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Sample records for candu fuel bundle

  1. Dimensional Measurements of Fresh CANDU Fuel Bundle

    Energy Technology Data Exchange (ETDEWEB)

    Jun, Ji Su; Jo, Chang Keun; Jung, Jong Yeob; Koo, Dae Seo; Cho, Moon Sung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    This paper intends to provide the dimensional measurements of fresh CANDU fuel (37-element) bundle for the estimation of deformation of post-irradiated (PI) bundle. It is expensive and difficult to measure the fretting wear of bearing pad, the element bowing and the waviness of endplate at the two-phase high flow condition (above 24 kg/s) of out-of-reactor test. So, it is recommended to compare the geometry of fresh bundle with that of PI bundle to estimate the integrity of fuel bundle in the CANDU-6 fuel channel with two-phase flow condition. The measurement system has been developed to provide the visual inspection and the dimensional measurements within the accuracy of 10 {mu}m. It is applicable in-air and underwater to the CANDU bundle as well as the CANFLEX bundle. The in-air measurements of the 36 fresh CANDU bundles (S/N: B400892 {approx} B400927) are done by this system from February 2004 to March 2004 in the PHWR fresh fuel storage building of KNFC. These bundles are produced by KNFC manufacturing procedure and are waiting for the delivery to the Wolsong-3 plant, and are planned to load into the proposed test channels. The detail measurements contain the outer rod profile (including the bearing pad), the diameter of bundle, the bowing of bundle, the rod length and the surface profile of end plate (waviness)

  2. CURRENT STATUS OF INTEGRITY ASSESSMENT BY SIPPING SYSTEM OF SPENT FUEL BUNDLES IRRADIATED IN CANDU REACTOR

    Directory of Open Access Journals (Sweden)

    JONG-YOUL PARK

    2014-12-01

    Full Text Available In terms of safety and the efficient management of spent fuel storage, detecting failed fuel is one of the most important tasks in a CANada Deuterium Uranium (CANDU reactor operation. It has been successfully demonstrated that in a CANDU reactor, on-power failed fuel detection and location systems, along with alarm area gamma monitors, can detect and locate defective and suspect fuel bundles before discharging them from the reactor to the spent fuel storage bay. In the reception bay, however, only visual inspection has been used to identify suspect bundles. Gaseous fission product and delayed neutron monitoring systems cannot precisely distinguish failed fuel elements from each fuel bundle. This study reports the use of a sipping system in a CANDU reactor for the integrity assessment of spent fuel bundles. The integrity assessment of spent fuel bundles using this sipping system has shown promise as a nondestructive test for detecting a defective fuel bundle in a CANDU reactor.

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

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    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.

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

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    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.

  5. Advanced CFD simulations of turbulent flows around appendages in CANDU fuel bundles

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    Abbasian, F.; Hadaller, G.I.; Fortman, R.A., E-mail: fabbasian@sternlab.com [Stern Laboratories Inc., Hamilton, Ontario (Canada)

    2013-07-01

    Computational Fluid Dynamics (CFD) was used to simulate the coolant flow in a modified 37-element CANDU fuel bundle, in order to investigate the effects of the appendages on the flow field. First, a subchannel model was created to qualitatively analyze the capabilities of different turbulence models such as k.ε, Reynolds Normalization Group (RNG), Shear Stress Transport (SST) and Large Eddy Simulation (LES). Then, the turbulence model with the acceptable quality was used to investigate the effects of positioning appendages, normally used in CANDU 37-element Critical Heat Flux (CHF) experiments, on the flow field. It was concluded that the RNG and SST models both show improvements over the k.ε method by predicting cross flow rates closer to those predicted by the LES model. Also the turbulence effects in the k.ε model dissipate quickly downstream of the appendages, while in the RNG and SST models appear at longer distances similar to the LES model. The RNG method simulation time was relatively feasible and as a result was chosen for the bundle model simulations. In the bundle model simulations it was shown that the tunnel spacers and leaf springs, used to position the bundles inside the pressure tubes in the experiments, have no measureable dominant effects on the flow field. The flow disturbances are localized and disappear at relatively short streamwise distances. (author)

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

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    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)

  7. Verification tests for CANDU advanced fuel

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

    1997-07-01

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

  8. Experimental and numerical investigations of three-dimensional turbulent flow of water surrounding a CANDU simulation fuel bundle structure inside a channel

    Energy Technology Data Exchange (ETDEWEB)

    Abbasian, F.; Yu, S.D. [Department of Mechanical and Industrial Engineering, Ryerson University, 350 Victoria Street, Toronto, Ontario, M5B 2K3 (Canada); Cao, J. [Department of Mechanical and Industrial Engineering, Ryerson University, 350 Victoria Street, Toronto, Ontario, M5B 2K3 (Canada)], E-mail: jcao@ryerson.ca

    2009-11-15

    Computational fluid dynamics (CFD) is used to simulate highly turbulent coolant flows surrounding a simulation CANDU fuel bundle structure inside a flow channel. Three CFD methods are used: large eddy simulation (LES), detached eddy simulation (DES), and Reynolds stress model (RSM). The outcome of the simulations is compared with the experimental pressure data measured using an in-water microphone and a miniature pressure transducer placed at various locations in the vicinity of the bundle structure. Among all the three methods employed in developing computational models, LES provides the most accurate results for turbulent pressures.

  9. Parallel CFD simulations of turbulent flows inside a CANDU fuel bundle

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    Abbasian, F.; Yu, S.D.; Cao, J. [Ryerson Univ., Dept. of Mechanical and Industrial Engineering, Toronto, Ontario (Canada)], E-mail: fabbasia@ryerson.ca

    2008-07-01

    Large Eddy Simulation (LES) is used to study the turbulent flow inside a 43-rod bundle. The two LES models developed in this paper are of dynamic Smagorinsky type, featuring a satisfactory prediction of anisotropic turbulence intensity and frequency. The first model, by taking advantage of the geometric periodicity, deals with one seventh of a rod bundle; it is developed for studying the axial, lateral turbulence intensities and frequencies in the centers of subchannels and narrow-gap regions. The second model, dealing with the full rod bundle inside a pressure tube with nominal eccentricity, is developed for studying the turbulent fluid forces acting on the bundle. In order to accelerate the solution process for the two large CFD models, the parallelized CFD technique is utilized in connection with 24 processors. The numerical results, obtained for a test case (an eight-rod bundle), are in good agreement with those experimental data available in the literature. Numerical simulations of turbulent flow phenomena within subchannels are advantageous since true flow features are difficult or costly to reveal by experiments. (author)

  10. Experience on management of CANDU spent fuel in Korea

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    Lee, H.-Y.; Choi, B.-I.; Yoon, J.-H.; Seo, U.-S. [Korea Hydro and Nuclear Power Co. Ltd., Nuclear Environment Technology Inst. (KHNP/NETEC), Yusung-Gu, Daejeon (Korea, Republic of)

    2002-07-01

    In Korea, national policy on the management of spent fuel from both PWR and CANDU reactors demands that all the spent fuel be kept within reactor site in until 2016 the time spent fuel interim storage facility might open. Based on the end of 2001, KHNP has 4 CANDU reactors in operation generating approximately 5,000 bundles of spent fuels per each unit annually. The generation, accumulation, and management of CANDU spent fuel by KHNP in Korea are reviewed. CANDU spent fuel storage technology including pool storage in fuel building, concrete silo storage, and on going project for consolidating storage adapting modular vault type MACSTOR concept are outlined. Especially current joint development of storage of CANDU spent fuel for improving land usage is addressed. The explanation of the new consolidated dry storage system includes description of the storage facility, its safety evaluations, and final implementation. Finally future movement on management of spent fuel in Korea is also briefly introduced. (author)

  11. Evaluation of safety margins during dry storage of CANDU fuel in MACSTOR/KN-400 module

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

    2005-03-15

    This paper covers an evaluation of the available safety margin against fuel bundle degradation during dry storage of CANDU spent fuel bundles in a MACSTOR/KN-400 module, considering normal, off-normal and postulated accidental conditions. (author)

  12. Fuel condition in Canadian CANDU 6 reactors

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    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

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

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

    1997-12-31

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

  14. Basic research and industrialization of CANDU advanced fuel

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

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    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)

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

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    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.

  18. 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.

  19. Assessment of DUPIC fuel compatibility with CANDU-6

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  20. 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)

  1. 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.

  2. Input modelling for subchannel analysis of CANFLEX fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-06-01

    This report describs the input modelling for subchannel analysis of CANFLEX fuel bundle using CASS(Candu thermalhydraulic Analysis by Subchannel approacheS) code which has been developed for subchannel analysis of CANDU fuel channel. CASS code can give the different calculation results according to users' input modelling. Hence, the objective of this report provide the background information of input modelling, the accuracy of input data and gives the confidence of calculation results. (author). 11 refs., 3 figs., 4 tabs.

  3. Safety assessment for the CANFLEX-NU fuel bundles with respect to the 37-element fuel bundles

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    Suk, H. C.; Lim, H. S. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-11-01

    The KAERI and AECL have jointly developed an advanced CANDU fuel, called CANFLEX-NU fuel bundle. CANFLEX 43-element bundle has some improved features of increased operating margin and enhanced safety compared to the existing 37-element bundle. Since CANFLEX fuel bundle is designed to be compatible with the CANDU-6 reactor design, the behaviour in the thermalhydraulic system will be nearly identical with 37-element bundle. But due to different element design and linear element power distribution between the two bundles, it is expected that CANFLEX fuel behaviour would be different from the behaviour of the 37-element fuel. Therefore, safety assessments on the design basis accidents which result if fuel failures are performed. For all accidents selected, it is observed that the loading of CANFLEX bundle in an existing CANDU-6 reactor would not worsen the reactor safety. It is also predicted that fission product release for CANFLEX fuel bundle generally is lower than that for 37-element bundle. 3 refs., 2 figs., 2 tabs. (Author)

  4. Luncheon address: Early days of CANDU fuel

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    Page, R.D. [Atomic Energy of Canada Limited (Canada)

    1997-07-01

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

  5. The travesty of discarding used CANDU fuel

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

    2016-09-15

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

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

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    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)

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

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    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).

  8. Overview of activities on CANDU fuel in Argentina

    Energy Technology Data Exchange (ETDEWEB)

    Alvarez, L.; Valesi, J., E-mail: lalvarez@cnea.gov.ar [National Commission on Atomic Energy, Fuel Engineering Department (Argentina)

    2013-07-01

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

  9. 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).

  10. CANFLEX fuel bundle cross-flow endurance test (test report)

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Sung Deok; Chung, C. H.; Chang, S. K.; Kim, B. D.

    1997-04-01

    As part of the normal refuelling sequence of CANDU nuclear reactor, both new and irradiated bundles can be parked in the cross-flow region of the liner tubes. This situation occurs normally for a few minutes. The fuel bundle which is subjected to the cross-flow should be capable of withstanding the consequences of cross flow for normal periods, and maintain its mechanical integrity. The cross-flow endurance test was conducted for CANFLEX bundle, latest developed nuclear fuel, at CANDU-Hot Test Loop. The test was carried out during 4 hours at the inlet cross-flow region. After the test, the bundle successfully met all acceptance criteria after the 4 hours cross-flow test. (author). 2 refs., 3 tabs.

  11. Estimation of CANDU spent fuel disposal canister lifetime

    Energy Technology Data Exchange (ETDEWEB)

    Kook, Dong Hak; Lee, Min Soo; Hwang, Yong Soo; Choi, Heui Joo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-05-15

    Active nuclear energy utilization causes significant spent fuel accumulation problem. The cumulative amount of spent fuel is about 10,083 ton as of Dec. 2008, and is expected to increase up to 19,000 ton by 2020. Of those, CANDU spent fuels account for more than 60% of the total amounts. CANDU spent fuels had been stored in dry concrete silos since 1991 and during the past 15 years, 300 silos were constructed and {approx}3,200 ton of spent fuels are stored now. Another dry storage facility MACSTOR /KN-400 will store new-coming CANDU spent fuels from 2009. But, after intermediate storage ends, all CANDU spent fuels have to be disposed within multi-layer metallic canister which is composed of cast iron inside and copper outside. Canister lifetime estimation, therefore, is very important for the final disposal safety analysis. The most significant factor of lifetime is copper corrosion, and Y. S. Hwang developed a corrosion model in order to predict the general corrosion effect on copper canister lifetime during the final disposal period. This research applied his model to KURT1 where many disposal researches are being performed actively and the results shows safe margin of the copper canister for the very long-term disposal.

  12. 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.

  13. A step towards closing the CANDU fuel cycle: an innovative scheme for reprocessing used CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Collins, F.; Lister, D. [Univ. of New Brunswick, UNB Nuclear, Dept. of Chemical Engineering, Fredericton, New Brunswick (Canada)

    2011-07-01

    Disposal versus reprocessing costs for used CANDU fuel was recently discussed by Rozon and Lister in a report produced for the Nuclear Waste Management Organization (NWMO). Their study discussed the economic incentives for reprocessing, not for the recovery of fissile uranium but for the recovery of plutonium ash. A $370/kg break-even price of uranium was calculated, and their model was found to be very sensitive to the reprocessing costs of the chosen technology. Findings were consistent with earlier studies done by Harvard University. Various reprocessing technologies (most based on solvent extraction) have been in use for many decades, but there appears to be no conceptual engineering study available in the open literature for a spent fuel reprocessing facility - one that includes process flows, operating costs and economic analysis. A deeper engineering study of the design and economics of re-processing technologies has since been undertaken by the nuclear group at the University of New Brunswick. An improved fluorination process was developed and modeled using ASPEN process simulation software. This study examines the impact of chosen technology on the spent fuel re-processing costs. (author)

  14. Evaluation of safety margins during dry storage of CANDU fuel in MACSTOR/KN-400 module

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

    This paper covers an evaluation of the available safety margin against fuel bundle degradation during dry storage of CANDU spent fuel bundles in a MACSTOR/KN-400 module, considering normal, off-normal and postulated accidental conditions. Korea Hydro and Nuclear Power (KHNP), in collaboration with Atomic Energy of Canada Limited (AECL), are developing a new module for the dry storage of spent fuel from the four CANDU 6 nuclear reactors at the Wolsong site in South Korea. The module provides the benefit of occupying significantly less area than the concrete canisters presently used. The modules are designed for a minimum service life of 50 years. During that period, the spent fuel bundles shall be safely stored. This imposes that failure of a fuel bundle element or unacceptable degradation of an existing defect (from reactor operation) does not occur during the dry storage period. The fuel bundles are stored in an air-filled fuel basket that releases 365 Watts on average and a maximum of 390 Watts when rare fuel loading conditions are postulated. In addition, specific accidental air flow cooling conditions are postulated that consist of 100% blockage of all air inlets on one side of the module. These conditions can generate a peak daily fuel temperature of up to 155{sup o}C during a reference hot summer day during the first year of operation. The fuel temperature decreases over the years and also fluctuates due to daily and seasonal temperature variations. At this temperature, fuel elements with intact Zircaloy sheathing will not experience damage. However, for the few fuel bundle elements that are non-leaktight (less than 1 per 37,000), some re-oxidation of UO{sub 2} into higher oxides such as U{sub 3}O{sub 7} / U{sub 4}O{sub 9} and U{sub 3}O{sub 8} will occur. This latter form of Uranium oxide is undesirable due to its lower density that results in a volumetric increase of the pellet that can overstress the fuel element sheathing. The level of fuel pellet

  15. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    St-Pierre, J., E-mail: joe.st-pierre@amec.com [AMEC NSS, Toronto, Ontario (Canada); Simons, B. [Stern Laboratories Incorporated, Hamilton, Ontario (Canada)

    2013-07-01

    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. Development of Bundle Position-Wise Linear Model for Predicting the Pressure Tube Diametral Creep in CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Yong [Korea Electric Power Corporation Research Institute, Daejeon (Korea, Republic of); Na, Man Gyun [Chosun University, Gwangju (Korea, Republic of)

    2011-08-15

    Diametral creep of the pressure tube (PT) is one of the principal aging mechanisms governing the heat transfer and hydraulic degradation of a heat transport system. PT diametral creep leads to diametral expansion that affects the thermal hydraulic characteristics of the coolant channels and the critical heat flux. Therefore, it is essential to predict the PT diametral creep in CANDU reactors, which is caused mainly by fast neutron irradiation, reactor coolant temperature and so forth. The currently used PT diametral creep prediction model considers the complex interactions between the effects of temperature and fast neutron flux on the deformation of PT zirconium alloys. The model assumes that long-term steady-state deformation consists of separable, additive components from thermal creep, irradiation creep and irradiation growth. This is a mechanistic model based on measured data. However, this model has high prediction uncertainty. Recently, a statistical error modeling method was developed using plant inspection data from the Bruce B CANDU reactor. The aim of this study was to develop a bundle position-wise linear model (BPLM) to predict PT diametral creep employing previously measured PT diameters and HTS operating conditions. There are twelve bundles in a fuel channel and for each bundle, a linear model was developed by using the dependent variables, such as the fast neutron fluxes and the bundle temperatures. The training data set was selected using the subtractive clustering method. The data of 39 channels that consist of 80 percent of a total of 49 measured channels from Units 2, 3 and 4 were used to develop the BPLM models. The remaining 10 channels' data were used to test the developed BPLM models. The BPLM was optimized by the maximum likelihood estimation method. The developed BPLM to predict PT diametral creep was verified using the operating data gathered from the Units 2,3 and 4 in Korea. Two error components for the BPLM, which are the

  18. BEAM 1.7: development for modelling fuel element and bundle buckling strength

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, G.; Xu, S.; Xu, Z.; Paul, U.K. [Atomic Energy of Canada, Mississauga, Ontario (Canada)

    2010-07-01

    This paper describes BEAM, an AECL developed computer program, used to assess mechanical integrity of CANDU fuel bundles. The BEAM code has been developed to satisfy the need for buckling strength analysis of fuel bundles. Buckling refers to the phenomenon where a compressive axial load is large enough that a small lateral load can cause large lateral deflections. The buckling strength refers to the critical compressive axial load at which lateral instability is reached. The buckling strength analysis has practical significance for the design of fuel bundles, where the buckling strength of a fuel element/bundle is assessed so that the conditions leading to bundle jamming in the pressure tube are excluded. This paper presents the development and qualification of the BEAM code, with emphasis on the theoretical background and code implementation of the newly developed fuel element/bundle buckling strength model. (author)

  19. Development of a System Dynamics Model for Evaluating the Economics of an Advanced CANDU Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Jong Yeob; Park, Joo Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-10-15

    Since the early 1990's, the Korea Atomic Energy Research Institute (KAERI) and the Atomic Energy of Canada Limited (AECL) have cooperated to develop, verify, and demonstrate the advanced CANDU fuel, so called CANFLEX-NU (Natural Uranium). The CANFLEX-NU fuel bundle consists of 43 fuel elements and has the buttons on the outer surface of the fuel elements for improving the CHF (Critical-Heat-Flux) characteristics. Because of this features of CANFLEXNU fuel, it offers higher operating and safety margins than current 37-element fuel. Recently, the interest for a CANFLEX-NU has been increased because of the power de-rating due to aging of CANDU reactors. Wolsong Unit 1 CANDU reactor has been operated over 25 years and the operating power at the present time is less than 90% of a full power because of a reduction of the margin of ROP trip set point. The most appropriate way to overcome such a power de-rating due to a crept pressure tube is the introduction of a CANFLEX-NU fuel into a CANDU reactor. Now, a CANFLEX-NU fuel is ready to be commercialized in a CANDU-6 reactor because the design and demonstration irradiation have been completed in both Korea and Canada. Economic evaluation for commercializing a CANFLEX-NU fuel in Wolsong Units was carried out by calculating the unit prime cost of electricity production. Throughout the economic evaluation, it was found that the introduction of CANFLEX-NU fuel into Wolsong Units would have much economic benefits due to a better operating performance. However, the amount of economic profit due to introducing CANFLEX-NU fuel depends on several parameters such as the required time to get license from regulatory institute before commercializing, licensing cost, failure probability of commercializing etc. Therefore, it is necessary to determine the optimum condition to get the highest economic profit. In this paper, an economic evaluation was carried out based on the starting year of the licensing study with considering the

  20. 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.

