogen control. The reactor building layout and hydrogen control system are designed to enhance atmospheric mixing and prevent unacceptably high local and global hydrogen concentrations. Preliminary safety analysis shows that a maximum of 300 kg of hydrogen is produced during the metal-water reaction phase of severe accident: a large LOCA with emergency core cooling unavailable. Even though the hydrogen production rate is conservatively overestimated, preliminary containment thermalhydraulic analysis predicts that the maximum hydrogen concentration in the accident vault peaks at 6.8% by volume without crediting hydrogen igniters and recombiners. After about one hour, the concentration throughout the reactor building is about 2.4%. (author)

  4. Experimental modeling of flow-induced vibration of multi-span U-tubes in a CANDU steam generator

    International Nuclear Information System (INIS)

    Flow-induced vibration of the tubes in a nuclear steam generator is a concern for designers who are trying to increase the life span of these units. The dominant excitation mechanisms are fluidelastic instability and random turbulence excitation. The outermost U-bend region of the tubes is of greatest concern because the flow is almost perpendicular to the tube axis and the unsupported span is relatively long. The support system in this region must be well designed in order to minimize fretting wear of the tubes at the support locations. Much of the previous testing was conducted on straight single-span or cantilevered tubes in cross-flow. However, the dynamic response of steam generator multi-span U-tubes with clearance supports is expected to be different. Accurate modeling of the tube dynamics is important to properly simulate the dynamic interaction of the tube and supports. This paper describes a test program that was developed to measure the dynamic response of a bundle of steam generator U-tubes with Anti-Vibration Bar (AVB) supports, subjected to Freon two-phase cross-flow. The tube bundle has similar geometrical conditions to those expected for future CANDU steam generators. Future steam generators will be larger than previous CANDU steam generators, nearly twice the heat transfer area, with significant changes in process conditions in the U-bend region, such as increased steam quality and a broader range of flow velocities. This test program was initiated at AECL to demonstrate that the tube support design for future CANDU steam generators will meet the stringent requirements associated with a 60 year design life. The main objective of the tests is to address the issue of in-plane and out-of-plane fluidelastic instability and random turbulent excitation of a U-tube bundle with Anti-Vibration Bar (AVB) supports. Details of the test rig, measurement techniques and preliminary instrumentation results are described in the paper. (author)

  5. Independence and diversity in CANDU shutdown systems

    International Nuclear Information System (INIS)

    Atomic Energy Control Board regulations state that Canadian CANDU reactors shall have two fully effective, independent and diverse shutdown systems. The Darlington Nuclear Generating Station is the first power plant operated by Ontario Hydro to make use of software-based computer control in its shutdown systems. By virtue of the reliance placed on these systems to prevent exposure of the public to harmful radioactivity in the event of an accident, the shutdown system software has been categorized as safety critical software. An important issue that was considered in the design of the Darlington shutdown systems was how the software should be designed and incorporated into the systems to comply with the independence and diversity requirement. This paper describes how the independence and diversity requirement was complied with in previous CANDU shutdown system designs utilizing hardware components. The difference between systems utilizing hardware alone, and those utilizing both hardware and software are discussed. The results of a literature search into the issue of software diversity, the behaviour of multi-version software systems, and experience in other industries utilizing safety critical software are referred to. This paper advocates a systems approach to designing independent shutdown systems utilizing software. Opportunities exist at the system level for design decisions that can enhance software diversity and can reduce the likelihood of common mode faults in the systems. In the light of recent experience in implementing diverse safety critical software, potential improvements to the design process for CANDU shutdown systems are identified. (Author) 19 refs

  6. Once-through CANDU reactor models for the ORIGEN2 computer code

    International Nuclear Information System (INIS)

    Reactor physics calculations have led to the development of two CANDU reactor models for the ORIGEN2 computer code. The model CANDUs are based on (1) the existing once-through fuel cycle with feed comprised of natural uranium and (2) a projected slightly enriched (1.2 wt % 235U) fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models, as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST, are given

  7. Once-through CANDU reactor models for the ORIGEN2 computer code

    Energy Technology Data Exchange (ETDEWEB)

    Croff, A.G.; Bjerke, M.A.

    1980-11-01

    Reactor physics calculations have led to the development of two CANDU reactor models for the ORIGEN2 computer code. The model CANDUs are based on (1) the existing once-through fuel cycle with feed comprised of natural uranium and (2) a projected slightly enriched (1.2 wt % /sup 235/U) fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models, as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST, are given.

  8. R and D in support of CANDU plant life management

    International Nuclear Information System (INIS)

    One of the keys to the long-term success of CANDUs is a high capacity factor over the station design life. Considerable R and D in underway at AECL to develop technologies for assessing, monitoring and mitigating the effect of plant ageing and for improving plant performance and extending plant life. To achieve longer service life and to realize high capacity factor from CANDU stations, AECL is developing new technologies to enhance fuel channel and steam generator inspection capabilities, to monitor system health, and to allow preventive maintenance and cleaning (e.g., on-line chemical cleaning processes that produce small volumes of wastes). The life management strategy for fuel channels and steam generators requires a program to inspect components on a routine basis to identify mechanisms that could potentially affect fitness-for-service. In the case of fuel channels, the strategy includes inspections for dimensional changes, flaw detection, and deuterium concentration. New techniques are been developed to enhance these inspection capabilities; examples include accurate measurement of the gap between a pressure tube and its calandria tube and rapid full-length inspections of steam generator tubes for all known flaw types. Central to life management of components are Fitness-for-Service Guidelines (FFSG) that have been developed with the CANDU Owners Group (COG) that provide a standardized method to assess the potential for propagation of flaws detected during in-service inspections, and assessment of any change in fracture characteristics of the material. FFSG continue to be improved with the development of new technologies such as the capability to credit relaxation of stresses due to creep and non-rejectable flaws in pressure tubes. Effective management of plant systems throughout their lifetime requires much more than data acquisition and display - it requires that system health is continually monitored and managed. AECL has developed a system Health Monitor

  9. Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Progress report

    Energy Technology Data Exchange (ETDEWEB)

    Todreas, N.E.; Golay, M.W.; Wold, L.

    1981-02-01

    Four tasks are reported on: bundle geometry (wrapped and bare rods), subchannel geometry (bare rods), LMFBR outlet plenum flow mixing, and theoretical determination of local temperature fields in LMFBR fuel rod bundles. (DLC)

  10. CANDU co-generation opportunities

    International Nuclear Information System (INIS)

    Modern technology makes use of natural energy 'wealth' (uranium) to produce useful energy 'currency' (electricity) that can be used to society's benefit. This energy currency can be further applied to help solve a difficult problem faced by mankind. Within the next few years we must reduce our use of the same fuels which have made many countries wealthy - fossil fuels. Fortunately, electricity can be called upon to produce another currency, namely hydrogen, which has some distinct advantages. Unlike electricity, hydrogen can be stored and can be recovered for later use as fuel. It also is extremely useful in chemical processes and refining. To achieve the objective of reducing greenhouse gas emissions hydrogen must, of course, be produced using a method which does not emit such gases. This paper summarizes four larger studies carried out in Canada in the past few years. From these results we conclude that there are several significant opportunities to use nuclear fission for various co-generation technologies that can lead to more appropriate use of energy resources and to reduced emissions. (author)

  11. Pressure tube life management in CANDU-6 nuclear plant

    International Nuclear Information System (INIS)

    Operating parameters of pressure tube in CANDU-6 reactor, the relation between pressure tube life and plant life improvement of pressure tube by AECL in past years were summarized, and the factors affecting pressure tube life, idea and main measures of pressure tube life management in QINSHAN CANDU-6 power plant introduced

  12. Validation of WIMS-AECL reactivity device calculations for CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Min, Byung Joo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Donnelly, J. V. [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)

    1997-06-01

    An important component of the overall program to validate WIMS-AECL for use with RFSP in the analysis of CANDU-6 reactors for design and safety analysis calculations is the validation of calculations of incremental cross sections used to represent reactivity devices. A method has been developed for the calculation of the three-dimensional neutron flux distribution in and around CANDU reactor fuel channels and reactivity control devices. The methods is based on one- and two dimensional transport calculations with the WIMS-AECL lattice cell code, SPH homogenization, and three-dimensional flux calculations with finite-difference diffusion theory using the MULTICELL code. Simulations of Wolsung 1 Phase-B commissioning measurements and Point Lepreau restart tests have been performed, as a part of the program to validate WIMS-AECL lattice cell calculations for application to CANDU reactor simulations in RFSP. The incremental cross section properties of the Wolsung 1 and Point Lepreau adjusters, Mechanical Control Absorbers(MCA) and liquid Zone Control Unit (ZCU) is based on the WIMS-AECL/MULTICELL modelling methods and the results are compared with those of WIMS-AECL/DRAGON-2 modelling methods. (author). 13 tabs., 4 figs., 11 refs.