  1. Procurement and supply of CANDU fuels

    Energy Technology Data Exchange (ETDEWEB)

    Bazeley, E.G. [E.G. Bazeley and Associates, Whitby, Ontario (Canada)

    2002-11-01

    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)

  2. SARAPAN—A Simulated-Annealing-Based Tool to Generate Random Patterned-Channel-Age in CANDU Fuel Management Analyses

    Directory of Open Access Journals (Sweden)

    Doddy Kastanya

    2017-02-01

    Full Text Available In any reactor physics analysis, the instantaneous power distribution in the core can be calculated when the actual bundle-wise burnup distribution is known. Considering the fact that CANDU (Canada Deuterium Uranium utilizes on-power refueling to compensate for the reduction of reactivity due to fuel burnup, in the CANDU fuel management analysis, snapshots of power and burnup distributions can be obtained by simulating and tracking the reactor operation over an extended period using various tools such as the *SIMULATE module of the Reactor Fueling Simulation Program (RFSP code. However, for some studies, such as an evaluation of a conceptual design of a next-generation CANDU reactor, the preferred approach to obtain a snapshot of the power distribution in the core is based on the patterned-channel-age model implemented in the *INSTANTAN module of the RFSP code. The objective of this approach is to obtain a representative snapshot of core conditions quickly. At present, such patterns could be generated by using a program called RANDIS, which is implemented within the *INSTANTAN module. In this work, we present an alternative approach to derive the patterned-channel-age model where a simulated-annealing-based algorithm is used to find such patterns, which produce reasonable power distributions.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-06-15

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

  4. Leaching of used CANDU fuel: Results from a 19-year leach test under oxidizing conditions

    Energy Technology Data Exchange (ETDEWEB)

    Stroes-Gascoyne, S.; Johnson, L.H.; Tait, J.C.; McConnell, J.L.; Porth, R.J. [AECL, Pinawa, Manitoba (Canada). Whiteshell Labs.

    1997-12-31

    A fuel leaching experiment has been in progress since 1977 to study the dissolution behavior of used CANDU fuel in aerated aqueous solution. The experiment involves exposure of 50-mm clad segments of an outer element of a Pickering fuel bundle (burnup 610 GJ/kg U; linear and peak power ratings 53 and 58 kW/m, respectively), to deionized distilled water (DDH{sub 2}O, {approximately}2 mg/L carbonate) and tapwater ({approximately}50 mg/L carbonate). In 1992, it was observed that the fuel in at least one of the leaching solutions showed some signs of deterioration and, therefore, in 1993, parts of the fuel samples were sacrificed for a detailed analysis of the physical state of the fuel, using SEM and optical microscopy. Leaching results to date show that even after >6900 days only 5 to 7.7% of the total calculated inventory of {sup 137}Cs has leached out preferentially and that leach rates suggest a development towards congruent dissolution. Total amounts of {sup 137}Cs and {sup 90}Sr leached are slightly larger in tapwater than in DDH{sub 2}O. SEM examinations of leached fuel surface fragments indicate that the fuel surface exposed to DDH{sub 2}O is covered in a needle-like precipitate. The fuel surface exposed to tapwater shows evidence of leaching but no precipitate, likely because uranium is kept in solution by carbonate. Detailed optical and SEM microscopy examinations on fuel cross sections suggest that grain-boundary dissolution in DDH{sub 2}O is not prevalent, and in tapwater appears to be limited to the outer {approximately}0.5 mm (pellet/cladding) region of the fuel. Grain boundary attack seems to be limited to microcracks at or near the surface of the fuel. It thus appears that grain-boundary attack occurs only near the fuel pellet surface and is prevalent only in the presence of carbonate in solution.

  5. ENHANCING ADVANCED CANDU PROLIFERATION RESISTANCE FUEL WITH MINOR ACTINIDES

    Energy Technology Data Exchange (ETDEWEB)

    Gray S. Chang

    2010-05-01

    The advanced nuclear system will significantly advance the science and technology of nuclear energy systems and to enhance the spent fuel proliferation resistance. Minor actinides (MA) are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. MAs can play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. In this work, an Advanced CANDU Reactor (ACR) fuel unit lattice cell model with 43 UO2 fuel rods will be used to investigate the effectiveness of a Minor Actinide Reduction Approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance. The main MARA objective is to increase the 238Pu / Pu isotope ratio by using the transuranic nuclides (237Np and 241Am) in the high burnup fuel and thereby increase the proliferation resistance even for a very low fuel burnup. As a result, MARA is a very effective approach to enhance the proliferation resistance for the on power refueling ACR system nuclear fuel. The MA transmutation characteristics at different MA loadings were compared and their impact on neutronics criticality assessed. The concept of MARA, significantly increases the 238Pu/Pu ratio for proliferation resistance, as well as serves as a burnable absorber to hold-down the initial excess reactivity. It is believed that MARA can play an important role in atoms for peace and the intermediate term of nuclear energy reconnaissance.

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

    Energy Technology Data Exchange (ETDEWEB)

    Marino, A.C. [Comison Nacional de Energia Atomica (Argentina)

    1997-07-01

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

  7. 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.

  8. 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)

  9. Development of CANDU advanced fuel fabrication technology - A development of amorphous alloys for the solder of nuclear reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jai Young; Lee, Ki Young; Kim, Yoon Kee; Jung, Jae Han; Yu, Ji Sang; Kim, Hae Yeol; Han, Young Su [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1996-07-01

    In the case of advanced CANDU fuel being useful in future, the fabrication processes for soundness insurance of a improved nuclear fuel bundle must be developed at the same time because it have three times combustibility as existing fuel. In particular, as the improved nuclear fuel bundle in which a coated layer thickness is thinner than existing that, firmity of a joint part is very important. Therefore, we need to develop a joint technique using new solder which can settle a potential problem in current joining method. As the Zr-Be alloy system and the Ti-Be system are composed with the elements having high neutron permeability, they are suitable for joint of nuclear fuel pack. The various compositions Zr-Be and Ti-Be binary metallic glass alloys were applicable to the joining the nuclear fuel bundles. The thickness of joint layer using the Zr{sub 1-x} Be{sub x} amorphous ribbon as a solder is thinner than that using physical vapor deposited Be. Among the Zr{sub 1-x} Be{sub x} amorphous binary alloys, Zr{sub 0.7} Be{sub 0.3} binary alloy is the most appropriated for joint of nuclear fuel bundle because its joint layer is smooth and thin due to low degree of Be diffusion. The microstructures of brazed layer using Ti{sub 1-y} Be{sub y} alloy, however, a solid-solution layer composed with Zr and Ti is formed toward the Zr cladding sheath and many of Zr is detected in the joint lever. 20 refs., 8 tabs., 23 figs. (author)

  10. 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)

  11. Extending the world's uranium resources through advanced CANDU fuel cycles

    Energy Technology Data Exchange (ETDEWEB)

    De Vuono, Tony; Yee, Frank; Aleyaseen, Val; Kuran, Sermet; Cottrell, Catherine

    2010-09-15

    The growing demand for nuclear power will encourage many countries to undertake initiatives to ensure a self-reliant fuel source supply. Uranium is currently the only fuel utilized in nuclear reactors. There are increasing concerns that primary uranium sources will not be enough to meet future needs. AECL has developed a fuel cycle vision that incorporates other sources of advanced fuels to be adaptable to its CANDU technology.

  12. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  14. A controllability study of TRUMOX fuel for load following operations in a CANDU-900 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Trudell, D.A., E-mail: trudelda@mcmaster.ca [McMaster Univ., Hamilton, Ontario (Canada)

    2012-07-01

    Using a core model of a generic CANDU-900 reactor in RFSP-IST, load following simulations have been performed to assess the controllability of the reactor due to Xenon transients. Week long load following simulations have been performed with daily power cycles 12 hours in duration. Simulations have shown that Natural Uranium fuel can be safely cycled between 100 and 90% Full Power without adjuster rod movement while TRUMOX fuel can be safely cycled between 100 and 85% Full Power. (author)

  15. LWR nuclear fuel bundle data for use in fuel bundle handling

    Energy Technology Data Exchange (ETDEWEB)

    Weihermiller, W.B.; Allison, G.S.

    1979-09-01

    Although increasing numbers of spent light water reactor (LWR) fuel bundles are moved into storage, no handling equipment is set up to manipulate all of the various types of fuel bundles. This report summarizes fuel bundle information of interest to the designer of such handling equipment. Dimensional descriptions are included with discussions of assembly procedure and manufacturer provisions for handling equipment. No attempt is made to make a complete compilation of dimensional information; the number of fuel bundle designs and design revisions makes it impractical. Because the fuel bundle designs are so varied, any equipment intended for handling all types of bundles will have to be designed with flexibility in mind. Besides the ability to manipulate fuel bundles in space, handling equipment may be required to locate an external surface or to position a cutting operation to avoid breaking a fuel rod pressure boundary. Even with the most sophisticated and flexible handling equipment, some situations will require use of the manufacturers' as-built descriptions of individual fuel bundles.

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

    Energy Technology Data Exchange (ETDEWEB)

    Sellers, M.T.; Corcoran, E.C.; Kelly, D.G., E-mail: David.Kelly@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada)

    2012-07-01

    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 {sup 233}U and {sup 235}U 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 UO{sub 2}, thoria and mixed oxide (MOX) based) by DNC analysis is evaluated through Monte Carlo simulations. (author)

  17. Measurement of gap and grain-boundary inventories of {sup 129}I in used CANDU fuels

    Energy Technology Data Exchange (ETDEWEB)

    Stroes-Gascoyne, S.; Moir, D.L.; Kolar, M.; Porth, R.J.; McConnell, J.L.; Kerr, A.H. [AECL Research, Pinawa, Manitoba (Canada). Whiteshell Labs.

    1995-12-31

    Combined gap and grain-boundary inventories of {sup 129}I in 14 used CANDU fuel elements were measured by crushing and simultaneously leaching fuel segments for 4 h in a solution containing KI carrier. From analogy with previous work a near one-to-one correlation was anticipated between the amount of stable Xe and the amount of {sup 128}I in the combined gap and grain-boundary regions of the fuel. However, the results showed that such a correlation was only apparent for low linear power rating (LLPR) fuels with an average linear power rating of < 42 kW/m. For high linear power rating (HLPR) fuels (> 44 kW/m), the {sup 129}I values were considerably smaller than expected. The combined gap and grain-boundary inventories of {sup 129}I in the 14 fuels tested varied from 1.8 to 11.0%, with an average value of 3.6 {+-} 2.4% which suggests that the average value of 8.1 {+-} 1% used in safety assessment calculations overestimates the instant release fraction for {sup 129}I. Segments of used CANDU fuels were leached for 92 d (samples taken at 5, 28 and 92 d) to determine the kinetics of {sup 129}I release. Results could be fitted tentatively to half-order reaction kinetics, implying that {sup 129}I release is a diffusion-controlled process for LLPR fuels, and also for HLPR fuels, once the gap inventory has been leached. However, more data are needed over longer leaching periods to gain more understanding of the processes that control grain-boundary release of {sup 129}I from used CANDU fuel.

  18. FAST: A Fuel And Sheath Modeling Tool for CANDU Reactor Fuel

    Science.gov (United States)

    Prudil, Andrew Albert

    before for any Canadian fuel performance code). This thesis documents the theory employed by the model, its implementation, and the results of a proof of concept validation. The validation compared model predictions against both experimental data and results obtained from the ELESTRES and ELOCA fuel performance codes. Overall, the results show excellent model performance except in cases of a strong axial dependence. An analysis of the sensitivity of the model to the uncertainty in input parameters and the material properties is also presented. Finally, this thesis includes a discussion of the limitations, applications, and potential for future development of code. Key words: nuclear fuel, CANDU fuel, fuel modeling, multiphysics modeling, Comsol

  19. Measurements of turbulent pressures of flow in a water-conveying pipe containing a simulation fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Abbasian, F.; Cao, J.; Yu, S.D. [Ryerson Univ., Dept. of Mechanical and Industrial Engineering, Toronto, Ontario (Canada)

    2008-07-01

    A test apparatus was set up to investigate the turbulent flows and flow induced vibrations in a fluid-conveying pipe containing a CANDU 43-element simulation fuel bundle. The fuel bundle is immersed in test pipe of 4-inch in diameter. A centrifugal pump circulates fresh water with a maximum velocity of 9 m/s at full pump power. The pressure fluctuation near the inner surface of the flow channel was measured at various locations using a pressure transducer and a data acquisition system. It was found that the turbulence away from the test section containing the simulation fuel bundle is largely caused by the pipe flow of high Reynolds number; the turbulence near and inside the bundle structures is the result of pipe flow and fluid-solid interactions. The measurements of pressures near the fuel bundle structure showed that the power spectral density (PSD) of pressure fluctuation has a frequency range of 1-300 Hz, and a normalized maximum pressure range of 0.04 to 0.05 times dynamic pressure. The effects of bundle angular alignments and subchannels on the pressure spectra, Strouhal number range, and streamwise pressure drop are also investigated in this paper. Results presented in this paper are useful in validating the computational models for flow-induced fluid forces that cause the fuel bundle structure to rock and fret. (author)

  20. A feasibility study on the use of the MOOSE computational framework to simulate three-dimensional deformation of CANDU reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, Kyle A., E-mail: Kyle.Gamble@inl.gov [Royal Military College of Canada, Chemistry and Chemical Engineering, 13 General Crerar Crescent, Kingston, Ontario, Canada K7K 7B4 (Canada); Williams, Anthony F., E-mail: Tony.Williams@cnl.ca [Canadian Nuclear Laboratories, Fuel and Fuel Channel Safety, 1 Plant Road, Chalk River, Ontario, Canada K0J 1J0 (Canada); Chan, Paul K., E-mail: Paul.Chan@rmc.ca [Royal Military College of Canada, Chemistry and Chemical Engineering, 13 General Crerar Crescent, Kingston, Ontario, Canada K7K 7B4 (Canada); Wowk, Diane, E-mail: Diane.Wowk@rmc.ca [Royal Military College of Canada, Mechanical and Aerospace Engineering, 13 General Crerar Crescent, Kingston, Ontario, Canada K7K 7B4 (Canada)

    2015-11-15

    Highlights: • This is the first demonstration of using the MOOSE framework for modeling CANDU fuel. • Glued and frictionless contact algorithms behave as expected for 2D and 3D cases. • MOOSE accepts and correctly interprets functions of arbitrary form. • 3D deformation calculations accurately compare against analytical solutions. • MOOSE is a viable simulation tool for modeling accident reactor conditions. - Abstract: Horizontally oriented fuel bundles, such as those in CANada Deuterium Uranium (CANDU) reactors present unique modeling challenges. After long irradiation times or during severe transients the fuel elements can laterally deform out of plane due to processes known as bow and sag. Bowing is a thermally driven process that causes the fuel elements to laterally deform when a temperature gradient develops across the diameter of the element. Sagging is a coupled mechanical and thermal process caused by deformation of the fuel pin due to creep mechanisms of the sheathing after long irradiation times and or high temperatures. These out-of-plane deformations can lead to reduced coolant flow and a reduction in coolability of the fuel bundle. In extreme cases element-to-element or element-to-pressure tube contact could occur leading to reduced coolant flow in the subchannels or pressure tube rupture leading to a loss of coolant accident. This paper evaluates the capability of the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework developed at the Idaho National Laboratory to model these deformation mechanisms. The material model capabilities of MOOSE and its ability to simulate contact are also investigated.

  1. Waste management issues and their potential impact on technical specifications of CANDU fuel materials

    Energy Technology Data Exchange (ETDEWEB)

    Tait, J.C.; Johnson, L.H. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada)

    1997-07-01

    The technical specifications for the composition of nuclear fuels and materials used in Canada's CANDU reactors have been developed by AECL and materials manufacturers, taking into account considerations specific to their manufacture and the effect of minor impurities on fuel behaviour in reactor. Nitrogen and chlorine are examples of UO{sub 2} impurities, however, where there is no technical specification limit. These impurities are present in the source materials or introduced in the fabrication process and are neutron activated to {sup 14}C and {sup 36}C1, which after {sup 129}I , are the two most significant contributors to dose in safety assessments for the disposal of used fuel. For certain impurities, environmental factors, particularly the safety of the disposal of used fuels, should be taken into consideration when deriving 'allowable' impurity limits for nuclear fuel materials. (author)

  2. Effect of DUPIC cycle on CANDU reactor safety parameters

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

  3. Design requirements of a consolidating dry storage module for CANDU spent fuels

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kyung Ho; Yoon, Jeong Hyoun; Yang, Ke Hyung; Choi, Byung Il; Lee, Heung Young [KHNP/NETEC, Taejon (Korea, Republic of); Cho, Gyu Seong [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2003-10-01

    This paper presents a technical description of design requirement document covers the requirements of the MACSTOR/KN-400 module, which is under development to densely accommodate CANDU spent fuels with more efficient way. The design requirement is for the module that will be constructed within a dry storage site after successfully licensed by the regulatory body. This temporary outdoor spent fuel dry storage facility provides for safe storage of spent nuclear fuel after it has been removed from the plant's storage pool after being allowed to decay for a period of at least 6 years. The MACSTOR/KN-400 module is being designed to the envelope of site environmental conditions encountered at the Wolsong station. The design requirements of MACSTOR/KN-400 module meets the requirements of the appropriate Codes and Standards for dry storage of spent fuel from nuclear power reactors such as lOCFR72, and Korea Atomic Energy Act and relevant technical standard.

  4. 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.

  5. 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)

  6. FGR Evaluation Code for CANDU Fuel under an Accidents: REDOU

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Jung, Jong Yeob

    2006-12-15

    When an end fitting failure is occurred, the sheath is broken and the fission gases are released promptly from the gap between the pellet and sheath and also released continuously from the inside of the pellet due to the oxidation of the pellet. Thus, the accurate calculation of the fission gas release from the gap and from the inside of the pellet is an essential to the accident analysis of the fuel behavior and to prepare the safety strategy. In this report, among the performance analysis or the transient behavior prediction computer codes, the REDOU code which is the fission gas calculating code for the postulated accident scenario such as an end fitting failure is introduced. Then, the user manual for REDOU code is provided so that it can be the guidance to the potential users of the code and save the time and economic loss by reducing the trial and error.

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

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, Grigore; Ionescu, Drags; Pauna, Eduard [Institute for Nuclear Research, Pitesti (Romania). Nuclear Fuel Engineering Lab.

    2012-03-15

    When nuclear power reactors are operated in a load following (LF) mode, the nuclear fuel may be subjected to step changes in power on weekly, daily, or even hourly basis, depending on the grid's needs. Two load following tests performed in TRIGA Research Reactor of Institute for Nuclear Research (INR) Pitesti were simulated with finite elements computer codes in order to evaluate Stress Corrosion Fatigue (SCF) of the sheath arising from expansion and contraction of the pellets in the corrosive environment. The 3D finite element analyses show that the cyclic strains give highly multiaxial stresses in the sheath at ridge region. This paper summarizes the results of the analytical assessment for SCF and their relation to CANDU fuel performance in LF tests conditions. (orig.)

  8. 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.

  9. Design and Development of a Robotic Crawler for CANDU Fuel Channel Inspection

    Science.gov (United States)

    Shukla, Shivam

    For the design of a new robotic crawler drive unit for CANDU fuel channel inspection, a complete design and screening process was done in order to fulfil the objective of this research. A brief explanation of CANDU reactors is provided along with a discussion of the inspection systems that are currently in use. A study of some existing inspection systems is presented which was used for the development of the new robotic crawler design. A number of concepts were generated which underwent a screening process with the help of two design tools. With the help of these tools, a concept was chosen as the final design and details of it are presented. To demonstrate a proof-of-concept, the physical prototype of the robotic crawler was manufactured and assembled. A speed controller was implemented in the final design of the robotic crawler. A set of test procedures were performed on the final design and the results are discussed. Some improvements that can be done on the final design of the robotic crawler are also discussed in the final section of this thesis.

  10. 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

  11. 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)

  12. Evaluation of maximum allowable temperature inside basket of dry storage module for CANDU spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kyung Ho; Yoon, Jeong Hyoun; Chae, Kyoung Myoung; Choi, Byung Il; Lee, Heung Young; Song, Myung Jae [Nuclear Environment Technology Institute, Taejon (Korea, Republic of); Cho, Gyu Seong [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2002-10-01

    This study provides a maximum allowable fuel temperature through a preliminary evaluation of the UO{sub 2} weight gain that may occur on a failed (breached sheathing) element of a fuel bundle. Intact bundles would not be affected as the UO{sub 2} would not be in contact with the air for the fuel storage basket. The analysis is made for the MACSTOR/KN-400 to be operated in Wolsong ambient air temperature conditions. The design basis fuel is a 6-year cooled fuel bundle that, on average has reached a burnup of 7,800 MWd/MTU. The fuel bundle considered for analysis is assumed to have a high burnup of 12,000 MWd/MTU and be located in a hot basket. The MACSTOR/KN-400 has the same air circuit as the MACSTOR and the air circuit will require a slightly higher temperature difference to exit the increased heat load. The maximum temperature of a high burnup bundle stored in the new MACSTOR/KN-400 is expected to be about 9 .deg. C higher than the fuel temperature of the MACSTOR at an equivalent constant ambient temperature. This temperature increase will in turn increase the UO{sub 2} weight gain from 0.06% (MACSTOR for Wolsong conditions) to an estimated 0.13% weight gain for the MACSTOR/KN-400. Compared to an acceptable UO{sub 2} weight gain of 0.6%, we are thus expecting to maintain a very acceptable safety factor of 4 to 5 for the new module against unacceptable stresses in the fuel sheathing. For the UO{sub 2} weight gain, the maximum allowable fuel temperature was shown by 164 .deg. C.

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

    Energy Technology Data Exchange (ETDEWEB)

    Choi, K. S.; Yang, K. H.; Paek, C. R.; Jung, J. S.; Lee, H. Y. [Korea Hydro and Nuclear Power Company, Taejon (Korea, Republic of)

    2003-07-01

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

  14. 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.

  15. 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.

  16. Localization of CANDU technology

    Energy Technology Data Exchange (ETDEWEB)

    Alizadeh, Ala

    2010-09-15

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

  17. 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.

  18. 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.

  19. A comprehensive model for in-plane and out-of-plane vibration of CANDU fuel endplate rings

    Energy Technology Data Exchange (ETDEWEB)

    Yu, S.D., E-mail: syu@ryerson.ca; Fadaee, M.

    2016-08-01

    Highlights: • Proposed an effective method for modelling bending and torsional vibration of CANDU fuel endplate rings. • Applied successfully the thick plate theory to curved structural members by accounting for the transverse shear effect. • The proposed method is computationally more efficient compared to the 3D finite element. - Abstract: In this paper, a comprehensive vibration model is developed for analysing in-plane and out-of-plane vibration of CANDU fuel endplate rings by taking into consideration the effects of in-plane extension in the circumferential and radial directions, shear, and rotatory inertia. The model is based on Reddy’s thick plate theory and the nine-node isoparametric Lagrangian plate finite elements. Natural frequencies of various modes of vibration of circular rings obtained using the proposed method are compared with 3D finite element results, experimental data and results available in the literature. Excellent agreement was achieved.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

  1. Heat transfer analysis of the MACSTOR/KN-400 storage module for CANDU spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. H.; Youn, J. H.; Choi, B. I.; Lee, H. Y. [Nuclear Environment Technology Institute, Taejon (Korea, Republic of)

    2003-10-01

    It was verified through heat transfer analysis that a consolidated dry storage system for CANDU spent fuel, MACSTOR/KN-400 was safe in thermal aspect. In order to validate the computer code of CATHENA which was employed to perform the analysis, the comparison between actual measurement data of MACSTOR-200 at Getilly-2 NPP in Canada and computed values from the code has been carried out. The comparison represented that the computed values acceptably agreed to the measurement data and thus the computer code was verified for its application to MACSTOR/KN-400. The identical K-values(parameter to describe head loss inside the module) and convective heat transfer coefficient of the module obtained by the validation was applied to the heat transfer analysis modelling of MACSTOR/KN-400. The result from the analysis showed that under 40 .deg. C of ambient temperature, maximum average and local temperatures of the concrete module were represented by 53 .deg. C and 69 .deg. C, respectively, which fulfilled well the allowable temperature limit of the concrete structure given by ACI349(American Concrete Institute)

  2. Release of [sup 14]C from the gap and grain-boundary regions of used CANDU fuels to aqueous solutions

    Energy Technology Data Exchange (ETDEWEB)

    Stroes-Gascoyne, S.; Tait, J.C.; Porth, R.J.; McConnell, J.L.; Lincoln, W.J. (Whiteshell Lab., Pinawa, Manitoba (Canada). AECL Research)

    1994-01-01

    This study was undertaken as part of the Canadian Nuclear Fuel Waste Management Program (CNFWMP), to measure [sup 14]C inventories of used CANDU fuel. Other objectives were to measure the fraction of the total [sup 14]C 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 [sup 14]C were correct. Results showed that the measured [sup 14]C inventories were a factor of 11.5 [+-] 3.9 lower than the estimated [sup 14]C inventory values used in safety assessment calculations. Measured instant release values for [sup 14]C ranged from 0.06 to 5.04% (of total [sup 14]C inventories) with an average of 2.7 [+-] 1.6%, indicating that instant release fractions for [sup 14]C used in safety assessment calculations (1.2--25%) were overestimated.

  3. 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.

  4. Pre-licensing of the Advanced CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ion, R.; Popov, N.; Snell, V.; Xu, C. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); West, J. [Candesco Co., Toronto, Ontario (Canada)

    2006-09-15

    Atomic Energy of Canada Limited (AECL) developed the Advanced CANDU Reactor-700 (ACR-700) as an evolutionary advancement of the current CANDU 6 reactor. As further advancement of the ACR design, AECL is currently developing the ACR-1000 for the Canadian and international market. The ACR-1000 is aimed at producing electrical power for a capital cost and a unit-energy cost significantly less than that of the current generation of operating nuclear plants, while achieving shorter construction schedule, high plant capacity factor, improved operations and maintenance, increased operating life, and enhanced safety features. The reference ACR-1000 plant design is based on an integrated two-unit plant, using enriched fuel and light-water coolant, with each unit having a nominal gross electrical output of 1165 MWe. The ACR-1000 design has evolved from AECL's in-depth knowledge of CANDU systems, components, and materials, as well as the experience and feedback received from owners and 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. It also features major improvements in economics, inherent safety characteristics, and performance, while retaining the proven benefits of the CANDU family of nuclear power plants. The CANDU system is ideally suited to this evolutionary approach since the modular fuel channel reactor design can be modified, through a series of incremental changes in the reactor core design, to increase the power output and improve the overall safety, economics, and performance. The safety enhancements made in ACR-1000 encompass improved safety margins, performance and reliability of safety related systems. In particular, the use of the CANFLEX-ACR fuel bundle, with lower linear rating and higher critical heat flux, provides increased operating and safety margins. Safety features draw from those of the existing CANDU plants (e.g., the two

  5. FAST: A combined NOC and transient fuel model for CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Prudil, A.; Lewis, B.J.; Chan, P.K., E-mail: Paul.Chan@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada); Baschuk, J.J. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2013-07-01

    The Fuel And Sheath modelling Tool (FAST) is a fuel performance code that is being developed for both normal and transient operating conditions. FAST includes models for heat generation and transport, thermal-expansion, elastic strain, densification, fission product swelling, pellet relocation, contact, grain growth, fission gas release, gas and coolant pressure and sheath creep. These models have been implemented using the Comsol finite-element platform. The equations are solved on a two-dimensional (radial-axial) geometry of a fuel pellet and sheath. FAST has undergone a proof of concept validation against experimental data and comparison to the ELESTRES and ELOCA fuel performance codes. The results show excellent agreement with experimental measurements and the above stated IST- codes. (author)

  6. Evaluation of CANDU6 PCR (power coefficient of reactivity) with a 3-D whole-core Monte Carlo Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Motalab, Mohammad Abdul; Kim, Woosong; Kim, Yonghee, E-mail: yongheekim@kaist.ac.kr

    2015-12-15

    Highlights: • The PCR of the CANDU6 reactor is slightly negative at low power, e.g. <80% P. • Doppler broadening of scattering resonances improves noticeably the FTC and make the PCR more negative or less positive in CANDU6. • The elevated inlet coolant condition can worsen significantly the PCR of CANDU6. • Improved design tools are needed for the safety evaluation of CANDU6 reactor. - Abstract: The power coefficient of reactivity (PCR) is a very important parameter for inherent safety and stability of nuclear reactors. The combined effect of a relatively less negative fuel temperature coefficient and a positive coolant temperature coefficient make the CANDU6 (CANada Deuterium Uranium) PCR very close to zero. In the original CANDU6 design, the PCR was calculated to be clearly negative. However, the latest physics design tools predict that the PCR is slightly positive for a wide operational range of reactor power. It is upon this contradictory observation that the CANDU6 PCR is re-evaluated in this work. In our previous study, the CANDU6 PCR was evaluated through a standard lattice analysis at mid-burnup and was found to be negative at low power. In this paper, the study was extended to a detailed 3-D CANDU6 whole-core model using the Monte Carlo code Serpent2. The Doppler broadening rejection correction (DBRC) method was implemented in the Serpent2 code in order to take into account thermal motion of the heavy uranium nucleus in the neutron-U scattering reactions. Time-average equilibrium core was considered for the evaluation of the representative PCR of CANDU6. Two thermal hydraulic models were considered in this work: one at design condition and the other at operating condition. Bundle-wise distributions of the coolant properties are modeled and the bundle-wise fuel temperature is also considered in this study. The evaluated nuclear data library ENDF/B-VII.0 was used throughout this Serpent2 evaluation. In these Monte Carlo calculations, a large number

  7. Feasibility study in aspect of thermal integrity on the dry storage expansion options for CANDU spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. H.; Yoon, J. H.; Choi, B. I.; Lee, H. Y.; Song, M. J. [Nuclear Environment Technology Institute, Taejon (Korea, Republic of); Cho, K. S. [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2002-10-01

    In order to expand the capability of the CANDU spent fuel dry storage facilities of the at Wolsong, the alternative concepts based on MACSTOR are suggested to replace with existing concrete silo of Wolsong. For this, the feasibility of its design changes from original MACSTOR is examined in term of heat transfer and thermal hydraulic. In this study, the configuration of the module was conceptually changed from its original 2 rows to 3 and 4 rows for review. Under normal operation, the results of heat transfer and thermal hydraulic shows that storage module can feasibly accomodate four rows of storage cylinders within allowable range in terms of maximum allowable temperature of the fuel basket.

  8. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Abbasian, F.; Hadaller, G.I.; Fortman, R.A. [Stern Laboratories, Hamilton, Ontario (Canada)

    2010-07-01

    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)

  10. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  12. 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)

  13. 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.

  14. 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)

  15. 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

  16. Diagnostic technology for degradation of feeder pipes and fuel channels in CANDU reactor; development of aging assessment technology for CANDU pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jin; Kim, Yun Jae; Huh, Nam Su; Kwak, Sang Log; Lee, Kyu Ho [Sungkyunkwan University, Seoul (Korea)

    2002-04-01

    This research project attempts to resolve two issues related to integrity assessment of CANDU pressure tubes; (1) FE analysis of blister formation and growth, and (2) engineering estimation scheme to predict creep deflection of pressure tubes. Results for blister formation and growth can be summarised as follows. Comparing the results from the FE analysis, developed within this project, with experimental data shows some differences ranging from 10-57%. Such difference results from two possible sources. One source is neglecting two phase diffusion. The present FE analysis considers only single phase diffusion, and thus blister growth can not be accurately modeled. The other source would be inherent errors associated with experimental measurement. Thus it has been concluded that further efforts should be made on two phase diffusion modeling. For developing mechanistic model of creep deflection, the proposed reference stress based model is simple to use. Extensive validation against creep FE results shows that the proposed model is also quite accurate. More important aspect of the proposed method is that it can be easily generalized to more complex problems. Thus it is believed that the present results provide a sound basis for sagging assessment of CANDU pressure tubes. 16 refs., 12 figs., 6 tabs. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. H.; Yoon, J. H.; Choi, B. I.; Lee, H. Y. [Korea Hydraulic and Nuclear Power Company, Taejon (Korea, Republic of)

    2004-07-01

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

  18. 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

  19. 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)

  20. 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.

  1. 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.

  2. 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)

  3. A Deformation Analysis Code of CANDU Fuel under the Postulated Accident: ELOCA

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Jung, Jong Yeob

    2006-11-15

    Deformations of the fuel element or fuel channel might be the main cause of the fuel failure. Therefore, the accurate prediction of the deformation and the analysis capabilities are closely related to the increase of the safety margin of the reactor. In this report, among the performance analysis or the transient behavior prediction computer codes, the analysis codes for deformation such as the ELOCA, HOTSPOT, CONTACT-1, and PTDFORM are briefly introduced and each code's objectives, applicability, and relations are explained. Especially, the user manual for ELOCA code which is the analysis code for the fuel deformation and the release of fission product during the transient period after the postulated accidents is provided so that it can be the guidance to the potential users of the code and save the time and economic loss by reducing the trial and err000.

  4. 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)

  5. 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.

  6. 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)

  7. 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)

  8. 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)

  9. 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.

  10. 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.

  11. Development of CANDU pressure tube integrity evaluation system

    Energy Technology Data Exchange (ETDEWEB)

    Kwac, S. L.; Kim, Y. J. [Sungkyunkwan Univ., Seoul (Korea, Republic of); Lee, J. S. [Kyonggi Univ., Suwon (Korea, Republic of); Park, Y. W. [KINS, Taejon (Korea, Republic of)

    1999-05-01

    The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle and it's containment vessel. If a flaw or contact with their calandria tubes is found during the periodic inspection, the integrity evaluation must be carried out, and the safety requirements must be satisfied for continued service. In order to perform the integrity evaluation, complicated and iterative calculation procedures are required. Besides, a lot of data and knowledge for the evaluation are required for the integrity evaluation process. For this reason, an integrity evaluation system was developed. The developed system was built on the basis of ASME Sec. XI and FFSG(Fitness For Service Guidelines for zirconium alloy pressure tubes in operating CANDU reactors) issued by the AECL. The evaluation procedure includes the crack growth calculation both by DHC and by fatigue. It also provides the prediction of fracture initiation, plastic collapse and leak-before-break(LBB), blister formation and blister growth. This system provides various databases including the 3-D finite element analysis results on pressure tubes, inspection data and design specifications. In order to verify the developed system, several case studies have been performed and the results were compared with those from AECL. A good agreement was observed between those two results.