  13. Measuring velocity profile in scaled CANDU6 moderator tank using particle image velocimetry

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Han; Bang, In Cheol [UNIST, Ulsan, (Korea, Republic of); Kim, Hyoung Tae; Cha, Jae Eun [KAISt, Daejeon (Korea, Republic of)

    2012-10-15

    The Calandria vessel, which is called as CANDU6 moderator tank, is the actual reactor core including fuel channels and moderator in PHWR. The understanding of circulation patterns in Calandria vessel is important because the cooling capability is related to the moderator tank. Therefore, measuring velocity and temperature patterns in the Calandria vessel is the key factor for the safety of CANDU reactor. Various experimental and numerical efforts to predict and analyze the thermal hydraulic characteristic of Calandria vessel have been made. Yoon et al. developed a CFD model for the CANDU6 moderator analyzing the velocity profile and the temperature distribution. Khartabil et al. did an experiment to measure 3 dimensional velocity and temperature distribution in moderator circulation tests at a 1/4 scaled down facility. Laser Doppler Anememetry (LDA) was used to detect the velocity profile and thermocouples detect the temperature distribution. Previous study presented the velocity profile in a 1/40 scaled-down Calandria vessel and compared with CFD analysis. In the present work, Particle Image Velocimetry (PIV) technique is used to obtain velocity profiles in a 1/8 scaled down Calandria vessel.

  14. Operation of the nuclear fuel cycle test facilities -Operation of the hot test loop facilities

    Energy Technology Data Exchange (ETDEWEB)

    Chun, S. Y.; Jeong, M. K.; Park, C. K.; Yang, S. K.; Won, S. Y.; Song, C. H.; Jeon, H. K.; Jeong, H. J.; Cho, S.; Min, K. H.; Jeong, J. H.

    1997-01-01

    A performance and reliability of a advanced nuclear fuel and reactor newly designed should be verified by performing the thermal hydraulics tests. In thermal hydraulics research team, the thermal hydraulics tests associated with the development of an advanced nuclear fuel and reactor haven been carried out with the test facilities, such as the Hot Test Loop operated under high temperature and pressure conditions, Cold Test Loop, RCS Loop and B and C Loop. The objective of this project is to obtain the available experimental data and to develop the advanced measuring techniques through taking full advantage of the facilities. The facilities operated by the thermal hydraulics research team have been maintained and repaired in order to carry out the thermal hydraulics tests necessary for providing the available data. The performance tests for the double grid type bottom end piece which was improved on the debris filtering effectivity were performed using the PWR-Hot Test Loop. The CANDU-Hot Test Loop was operated to carry out the pressure drop tests and strength tests of CANFLEX fuel. The Cold Test Loop was used to obtain the local velocity data in subchannel within HANARO fuel bundle and to study a thermal mixing characteristic of PWR fuel bundle. RCS thermal hydraulic loop was constructed and the experiments have been carried out to measure the critical heat flux. In B and C Loop, the performance tests for each component were carried out. (author). 19 tabs., 78 figs., 19 refs.

  15. Operation of the nuclear fuel cycle test facilities -Operation of the hot test loop facilities

    International Nuclear Information System (INIS)

    A performance and reliability of a advanced nuclear fuel and reactor newly designed should be verified by performing the thermal hydraulics tests. In thermal hydraulics research team, the thermal hydraulics tests associated with the development of an advanced nuclear fuel and reactor haven been carried out with the test facilities, such as the Hot Test Loop operated under high temperature and pressure conditions, Cold Test Loop, RCS Loop and B and C Loop. The objective of this project is to obtain the available experimental data and to develop the advanced measuring techniques through taking full advantage of the facilities. The facilities operated by the thermal hydraulics research team have been maintained and repaired in order to carry out the thermal hydraulics tests necessary for providing the available data. The performance tests for the double grid type bottom end piece which was improved on the debris filtering effectivity were performed using the PWR-Hot Test Loop. The CANDU-Hot Test Loop was operated to carry out the pressure drop tests and strength tests of CANFLEX fuel. The Cold Test Loop was used to obtain the local velocity data in subchannel within HANARO fuel bundle and to study a thermal mixing characteristic of PWR fuel bundle. RCS thermal hydraulic loop was constructed and the experiments have been carried out to measure the critical heat flux. In B and C Loop, the performance tests for each component were carried out. (author). 19 tabs., 78 figs., 19 refs

  16. Parabolic k-ample bundles

    CERN Document Server

    Biswas, Indranil

    2011-01-01

    We construct projectivization of a parabolic vector bundle and a tautological line bundle over it. It is shown that a parabolic vector bundle is ample if and only if the tautological line bundle is ample. This allows us to generalize the notion of a k-ample bundle, introduced by Sommese, to the context of parabolic bundles. A parabolic vector bundle $E_*$ is defined to be k-ample if the tautological line bundle ${\\mathcal O}_{{\\mathbb P}(E_*)}(1)$ is $k$--ample. We establish some properties of parabolic k-ample bundles.

  17. The Impact of Power Coefficient of Reactivity on CANDU 6 Reactors

    International Nuclear Information System (INIS)

    The combined effects of reactivity coefficients, along with other core nuclear characteristics, determine reactor core behavior in normal operation and accident conditions. The Power Coefficient of Reactivity (PCR) is an aggregate indicator representing the change in reactor core reactivity per unit change in reactor power. It is an integral quantity which captures the contributions of the fuel temperature, coolant void, and coolant temperature reactivity feedbacks. All nuclear reactor designs provide a balance between their inherent nuclear characteristics and the engineered reactivity control features, to ensure that changes in reactivity under all operating conditions are maintained within a safe range. The CANDU reactor design takes advantage of its inherent nuclear characteristics, namely a small magnitude of reactivity coefficients, minimal excess reactivity, and very long prompt neutron lifetime, to mitigate the demand on the engineered systems for controlling reactivity and responding to accidents. In particular, CANDU reactors have always taken advantage of the small value of the PCR associated with their design characteristics, such that the overall design and safety characteristics of the reactor are not sensitive to the value of the PCR. For other reactor design concepts a PCR which is both large and negative is an important aspect in the design of their engineered systems for controlling reactivity. It will be demonstrated that during Loss of Regulation Control (LORC) and Large Break Loss of Coolant Accident (LBLOCA) events, the impact of variations in power coefficient, including a hypothesized larger than estimated PCR, has no safety-significance for CANDU reactor design. Since the CANDU 6 PCR is small, variations in the range of values for PCR on the performance or safety of the reactor are not significant

  18. Estimation of Aging Effects on LOHS for CANDU-6

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Yong Ki; Moon, Bok Ja; Kim, Seoung Rae [Nuclear Engineering Service and Solution Co. Ltd., Daejeon (Korea, Republic of)

    2014-05-15

    To evaluate the Wolsong Unit 1's capacity to respond to large-scale natural disaster exceeding design, the loss of heat sink(LOHS) accident accompanied by loss of all electric power is simulated as a beyond design basis accident. This analysis is considered the aging effects of plant as the consequences of LOHS accident. Various components of primary heat transport system(PHTS) get aged and some of the important aging effects of CANDU reactor are pressure tube(PT) diametral creep, steam generator(SG) U-tube fouling, increased feeder roughness, and feeder orifice degradation. These effects result in higher inlet header temperatures, reduced flows in some fuel channels, and higher void fraction in fuel channel outlets. Fresh and aged models are established for the analysis where fresh model is the circuit model simulating the conditions at retubing and aged model corresponds to the model reflecting the aged condition at 11 EFPY after retubing. CATHENA computer code[1] is used for the analysis of the system behavior under LOHS condition. The LOHS accident is analyzed for fresh and aged models using CATHENA thermal hydraulic computer code. The decay heat removal is one of the most important factors for mitigation of this accident. The major aging effect on decay heat removal is the reduction of heat transfer efficiency by steam generator. Thus, the channel failure time cannot be conservatively estimated if aged model is applied for the analysis of this accident.

  19. A CANDU-type small/medium power reactor

    International Nuclear Information System (INIS)

    The assembly known as a CANDU power reactor consists of a number of standardized fuel channels or 'power modules'. Each of these channels produces about 5 thermal megawatts on average. Within practical limitations on fuel enrichment and ultimately on economics, the number of these channels is variable between about 50 and approximately 700. Small reactors suffer from inevitable disadvantages in terms of specific cost of design/construction as well as operating cost. Their natural 'niche' for application is in remote off-grid locations. At the same time this niche application imposes new and strict requirements for staff complement, power system reliability, and so on. The distinct advantage of small reactors arises if the market requires installation of several units in a coordinated installation program - a feature well suited to power requirements in Canada's far North. This paper examines several of the performance requirements and constraints for installation of these plants and presents means for designers to overcome the consequent negative feasibility factors.

  20. Multi-step approach to Code-coupling for progression induced severe accidents in CANDU NPPs (MACPISA-CANDU)

    Energy Technology Data Exchange (ETDEWEB)

    Pohl, D.J.; Luxat, J.C. [McMaster Univ., Dept. of Engineering Physics, Hamilton, Ontario (Canada); Giannotti, W.; D' Auria, F. [Univ. of Pisa, Dept. of Mechanical, Nuclear and Production Engineering, Pisa (Italy)

    2009-07-01

    This paper reviews the progression of severe accidents, describes computer codes currently employed for analysis of severe accidents and outlines a new methodology to modelling the progression of severe accidents in CANDU nuclear power plants (NPPs) called the Multi-step Approach to Code-coupling for Progression Induced Severe Accidents in CANDU NPPs (MACPISA-CANDU). The MACPISA-CANDU methodology was used to couple the U.S. NRC codes SCDAP/RELAP5 (RELAP/SCDAPSIM Mod 3.4) and MELCOR (1.8.5) in order to model a small break loss of coolant accident with loss of emergency coolant injection (SBLOCA-LOECI) under natural circulation in a CANDU 6 NPP. Using this model it was shown that the sheath temperature did not exceed the zirconium melting temperature of 2098 K and hence the progression of the severe accident was terminated as expected. (author)

  1. Proof testing of CANDU concrete containment structures

    International Nuclear Information System (INIS)

    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

  2. Configuration management for CANDU feeder refurbishment

    International Nuclear Information System (INIS)

    The Canada Deuterium Uranium Reactor CANDU was originally designed to last twenty-five years. In 2005, Atomic Energy of Canada (AECL) made the decision to extend the life of the reactor by thirty years. One of the most critical elements of the life extension project was determining how to refurbish the Primary Heat Transport System. It was determined that the refurbishment required replacing the entire length of inlet and outlet feeders, from the end fittings to the header. The use of a robust Configuration Management program would have added significant value to the life extension project. (author)

  3. 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)

  4. Contact fiber bundles

    OpenAIRE

    Lerman, Eugene

    2003-01-01

    We define contact fiber bundles and investigate conditions for the existence of contact structures on the total space of such a bundle. The results are analogous to minimal coupling in symplectic geometry. The two applications are construction of K-contact manifolds generalizing Yamazaki's fiber join construction and a cross-section theorem for contact moment maps

  5. Principal noncommutative torus bundles

    DEFF Research Database (Denmark)

    Echterhoff, Siegfried; Nest, Ryszard; Oyono-Oyono, Herve

    2008-01-01

    action) and give necessary and sufficient conditions for any non-commutative principal torus bundle being RKK-equivalent to a commutative one. As an application of our methods we shall also give a K-theoretic characterization of those principal torus-bundles with H-flux, as studied by Mathai...... and Rosenberg which possess "classical" T-duals....

  6. Radiological Characteristics of decommissioning waste from a CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Dong Keun; Choi, Heui Joo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Ahmed, Rizwan; Heo, Gyun Young [Dept. of Nuclear Engineering, Kyung Hee University, Yongin (Korea, Republic of)

    2011-11-15

    The radiological characteristics for waste classification were assessed for neutron-activated decommissioning wastes from a CANDU reactor. The MCNP/ORIGEN2 code system was used for the source term analysis. The neutron flux and activation cross-section library for each structural component generated by MCNP simulation were used in the radionuclide buildup calculation in ORIGEN2. The specific activities of the relevant radionuclides in the activated metal waste were compared with the specified limits of the specific activities listed in the Korean standard and 10 CFR 61. The time-average full-core model of Wolsong Unit 1 was used as the neutron source for activation of in-core and ex-core structural components. The approximated levels of the neutron flux and cross-section, irradiated fuel composition, and a geometry simplification revealing good reliability in a previous study were used in the source term calculation as well. The results revealed the radioactivity, decay heat, hazard index, mass, and solid volume for the activated decommissioning waste to be 1.04 x 10{sup 16} Bq, 2.09 x 10{sup 3} W, 5.31 x 10{sup 14} m{sup 3}-water, 4.69 x 10{sup 5} kg, and 7.38 x 10{sup 1} m{sup 3}, respectively. According to both Korean and US standards, the activated waste of the pressure tubes, calandria tubes, reactivity devices, and reactivity device supporters was greater than Class C, which should be disposed of in a deep geological disposal repository, whereas the side structural components were classified as low- and intermediate-level waste, which can be disposed of in a land disposal repository. Finally, this study confirmed that, regardless of the cooling time of the waste, 15% of the decommissioning waste cannot be disposed of in a land disposal repository. It is expected that the source terms and waste classification evaluated through this study can be widely used to establish a decommissioning/disposal strategy and fuel cycle analysis for CANDU reactors.