  12. Measurements of grain-boundary inventories of {sup 137}Cs, {sup 90}Sr and {sup 99}Tc in used CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Stroes-Gascoyne, S.; Tait, J.C.; Porth, R.J.; McConnell, J.L.; Barnsdale, T.R.; Watson, S. [Whiteshell Labs., Pinawa, Manitoba (Canada)

    1993-12-31

    Two methods were used to measure grain-boundary inventories of {sup 137}Cs, {sup 90}Sr and {sup 99}Tc in used CANDU fuel, to corroborate source term estimates based on a fission gas release code. Used fuels were partially oxidized at 200{degrees}C in air to overall compositions of UO{sub 2+x} (0.15{<=} {times} {<=}0.25) to expose UO{sub 2} grain boundaries, followed by leaching in aqueous solution. Only a fraction (2 to 18%) of the calculated gap + grain-boundary inventories for {sup 137}Cs was released. This suggests that the calculations overestimate Cs release or that oxidation does not expose all grain boundaries, or that Cs release from grain boundaries is slow. Release of {sup 90}Sr (0.01 to 0.7%) agreed reasonably well with the source term estimates (0.001 to 0.3%). Release of {sup 99}Tc (0.3 to 1.5%) suggests that the source term estimate for the upper involved leaching of crushed and side-fractionated used fuel in either a static or dynamic system. A direct one-to-one correlation between calculated and measured gap + grain-boundary inventories for {sup 137}Cs was found for low- and medium-power fuels.

  13. 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

  14. 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.

  15. Failure probability estimation of flaw in CANDU pressure tube considering the dimensional change

    Energy Technology Data Exchange (ETDEWEB)

    Kwak, Sang Log; Kim, Young Jin [Sungkyunkwan Univ., Suwon (Korea, Republic of); Lee, Joon Seong [Kyonggi Univ., Suwon (Korea, Republic of); Park, Youn Won [KINS, Taejon (Korea, Republic of)

    2002-11-01

    The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle and heavy water coolant. Pressure tubes are installed horizontally inside the reactor and only selected samples are periodically examined during in-service inspection. In this respect, a probabilistic safety assessment method is more appropriate for the assessment of overall pressure tube safety. The failure behavior of CANDU pressure tubes, however, is governed by delayed hydride cracking which is the major difference from pipings and reactor pressure vessels. Since the delayed hydride cracking has more widely distributed governing parameters, it is impossible to apply a general PFM methodology directly. In this paper, a PFM methodology for the safety assessment of CANDU pressure tubes is introduced by applying Monte Carlo simulation in determining failure probability. Initial hydrogen concentration, flaw shape and depth, axial and radial crack growth rate and fracture toughness were considered as probabilistic variables. Parametric study has been done under the base of pressure tube dimension and hydride precipitation temperature in calculating failure probability. Unstable fracture and plastic collapse are used for the failure assessment. The estimated failure probability showed about three-order difference with changing dimensions of pressure tube.

  16. 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.

  17. 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.

  18. 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)

  19. 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.

  20. Simulations and measurements of adiabatic annular flows in triangular, tight lattice nuclear fuel bundle model

    Energy Technology Data Exchange (ETDEWEB)

    Saxena, Abhishek, E-mail: asaxena@lke.mavt.ethz.ch [ETH Zurich, Laboratory for Nuclear Energy Systems, Department of Mechanical and Process Engineering, Sonneggstrasse 3, 8092 Zürich (Switzerland); Zboray, Robert [Laboratory for Thermal-hydraulics, Nuclear Energy and Safety Department, Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Prasser, Horst-Michael [ETH Zurich, Laboratory for Nuclear Energy Systems, Department of Mechanical and Process Engineering, Sonneggstrasse 3, 8092 Zürich (Switzerland); Laboratory for Thermal-hydraulics, Nuclear Energy and Safety Department, Paul Scherrer Institute, 5232 Villigen PSI (Switzerland)

    2016-04-01

    High conversion light water reactors (HCLWR) having triangular, tight-lattice fuels bundles could enable improved fuel utilization compared to present day LWRs. However, the efficient cooling of a tight lattice bundle has to be still proven. Major concern is the avoidance of high-quality boiling crisis (film dry-out) by the use of efficient functional spacers. For this reason, we have carried out experiments on adiabatic, air-water annular two-phase flows in a tight-lattice, triangular fuel bundle model using generic spacers. A high-spatial-resolution, non-intrusive measurement technology, cold neutron tomography, has been utilized to resolve the distribution of the liquid film thickness on the virtual fuel pin surfaces. Unsteady CFD simulations have also been performed to replicate and compare with the experiments using the commercial code STAR-CCM+. Large eddies have been resolved on the grid level to capture the dominant unsteady flow features expected to drive the liquid film thickness distribution downstream of a spacer while the subgrid scales have been modeled using the Wall Adapting Local Eddy (WALE) subgrid model. A Volume of Fluid (VOF) method, which directly tracks the interface and does away with closure relationship models for interfacial exchange terms, has also been employed. The present paper shows first comparison of the measurement with the simulation results.

  1. 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.

  2. Conceptional design of test loop for FIV in fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Sim, W. G.; Yang, J. S.; Kim, S. W. [Hannam Univ., Taejeon (Korea)

    2001-01-01

    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., 23 figs., 2 tabs. (Author)

  3. Radionuclide Release after End Fitting Failure Accident in CANDU-6 Plant

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hoon; Kim, Yun Ho; Lee, Kwang Ho [Korea Electric Power Corporation Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    The total amount of mass and energy discharged into containment building from primary heat transport system in the event of end fitting failure accident in CANDU-6 plant is similar to that of small loss of coolant accident. But the ejection of fuel bundles into fuelling machine room is unique phenomenon and causes radio nuclides release from the physically broken fuel rod to outside containment building. The only objective of containment behavior analysis for end fitting failure event is to assess the amount of radio nuclides release to the ambient atmosphere. Radionuclide release rates in case of end fitting failure with all safety system available, that is containment building is intact, as well as with containment system impairment are analyzed with GOTHIC and SMART code

  4. 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

  5. Evaluation of the magnitude and effects of bundle duct interaction in fuel assemblies at developmental plant conditions

    Energy Technology Data Exchange (ETDEWEB)

    Serell, D.C.; Kaplan, S.

    1980-09-01

    Purpose of this evaluation is to estimate the magnitude and effects of irradiation and creep induced fuel bundle deformations in the developmental plant. This report focuses on the trends of the results and the ability of present models to evaluate the assembly temperatures in the presence of bundle deformation. Although this analysis focuses on the developmental plant, the conclusions are applicable to LMFBR fuel assemblies in general if they have wire spacers.

  6. 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

  7. 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.

  8. 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)

  9. 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.

  10. 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.

  11. Thermal hydraulic investigations on porous blockage in a prototype sodium cooled fast reactor fuel pin bundle

    Energy Technology Data Exchange (ETDEWEB)

    Raj, M.Naveen; Velusamy, K., E-mail: kvelu@igcar.gov.in; Maity, Ram Kumar

    2016-07-15

    Highlights: • We simulate flow and temperature fields in FBR fuel bundle with porous blockage. • We perform RANS-based CFD simulation for 217 pin bundle of 7 axial pitch lengths. • Flow reduction in fuel bundle due to porous internal blockage is estimated. • Monitoring bulk sodium outlet temperature does not guarantee blockage detection. • Admissible blockage length to avoid sodium boiling is determined. - Abstract: Thermal hydraulic characteristics of sodium flow in a prototype fuel subassembly with porous internal blockage have been investigated by computational fluid dynamics (CFD) simulations. CFD solutions for a subassembly having 217 pin bundle with seven helical pitch length were obtained by parallel processing. The CFD model has been validated against benchmark blockage experiment reported in literature. Wide parametric ranges for blockage radius, porosity, mean particle diameter and location of blockage have been considered. Critical length of blockage that would result in local sodium boiling as a function of aforementioned blockage parameters has been estimated and the parametric zone posing risk of sodium boiling has been identified. Attention has been paid to coolant mixing and flow and temperature fields downstream of the blockage zone. It is seen that for a prototype subassembly with various sections contributing to pressure loss, the total flow reduction is <2.5% for all blockages that can lead to local sodium boiling. This suggests, that global bulk sodium temperature monitoring at subassembly outlet is unlikely to detect slowly growing blockages. Comparing the sodium flow and temperature fields in unblocked and blocked bundles, it is found that the wake-induced temperature non-uniformity persist even upto 3 helical pitch length, highlighting that the sodium temperature non-uniformity at the bundle exit can serve as an efficient blockage indicator, provided that the cross-section temperature is mapped by a proper instrumentation. The peak

  12. 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

  13. 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.

  14. 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)

  15. Some applications related to the structural integrity analyses of CANDU 6 pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Radu, V.S. [Inst. for Nuclear Research Pitesti, Pitesti (Romania)]. E-mail: vradu@nuclear.ro

    2006-07-01

    The flaws found during in-service inspection of Zr-2.5%Nb CANDU pressure tubes include fuel bundle bearing pad fretting flaws and debrise fretting flaws. In-service flaws are evaluated using fitness-for-service procedures to justify continued operation of pressure tube containing the flaw. The flaw evaluation procedures address crack initiation due to Delayed Hydride Cracking (DHC) under constant loading and also address fracture initiation and plastic collapse. The paper presents some applications related to the influence of the residual hoop stresses at roll-expanded joint region into stainless steel fittings at both ends on the structural integrity evaluation in the presence of blunt flaws. Two cases of blunt flaws were considered for evaluation: fuel bundle bearing pad fretting flaws and debrise fretting flaws. The blunt flaw geometry modeling and stress-strain analyses were performed by means finite element method (FEM) with FEA-Flaw computer code. The allowable peak flaw-tip stress and the Failure Assessment Diagram (FAD) for DHC initiation criterion were used for integrity assessment for the mentioned blunt flaws. Applications are performed as part of the research program address to evaluation of the in-service inspection results of the fuel channels from Cernavoda NPP. (author)

  16. Probabilistic fracture mechanics applied for DHC assessment in the cool-down transients for CANDU pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Radu, Vasile, E-mail: vasile.radu@nuclear.ro [Institute for Nuclear Research Pitesti, 1st Campului Street, 115400 Mioveni, Arges, P.O. Box 78, Mioveni (Romania); Roth, Maria [Institute for Nuclear Research Pitesti, 1st Campului Street, 115400 Mioveni, Arges, P.O. Box 78, Mioveni (Romania)

    2012-12-15

    For CANDU pressure tubes made from Zr-2.5%Nb alloy, the mechanism called delayed hydride cracking (DHC) is widely recognized as main mechanism responsible for crack initiation and propagation in the pipe wall. Generation of some blunt flaws at the inner pressure tube surface during refueling by fuel bundle bearing pad or by debris fretting, combined with hydrogen/deuterium up-take (20-40 ppm) from normal corrosion process with coolant, may lead to crack initiation and growth. The process is governed by hydrogen hysteresis of terminal solid solubility limits in Zirconium and the diffusion of hydrogen atoms in the stress gradient near to a stress spot (flaw). Creep and irradiation growth under normal operating conditions promote the specific mechanisms for Zirconium alloys, which result in circumferential expansion, accompanied by wall thinning and length increasing. These complicate damage mechanisms in the case of CANDU pressure tubes that are also are affected by irradiation environment in the reactor core. The structural integrity assessment of CANDU fuel channels is based on the technical requirements and methodology stated in the Canadian Standard N285.8. Usually it works with fracture mechanics principles in a deterministic manner. However, there are inherent uncertainties from the in-service inspection, which are associated with those from material properties determination; therefore a necessary conservatism in deterministic evaluation should be used. Probabilistic approach, based on fracture mechanics principle and appropriate limit state functions defined as fracture criteria, appears as a promising complementary way to evaluate structural integrity of CANDU pressure tubes. To perform this, one has to account for the uncertainties that are associated with the main parameters for pressure tube assessment, such as: flaws distribution and sizing, initial hydrogen concentration, fracture toughness, DHC rate and dimensional changes induced by long term

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

    Energy Technology Data Exchange (ETDEWEB)

    Tapping, R.L. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Nickerson, J.H.; Subash, N. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Wright, M.D

    2001-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Jayalal, M.L., E-mail: jayalal@igcar.gov.in [Electronics, Instrumentation and Radiological Safety Group (EIRSG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India); Ramachandran, Suja [Electronics, Instrumentation and Radiological Safety Group (EIRSG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India); Rathakrishnan, S. [Reactor Physics Section, Madras Atomic Power Station (MAPS), Kalpakkam, Tamil Nadu (India); Satya Murty, S.A.V. [Electronics, Instrumentation and Radiological Safety Group (EIRSG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India); Sai Baba, M. [Resources Management Group (RMG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India)

    2015-01-15

    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

  19. Study on the requirement for the fuel test loop performance in the HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byung Chul; Kim, Hark Rho

    2000-06-01

    The requirement of the FTL (Fuel Test Loop) performance were investigated to reaffirm the technical feasibility of the FTL facility which is in consideration to install at the HANARO. LH hole in the reflector region and OR3 in the outer core region are considered as candidate sites for the IPS (In-Pile Section) for the analysis purpose. The achievable linear power at test fuel pin(s) and neutron flux levels at the cladding are analyzed in the IPS, which accommodates CANDU or PWR test fuel. The enrichment of test fuels is assumed as natural uranium for CANDU and 3.5% or 5% for PWR. The test fuel configuration is bundle or 7-pin in LH hole but 1-pin in OR site. For the CANDU test fuel, the target linear power of 60kW/m can not be achieved for all cases. For the PWR test fuel, the target linear power of 40kW/m is obtained at fuels located in the direction of the core for only the case of 5% bundle irradiation. From a sensitivity study, the linear power at test fuel is expected to increase at least 30% of the present results if the core bumup effect, optimization of the ratio of fuel-to-moderator number density, etc., are considered in the detail design. Thus, the linear power for PWR fuel is expected to reach the target value, and that for CANDU fuel will reach the target value if enriched fuel is used. The fast neutron flux at the test fuel cladding is estimated for most cases to be lower than one-third of the target value of 10{sup 1}4 n/cm{sup 2}-sec and expected not to reach the target value.

  20. Laminar simulation of intersubchannel mixing in a triangular nuclear fuel bundle geometry

    Energy Technology Data Exchange (ETDEWEB)

    Zaretsky, A.; Lightstone, M.F., E-mail: lightsm@mcmaster.ca; Tullis, S.

    2015-12-15

    Highlights: • Quasi-periodic flow was observed through rod-to-wall gaps. • Triangular subchannel flows were fundamentally irregular. • Cross-gap flow was influenced both by local and adjacent cross-gap intensity. • Phase-linking between gaps induced cross-plane peripheral circulation through rod–wall gaps. • Cross-gap flow structure was dependent on subchannel geometry. - Abstract: Predicting temperature distributions in fuel rod bundles is an important component of nuclear reactor safety analysis. Intersubchannel mixing acts to homogenize coolant temperatures thus reducing the likelihood of localized regions of high fuel temperature. Previous research has shown that intersubchannel mixing in nuclear fuel rod bundles is enhanced by a large-scale quasi-periodic energetic fluid motion, which transports fluid on the cross-plane between the narrow gaps connecting subchannels. This phenomenon has also been observed in laminar flows. Unsteady laminar flow simulations were performed in a simplified bundle of three rods with a pipe. Three similar geometries of varying gap width were examined, and a thermal trace was implemented on the first geometry. Thermal mixing was driven by the advection of energy between subchannels by the cross-plane flow. Flow through the rod-to-wall gaps in the wall subchannels alternated with a dominant frequency, particularly when rod-to-wall gaps were smaller than rod-to-rod gaps. Significant phase-linking between rod-to-wall gaps was also observed such that a peripheral circulation occurred through each gap simultaneously. Cross-plane flow through the rod-to-rod gaps in the triangular subchannel was irregular in each case. This was due to the fundamental irregularity of the triangular subchannel geometry. Vortices were continually broken up by cross-plane flow from other gaps due to the odd number of fluid pathways within the central subchannel. Cross-plane flow in subchannel geometries is highly interconnected between gaps. The

  1. 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.

  2. Full scale stability and void fraction measurements for the ATRIUM trademark 10XM BWR fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Wehle, Franz; Velten, Roger; Kronenberg, Juris; Beisiegel, Achim [AREVA NP GmbH, Erlangen (Germany); Pruitt, D.W.; Greene, K.R. [AREVA NP Inc., Lynchburg, VA (United States); Farawila, Y.M. [Farawila et al., Inc., Richland, WA (United States)

    2011-07-01

    This paper describes recent advances in BWR fuel testing at AREVA NP's KATHY loop including stability and void fraction measurements. The stability tests for the ATRIUM trademark 10XM bundle with corner PLFR's were expanded in scope compared with previous campaigns to include simulated reactivity and power feedback essentially reproducing BWR operational environment. The oscillation magnitude was allowed to grow to explore inlet flow reversal and cyclical dryout and rewetting. The void fraction measurements employed a gamma ray computed tomography technique that reveals not only the average but the detailed sub-channel void distribution, and the range of measured void fraction has been expanded to higher values than was previously attained. With the completion of the required licensing tests and stability performance demonstration, the ATRIUM trademark 10XM is available and fully qualified for reload supply. (orig.)

  3. 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.

  4. 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).

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

    Energy Technology Data Exchange (ETDEWEB)

    Hummel, D.W.; Novog, D.R. [McMaster Univ., Hamilton, Ontario (Canada)

    2012-07-01

    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 PuO{sub 2} in ThO{sub 2}) 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% PuO{sub 2} (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)

  6. 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.

  7. Heat transfer on HLM cooled wire-spaced fuel pin bundle simulator in the NACIE-UP facility

    Energy Technology Data Exchange (ETDEWEB)

    Di Piazza, Ivan, E-mail: ivan.dipiazza@enea.it [Italian National Agency for New Technologies, Energy and Sustainable Economic Development, C.R. ENEA Brasimone, Camugnano (Italy); Angelucci, Morena; Marinari, Ranieri [University of Pisa, Dipartimento di Ingegneria Civile e Industriale, Pisa (Italy); Tarantino, Mariano [Italian National Agency for New Technologies, Energy and Sustainable Economic Development, C.R. ENEA Brasimone, Camugnano (Italy); Forgione, Nicola [University of Pisa, Dipartimento di Ingegneria Civile e Industriale, Pisa (Italy)

    2016-04-15

    Highlights: • Experiments with a wire-wrapped 19-pin fuel bundle cooled by LBE. • Wall and bulk temperature measurements at three axial positions. • Heat transfer and error analysis in the range of low mass flow rates and Péclet number. • Comparison of local and section-averaged Nusselt number with correlations. - Abstract: The NACIE-UP experimental facility at the ENEA Brasimone Research Centre (Italy) allowed to evaluate the heat transfer coefficient of a wire-spaced fuel bundle cooled by lead-bismuth eutectic (LBE). Lead or lead-bismuth eutectic are very attractive as coolants for the GEN-IV fast reactors due to the good thermo-physical properties and the capability to fulfil the GEN-IV goals. Nevertheless, few experimental data on heat transfer with heavy liquid metals (HLM) are available in literature. Furthermore, just a few data can be identified on the specific topic of wire-spaced fuel bundle cooled by HLM. Additional analysis on thermo-fluid dynamic behaviour of the HLM inside the subchannels of a rod bundle is necessary to support the design and safety assessment of GEN. IV/ADS reactors. In this context, a wire-spaced 19-pin fuel bundle was installed inside the NACIE-UP facility. The pin bundle is equipped with 67 thermocouples to monitor temperatures and analyse the heat transfer behaviour in different sub-channels and axial positions. The experimental campaign was part of the SEARCH FP7 EU project to support the development of the MYRRHA irradiation facility (SCK-CEN). Natural and mixed circulation flow regimes were investigated, with subchannel Reynolds number in the range Re = 1000–10,000 and heat flux in the range q″ = 50–500 kW/m{sup 2}. Local Nusselt numbers were calculated for five sub-channels in different ranks at three axial positions. Section-averaged Nusselt number was also defined and calculated. Local Nusselt data showed good consistency with some of the correlation existing in literature for heat transfer in liquid metals

  8. 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)

  9. 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.