  7. COMET solutions to whole core CANDU-6 benchmark problems

    International Nuclear Information System (INIS)

    In this paper, the coarse mesh transport code COMET is used to solve CANDU-6 benchmark problems in two and three dimensional geometry. These problems are representative of a simplified quarter core reactor model. The COMET solutions, the core eigenvalue and the fuel pin fission density distribution, are compared to those from the Monte Carlo code MCNP using two-group cross sections. COMET decomposes the core volume into a set of non-overlapping sub-volumes (coarse meshes) and uses pre-computed heterogeneous response functions that are constructed using Legendre polynomials as boundary conditions to generate a user selected whole core solution (e.g., the core eigenvalue and fuel pin fission density distribution). These response functions are pre-computed by performing fixed source calculations with a modified version of MCNP in only the unique coarse meshes in the core. Reference solutions are calculated by MCNP5 with a two-group energy library generated with the HELIOS lattice code. In the 2-D problem, the angular current on the coarse mesh interfaces in COMET is expanded to 2. order in both spatial and angular variables. The COMET eigenvalue error is 0.09%. The corresponding average error in the fission density over all 3515 fuel pins is 0.5%. The maximum error observed is 2.0%. For the 3-D case, with 4. order expansion in space and azimuthal angle and 2. order expansion in the cosine of the polar angle, the eigenvalue differs from the reference solution by 0.05%. The average fission density error over the 42180 fuel pins is 0.7% with a maximum error of 3.3%. (authors)

  8. Simulation-based reactor control design methodology for CANDU 9

    Energy Technology Data Exchange (ETDEWEB)

    Kattan, M.K.; MacBeth, M.J. [Atomic Energy of Canada Limited, Saskatoon, Saskatchewan (Canada); Chan, W.F.; Lam, K.Y. [Cassiopeia Technologies Inc., Toronto, Ontario (Canada)

    1996-07-01

    The next generation of CANDU nuclear power plant being designed by AECL is the 900 MWe CANDU 9 station. This design is based upon the Darlington CANDU nuclear power plant located in Ontario which is among the world leading nuclear power stations for highest capacity factor with the lowest operation, maintenance and administration costs in North America. Canadian-designed CANDU pressurized heavy water nuclear reactors have traditionally been world leaders in electrical power generation capacity performance. This paper introduces the CANDU 9 design initiative to use plant simulation during the design stage of the plant distributed control system (DCS), plant display system (PDS) and the control centre panels. This paper also introduces some details of the CANDU 9 DCS reactor regulating system (RRS) control application, a typical DCS partition configuration, and the interfacing of some of the software design processes that are being followed from conceptual design to final integrated design validation. A description is given of the reactor model developed specifically for use in the simulator. The CANDU 9 reactor model is a synthesis of 14 micro point-kinetic reactor models to facilitate 14 liquid zone controllers for bulk power error control, as well as zone flux tilt control. (author)

  9. Trends in the capital costs of CANDU generating stations

    International Nuclear Information System (INIS)

    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

  10. Annular burnout data from rod-bundle experiments

    International Nuclear Information System (INIS)

    Burnout data for annular flow in a rod bundle are presented for both transient and steady-state conditions. Tests were performed at the Oak Ridge National Laboratory in the Thermal Hydraulic Test Facility (THTF), a pressurized-water loop containing an electrically heated 64-rod bundle. The bundle configuration is typical of later generation pressurized-water reactors with 17 x 17 fuel arrays. Both axial and radial power profiles are flat. All experiments were carried out in upflow with subcooled inlet conditions, insuring accurate flow measurement. Conditions within the bundle were typical of those which could be encountered during a nuclear reactor loss-of-coolant accident

  11. Annular burnout data from rod bundle experiments

    International Nuclear Information System (INIS)

    Burnout data for annular flow in a rod bundle are presented for both transient and steady-state conditions. Tests were performed at the Oak Ridge National Laboratory in the Thermal Hydraulic Test Facility (THTF), a pressurized-water loop containing an electrically heated 64-rod bundle. The bundle configuration is typical of later generation pressurized-water reactors with 17 x 17 fuel arrays. Both axial and radial power profiles are flat. All experiments were carried out in upflow with subcooled inlet conditions, insuring accurate flow measurement. Conditions within the bundle were typical of those which could be encountered during a nuclear reactor loss-of-coolant accident. Level average fluid conditions within the test section were calculated using steady-state mass and energy conservation considerations for the steady-state tests and a transient, homogeneous, equilibrium computer code for the transient tests. Unlike tube dryout, burnout within a rod bundle does not necessarily occur at one distinct axial level. The location of individual rod dryout was determined by scanning rods axially and locating the position where rod superheat increased from approx. =0 to 30 K or greater. Thermocouple instrumentation within the bundle allows the location of dryout to be determined to within approximately +.5 cm for many of the tests

  12. The Candu suppliers: Growing beyond Canada

    International Nuclear Information System (INIS)

    Canada has demonstrated its ability to successfully manage Candu projects. It has been shown domestically by Ontario Hydro over a period of many years. It has been demonstrated abroad by private sector service companies, under Nuclear Construction Managers, on the 600 MW Wolsung project in South Korea, where NPM companies carried out virtually all project and construction management, the plant was in service only 60 months after the first concrete was poured for the reactor building. In the past, however, the project management team for Candu nuclear plants overseas, and for some at home, was often assembled ad hoc. Though this approach can be successful, it does not guarantee success in project and construction management. Because of the size, diversity and flexibility of the private owners of NPM, they have been able to keep their experienced nuclear power teams on staff even though the nuclear power industry has been going through slow times, assigning these people to large projects in other industries that call for the same high level of project, construction and commissioning management skills. Yet they are in a position to reassemble this team at very short notice, and have them ready to go to work on a major project within weeks of the green light

  13. LOCA power pulse analysis for CANDU-6 CANFLEX-RU core

    International Nuclear Information System (INIS)

    The power pulses following a large LOCA are analyzed for CANDU-6 reactor core fuelled with CANFLEX-RU fuel. The coupled simulations for reactor physics and channel thermal-hydraulic phenomena are done using RFSP and CATHENA codes. The 55% pump suction, 35% reactor inlet header and 100% reactor outlet header breaks are selected. The highest power pulse is predicted for 100% reactor outlet header break and it is higher than that for the standard 37-element natural fuel. However, the summation of initial stored energy and transient pulse energy of hottest pin has the minimum 17% margin to the fuel break up. Therefore, it is expected that there is no fuel breakup during the LOCA for CANFLEX-RU core

  14. Reactor physics innovations in ACR-700 design for next CANDU generation

    International Nuclear Information System (INIS)

    ACR-700 is the 'Next Generation' CANDU reactor, aimed at producing electrical power at a capital cost significantly less than that of the current reactor designs. A key element of cost reduction is the use of H20 as coolant and Slightly Enriched Uranium fuel in a tight D20- moderated lattice. The innovations in the ACR core physics result in substantial improvements in economics, as well as significant enhancements in reactor controllability and waste reduction. Fuel design is chosen to balance fuel performance, cost, and reactor-physics characteristics. Full-core coolant void reactivity in ACR-700 is about 3 mk. Power coefficient is substantially negative. Discharge fuel burnup is about three times the current natural-uranium discharge burn-up. The result is a core design which provides a high degree of inherent safety with attractive power-production efficiency and stability. (author)

  15. Licensing assessment of the Candu Pressurized Heavy Water Reactor. Preliminary safety information document. Volume II. [USA

    Energy Technology Data Exchange (ETDEWEB)

    1977-06-01

    ERDA has requested United Engineers and Constructors (UE and C) to evaluate the design of the Canadian natural uranium fueled, heavy water moderated (CANDU) nuclear reactor power plant to assess its conformance with the licensing criteria and guidelines of the U.S. Nuclear Regulatory Commission (USNRC) for light water reactors. This assessment was used to identify cost significant items of nonconformance and to provide a basis for developing a detailed cost estimate for a 1140 MWe, 3-loop Pressurized Heavy Water Reactor (PHWR) located at the Middletown, USA Site.

  16. Licensing assessment of the Candu Pressurized Heavy Water Reactor. Preliminary safety information document. Volume II

    International Nuclear Information System (INIS)

    ERDA has requested United Engineers and Constructors (UE and C) to evaluate the design of the Canadian natural uranium fueled, heavy water moderated (CANDU) nuclear reactor power plant to assess its conformance with the licensing criteria and guidelines of the U.S. Nuclear Regulatory Commission (USNRC) for light water reactors. This assessment was used to identify cost significant items of nonconformance and to provide a basis for developing a detailed cost estimate for a 1140 MWe, 3-loop Pressurized Heavy Water Reactor (PHWR) located at the Middletown, USA Site

  17. Pressure tube creep impact on the physics parameters for CANDU-6 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, W. Y.; Min, B. J. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Kam, S. C.; Kim, M. E. [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

    2004-07-01

    The lattice cell calculations are performed to assess the sensitivity of the reactor physics parameters to pressure tube creep resulting from radiation aging. The physics parameters of the lattice cell are calculated by using WIMSD-5B code, WIMS- AECL code, and MCNP code. The reference model(normal state) and two perturbed models accounting for the pressure tube creep are developed on the basis of CANDU-6 lattice cell. The 2.5% and 5% values of pressure tube diameter creep are considered. Also, The effects of the analyzed lattice parameters which are the coolant void reactivity, the fuel fission density and the atom density of Pu isotopes on the lattice.