  10. 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.

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

    Science.gov (United States)

    Zafred, Paolo R [Murrysville, PA; Gillett, James E [Greensburg, PA

    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.

  12. 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.

  13. 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.

  14. Verification of the FBR fuel bundle–duct interaction analysis code BAMBOO by the out-of-pile bundle compression test with large diameter pins

    Energy Technology Data Exchange (ETDEWEB)

    Uwaba, Tomoyuki, E-mail: uwaba.tomoyuki@jaea.go.jp [Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Ibaraki 311-1393 (Japan); Ito, Masahiro; Nemoto, Junichi [Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Ibaraki 311-1393 (Japan); Ichikawa, Shoichi [Japan Atomic Energy Agency, 2-1, Shiraki, Tsuruga-shi, Fukui 919-1279 (Japan); Katsuyama, Kozo [Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Ibaraki 311-1393 (Japan)

    2014-09-15

    The BAMBOO computer code was verified by results for the out-of-pile bundle compression test with large diameter pin bundle deformation under the bundle–duct interaction (BDI) condition. The pin diameters of the examined test bundles were 8.5 mm and 10.4 mm, which are targeted as preliminary fuel pin diameters for the upgraded core of the prototype fast breeder reactor (FBR) and for demonstration and commercial FBRs studied in the FaCT project. In the bundle compression test, bundle cross-sectional views were obtained from X-ray computer tomography (CT) images and local parameters of bundle deformation such as pin-to-duct and pin-to-pin clearances were measured by CT image analyses. In the verification, calculation results of bundle deformation obtained by the BAMBOO code analyses were compared with the experimental results from the CT image analyses. The comparison showed that the BAMBOO code reasonably predicts deformation of large diameter pin bundles under the BDI condition by assuming that pin bowing and cladding oval distortion are the major deformation mechanisms, the same as in the case of small diameter pin bundles. In addition, the BAMBOO analysis results confirmed that cladding oval distortion effectively suppresses BDI in large diameter pin bundles as well as in small diameter pin bundles.

  15. Development of advanced BWR fuel bundle with spectral shift rod - BWR core characteristics with SSR

    Energy Technology Data Exchange (ETDEWEB)

    Hino, T.; Kondo, T.; Chaki, M.; Ohga, Y. [Hitachi-GE Nuclear Energy, Ltd., 1-1, Saiwai-cho, 3-chome, Hitachi-shi, Ibaraki-ken, 317-0073 (Japan); Makigami, T. [Tokyo Electric Power Company Inc., 1-1-3, Uchisaiwai-cho, Chiyoda-ku, Tokyo, 100-0011 (Japan)

    2012-07-01

    The neutron energy spectrum can be varied during an operation cycle to generate and utilize more plutonium from the non-fissile {sup 238}U by changing the void fraction in the core through control of the core coolant flow rate. This operation method, which is called a spectral shift operation, is practiced in BWRs to save natural uranium. A new component called a spectral shift rod (SSR), which is utilized instead of a conventional water rod, has been introduced to amplify the void fraction change and increase the spectral shift effect. In this study, fuel bundle design with the SSR and core design were carried out for the ABWR and the next generation BWR, HP-ABWR (High-Performance ABWR).The core characteristics with the SSR were evaluated and compared with those when using the conventional water rod. Influences of uncertainty of the water level in the SSR on the safety limit minimum critical power ratio (SLMCPR) were considered for evaluation of the uranium saving effect attained by the SSR. As a result, it was found that the amount of natural uranium needed for an operation cycle could be reduced more than 3% with 20% core coolant flow change and more than 5% with 30% core coolant flow change, in the form of increased discharge exposure by using the SSR compared with the conventional water rod use. (authors)

  16. CFD Validation Benchmark Dataset for Natural Convection in Nuclear Fuel Rod Bundles

    Science.gov (United States)

    Smith, Barton; Jones, Kyle

    2016-11-01

    The present study provide CFD validation benchmark data for coupled fluid flow/convection heat transfer on the exterior of heated rods arranged in a 2 × 2 array. The rod model incorporates grids with swirling veins to resemble a nuclear fuel bundle. The four heated aluminum rods are suspended in an open-circuit wind tunnel. Boundary conditions (BCs) are measured and uncertainties calculated to provide all quantities necessary to successfully conduct a CFD validation exercise. System response quantities (SRQs) are measured for comparing the simulation output to the experiment. Stereoscopic Particle Image Velocimetry (SPIV) is used to non-intrusively measure 3-component velocity fields. A through-plane measurement is used for the inflow while laser sheet planes aligned with the flow direction at several downstream locations are used for system response quantities. Two constant heat flux rod surface conditions are presented (400 W/m2 and 700 W/m2) achieving a peak Rayleigh number of 1010 . Uncertainty for all measured variables is reported. The boundary conditions, system response, and all material properties are now available online for download. The U.S. Department of Energy Nuclear Engineering University Program provided the funding for these experiments under Grant 00128493.

  17. 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)

  18. Bundled Slash: A Potential New Biomass Resource for Fuels and Chemicals

    Science.gov (United States)

    Steele, Philip H.; Mitchell, Brian K.; Cooper, Jerome E.; Arora, S.

    Postharvest residues for southern pine species have not previously been quantified to compare volumes produced from both thinnings and clearcut volumes. A John Deere 1490 Slash Bundler bundled postharvest residues following a first thinning of a 14-year-old stand, a second thinning of a 25-year-old stand, and a clearcut of a naturally regenerated mature stand of 54 years of age. Regardless of stand type, nearly one-fifth of merchantable volume harvested was collected as postharvest residue. Initial bundle moisture contents were 127.3, 81.1, and 49.4% dry basis (db) for the first and second thinnings and mature stands, respectively. Bundle needles content was found to significantly influence the relative moisture contents of the bundles by stand type due to the high moisture content of needles compared to other bundle components. Bundles were stored outside and exposed to very hot and dry conditions and dried very rapidly to lowest moisture contents of 22.8, 14.5, and 13.5% (db) for first and second thinnings and mature stands, respectively. Response to moderating temperatures and higher precipitation resulted in rapid moisture content increase to 69.9, 46.2, and 38.1% (db) for the first and second thinnings and mature stand bundles by the end of the study. Temperature and precipitation and bundle percentage needles content all significantly influenced the rapid moisture content variations observed over the study periods.

  19. 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.

  20. 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.

  1. 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.

  2. Electrolytic/fuel cell bundles and systems including a current collector in communication with an electrode thereof

    Science.gov (United States)

    Hawkes, Grant L.; Herring, James S.; Stoots, Carl M.; O& #x27; Brien, James E.

    2013-03-05

    Electrolytic/fuel cell bundles and systems including such bundles include an electrically conductive current collector in communication with an anode or a cathode of each of a plurality of cells. A cross-sectional area of the current collector may vary in a direction generally parallel to a general direction of current flow through the current collector. The current collector may include a porous monolithic structure. At least one cell of the plurality of cells may include a current collector that surrounds an outer electrode of the cell and has at least six substantially planar exterior surfaces. The planar surfaces may extend along a length of the cell, and may abut against a substantially planar surface of a current collector of an adjacent cell. Methods for generating electricity and for performing electrolysis include flowing current through a conductive current collector having a varying cross-sectional area.

  3. CFD modelling of supercritical water flow and heat transfer in a 2 × 2 fuel rod bundle

    Energy Technology Data Exchange (ETDEWEB)

    Podila, Krishna, E-mail: krishna.podila@cnl.ca; Rao, Yanfei, E-mail: yanfei.rao@cnl.ca

    2016-05-15

    Highlights: • Bare and wire wrapped 2 × 2 fuel rod bundles were modelled with CFD. • Sensitivity of predictions to SST k–ω, v{sup 2}–f and turbulent Prandtl number was tested. • CFD predictions were assessed with experimentally reported fuel wall temperatures. - Abstract: In the present assessment of the CFD code, two heat transfer experiments using water at supercritical pressures were selected: a 2 × 2 rod bare bundle; and a 2 × 2 rod wire-wrapped bundle. A systematic 3D CFD study of the fluid flow and heat transfer at supercritical pressures for the rod bundle geometries was performed with the key parameter being the fuel rod wall temperature. The sensitivity of the prediction to the steady RANS turbulence models of SST k–ω, v{sup 2}–f and turbulent Prandtl number (Pr{sub t}) was tested to ensure the reliability of the predicted wall temperature obtained for the current analysis. Using the appropriate turbulence model based on the sensitivity analysis, the mesh refinement, or the grid convergence, was performed for the two geometries. Following the above sensitivity analyses and mesh refinements, the recommended CFD model was then assessed against the measurements from the two experiments. It was found that the CFD model adopted in the current work was able to qualitatively capture the trends reported by the experiments but the degree of temperature rise along the heated length was underpredicted. Moreover, the applicability of turbulence models varied case-by-case and the performance evaluation of the turbulence models was primarily based on its ability to predict the experimentally reported fuel wall temperatures. Of the two turbulence models tested, the SST k–ω was found to be better at capturing the measurements at pseudo-critical and supercritical test conditions, whereas the v{sup 2}–f performed better at sub-critical test conditions. Along with the appropriate turbulence model, CFD results were found to be particularly sensitive to

  4. 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.

  5. The DUPIC fuel development program in KAERI

    Energy Technology Data Exchange (ETDEWEB)

    Yang, M. S.; Park, H. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-07-01

    This study describes the DUPIC fuel development program in KAERI as follows; Burning spent PWR fuel again in CANDU by DUPIC, Compatibility with existing CANDU system, Feasibility of DUPIC fuel fabrication, Waste reduction, Safeguard ability, Economics of DUPIC fuel cycle, The DUPIC fuel development program, and International prospective. 5 refs., 10 figs.

  6. CFX Analysis of the CANDU Moderator Thermal-Hydraulics in the Stern Lab. Test Facility

    Science.gov (United States)

    Kim, Hyoung Tae

    2014-06-01

    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.

  7. 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'.

  8. Turbulet flow in a model nuclear fuel rod bundle containing partial flow blockages

    Energy Technology Data Exchange (ETDEWEB)

    Creer, J.M.; Rowe, D.S.; Bates, J.M.; Sutey, A.M.

    1977-03-01

    Local velocity and turbulence intensity measurements were obtained with a laser Doppler anemometer near flow blockages in an unheated 7 x 7 rod bundle. Sleeve blockages were positioned on the center nine rods to create area reductions of 70 and 90 percent in the center four subchannels of the bundle. Experimental results indicated that severe flow disturbances existed downstream from the blockage clusters and showed that only minor disturbances can be expected upstream from the blockages. Recirculation zones for both 70 and 90 percent blockages were detected downstream from the blockage clusters and persisted for approximately three to five subchannel hydraulic diameters depending on blockage severity. The experimental velocity results obtained with blockage clusters located midway between grid spacers were successfully predicted using the COBRA computer program.

  9. Development of modern CANDU PHWR cross-section libraries for SCALE

    Energy Technology Data Exchange (ETDEWEB)

    Shoman, Nathan T., E-mail: nshoman@vols.utk.edu; Skutnik, Steven E., E-mail: sskutnik@utk.edu

    2016-06-15

    Highlights: • New ORIGEN libraries for CANDU 28 and 37-element fuel assemblies have been created. • These new reactor data libraries are based on modern ENDF/B-VII.0 cross-section data. • The updated CANDU data libraries show good agreement with radiochemical assay data. • Eu-154 overestimated when using ENDF-VII.0 due to a lower thermal capture cross-section. - Abstract: A new set of SCALE fuel lattice models have been developed for the 28-element and 37-element CANDU fuel assembly designs using modern cross-section data from ENDF-B/VII.0 in order to produce new reactor data libraries for SCALE/ORIGEN depletion analyses. These new libraries are intended to provide users with a convenient means of evaluating depletion of CANDU fuel assemblies using ORIGEN through pre-generated cross sections based on SCALE lattice physics calculations. The performance of the new CANDU ORIGEN libraries in depletion analysis benchmarks to radiochemical assay data were compared to the previous version of the CANDU libraries provided with SCALE (based on WIMS-AECL models). Benchmark comparisons with available radiochemical assay data indicate that the new cross-section libraries perform well at matching major actinide species (U/Pu), which are generally within 1–4% of experimental values. The library also showed similar or better results over the WIMS-AECL library regarding fission product species and minor actinoids (Np, Am, and Cm). However, a notable exception was in calculated inventories of {sup 154}Eu and {sup 155}Eu, where the new library employing modern nuclear data (ENDF/B-VII.0) performed substantially poorer than the previous WIMS-AECL library (which used ENDF-B/VI.8 cross-sections for these species). The cause for this discrepancy appears to be due to differences in the {sup 154}Eu thermal capture cross-section between ENDF/B-VI.8 and ENDF/B-VII.0, an effect which is exacerbated by the highly thermalized flux of a CANDU heavy water reactor compared to that of a

  10. 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.

  11. Validation of physics and thermalhydraulic computer codes for advanced Candu reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    Wren, D.J.; Popov, N.; Snell, V.G. [Atomic Energy of Canada Ltd, (Canada)

    2004-07-01

    Atomic Energy of Canada Ltd. (AECL) is developing an Advanced Candu Reactor (ACR) that is an evolutionary advancement of the currently operating Candu 6 reactors. The ACR is being designed to produce electrical power for a capital cost and at a unit-energy cost significantly less than that of the current reactor designs. The ACR retains the modular Candu concept of horizontal fuel channels surrounded by a heavy water moderator. However, ACR uses slightly enriched uranium fuel compared to the natural uranium used in Candu 6. This achieves the twin goals of improved economics (via large reductions in the heavy water moderator volume and replacement of the heavy water coolant with light water coolant) and improved safety. AECL has developed and implemented a software quality assurance program to ensure that its analytical, scientific and design computer codes meet the required standards for software used in safety analyses. Since the basic design of the ACR is equivalent to that of the Candu 6, most of the key phenomena associated with the safety analyses of ACR are common, and the Candu industry standard tool-set of safety analysis codes can be applied to the analysis of the ACR. A systematic assessment of computer code applicability addressing the unique features of the ACR design was performed covering the important aspects of the computer code structure, models, constitutive correlations, and validation database. Arising from this assessment, limited additional requirements for code modifications and extensions to the validation databases have been identified. This paper provides an outline of the AECL software quality assurance program process for the validation of computer codes used to perform physics and thermal-hydraulics safety analyses of the ACR. It describes the additional validation work that has been identified for these codes and the planned, and ongoing, experimental programs to extend the code validation as required to address specific ACR design

  12. Operating Experience of MACSTOR Modules at CANDU 6 Stations

    Energy Technology Data Exchange (ETDEWEB)

    Beaudoin, Robert R. [Atomic Energy Canada Ltd., Chalk River (Canada)

    2005-11-15

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

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. Pre-licensing of the Advanced CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ion, R.; Popov, N.K.; Snell, V. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)]. E-mail: popovn@aecl.ca; West, J. [Candesco, Toronto, Ontario (Canada); Xu, C. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)

    2006-07-01

    Atomic Energy of Canada Limited (AECL) developed the Advanced CANDU Reactor-700 (ACR-700) as an evolutionary advancement of the current CANDU 6 reactor. As further advancement of the ACR design, AECL is currently developing the ACR-1000 for the Canadian and international market. The ACR-1000 is aimed at producing electrical power for a capital cost and a unit-energy cost significantly less than that of the current generation of operating nuclear plants, while achieving shorter construction schedule, high plant capacity factor, improved operations and maintenance, increased operating life, and enhanced safety features. The reference ACR-1000 plant design is based on an integrated two-unit plant, using enriched fuel and light-water coolant, with each unit having a nominal gross output of about 1200 MWe. AECL initiated pre-licensing reviews of the ACR reactor design in Canada, US and China, with an objective to take into account regulatory feedback early in the design process. The Canadian Nuclear Safety Commission (CNSC) is performing a pre-project pre-licensing assessment of the ACR design. The objective of the assessment is to issue a formal statement as to whether there are any fundamental barriers that would prevent the licensing of the new CANDU reactor design in Canada under the Nuclear Safety and Control Act. The CNSC review is being conducted in four phases. In Phase 1 (September 2003 to September 2004) CNSC performed a pre-licensing review of the ACR-700, and focused on the design process, methodology, design concepts and R and D. CNSC staff reviewed about 100 reports, and submitted to AECL questions and comments. In Phase 2 (September 2004 to August 2005) AECL provided responses and additional information to CNSC on their comments and questions in Phase 1. Phase 3 is the Transition Phase (September 2005 to May 2006), bridging the transition from the ACR-700 to the ACR-1000 design. Phase 3 focused on review of generic aspects of the ACR design, on the Safety

  18. 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.

  19. 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.)

  20. 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.

  1. ACR fuel storage analysis: finite element heat transfer analysis of dry storage

    Energy Technology Data Exchange (ETDEWEB)

    Khair, K.; Baset, S.; Millard, J. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)

    2006-07-01

    Over the past decade Atomic Energy of Canada Limited (AECL) has designed and licensed air-cooled concrete structures used as above ground dry storage containers (MACSTOR) to store irradiated nuclear fuel from CANDU plants. A typical MACSTOR 200 module is designed to store 12,000 bundles in 20 storage cylinders. MACSTOR 200 modules are in operation at Gentilly-2 in Canada and at Cernavoda in Romania. The MACSTOR module is cooled passively by natural convection and by conduction through the concrete walls and roof. Currently AECL is designing the Advanced Candu Reactor (ACR) with CANFLEX slightly enriched uranium fuel to be used. AECL has initiated a study to explore the possibility of storing the irradiated nuclear fuel from ACR in MACSTOR modules. This included work to consider ways of minimizing footprint both in the spent fuel storage bay and in the dry storage area. The commercial finite element code ANSYS has been used in this study. The FE model is used to complete simulations with the higher heat source using the same concrete structural dimensions to assess the feasibility of using the MACSTOR design for storing the ACR irradiated fuel. This paper presents the results of the analysis. The results are used to confirm the possibility of using, with minimal changes to the design of the storage baskets and the structure, the proven design of the MACSTOR 200 containment to store the ACR fuel bundles with higher enrichment and burnup. This has thus allowed us to confirm conceptual feasibility and move on to investigation of optimization. (author)

  2. 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.

  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. 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.

  5. 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.

  6. 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.

  7. EC6{sup TM} - Enhanced Candu 6{sup TM} reactor safety characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Popov, N.; Snell, V.; Cormier, M.; Hopwood, J. [Atomic Energy of Canada Ltd., 2251 Speakman Drive, Mississauga, Ontario, L5K 1B2 (Canada)

    2010-07-01

    The EC6 is a 740 MWe-class natural-uranium-fuelled, heavy-water-cooled and -moderated pressure-tube reactor, which has evolved from the eleven (11) CANDU{sup R} 6 plants operating in five countries (on four continents). CANDU 6 has over 150 reactor-years of safe operation. The most recent CANDU 6 - at Qinshan, in China - is the Reference Design for EC6. The EC6 shares many inherent, passive and engineered safety characteristics with the Reference Design. However EC6 has been designed to meet modern regulatory requirements and safety expectations. The resulting design changes have improved these safety characteristics, and this paper provides a convenient summary. The paper addresses the safety functions of reactivity control, heat removal, and containment of radioactive material. For each safety function, the EC6 characteristics are categorized as inherent, passive, or engineered. The paper focuses mostly on the first two. The Enhanced CANDU 6 uses an appropriate mix of passive, inherent, and engineered safety functions. Reactivity transients are generally slow, mild and inherently limited due to the natural uranium core and use of on-power refuelling. Only the coolant void coefficient can cause a large reactivity insertion, particularly in a large LOCA. This is mitigated by the long prompt neutron lifetime and the large delayed neutron fraction, and terminated by either of the two shutdown systems. For EC6, the large LOCA power transient has been reduced significantly by speeding up the slower of the two shutdown systems. Redundant shutdown and the LOCA power pulse improvements mitigate the limiting large positive reactivity insertion. Decay heat removal shows a very high component of passive safety, from thermo-siphoning in the Reactor Coolant System to passive heat removal in severe accidents via the moderator or reactor vault. The latter two can maintain the fuel in a more predictable and favourable geometry than 'core on the floor'. The containment

  8. 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

  9. Development of CANFLEX fuel fabrication technology

    Energy Technology Data Exchange (ETDEWEB)

    Kang, M. S.; Choi, C. B.; Park, C. H.; Kwon, W. J.; Kim, C. H.; Kim, B. J.; Koo, C. H.; Cho, D. S.; So, D. Y.; Suh, S. W.; Park, C. J.; Chang, D. H.; Yun, S. H. [KEPCO Nuclear Fuel Company, Taejeon (Korea)

    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. One of the improvements in CANDU fuel fabrication technology, and advanced method of Zr-Be brazing was developed. For the formation of Zr-Be alloy, preheating and main heating temperature in the furnace is 700 deg C, 1200 deg C respectively. In order to find an appropriate material for the brazing joints in the CANDU fuel, the composition of Zr based amorphous metals were designed. And, the effect of hydrogen on the mechanical properties of cladding sheath and feasibility of the eddy current test to evaluate quality of end cap weld were also studied for the fundamental research purpose. As a preliminary study to suggest optimal way for the mass production of CANFLEX-NU fuel at KNFC the existing CANDU fuel facilities and fabrication/inspection processes were reviewed. The best way is that the current CANDU facility shall be modified to produce small diametrial CANFLEX elements and a new facility shall be constructed to produce large diametrial CANFLEX fuel elements. 46 refs., 99 figs., 10 tabs. (Author)

  10. Development of an Integrity Assessment Procedure for CANDU Pressure Tubes

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Han Sub [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    The pressure tubes used in a CANDU reactor are made from Zr-2.5Nb. During service the pressure tubes operate at temperatures between about 150 and 310 .deg. C, and with variable coolant pressures up to 11MPa corresponding to hoop stress of up to 130MPa. The maximum flux of fast neutrons (E>1MeV) from the fuel is about 4X10{sup 17}nm{sup -2}{sub s}{sup -1}. The pressure tubes are exposed to very severe degradation environment. The aging degradation of the pressure tubes are summarized as below. - Geometric deformation; axial elongation, diametric creep, and wall thinning. - Deuterium uptake; some fraction of the deuterium generated by the corrosion of pressure tubes is absorbed into the pressure tubes. Total equivalent hydrogen content in the pressure tube is the sum of the initial hydrogen content before operation and the deuterium uptake during operation. High concentration of hydrogen inside the pressure tubes makes the metal susceptible to Delayed Hydride Cracking. The DHC is a degradation mechanism of prime importance for CANDU pressure tubes. Mechanical properties, in particular fracture toughness, are deteriorated by high concentration of dissolved hydrogen. - Flaws; volumetric flaws are generated during operation. Wear scars by debris fretting, and bearing pad fretting are common. These volumetric flaws can be a site of crack initiation by fatigue or DHC. Cracks can propagate by DHC or fatigue crack propagation if conditions are met. - Material properties degradation; mechanical properties are affected by neutron irradiation. Yield strength and tensile strength are increased, and fracture toughness is deteriorated. The susceptibility to DHC is also affected. The integrity assessment of the pressure tube is a procedure to determine if the risk of pressure tube failure is controlled to maintain acceptably low. CSA N285.4 and 285.8 are two important guidelines regarding the integrity of pressure tubes. N285.4 is to guide in-service inspection, and N285

  11. Advanced CANDU reactor pre-licensing progress

    Energy Technology Data Exchange (ETDEWEB)

    Popov, N.K.; West, J.; Snell, V.G.; Ion, R.; Archinoff, G.; Xu, C. [Atomic Energy of Canada Limited., Mississauga, Ontario (Canada)]. E-mail: popovn@aecl.ca

    2005-07-01

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

  12. Post-test calculation of the QUENCH-17 bundle experiment with debris formation and bottom water reflood using thermal hydraulic and severe fuel damage code SOCRAT/V3

    Energy Technology Data Exchange (ETDEWEB)

    Vasiliev, A., E-mail: vasil@ibrae.ac.ru [Nuclear Safety Institute (IBRAE), B. Tulskaya 52, 115191 Moscow (Russian Federation); Stuckert, J., E-mail: juri.stuckert@kit.edu [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2015-03-15

    Highlights: • Modeling of processes in porous debris regions. • Analysis of coolability of massive debris bed. • Complexity of simulation of flow regime near boiling curve. - Abstract: The thermal hydraulic and SFD (Severe Fuel Damage) best estimate computer modeling code SOCRAT/V3 was used for the post-test analysis of the QUENCH-17 experiment performed at KIT on January 2013. The objective of this test was to examine the formation of a debris bed inside the completely oxidized region of the bundle without melt formation and to investigate the coolability behavior during the reflood. The test bundle for QUENCH-17 test was intentionally changed in comparison to basic QUENCH bundles (usually 21 heated rod simulators) with the emphasis to investigate debris behavior phenomena. Only 12 periphery fuel rod simulators were heated by centerline tungsten heaters. 9 unheated fuel rod simulators were located in the inner part of the test bundle. This is why the massive porous debris formation in the inner part of the bundle was not influenced by the presence of tungsten heaters. The QUENCH-17 test conditions simulated a hypothetical scenario of nuclear power plant severe accident sequence with debris bed formation in which the overheated up to 1800 K core would be flooded from the bottom by ECCS (Emergency Core Cooling System). The QUENCH-17 test included the following phases: (1) heat-up phase (heat-up rate up to 0.25 K/s); (2) oxidation phase (the cladding temperature about 1800 K in hottest region, steam mass flow rate 2 g/s); (3) bottom flood phase (characteristic cooling time about 600 s, water mass flow rate 10 g/s). SOCRAT/V3 computer modeling code was used for calculation of basic thermal hydraulic, oxidation and thermal mechanical behavior during all phases of the experiment. The calculated results are in a good agreement with experimental data which justifies the adequacy of modeling capabilities of SOCRAT code system.

  13. 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.

  14. 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.

  15. Thermal Hydraulic Performance of Tight Lattice Bundle

    Science.gov (United States)

    Yamamoto, Yasushi; Akiba, Miyuki; Morooka, Shinichi; Shirakawa, Kenetsu; Abe, Nobuaki

    Recently, the reduced moderation spectrum BWR has been studied. The fast neutron spectrum is obtained through triangular tight lattice fuel. However, there are few thermal hydraulic test data and thermal hydraulic correlation applicable to critical power prediction in such a tight lattice bundle. This study aims to enhance the database of the thermal hydraulic performance of the tight lattice bundle whose rod gap is about 1mm. Therefore, thermal hydraulic performance measurement tests of tight lattice bundles for the critical power, the pressure drop and the counter current flow limiting were performed. Moreover, the correlations to evaluate the thermal-hydraulic performance of the tight lattice bundle were developed.

  16. 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.)

  17. Large-break loss-of-coolant accident phenomena identification and ranking table (PIRT) for the advanced Candu reactor

    Energy Technology Data Exchange (ETDEWEB)

    Popov, N.; Snell, V.G.; Sills, H.E.; Langman, V.J.; Boyack, B. [Atomic Energy of Canada Ltd (Canada)

    2004-07-01

    The Advanced Candu Reactor (ACR) is an evolutionary advancement of the current Candu-6 reactor, aimed at producing electrical power for a capital cost and unit-energy cost significantly less than that of current reactor designs. The ACR retains the modular concept of horizontal fuel channels surrounded by heavy water moderator, as with all Candu reactors. However, ACR uses slightly enriched uranium (SEU) fuel, compared to the natural uranium used in Candu 6. This achieves the twin goals of improved economics (e.g., via reductions in the heavy water requirements and the use of a light water coolant), as well as improved safety. This paper is focused on the double-ended guillotine critical inlet header break (CRIHB) loss-of-coolant accident (LOCA) in an ACR reactor, which is considered as a large break LOCA. Large Break LOCA in water-cooled reactors has been used historically as a design basis event by regulators, and it has attracted a very large share of safety analysis and regulatory review. The LBLOCA event covers a wide range of system behaviours and fundamental phenomena. The Phenomena Identification and Ranking Table (PIRT) for LBLOCA therefore provides a good understanding of many of the safety characteristics of the ACR design. The paper outlines the design characteristics of the ACR reactor that impact the PIRT process and computer code applicability. It also describes the LOCA phenomena, lists all components and systems that have an important role during the event, discusses the PIRT process and results, and presents the final PIRT summary table. (authors)

  18. 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.

  19. THE IMPACT OF POWER COEFFICIENT OF REACTIVITY ON CANDU 6 REACTORS

    Directory of Open Access Journals (Sweden)

    D. KASTANYA

    2013-10-01

    Full Text Available 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.

  20. 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.

  1. Incorporating single detector failure into the ROP detector layout optimization for CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kastanya, Doddy, E-mail: Doddy.Kastanya@snclavalin.com

    2015-12-15

    Highlights: • ROP TSP value needs to be adjusted when any detector in the system fails. • Single detector failure criterion has been incorporated into the detector layout optimization as a constraint. • Results show that the optimized detector layout is more robust with respect to its vulnerability to a single detector failure. • An early rejection scheme has been introduced to speed-up the optimization process. - Abstract: In CANDU{sup ®} reactors, the regional overpower protection (ROP) systems are designed to protect the reactor against overpower in the fuel which could reduce the safety margin-to-dryout. In the CANDU{sup ®} 600 MW (CANDU 6) design, there are two ROP systems in the core, each of which is connected to a fast-acting shutdown system. Each ROP system consists of a number of fast-responding, self-powered flux detectors suitably distributed throughout the core within vertical and horizontal flux detector assemblies. The placement of these ROP detectors is a challenging discrete optimization problem. In the past few years, two algorithms, DETPLASA and ADORE, have been developed to optimize the detector layout for the ROP systems in CANDU reactors. These algorithms utilize the simulated annealing (SA) technique to optimize the placement of the detectors in the core. The objective of the optimization process is typically either to maximize the TSP value for a given number of detectors in the system or to minimize the number of detectors in the system to obtain a target TSP value. One measure to determine the robustness of the optimized detector layout is to evaluate the maximum decrease (penalty) in TSP value when any single detector in the system fails. The smaller the penalty, the more robust the design is. Therefore, in order to ensure that the optimized detector layout is robust, the single detector failure (SDF) criterion has been incorporated as an additional constraint into the ADORE algorithm. Results from this study indicate that there

  2. Coolability of ballooned VVER bundles with pellet relocation

    Energy Technology Data Exchange (ETDEWEB)

    Hozer, Z.; Nagy, I.; Windberg, P.; Vimi, A. [AEKI, P.O.box 49, Budapest, H-1525 (Hungary)

    2009-06-15

    During a LOCA incident the high pressure in the fuel rods can lead to clad ballooning and the debris of fuel pellets can fill the enlarged volume. The evaluation of the role of these two effects on the coolability of VVER type fuel bundles was the main objective of the experimental series. The tests were carried out in the modified configuration of the CODEX facility. 19-rod electrically heated VVER type bundle was used. The test section was heated up to 600 deg. C in steam atmosphere and the bundle was quenched from the bottom by cold water. Three series of tests were performed: 1. Reference bundle with fuel rods without ballooning, with uniform power profile. 2. Bundle with 86% blockage rate and with uniform power profile. The blockage rate was reached by superimposing hollow sleeves on all 19 fuel rods. 3. Bundle with 86% blockage rate and with local power peak in the ballooned area. The local power peak was produced by the local reduction the cross section of the internal heater bar inside of the fuel rods. In all three bundle configurations three different cooling water flow-rates were applied. The experimental results confirmed that a VVER bundle with even 86% blockage rate remains coolable after a LOCA event. The ballooned section creates some obstacles for the cooling water during reflood of the bundle, but this effect causes only a short delay in the cooling down of the hot fuel rods. Earlier tests on the coolability of ballooned bundles were performed only with Western type bundles with square fuel lattice. The present test series was the first confirmation of the coolability of VVER type bundles with triangular lattice. The accumulation of fuel pellet debris in the ballooned volume results in a local power peak, which leads to further slowing down of quench front. The first tests indicated that the effect of local power peak was less significant on the delay of cooling down than the effect of ballooning. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-03-15

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

  4. On Double Vector Bundles

    Institute of Scientific and Technical Information of China (English)

    Zhuo CHEN; Zhang Ju LIU; Yun He SHENG

    2014-01-01

    In this paper, we construct a category of short exact sequences of vector bundles and prove that it is equivalent to the category of double vector bundles. Moreover, operations on double vector bundles can be transferred to operations on the corresponding short exact sequences. In particular, we study the duality theory of double vector bundles in term of the corresponding short exact sequences. Examples including the jet bundle and the Atiyah algebroid are discussed.

  5. On Double Vector Bundles

    OpenAIRE

    Chen, Zhuo; Liu, Zhangju; Sheng, Yunhe

    2011-01-01

    In this paper, we construct a category of short exact sequences of vector bundles and prove that it is equivalent to the category of double vector bundles. Moreover, operations on double vector bundles can be transferred to operations on the corresponding short exact sequences. In particular, we study the duality theory of double vector bundles in term of the corresponding short exact sequences. Examples including the jet bundle and the Atiyah algebroid are discussed.

  6. A review of CANDU feeder wall thinning

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Han Sub [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    Flow Accelerated Corrosion is an active degradation mechanism of CANDU feeder. The tight bend downstream to Grayloc weld connection, close to reactor face, suffers significant wall thinning by FAC. Extensive in-service inspection of feeder wall thinning is very difficult because of the intense radiation field, complex geometry, and space restrictions. Development of a knowledge-based inspection program is important in order to guarantee that adequate wall thickness is maintained throughout the whole life of feeder. Research results and plant experiences are reviewed, and the plant inspection databases from Wolsong Units One to Four are analyzed in order to support developing such a knowledge-based inspection program. The initial thickness before wall thinning is highly non-uniform because of bending during manufacturing stage, and the thinning rate is non-uniform because of the mass transfer coefficient distributed non-uniformly depending on local hydraulics. It is obvious that the knowledge-based feeder inspection program should focus on both fastest thinning locations and thinnest locations. The feeder wall thinning rate is found to be correlated proportionately with QV of each channel. A statistical model is proposed to assess the remaining life of each feeder using the QV correlation and the measured thicknesses. W-1 feeder suffered significant thinning so that the shortest remaining life barely exceeded one year at the end of operation before replacement. W-2 feeder showed far slower thinning than W-1 feeder despite the faster coolant flow. It is believed that slower thinning in W-2 is because of higher chromium content in the carbon steel feeder material. The average Cr content of W-2 feeder is 0.051%, while that value is 0.02% for W-1 feeder. It is to be noted that FAC is reduced substantially even though the Cr content of W-2 feeder is still very low

  7. Optimisation de la gestion du combustible dans les reacteurs CANDU refroidis a l'eau legere

    Science.gov (United States)

    Chambon, Richard

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

  8. Update on use of AECL's MACSTOR module at CANDU 6 stations

    Energy Technology Data Exchange (ETDEWEB)

    Beaudoin, R.; Moussalam, G.; Kachef, I. [Atomic Energy of Canada Limited, Montreal, Quebec (Canada)

    2006-07-01

    AECL has contributed to the technology development and implementation of dry spent fuel management facilities in Canada and internationally over the last three decades. During that period, AECL has designed a number of concrete canister models and the MACSTOR module; a medium size air-cooled vault. AECL's dry storage technology was used in Canada, Korea and Romania for the construction of eight large-scale above ground dry storage facilities for CANDU spent fuel. These projects add up to a constructed capacity in excess of 5,000 MgU, that represents a significant share of the total worldwide dry storage capacity. This paper describes basic research and technology developments made at AECL's facilities to develop those dry storage technologies for its own reactors and for the operating CANDU 6 reactors. The current operating status of the facilities using concrete canisters is provided. A description of the MACSTOR 200 modules each having a capacity of 228 MgU that is in use at the Gentilly 2 and Cernavoda stations is provided. The Cernavoda spent fuel management facility was commissioned in 2003. The organisational, licensing, equipment supply and construction aspects that were necessary to deliver this turnkey project by AECL and its Romanian partners in 25 months are described. The paper also provides an outline of the joint program between AECL and KHNP-NETEC to develop the new MACSTOR/KN-400 and provides a description of this module having a capacity of 456 MgU (thus twice the MACSTOR 200 capacity) to be deployed by 2007 at the Wolsong site in Korea. (author)

  9. Principal noncommutative torus bundles

    DEFF Research Database (Denmark)

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

    2008-01-01

    In this paper we study continuous bundles of C*-algebras which are non-commutative analogues of principal torus bundles. We show that all such bundles, although in general being very far away from being locally trivial bundles, are at least locally trivial with respect to a suitable bundle version...... of bivariant K-theory (denoted RKK-theory) due to Kasparov. Using earlier results of Echterhoff and Williams, we shall give a complete classification of principal non-commutative torus bundles up to equivariant Morita equivalence. We then study these bundles as topological fibrations (forgetting the group...... 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...