  18. DUPIC fuel compatibility assessment

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok; Rho, G. H.; Park, J. W. [and others

    2000-03-01

    The purpose of this study is to assess the compatibility of DUPIC(Direct Use of Spent PWR Fuel in CANDU Reactors) fuel with the current CANDU 6 reactor, which is one of the technology being developed to utilize the spent PWR fuel in CANDU reactors. The phase 1 study of this project includes the feasibility analysis on applicability of the current core design method, the feasibility analysis on operation of the DUPIC fuel core, the compatibility analysis on individual reactor system, the sensitivity analysis on the fuel composition, and the economic analysis on DUPIC fuel cycle. The results of the validation calculations have confirmed that the current core analysis system is acceptable for the feasibility study of the DUPIC fuel compatibility analysis. The results of core simulations have shown that both natural uranium and DUPIC fuel cores are almost the same from the viewpoint of the operational performance. For individual reactor system including reactively devices, the functional requirements of each system are satisfied in general. However, because of the pronounced power flattening in the DUPIC core, the radiation damage on the critical components increases, which should be investigated more in the future. The DUPIC fuel composition heterogeneity dose not to impose any serious effect on the reactor operation if the fuel composition is adjusted. The economics analysis has been performed through conceptual design studies on the DUPIC fuel fabrication, fuel handling in a plant, and spent fuel disposal, which has shown that the DUPIC fuel cycle is comparable to the once-trough fuel cycle considering uncertainties associated with unit costs of the fuel cycle components. The results of Phase 1 study have shown that it is feasible to use the DUPIC fuel in CANDU reactors without major changes in hardware. However further studies are required to confirm the safety of the reactor under accident condition.

  19. COBRA-IV-I: an interim version of COBRA for thermal-hydraulic analysis of rod bundle nuclear fuel elements and cores

    International Nuclear Information System (INIS)

    The COBRA-IV-I computer code uses the subchannel analysis approach to determine the enthalpy and flow distribution in rod bundles for both steady-state and transient conditions. The steady-state and transient solution schemes used in COBRA-IIIC are still available in COBRA-IV-I as the implicit solution scheme option. In addition to these techniques, a new explicit solution scheme is now available which allows the calculation of severe transients involving flow reversals, recirculations, expulsion and reentry flows, with a pressure or flow boundary condition specified. Significant storage compaction and reduced running times have been achieved to allow the calculation of problems involving hundreds of subchannels

  20. Speciation of iodine (I-127) in the natural environment around Canadian CANDU sites

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, S.J.; Kotzer, T.G.; Chant, L.A

    2001-06-01

    In Canada, very little data is available regarding the concentrations and chemical speciation of iodine in the environment proximal and distal to CANDU Nuclear Power Generating Stations (NPGS). In the immediate vicinity of CANDU reactors, the short-lived iodine isotope {sup 131}I (t{sub 1/2} = 8.04 d), which is produced from fission reactions, is generally below detection and yields little information about the environmental cycling of iodine. Conversely, the fission product {sup 129}I has a long half-life (t{sub 1/2} = 1.57x10{sup 7} y) and has had other anthropogenic inputs (weapons testing, nuclear fuel reprocessing) other than CANDU over the past 50 years. As a result, the concentrations of stable iodine ({sup 127}I) have been used as a proxy. In this study, a sampling system was developed and tested at AECL's Chalk River Laboratories (CRL) to collect and measure the particulate and gaseous inorganic and organic fractions of stable iodine ({sup 127}I) in air and associated organic and inorganic reservoirs. Air, vegetation and soil samples were collected at CRL, and at Canadian CANDU Nuclear Power Generating Stations (NPGS) at OPG's (Ontario Power Generation) Pickering (PNGS) and Darlington NPGS (DNGS) in Ontario, as well as at NB Power's Pt. Lepreau NPGS in New Brunswick. The concentrations of particulate and inorganic iodine in air at CRL were extremely low, and were often found to be below detection. The concentrations are believed to be at this level because the sediments in the CRL area are glacial fluvial and devoid of marine ionic species, and the local atmospheric conditions at the sampling site are very humid. Concentrations of a gaseous organic species were comparable to worldwide levels. The concentrations of particulate and inorganic iodine in air were also found to be low at PNGS and DNGS, which may be attributed to reservoir effects of the large freshwater lakes in southern Ontario, which might serve to dilute the atmospheric iodine

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

    International Nuclear Information System (INIS)

    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

  2. Restrictions of stable bundles

    CERN Document Server

    Balaji, V

    2011-01-01

    The Mehta-Ramanathan theorem ensures that the restriction of a stable vector bundle to a sufficiently high degree complete intersection curve is again stable. We improve the bounds for the "sufficiently high degree" and propose a possibly optimal conjecture.

  3. Proceedings of the Canadian Nuclear Society CANDU maintenance conference

    International Nuclear Information System (INIS)

    The conference proceedings comprise 51 papers on the following aspects of maintenance of CANDU reactors: Major maintenance projects, maintenance planning and preparation, maintenance effectiveness, future maintenance issues, safety and radiation protection. The individual papers have been abstracted separately

  4. Korea signs for 2nd CANDU at Wolsong

    International Nuclear Information System (INIS)

    The sale of a second CANDU 6 reactor to Korea for the Wolsong site is discussed in relation to nuclear power in Korea, the Korean economy generally, Canadian trade with Korea, and cooperation between AECL and KAERI

  5. Periodic inspection of CANDU nuclear power plant containment components

    International Nuclear Information System (INIS)

    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

  6. Severe Accident Progression and Consequence Assessment Methodology Upgrades in ISAAC for Wolsong CANDU6

    International Nuclear Information System (INIS)

    Amongst the applications of integrated severe accident analysis codes like ISAAC, the principal are to a) help develop an understanding of the severe accident progression and its consequences; b) support the design of mitigation measures by providing for them the state of the reactor following an accident; and c) to provide a training platform for accident management actions. After Fukushima accident there is an increased awareness of the need to implement effective and appropriate mitigation measures and empower the operators with training and understanding about severe accident progression and control opportunities. An updated code with reduced uncertainties can better serve these needs of the utility making decisions about mitigation measures and corrective actions. Optimal deployment of systems such as PARS and filtered containment venting require information on reactor transients for a number of critical parameters. Thus there is a greater consensus now for a demonstrated ability to perform accident progression and consequence assessment analyses with reduced uncertainties. Analyses must now provide source term transients that represent the best in available understanding and so meaningfully support mitigation measures. This requires removal of known simplifications and inclusion of all quantifiable and risk significant phenomena. Advances in understanding of CANDU6 severe accident progression reflected in the severe accident integrated code ROSHNI are being incorporated into ISAAC using CANDU specific component and system models developed and verified for Wolsong CANDU 6 reactors. A significant and comprehensive upgrade of core behavior models is being implemented in ISAAC to properly reflect the large variability amongst fuel channels in feeder geometry, fuel thermal powers and burnup. The paper summarizes the models that have been added and provides some results to illustrate code capabilities. ISAAC is being updated to meet the current requirements and

  7. 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.