  10. Development of CANDU ECCS performance evaluation methodology and guides

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-03-15

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

  11. Quality assurance in CANDU-type fuel manufacturing

    Energy Technology Data Exchange (ETDEWEB)

    Afifi, Y.K.; El-Hakim, E. [Atomic Energy Authority, Metallurgy Dept., Nuclear Research Center, Cairo (Egypt)

    1997-07-01

    This paper is concerned with fabrication of UO{sub 2} pellets from UO{sub 2} powder by powder metallurgical methods (pelletizing and sintering). The effect of ejection to molding force ratio on the fired pellets properties was studied. It is observed that pellets cold-pressed at ratio less than 75% is cracked to two parts. The effect of the sintering temperature on the fired pellets properties was studied. It was found that the sintering of UO{sub 2} pellets at 1650 deg C leads to production of pelletswithin the qualification requirements. The data and information available in the ASTM for each step in UO{sub 2} pellets fabrication process and the technical experience (gained or published) are transformed into a group of logic flow charts (LFC'S). These logic flow charts are collected to form a module of a software to qualify the sintered pellets and also gives a technical assistance according to the ASTM for each step in the fabrication process. (author)

  12. A study to develop the domestic functional requirements of the specific safety systems of CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Man Woong [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Lee, Jae Young; Park, Kun Chul [Handong Global Univ., Pohang (Korea, Republic of)] (and others)

    2003-03-15

    The present research has been made to develop and review critically the functional requirements of the specific safety systems of CANDU such as SDS-1, SDS2, ECCS, and containment. Based on R documents for this, a systematic study was made to develop the domestic regulation statements. Also, the conventional laws are carefully reviewed to see the compatibility to CANDU. Also, the safety assessment method for CANDU was studied by reviewing C documents and recommendation of IAEA. Through the present works, the vague policy in the CANDU safety regulation is cleaning up in a systematic form and a new frame to measure the objective risk of nuclear power plants was developed.

  13. A study to develop the domestic functional requirements of the specific safety systems of CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Man Woong; Lee, Jae Young; Bang, Kwang Hyun [Handong Global Univ., Pohang (Korea, Republic of)] (and others)

    2001-03-15

    The present research has been made to develop and review critically the functional requirements of the specific safety systems of CANDU such as SOS-1, SOS-2, ECCS and containment. Based on R documents for this, a systematic study was made to develop the domestic regulation statements. Also, the conventional laws are carefully reviewed to see the compatibility to CANDU. Also, the safety assessment method for CANDU was studied by reviewing C documents and recommendation of IAEA. Through the present works, the vague policy in the CANDU safety regulation is cleaning up in a systematic form and a new frame to measure the objective risk of nuclear power plants was developed.

  14. Performance of pressure tubes in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rodgers, D.; Griffiths, M.; Bickel, G.; Buyers, A.; Coleman, C.; Nordin, H.; St Lawrence, S. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2016-06-15

    The pressure tubes in CANDU reactors typically operate for times up to about 30 years prior to refurbishment. The in-reactor performance of Zr-2.5Nb pressure tubes has been evaluated by sampling and periodic inspection. This paper describes the behavior and discusses the factors controlling the behaviour of these components. The Zr–2.5Nb pressure tubes are nominally extruded at 815{sup o}C, cold worked nominally 27%, and stress relieved at 400 {sup o}C for 24 hours, resulting in a structure consisting of elongated grains of hexagonal close-packed alpha-Zr, partially surrounded by a thin network of filaments of body-centred-cubic beta-Zr. These beta-Zr filaments are meta-stable and contain about 20% Nb after extrusion. The stress-relief treatment results in partial decomposition of the beta-Zr filaments with the formation of hexagonal close-packed alpha-phase particles that are low in Nb, surrounded by a Nb-enriched beta-Zr matrix. The material properties of pressure tubes are determined by variations in alpha-phase texture, alpha-phase grain structure, network dislocation density, beta-phase decomposition, and impurity concentration that are a function of manufacturing variables. The pressure tubes operate at temperatures between 250 {sup o}C and 310 {sup o}C with coolant pressures up to about 11 MPa in fast neutron fluxes up to 4 x 10{sup 17} n·m{sup -2}·s{sup -1} (E > 1 MeV) and the properties are modified by these conditions. The properties of the pressure tubes in an operating reactor are therefore a function of both manufacturing and operating condition variables. The ultimate tensile strength, fracture toughness, and delayed hydride-cracking properties (velocity (V) and threshold stress intensity factor (K{sub IH})) change with irradiation, but all reach a nearly limiting value at a fluence of less than 10{sup 25} n·m{sup -2} (E > 1 MeV). At this point the ultimate tensile strength is raised about 200 MPa, toughness is reduced by about 50%, V increases

  15. Qualification of inspection systems in the CANDU nuclear industry

    Energy Technology Data Exchange (ETDEWEB)

    Baron, J.A. [CANDU Owners Group, CANDU Inspection Qualification Bureau, Toronto, Ontario (Canada)

    2014-01-15

    Most jurisdictions that generate electricity through nuclear-electric plants have imposed requirements on inspection systems beyond the typical Level 1, 2 and 3 found in personnel qualification/certification schemes. The paper discusses the rationale for this obligation and describes how the requirement for inspection qualification has been implemented for CANDU plants. The paper discusses the qualification structure and process, including a brief overview of experience to-date in qualifying Inspection Procedures. (author)

  16. Advancement of safeguards inspection technology for CANDU nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Sung; Park, W. S.; Cha, H. R.; Ham, Y. S.; Lee, Y. G.; Kim, K. P.; Hong, Y. D

    1999-04-01

    The objectives of this project are to develop both inspection technology and safeguards instruments, related to CANDU safeguards inspection, through international cooperation, so that those outcomes are to be applied in field inspections of national safeguards. Furthermore, those could contribute to the improvement of verification correctness of IAEA inspections. Considering the level of national inspection technology, it looked not possible to perform national inspections without the joint use of containment and surveillance equipment conjunction with the IAEA. In this connection, basic studies for the successful implementation of national inspections was performed, optimal structure of safeguards inspection was attained, and advancement of safeguards inspection technology was forwarded. The successful implementation of this project contributed to both the improvement of inspection technology on CANDU reactors and the implementation of national inspection to be performed according to the legal framework. In addition, it would be an opportunity to improve the ability of negotiating in equal shares in relation to the IAEA on the occasion of discussing or negotiating the safeguards issues concerned. Now that the national safeguards technology for CANDU reactors was developed, the safeguards criteria, procedure and instruments as to the other item facilities and fabrication facilities should be developed for the perfection of national inspections. It would be desirable that the recommendations proposed and concreted in this study, so as to both cope with the strengthened international safeguards and detect the undeclared nuclear activities, could be applied to national safeguards scheme. (author)

  17. Analysis of Moderator System Failure Accidents by Using New Method for Wolsong-1 CANDU 6 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Dongsik; Kim, Jonghyun; Cho, Cheonhwey [Atomic Creative Technology Co., Ltd., Daejeon (Korea, Republic of); Kim, Sungmin [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of)

    2013-05-15

    To reconfirm the safety of moderator system failure accidents, the safety analysis by using the reactor physics code, RFSP-IST, coupled with the thermal hydraulics code, CATHENA is performed additionally. In the present paper, the newly developed analysis method is briefly described and the results obtained from the moderator system failure accident simulations for Wolsong-1 CANDU 6 reactor by using the new method are summarized. The safety analysis of the moderator system failure accidents for Wolsong-1 CANDU 6 reactor was carried out by using the new code system, i. e., CATHENA and RFSP-IST, instead of the non-IST old codes, namely, SMOKIN G-2 and MODSTBOIL. The analysis results by using the new method revealed as same with the results by using the old method that the fuel integrity is warranted because the localized power peak remained well below the limits and, most importantly, the reactor operation enters into the self-shutdown mode due to the substantial loss of moderator D{sub 2}O inventory from the moderator system. In the analysis results obtained by using the old method, it was predicted that the ROP trip conditions occurred for the transient cases which are also studied in the present paper. But, in the new method, it was found that the ROP trip conditions did not occur. Consequently, in the safety analysis performed additionally by using the new method, the safety of moderator system failure accidents was reassured. In the future, the new analysis method by using the IST codes instead of the non-IST old codes for the moderator system failure accidents is strongly recommended.

  18. Field measurements of beta ray energy spectra in CANDU nuclear generating stations

    Energy Technology Data Exchange (ETDEWEB)

    Horowitz, Y.S. (Ben-Gurion Univ. of the Negev, Beersheba (Israel). Dept. of Physics); Hirning, C.R. (Ontario Hydro, Whitby, ON (Canada)); Yuen, P.S.; Aikens, M.S. (AECL Research, Chalk River, ON (Canada). Chalk River Labs.)

    1994-01-01

    Field measurements of beta ray energy spectra have been carried out at various locations in CANDU nuclear generating stations operated by Ontario Hydro. The beta ray energy spectrometer consists of a 5 cm diameter x 2 cm thick BC-404 plastic scintillator situated behind a 100 [mu]m thick, totally depleted, silicon detector. Photon events are rejected by requiring a coincidence between the two detectors. The spectrometer is capable of measuring electron energies from 125 keV to 3.5 MeV. Beta ray energy spectra have been measured for uncontaminated and contaminated fueling machine components, fueling machine swipes and a reactor containment vault. The degree of protection afforded by various articles of protective clothing has also been investigated for the various fueling machine components. Monte Carlo calculations have been used to estimate beta factors for 100 mg.cm[sup -2] and 240 mg.cm[sup -2] LiF-TLD chips, which are used as 'skin-and 'extremity' dosemeters in the Ontario Hydro Radiation Dosimetry Programme. (Author).

  19. 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.

  20. Etude de l'influence du champ magnetique dans une section d'essais thermohydraulique d'un canal de reacteur nucleaire CANDU 6

    Science.gov (United States)

    Landry-Lavoie, Renaud

    This memoir deals with the effects of the magnetic fields present in a thermal hydraulic test section of the Canadian nuclear industry. This test section is used to determine the thermal hydraulic conditions that can lead to critical heat flux in a channel of a CANDU 6 nuclear reactor. To perform their series of experiments the STERN Company used strong electric currents to heat the simulation bundles with a thermal power similar to the one found in a channel of a CANDU reactor. The materials constituting the simulation channel and its supports are of ferromagnetic nature. The strong magnetic field generated by the bundles implies that they are subjected to a magnetostatic force due to the magnetization of the ferromagnetic materials. The nuclear industry wants to know if these efforts, combined with the force of gravity, are sufficient to maintain the bundles in place in the simulation channel. The question also arises whether or not the magnetic field present in the channel can affect the parameters of boiling heat transfer. To determine the magnetic field distribution in the simulation channel, we had recourse to the magnetostatic image method and the integral method of calculation of magnetization. The results of the calculations show that the magnetostatic forces exerted by the ferromagnetic elements of the test section are inferior in magnitude to the one estimated by the STERN laboratorie. We used the mechanistic model of Sullivan et al. (1964) to evaluate the possible influence of the magnetic fields on the departure diameter of the vapor bubbles. The deviation in the frequency of bubble emission was evaluated by using the correlations of Zuber et al. (1959) and Cole (1960). By introducing a magnetostatic force in the boiling model and in the correlations, we demonstrated that the magnetic field present in the STERN test section has a negligible effect on the bubble departure diameter and their emission frequency. We conclude that the conditions in the test

  1. Deployment of advanced MACSTOR dry spent fuel storage technology in Korea - A joint development program

    Energy Technology Data Exchange (ETDEWEB)

    Cobanoglu, M. M.; Pattantyus, P. [Atomic Energy Canada Limited, Ottawa (Canada); Song, M. J.; Lee, H. Y. [KHNP/NETEC, Daejeon (Korea, Republic of)

    2002-04-15

    KHNP/NETEC's (K/N) and Atomic Energy of Canada Limited (AECL) are undertaking to jointly develop a high capacity dry storage structure made of reinforced concrete that uses the MACSTOR storage module concept. This effort is based on AECL's experience and on the successful deployment of concrete canisters at Wolsong and on the deployment of air-cooled MACSTOR modules at the Gentilly 2 reactor in Canada. The proposed approach addresses the conditions specific to the Wolsong site: large yearly fuel throughput, space limitations and the need for an economical dry storage structure that can store lifetime spent fuel inventories expected from the four CANDU units. The selected configuration is a 4-row MACSTOR module with a capacity of 24,000 bundles stored in 400 baskets, each holding 60 spent fuel bundles. The module is thus termed MACSTOR/KN-400 and is expected to offer a repetitive storage density increase by a factor of approximately 3, compared to concrete canisters presently used. The four Wolsong units generate spent fuel bundles that, with the high capacity factors achieved, are in the order of 20,000 bundles or more per year. At all Korean nuclear facilities, space limitations dictate the need for storage structures having high storage density. Storage density increases have to be accomplished while maintaining safety parameters during the full term storage of nuclear fuel. During the early 1990's AECL has proceeded with the development of a 2-row MACSTOR storage module that offered a higher storage density and a more economical solution compared to the stand alone concrete canister used at Wolsong 1. These modules are in use at Gentilly since the mid 1990's and operate at a capacity of 200 baskets. The selection of a MACSTOR module with 4 rows of storage cylinders is the natural evolution of the already deployed configuration. It can be developed without additional thermal testing as the fuel is maintained within the existing licensing

  2. In-reactor performance of pressure tubes in CANDU reactors

    Science.gov (United States)

    Rodgers, D. K.; Coleman, C. E.; Griffiths, M.; Bickel, G. A.; Theaker, J. R.; Muir, I.; Bahurmuz, A. A.; Lawrence, S. St.; Resta Levi, M.

    2008-12-01

    The pressure tubes in CANDU reactors have been operating for times up to about 25 years. The in-reactor performance of Zr-2.5Nb pressure tubes has been evaluated by sampling and periodic inspection. This paper describes the behaviour and discusses the factors controlling the behaviour of these components in currently operating CANDU reactors. The mechanical properties (such as ultimate tensile strength, UTS, and fracture toughness), and delayed-hydride-cracking properties (crack growth rate Vc, and threshold stress intensity factor, KIH) change with irradiation; the former reach a limiting value at a fluence of Pressure tubes exhibit elongation and diametral expansion. The deformation behaviour is a function of operating conditions and material properties that vary from tube-to-tube and as a function of axial location. Semi-empirical predictive models have been developed to describe the deformation response of average tubes as a function of operating conditions. For corrosion and, more importantly deuterium pickup, semi-empirical predictive models have also been developed to represent the behaviour of an average tube. The effect of material variability on corrosion behaviour is less well defined compared with other properties. Improvements in manufacturing have increased fracture resistance by minimising trace elements, especially H and Cl, and reduced variability by tightening controls on forming parameters, especially hot-working temperatures.

  3. Gothic simulation of single-channel fuel heatup following a loss of forced flow

    Energy Technology Data Exchange (ETDEWEB)

    Chen, X-Q; Tahir, A. [NSS, Dept. of Thermal Hydraulics Analysis, Toronto, Ontario (Canada); Parlatan, Y. [Ontario Power Generation, NSATD, Pickering, Ontario (Canada); Kwee, M. [Bruce Power, NSASD, Toronto, Ontario (Canada)

    2011-07-01

    GOTHIC v7.2 was used to develop a computer model for the simulation of 28- and 37-element fuel heat-up at a loss of forced flow. The model has accounted for the non-uniformity of both axial and radial power distributions along the fuel channel for a typical CANDU reactor. In addition, the model has also accounted for the fuel rods, end-fittings, feeders and headers. Experimental test conditions for both 28- and 37-element bundles at either low or high powers were used for model validation. GOTHIC predictions of the rod and/or pressure-tube temperatures at a variety of test locations were compared with the corresponding experimental measurements. It is found that the numerical results agree well with the experimental measurements for most of the test locations. Results have also shown that the channel venting time is sensitive to the initial temperature distribution in the feeders and headers. An imposed temperature asymmetry at the beginning will cause the channel flow to vent earlier. (author)

  4. 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.

  5. 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, Taejon (Korea)

    1999-04-01

    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 second step of the whole project, and focus to the implementation of CANDU models based on the previous study. FORTRAN 90 language have been used for the development of RELAP5.MOD3/CANDU PC version. For the convenience of the previous Workstation users, the FOTRAN 77 version has been coded also and implanted into the original RELAP5 source file. The verification of model implementation has been performed through the simple verification calculations using the CANDU version. 6 refs., 15 figs., 7 tabs. (Author)

  6. 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.......60-1.62). The presence of IRBBB was not associated with any adverse outcome.ConclusionIn this cohort study, RBBB and IRBBB were two to three times more common among men than women. Right bundle branch block was associated with increased cardiovascular risk and all-cause mortality, whereas IRBBB was not. Contrary...

  7. 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.

  8. 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 .

  9. 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.