  8. CANDU energy for steam assisted gravity drainage

    International Nuclear Information System (INIS)

    Traditional open-pit mining has been used by industry for many years to remove oil sands from shallow deposits. To increase production capacity, the industry is looking for new technology to exploit bitumen from deep deposits. Among them, SAGD (Steam-Assisted Gravity Drainage) appears to be the most promising approach. It uses steam to remove bitumen from underground reservoirs. Recently, the SAGD recovery process has been put into commercial operation by major oil companies.Atomic Energy Canada Limited has assessed the use of the ACR-1000 as a source of heat and electricity for oil sand extraction and processing. The ACR-1000 design is an evolutionary development of the familiar CANDU technology, adding innovations to enhance economics, operations, and safety margins. The net electrical output from a standard ACR-1000 will be close to 1100 MWe, depending on local cooling water temperature

  9. Ballooning of CANDU pressure tubes. Model assessment

    International Nuclear Information System (INIS)

    The transient creep equations used to analyze the possible ballooning and failure of Zr-2.5% Nb pressure tubes during a loss-of-coolant accident (LOCA) were developed and verified using as-received Zr-2.5% Nb pressure tube material. But in a CANDU reactor, the pressure tubes absorb deuterium and are exposed to a continuous neutron fluence. Consequently, a literature survey was done to determine how irradiation damage and deuterium might affect the creep rate and ductility of Zr-2.5% Nb pressure tubes in the temperature range from 600 to 800 degrees C. It was found that irradiation damage, dissolved deuterium and deuteride blisters could possibly affect the creep rate and ductility of ZR-2.5% Nb pressure tubes in this temperature range, but deuteride platelets are expected to have little effect. Further tests are required to determine the effect of irradiation damage and deuterium on the creep rate and ductility of pressure tubes

  10. Leak detection capability in CANDU reactors

    International Nuclear Information System (INIS)

    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

  11. Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Progress report, September 1, 1980-November 30, 1980

    Energy Technology Data Exchange (ETDEWEB)

    Todreas, N.E.; Golay, M.W.; Wolf, L.

    1981-02-01

    Four tasks are reported: bundle geometry (wrapped and bare rods), subchannel geometry (bare rods), subchannel geometry (bare rods), LMFBR outlet plenum flow mixing, and theoretical determination of local temperature fields in LMFBR fuel rod bundles. (DLC)

  12. Subtleties Concerning Conformal Tractor Bundles

    CERN Document Server

    Graham, C Robin

    2012-01-01

    The realization of tractor bundles as associated bundles in conformal geometry is studied. It is shown that different natural choices of principal bundle with normal Cartan connection corresponding to a given conformal manifold can give rise to topologically distinct associated tractor bundles for the same inducing representation. Consequences for homogeneous models and conformal holonomy are described. A careful presentation is made of background material concerning standard tractor bundles and equivalence between parabolic geometries and underlying structures.

  13. Verification of revised EOP using the CATHENA code for Wolsong NPP- Candu 6

    International Nuclear Information System (INIS)

    EOP (Emergency Operating Procedure) is the operating procedure document which describes the operator's action needed for the maintenance of nuclear power plant (NPP) safety when a certain accident takes place. EOPs of Wolsong NPP Units 2, 3, 4 (Candu 6) have been revised aiming at the optimization of procedure to recover the NPP from an accident and the minimization of human errors. The revised overall EOPs are composed of 2 EOPs describing the accident diagnosis and recovery of critical safety parameters and 13 EOPs describing the operator actions needed for the individual accidents which include LOCA and MSLB and so on. The purpose of the present study is to verify the revised EOPs if the procedures and operator actions described in each EOP are fully appropriate and effective for the safe shutdown of NPP leading to temperature low enough to start the long term cooling using the shutdown cooling system and for the maintenance of fuel and pressure boundary integrity. The verification was carried out with the deterministic safety analysis methodology using the CATHENA computer program which is a Thermalhydraulic numerical code used in the safety analysis of Candu 6 reactors. CATHENA can simulate the various thermalhydraulic behaviors which are expected to happen during each accident and by various operator actions. From the verification results it can be assured that the revised individual EOPs have the enough appropriateness with respect to the engineering aspects and effectiveness in view of fuel and pressure boundary integrity. (authors)

  14. Chemistry aspects of the source term formation for a severe accident in a CANDU type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Constantin, A.; Constantin, M. [Institute for Nuclear Research, Pitesti (Romania)

    2013-07-15

    The progression of a severe accident in a CANDU type reactor is slow because the core is surrounded by a large quantity of heavy and light water which acts as a heat sink to remove the decay heat. Therefore, the source term formation is a complex and long process involving fission products transport and releasing in the fuel matrix, thermal hydraulics of the transport fluid in the primary heat system and containment, deposition and transport of fission products, chemistry including the interaction with the dousing system, structural materials and paints, etc. The source term is strongly dependent on initial conditions and accident type. The paper presents chemistry aspects for a severe accident in a CANDU type reactor, in terms of the retention in the primary heat system. After releasing from the fuel elements, the fission products suffer a multitude of phenomena before they are partly transferred into the containment region. The most important species involved in the deposition were identified. At the same time, the influence of the break position in the transfer fractions from the primary heat system to the containment was investigated. (orig.)

  15. Right bundle branch block

    DEFF Research Database (Denmark)

    Bussink, Barbara E; Holst, Anders Gaarsdal; Jespersen, Lasse;

    2013-01-01

    AimsTo determine the prevalence, predictors of newly acquired, and the prognostic value of right bundle branch block (RBBB) and incomplete RBBB (IRBBB) on a resting 12-lead electrocardiogram in men and women from the general population.Methods and resultsWe followed 18 441 participants included.......5%/2.3% in women, P Right bundle branch block was associated with significantly...... increased all-cause and cardiovascular mortality in both genders with age-adjusted hazard ratios (HR) of 1.31 [95% confidence interval (CI), 1.11-1.54] and 1.87 (95% CI, 1.48-2.36) in the gender pooled analysis with little attenuation after multiple adjustment. Right bundle branch block was associated...

  16. Evolution of the CANDU ICS-90+ control room design

    International Nuclear Information System (INIS)

    The design of the CANDU Control Room and the associated design process has evolved considerably over several generations of plants, from the first commercial scale demonstration CANDU at Douglas Point through to the large scale CANDUs at Darlington, and beyond, for the next generation of CANDU plant, ICS-90+, represented by new designs like CANDU 3. In the early plants, the control room configuration was based on designers' projections of control interface requirements. With succeeding generations, of designs, there has been an evolution towards: increasing attention to formal requirements definition, incorporation into the Human Machine Interface (HMI) of a larger base of operational experience, more systematic consideration of Human Factors (HF) aspects of the design and the application of a more powerful computer based HMI. For the newest plant, the CANDU 3, a Human Factors Engineering Program Plan (HFEPP) defines the overall HF engineering process, the associated requirements and HF engineering standards to be followed in each stage, and for all HMI aspects of the control room and plant design. The CANDU 3 control room also incorporates several new design innovations that will facilitate operating crew performance improvements. These are based on past experience with operating CANDU plants, incorporated with the use of formal design and validation methods plus results from Canadian research program to support control centre design and operation. For example, there are design improvements to facilitate: operator tracking of plant state, problem solving, alarm filtering, annunciation system interrogation, special safety system testing features, etc. The present paper will expand and elaborate on each of the above topics. (author). 7 refs, 2 figs, 1 tab

  17. An advancement in iterative solution schemes for three-dimensional, two-fluid modeling of two-phase flow in PWR fuel bundles

    International Nuclear Information System (INIS)

    Highlights: • A fully three-dimensional two-fluid model coupled with heat conduction was outlined. • Two-fluid numerical scheme capability was evaluated against NUPEC PSBT Benchmark. • GMRES, FGMRES, DQGMRES, CGNR, BCG, and TFQMR solvers were tested as iterative schemes. • Candidate Krylov solvers do not introduce deviations to the two-phase flow results. • GMRES, FGMRES, and DQGMRES have a more efficient and stable convergence performance. - Abstract: This paper outlines a fully three-dimensional two-fluid one-pressure model with a semi-implicit finite difference scheme coupled with heat conduction which can be applicable to thermal non-equilibrium two-phase flow field in subchannel geometry of Pressurized Water Reactors (PWR). The system of equations was linearized using the Newton–Raphson method and was collapsed into the pressure equations forming a system of the Poisson type. Then, two-phase flow modeling was combined with Krylov methods as advanced computing techniques to investigate the feasibility of implementing preconditioned Krylov subspace solvers as the numerical scheme to solve pressure equations. Six popular Krylov subspace solvers were considered: GMRES, FGMRES, DQGMRES, CGNR, BCG, and TFQMR combined with the block incomplete LU factorization with a dual truncation strategy (BILUT) preconditioner. These proposed iterative solvers were applied to the constructed linear pressure equations in the inner iteration in combination with the outer-Raphson iteration loop. Evaluation was performed in two stages. First, two-fluid numerical scheme capability was evaluated against OECD/NRC NUPEC PWR Bundle tests (PSBT Benchmark). The results for steady-state (PSBT) bundle show that an overall agreement can be found. At the second stage, convergency, stability, and accuracy of the proposed schemes were studied based on PSBT steady-state data through a comparison of utilized Krylov solvers and the direct inversion method as the pressure solution

  18. Principal -bundles on Nodal Curves

    Indian Academy of Sciences (India)

    Usha N Bhosle

    2001-08-01

    Let be a connected semisimple affine algebraic group defined over . We study the relation between stable, semistable -bundles on a nodal curve and representations of the fundamental group of . This study is done by extending the notion of (generalized) parabolic vector bundles to principal -bundles on the desingularization of and using the correspondence between them and principal -bundles on . We give an isomorphism of the stack of generalized parabolic bundles on with a quotient stack associated to loop groups. We show that if is simple and simply connected then the Picard group of the stack of principal -bundles on is isomorphic to ⊕ , being the number of components of .

  19. Bundles of Banach algebras

    Directory of Open Access Journals (Sweden)

    J. W. Kitchen

    1994-01-01

    Full Text Available We study bundles of Banach algebras π:A→X, where each fiber Ax=π−1({x} is a Banach algebra and X is a compact Hausdorff space. In the case where all fibers are commutative, we investigate how the Gelfand representation of the section space algebra Γ(π relates to the Gelfand representation of the fibers. In the general case, we investigate how adjoining an identity to the bundle π:A→X relates to the standard adjunction of identities to the fibers.

  20. Value added services to CANDU plants

    International Nuclear Information System (INIS)

    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

  1. On projective space bundle with nef normalized tautological line bundle

    CERN Document Server

    Yasutake, Kazunori

    2011-01-01

    In this paper, we study the structure of projective space bundles whose relative anti-canonical line bundle is nef. As an application, we get a characterization of abelian varieties up to finite etale covering.

  2. Operational improvements in the CANDU 9 control centre

    International Nuclear Information System (INIS)

    AECL has adopted an evolutionary approach to the design of the CANDU 9 control centre. Several factors have contributed to this decision including the desire to build on the successes of the current generation of CANDU stations, the changing roles and responsibilities of operations staff, an improved understanding of human error in operational situations, the opportunity for improved plant performance through the introduction of new technologies, and evolving customer and regulatory requirements. 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 a layout and information organization that is better matched to operational tasks, thereby leading to reduced operations, maintenance and administration (OM and A) costs. Significant design features that contribute to the improved operational capabilities of the CANDU 9 control centre include: a control centre layout with improved functionality; a new Plant Display System that is separated from the digital control computer system; and an enhanced computerized reactor shutdown system. The paper will present a summary of the design process, a detailed description of the CANDU 9 control centre layout and features, a description of the plant control and display systems design, including findings from a regulatory review, and other improvements to enhance operability. (author)

  3. Sensitivity of Candu-Scwr Reactor Physics Calculations to Nuclear Data Files

    Science.gov (United States)

    Kozier, K. S.; Dyck, G. R.

    2006-04-01

    A simplified MCNP model of a CANDU-SCWR lattice was used to test the sensitivity of the calculated reactivity to various nuclear data files involving issues of current interest. These tests were performed for cooled and voided conditions, with and without axial neutron leakage, for a uniform lattice of mid-life fuel and a mixed lattice with high-burnup and low-burnup fuel in alternate channels. Results were compared using different room-temperature data files for deuterium, various thermal-scattering-law data files for hydrogen bound in light water and deuterium bound in heavy water, and for pre-ENDF/B-VII and ENDF/B-VI.8 data for uranium. The reactivity differences observed were small (typically <1 mk) and increased with axial neutron leakage.

  4. Hybrid bundle divertor design

    International Nuclear Information System (INIS)

    A hybrid bundle divertor design is presented that produces <0.3% magnetic ripple at the center of the plasma while providing adequate space for the coil shielding and structure for a tokamak fusion test reactor similar to the International Tokamak Reactor and the Engineering Test Facility (with R = 5 m, B = 5 T, and a /SUB wall/ = 1.5 m, in particular). This hybrid divertor consists of a set of quadrupole ''wing'' coils running tangent to the tokamak plasma on either side of a bundle divertor. The wing coils by themselves pull the edge of the plasma out 1.5 m and spread the thickness of the scrape-off layer from 0.1 to 0.7 m at the midplane. The clear aperture of the bundle divertor throat is 1.0 m high and 1.8 m wide. For maintenance or replacement, the hybrid divertor can be disassembled into three parts, with the bundle divertor part pulling straight out between toroidal field coils and the wing coils then sliding out through the same opening

  5. ALUMINUM BOX BUNDLING PRESS

    Directory of Open Access Journals (Sweden)

    Iosif DUMITRESCU

    2015-05-01

    Full Text Available In municipal solid waste, aluminum is the main nonferrous metal, approximately 80- 85% of the total nonferrous metals. The income per ton gained from aluminum recuperation is 20 times higher than from glass, steel boxes or paper recuperation. The object of this paper is the design of a 300 kN press for aluminum box bundling.

  6. Universal Lagrangian bundles

    NARCIS (Netherlands)

    Sepe, D.

    2013-01-01

    The obstruction to construct a Lagrangian bundle over a fixed integral affine manifold was constructed by Dazord and Delzant (J Differ Geom 26:223–251, 1987) and shown to be given by ‘twisted’ cup products in Sepe (Differ GeomAppl 29(6): 787–800, 2011). This paper uses the topology of universal Lagr

  7. Industrial process heat from CANDU reactors

    International Nuclear Information System (INIS)

    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)

  8. On framed quantum principal bundles

    CERN Document Server

    Durdevic, M

    1995-01-01

    A noncommutative-geometric formalism of framed principal bundles is sketched, in a special case of quantum bundles (over quantum spaces) possessing classical structure groups. Quantum counterparts of torsion operators and Levi-Civita type connections are analyzed. A construction of a natural differential calculus on framed bundles is described. Illustrative examples are presented.

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

    International Nuclear Information System (INIS)

    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)

    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. Advancing CANDU experience to the world steam generator market

    International Nuclear Information System (INIS)

    Tube degradation in certain recirculating nuclear steam generators has provided a market for steam generator replacement. Prior to this need, B and W supplied over 200 steam generators for CANDU nuclear plants. With this experience, and implementing extensive research and development improvements in material selection, design enhancements, and new manufacturing and analytical methods, B and W has supplied or secured orders for the replacement of 26 steam generators. Along with plans for new replacement orders, B and W will continue to supply steam generators for future CANDU plants. This paper will review the progression of B and W's CANDU experience to meet the replacement steam generator market, and examine the continuous improvements required for today's increasingly demanding nuclear specifications. (author). 1 tab., 4 figs

  12. Cold-neutron tomography of annular flow and functional spacer performance in a model of a boiling water reactor fuel rod bundle

    International Nuclear Information System (INIS)

    Highlights: → Annular flows w/wo functional spacers are investigated by cold neutron imaging. → Liquid film thickness distribution on fuel pins and on spacer vanes is measured. → The influence of the spacers on the liquid film distributions has been quantified. → The cross-sectional averaged liquid hold-up significantly affected by the spacers. → The sapers affect the fraction of the entrained liquid hold up in the gas core. - Abstract: Dryout of the coolant liquid film at the upper part of the fuel pins of a boiling water reactor (BWR) core constitutes the type of heat transfer crisis relevant for the conditions of high void fractions. It is both a safety concern and a limiting factor in the thermal power and thus for the economy of BWRs. We have investigated adiabatic, air-water annular flows in a scaled-up model of two neighboring subchannels as found in BWR fuel assemblies using cold-neutron tomography. The imaging of the double suchannel has been performed at the ICON beamline at the neutron spallation source SINQ at the Paul Scherrer Institute, Switzerland. Cold-neutron tomography is shown here to be an excellent tool for investigating air-water annular flows and the influence of functional spacers of different geometries on such flows. The high-resolution, high-contrast measurements provide the spatial distributions of the coolant liquid film thickness on the fuel pin surfaces as well as on the surfaces of the spacer vanes. The axial variations of the cross-section averaged liquid hold-up and its fraction in the gas core shows the effect of the spacers on the redistribution of the two phases.

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

    International Nuclear Information System (INIS)

    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

  14. Heavy water detritiation to support CANDU station maintenance activities

    International Nuclear Information System (INIS)

    Heavy water and tritium control are important aspects of CANDU operation and have a major impact on station maintenance activities. Station personnel are trained to understand the importance of heavy water management and the economics and environmental impact of tritiated heavy water losses. This paper discusses new GE technology that can now make a major improvement in CANDU maintenance activities through significant reductions in station tritium levels. Tritium is of particular concern in the CANDU industry given the nature of heavy water reactors to build up high levels of tritium over time. High tritium levels in the reactor vault significantly slow down maintenance activities in the reactor vault due to the requirement for personnel protective equipment, including breathing apparatus and cumbersome plastic air suits. The difficulties increase as reactors age and tritium levels increase. Building upon GE's extensive operational experience in tritium management in CANDU reactors and its own tritium handling facility, GE. has developed a new large-scale diffusion based isotope separation process as an alternative to conventional cryogenic distillation. Having a tritium inventory an order of magnitude lower than conventional cryogenic distillation, this process is very attractive for heavy water detritiation, and applicable to single and multi-unit CANDU stations. This new process can now provide a step change reduction in CANDU heavy water tritium levels resulting in reduced environmental emissions and lowering reactor vault tritium MPC(a) levels. Reactor vault tritium can be reduced sufficiently for maintenance activities to be done without plastic suits, leading to shorter outages, improved station capacity factors, and improved station economics. (author)

  15. Development of Regulatory Requirements and Inspection Guides for CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, M. W.; Kim, K.; Ryu, Y. H. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Ro, H. Y.; Jin, T. E. [Korea Power Engineering Co., Yongin (Korea, Republic of)

    2009-05-15

    The first domestic CANDU power reactor, Wolsong unit 1, has been operated for about twenty years since commercial operation in 1983, and has been raised common aging issues of CANDU reactors in pressure tubes, calandria tubes, feeder pipes, etc. To solve these aging issues, utility is promoting the refurbishment activities for these major components. Therefore, confirmation and improvement for insufficient requirements considering the CNSC regulatory documents, regulatory principles between regulatory body and utilities related with refurbishment activities are required. These review contents are described herein, and representative review results are presented.

  16. Evolution of the CANDU control centre design process

    International Nuclear Information System (INIS)

    The design of the CANDU NPP control centre and the associated control centre design process has evolved considerably over several generations of plants, from Douglas Point through Darlington, and beyond, to new designs like CANDU 3. In the early plants, the control centre configuration had to be based on designers' projections of control interface requirements. With succeeding generations of designs, along with the introduction of advancing computer control technology, a larger based of operational experience has been factored into the control interface design, and increasing attention has been given to more formal requirements definition, and more systematic consideration of human factors aspects of the design

  17. Deformation quantization of principal bundles

    CERN Document Server

    Aschieri, Paolo

    2016-01-01

    We outline how Drinfeld twist deformation techniques can be applied to the deformation quantization of principal bundles into noncommutative principal bundles, and more in general to the deformation of Hopf-Galois extensions. First we twist deform the structure group in a quantum group, and this leads to a deformation of the fibers of the principal bundle. Next we twist deform a subgroup of the group of authomorphisms of the principal bundle, and this leads to a noncommutative base space. Considering both deformations we obtain noncommutative principal bundles with noncommutative fiber and base space as well.

  18. Helices and vector bundles

    CERN Document Server

    Rudakov, A N

    1990-01-01

    This volume is devoted to the use of helices as a method for studying exceptional vector bundles, an important and natural concept in algebraic geometry. The work arises out of a series of seminars organised in Moscow by A. N. Rudakov. The first article sets up the general machinery, and later ones explore its use in various contexts. As to be expected, the approach is concrete; the theory is considered for quadrics, ruled surfaces, K3 surfaces and P3(C).

  19. Canada's nuclear fuel industry: An overview. Background paper

    International Nuclear Information System (INIS)

    Canada was among the first countries to mine and process uranium-bearing ores. Such ores contain trace amounts of radium, which was in great demand for medical treatment and for use by research laboratories in the early part of the century. For the last half century, the same basic processes have been used to extract uranium from its ores and convert it to a form suitable for use in nuclear reactors. The process described here is that currently in use in Canada. Mining can take a variety of forms, from open-pit to deep, hard-rock. Mining is typically the most costly step in the process, particularly for lower-grade ores. The ore is crushed and ground in the mill to the consistency of fine sand from which the uranium is extracted chemically to produce the impure concentrate known as yellowcake. In the next step, the impure uranium concentrate is chemically refined into highly purified, nuclear-grade, uranium trioxide (UO3). Uranium trioxide is then converted, in two separate chemical processes, into uranium dioxide (UO2) which is destined for domestic consumption and uranium hexafluoride (UF6) which is exported. In Canada, fabrication is the final step of the fuel production process. Uranium dioxide powder is compressed and sintered into very dense ceramic pellets which are then sealed in zirconium tubes and assembled into fuel bundles for Candu reactors. This background paper will review the Canadian nuclear fuels industry. 1 fig

  20. Bundled monocapillary optics

    Science.gov (United States)

    Hirsch, Gregory

    2002-01-01

    A plurality of glass or metal wires are precisely etched to form the desired shape of the individual channels of the final polycapillary optic. This shape is created by carefully controlling the withdrawal speed of a group of wires from an etchant bath. The etched wires undergo a subsequent operation to create an extremely smooth surface. This surface is coated with a layer of material which is selected to maximize the reflectivity of the radiation being used. This reflective surface may be a single layer of material, or a multilayer coating for optimizing the reflectivity in a narrower wavelength interval. The collection of individual wires is assembled into a close-packed multi-wire bundle, and the wires are bonded together in a manner which preserves the close-pack configuration, irrespective of the local wire diameter. The initial wires are then removed by either a chemical etching procedure or mechanical force. In the case of chemical etching, the bundle is generally segmented by cutting a series of etching slots. Prior to removing the wire, the capillary array is typically bonded to a support substrate. The result of the process is a bundle of precisely oriented radiation-reflecting hollow channels. The capillary optic is used for efficiently collecting and redirecting the radiation from a source of radiation which could be the anode of an x-ray tube, a plasma source, the fluorescent radiation from an electron microprobe, a synchrotron radiation source, a reactor or spallation source of neutrons, or some other source.