  10. 棒束燃料组件特征栅元CFD方法研究%CFD Method Research on Characteristic Cells in Rod Bundle Fuel Assembly

    Institute of Scientific and Technical Information of China (English)

    陈杰; 陈炳德; 张虹

    2011-01-01

    Two characteristic cells are in AFA-3G fuel assembly, that is typical cell and control rod guide cell. And there are some rules on the arrangement of mixing vanes. For the two characteristic cells, mixing capability is evaluated axially from the point of the first and second kind of sub-channel with CFD method.Mass mixing and heat mixing are interaction but different with each other. Although the mass mixing in the first kind of sub-channel is stronger, the thermal capability of the two is to some tune from the point of heat transfer. In the experiment research on thermal-hydraulic performance of AFA-3G fuel assembly, the arrangements of mixing vanes should refer to the two spacer grids of characteristic cells.%AFA-3G燃料组件中存在典型栅元和控制棒导向管栅元两种特征栅元,定位格架搅混翼的排列也具有一定的规律性.本文采用计算流体力学(CFD)方法,分别针对两种特征栅元,从第一类子通道和第二类子通道的角度,沿程评价其交混性能.质量交混与热交混紧密联系又相互区别,第一类子通道质量交换较强,但从传热角度,二者性能相当.AFA-3G燃料组件热工水力性能的实验研究中,格架搅混翼的排列方式应分别参照两种特征栅元格架.

  11. 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.

  12. Kernel bundle EPDiff

    DEFF Research Database (Denmark)

    Sommer, Stefan Horst; Lauze, Francois Bernard; Nielsen, Mads

    2011-01-01

    In the LDDMM framework, optimal warps for image registration are found as end-points of critical paths for an energy functional, and the EPDiff equations describe the evolution along such paths. The Large Deformation Diffeomorphic Kernel Bundle Mapping (LDDKBM) extension of LDDMM allows scale space...

  13. 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

  14. Integrated evolution of the medium power CANDU{sup MD} reactors; Evolution integree des reacteurs CANDU{sup MD} de moyenne puissance

    Energy Technology Data Exchange (ETDEWEB)

    Nuzzo, F. [AECL Accelerators, Kanata, ON (Canada)

    2002-07-01

    The aim of this document is the main improvements of the CANDU reactors in the economic, safety and performance domains. The presentation proposes also other applications as the hydrogen production, the freshening of water sea and the bituminous sands exploitation. (A.L.B.)

  15. 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.

  16. Macstor system for spent fuel storage

    Energy Technology Data Exchange (ETDEWEB)

    Pattantyus, P. (Atomic Energy of Canada Ltd., Montreal, PQ (Canada). Power Projects)

    1993-01-01

    In 1989, Transnuclear Inc. and AECL jointly developed the conceptual design for the Modular Aircooled Canister Storage System (Macstor) for LWR fuel. The development effort has proceeded to the completion of successful full-scale thermal testing. In 1990, AECL adapted the Macstor System approach for use with Candu fuel. The adapted design, called Canstor, has also successfully completed full-scale thermal testing, and the final system design has been completed. (author) 1 fig.

  17. 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).

  18. Bundling harvester; Nippukorjausharvesteri

    Energy Technology Data Exchange (ETDEWEB)

    Koponen, K. [Eko-Log Oy, Kuopio (Finland)

    1996-12-31

    The staring point of the project was to design and construct, by taking the silvicultural point of view into account, a harvesting and processing system especially for energy-wood, containing manually driven bundling harvester, automatizing of the harvester, and automatized loading. The equipment forms an ideal method for entrepreneur`s-line harvesting. The target is to apply the system also for owner`s-line harvesting. The profitability of the system promotes the utilization of the system in both cases. The objectives of the project were: to construct a test equipment and prototypes for all the project stages, to carry out terrain and strain tests in order to examine the usability and durability, as well as the capacity of the machine, to test the applicability of the Eko-Log system in simultaneous harvesting of energy and pulp woods, and to start the marketing and manufacturing of the products. The basic problems of the construction of the bundling harvester have been solved using terrain-tests. The prototype machine has been shown to be operable. Loading of the bundles to form sufficiently economically transportable loads has been studied, and simultaneously, the branch-biomass has been tried to be utilized without loosing the profitability of transportation. The results have been promising, and will promote the profitable utilization of wood-energy

  19. Release of segregated nuclides from spent fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-10-01

    The potential release of fission and activation products from spent nuclear fuel into groundwater after container failure in the Swedish deep repository is discussed. Data from studies of fission gas release from representative Swedish BWR fuel are used to estimate the average fission gas release for the spent fuel population. Information from a variety of leaching studies on LWR and CANDU fuel are then reviewed as a basis for estimating the fraction of the inventory of key radionuclides that could be released preferentially (the Instant Release Fraction of IRF) upon failure of the fuel cladding. The uncertainties associated with these estimates are discussed. 33 refs, 6 figs, 3 tabs.

  20. Computer code applicability assessment for the advanced Candu reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wren, D.J.; Langman, V.J.; Popov, N.; Snell, V.G. [Atomic Energy of Canada Ltd (Canada)

    2004-07-01

    AECL Technologies, the 100%-owned US subsidiary of Atomic Energy of Canada Ltd. (AECL), is currently the proponents of a pre-licensing review of the Advanced Candu Reactor (ACR) with the United States Nuclear Regulatory Commission (NRC). A key focus topic for this pre-application review is the NRC acceptance of the computer codes used in the safety analysis of the ACR. These codes have been developed and their predictions compared against experimental results over extended periods of time in Canada. These codes have also undergone formal validation in the 1990's. In support of this formal validation effort AECL has developed, implemented and currently maintains a Software Quality Assurance program (SQA) to ensure that its analytical, scientific and design computer codes meet the required standards for software used in safety analyses. This paper discusses the SQA program used to develop, qualify and maintain the computer codes used in ACR safety analysis, including the current program underway to confirm the applicability of these computer codes for use in ACR safety analyses. (authors)

  1. Application of principal component analysis for the diagnosis of neutron overpower system oscillations in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Nasimi, Elnara; Gabbar, Hossam A., E-mail: Hossam.gabbar@uoit.ca

    2014-04-01

    Highlights: • Diagnosis of neutron overpower protection (NOP) in CANDU reactors. • Accurate reactor detector modeling. • NOP detectors response analysis. • Statistical methods for quantitative analysis of NOP detector behavior. - Abstract: An accurate fault modeling and troubleshooting methodology is required to aid in making risk-informed decisions related to design and operational activities of current and future generation of CANDU{sup ®} designs. This paper attempts to develop an explanation for the unanticipated detector response and overall behavior phenomena using statistical methods to compliment traditional engineering analysis techniques. Principal component analysis (PCA) methodology is used for pattern recognition using a case study of Bruce B zone-control level oscillations.

  2. A probabilistic method for leak-before-break analysis of CANDU reactor pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Puls, M.P.; Wilkins, B.J.S.; Rigby, G.L. [Whiteshell Labs., Pinawa (Canada)] [and others

    1997-04-01

    A probabilistic code for the prediction of the cumulative probability of pressure tube ruptures in CANDU type reactors is described. Ruptures are assumed to result from the axial growth by delayed hydride cracking. The BLOOM code models the major phenomena that affect crack length and critical crack length during the reactor sequence of events following the first indications of leakage. BLOOM can be used to develop unit-specific estimates of the actual probability of pressure rupture in operating CANDU reactors and supplement the existing leak before break analysis.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-03-15

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

  4. Managing bundled payments.

    Science.gov (United States)

    Draper, Andrew

    2011-04-01

    Results of Medicare's ACE demonstration project and Geisinger Health System's ProvenCare initiative provide insight into the challenges hospitals will face as bundled payment proliferates. An early analysis of these results suggests that hospitals would benefit from bringing full automation using clinical IT tools to bear in their efforts to meet these challenges. Other important factors contributing to success include board and physician leadership, organizational structure, pricing methodology for bidding, evidence-based medical practice guidelines, supply cost management, process efficiency management, proactive and aggressive case management, business development and marketing strategy, and the financial management system.

  5. Differential calculi on noncommutative bundles

    OpenAIRE

    Pflaum, Markus J.; Schauenburg, Peter

    1996-01-01

    We introduce a category of noncommutative bundles. To establish geometry in this category we construct suitable noncommutative differential calculi on these bundles and study their basic properties. Furthermore we define the notion of a connection with respect to a differential calculus and consider questions of existence and uniqueness. At the end these constructions are applied to basic examples of noncommutative bundles over a coquasitriangular Hopf algebra.

  6. 快堆燃料组件棒束通道内流动和传热现象分析与研究%Analysis and Investigation on Flow and Convective Heat Transfer Phenomena of Fast Reactor Fuel Bundle

    Institute of Scientific and Technical Information of China (English)

    刘洋; 喻宏; 刘一哲

    2014-01-01

    利用三维计算流体力学软件CFX 12.0对由7根带螺旋状定位绕丝的燃料棒组成的快堆燃料组件典型棒束通道内的流动和传热现象进行了数值模拟。模拟得到不同 Re下的压降系数曲线与 N u曲线,并将计算结果与经验公式的计算结果进行了比较,两者符合较好。研究了组件内3类典型子通道的横向流交混效应,分析了3类典型子通道的横向流分布特点,发现角子通道横向流交混强度沿轴向波动较大,而3类子通道的横向流交混强度均存在周期性。研究了中心燃料棒壁面上3个截面的局部换热效应,发现在燃料棒与绕丝接触处传热效果最差,在事故分析时应重点关注。%Commercial CFD code CFX 12.0 was used as an approach to investigate the flow and convective heat transfer phenomena of a typical 7-pin fuel bundle with wrapped-wire .The pressure drop coefficient and Nusselt number under different Re were obtained , and the simulation results are in good agreement with empirical correlations .The transverse flow effects in three kinds of typical subchannels were investigated ,and its features were analyzed .The results show that the corner channel’s transverse flow intensity varies heavily along the axial direction ,and the transverse flow intensities in three typical subchannels show periodic feature .The local Nusselt number distribution on three cross sections of the central rod surface was also analyzed .The results show that the minimum local Nusselt number occurs close to the contact location of wire and rod ,which means that the convective heat transfer condition is the worst in such area ,and the special attention in accident analysis is deserved .

  7. Safety design of next generation SUI of CANDU stations

    Energy Technology Data Exchange (ETDEWEB)

    Nasimi, Elnara [Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, 2000 Simcoe St. N., Oshawa, L1H 7K4 ON (Canada); Gabbar, Hossam A., E-mail: hossam.gabbar@uoit.ca [Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, 2000 Simcoe St. N., Oshawa, L1H 7K4 ON (Canada)

    2013-01-15

    Highlights: Black-Right-Pointing-Pointer Review of current SUI technologies and challenges. Black-Right-Pointing-Pointer Propose a new type of SUI detectors. Black-Right-Pointing-Pointer Propose a new SUI system architecture and layout. Black-Right-Pointing-Pointer Propose implementation procedure for SUI with reduced risks. - Abstract: Due to the age and operating experience of Nuclear Power Plants, equipment ageing and obsolescence has become one of the main challenges that need to be resolved for all systems, structures and components in order to ensure a safe and reliable production of energy. This paper summarizes the research into a methodology for modernization of Start-Up Instrumentation (SUI), both in-core and Control Room equipment, using a new generation of detectors and cables in order to manage obsolescence. The main objective of this research is to develop a new systematic approach to SUI installation/replacement procedure development and optimization. Although some additional features, such as real-time data monitoring and storage/archiving solutions for SUI systems are also examined to take full advantage of today's digital technology, the objectives of this study do not include detailed parametrical studies of detector or system performance. Instead, a number of technological, operational and maintenance issues associated with Start-Up Instrumentation systems at Nuclear Power Plants (NPPs) will be identified and a structured approach for developing a replacement/installation procedure that can be standardized and used across all of the domestic CANDU (Canadian Deuterium Uranium) stations is proposed.

  8. Signal processing system design for improved shutdown system of CANDU{sup ®} nuclear reactors in large break LOCA events

    Energy Technology Data Exchange (ETDEWEB)

    Gabbar, Hossam A., E-mail: hossam.gabbar@uoit.ca [Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, 2000 Simcoe Street North, Oshawa, ON, Canada L1H 7K4 (Canada); Faculty of Engineering and Applied Science, University of Ontario Institute of Technology, 2000 Simcoe Street North, Oshawa, ON, Canada L1H 7K4 (Canada); Xia, Lingzhi; Isham, Manir U. [Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, 2000 Simcoe Street North, Oshawa, ON, Canada L1H 7K4 (Canada); Ponomarev, Vladimir [Megawatt Solutions, 1235 Radom St., unit 68, Pickering, ON, Canada L1W 1J3 (Canada)

    2016-03-15

    Highlights: • Neutronic signal processing system design to improve CANDU SDS1 performance. • Reactor modeling for CANDU LLOCA transient. • MATLAB/Simulink system implementation for the SDS1 trip logic. • Increasing the SDS1 trip response. - Abstract: For CANDU reactors, several options to improve CANDU nuclear power plant operation safety margin have been investigated in this paper. A particular attention is paid to the response time of CANDU shutdown system number 1 (SDS1) in case of large break loss of coolant accident (LLOCA). Based on point kinetic method, a systematic fundamental analysis is performed to CANDU LLOCA event, and the power transient signal is generated. In order to improve the SDS1 response time during LLOCA events, an innovative power measurement and signal processing system is particularly designed. The new signal processing system is implemented with the input of the LLOCA power transient, and the simulation results of the reactor trip time and signal are compared to those of the existing system in CANDU power plants. It is demonstrated that the new signal processing system can not only achieve a shorter reactor trip time than the existing system, but also accommodate the spurious trip immunity. This will significantly enhance the safety margin for the power plant operation, or bring extra economical benefits to the power plant units.

  9. EVALUATION OF THE APPLICABLE REACTIVITY RANGE OF A REACTIVITY COMPUTER FOR A CANDU-6 REACTOR

    Directory of Open Access Journals (Sweden)

    EUN KI LEE

    2014-04-01

    Full Text Available Recently, a CANDU digital reactivity computer system (CDRCS to measure the worth of the liquid zone controller in a CANDU-6 was developed and successfully applied to a physics test of refurbished Wolsong Unit 1. In advance of using the CDRCS, its measureable reactivity range should be investigated and confirmed. There are two reasons for this investigation. First, the CANDU-6 has a larger reactor and smaller excore detectors than a general PWR and consequently the measured reactivity is likely to reflect the peripheral power variation only, not the whole core. The second reason is photo neutrons generated from the interaction of the moderator and gamma-rays, which are never considered in a PWR. To evaluate the limitations of the CDRCS, several tens of three-dimensional steady and transient simulations were performed. The simulated detector signals were used to obtain the dynamic reactivity. The difference between the dynamic reactivity and the static worth increases in line with the water level changes. The maximum allowable reactivity was determined to be 1.4 mk in the case of CANDU-6 by confining the difference to less than 1%.

  10. FMDP Reactor Alternative Summary Report: Volume 2 - CANDU heavy water reactor alternative

    Energy Technology Data Exchange (ETDEWEB)

    Greene, S.R.; Spellman, D.J.; Bevard, B.B. [and others

    1996-09-01

    The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 2 of a four volume report, summarizes the results of these analyses for the CANDU reactor based plutonium disposition alternative.

  11. Bundle Security Protocol for ION

    Science.gov (United States)

    Burleigh, Scott C.; Birrane, Edward J.; Krupiarz, Christopher

    2011-01-01

    This software implements bundle authentication, conforming to the Delay-Tolerant Networking (DTN) Internet Draft on Bundle Security Protocol (BSP), for the Interplanetary Overlay Network (ION) implementation of DTN. This is the only implementation of BSP that is integrated with ION.

  12. Bundle Formation in Biomimetic Hydrogels

    NARCIS (Netherlands)

    Jaspers, Maarten; Pape, A C H; Voets, Ilja K; Rowan, Alan E; Portale, Giuseppe; Kouwer, Paul H J

    2016-01-01

    Bundling of single polymer chains is a crucial process in the formation of biopolymer network gels that make up the extracellular matrix and the cytoskeleton. This bundled architecture leads to gels with distinctive properties, including a large-pore-size gel formation at very low concentrations and

  13. Vector Bundles And F Theory

    CERN Document Server

    Friedman, R; Witten, Edward

    1997-01-01

    To understand in detail duality between heterotic string and F theory compactifications, it is important to understand the construction of holomorphic G bundles over elliptic Calabi-Yau manifolds, for various groups G. In this paper, we develop techniques to describe these bundles, and make several detailed comparisons between the heterotic string and F theory.

  14. Vector Bundles And F Theory

    OpenAIRE

    Friedman, Robert; Morgan, John; Witten, Edward

    1997-01-01

    To understand in detail duality between heterotic string and F theory compactifications, it is important to understand the construction of holomorphic G bundles over elliptic Calabi-Yau manifolds, for various groups G. In this paper, we develop techniques to describe the bundles, and make several detailed comparisons between the heterotic string and F theory.

  15. Thermal hydraulics of rod bundles: The effect of eccentricity

    Energy Technology Data Exchange (ETDEWEB)

    Chauhan, Amit K., E-mail: amit_fmlab@yahoo.co.in [Fluid Mechanics Laboratory, Department of Applied Mechanics, Indian Institute of Technology Madras, Chennai 600036 (India); Prasad, B.V.S.S.S., E-mail: prasad@iitm.ac.in [Thermal Turbomachines Laboratory, Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Patnaik, B.S.V., E-mail: bsvp@iitm.ac.in [Fluid Mechanics Laboratory, Department of Applied Mechanics, Indian Institute of Technology Madras, Chennai 600036 (India)

    2013-10-15

    Highlights: • Present CFD investigation explores, whole bundle eccentricity for the first time. • Fluid flow and thermal characteristics in various subchannels are analyzed. • Mass flux distribution is particularly analyzed to study eccentricity effect. • Higher eccentricity resulted in a shoot up in rod surface temperature distribution. • Both tangential and radial flow in rod bundles has resulted due to eccentricity. -- Abstract: The effect of eccentricity on the fluid flow and heat transfer through a 19-rod bundle is numerically carried out. When the whole bundle shifts downwards with respect to the outer (pressure) tube, flow redistribution happens. This in turn is responsible for changes in mass flux, pressure and differential flow development in various subchannels. The heat flux imposed on the surface of the fuel rods and the mass flux through the subchannels determines the coolant outlet temperatures. The simulations are performed for a coolant flow Reynolds number of 4 × 10{sup 5}. For an eccentricity value of 0.7, the mass flux in the bottom most subchannel (l) was found to decrease by 10%, while the surface temperature of the fuel rod in the vicinity of this subchannel increased by 250% at the outlet section. Parameters of engineering interest including skin friction coefficient, Nusselt number, etc., have been systematically explored to study the effect of eccentricity on the rod bundle.

  16. Fiber bundle phase conjugate mirror

    Science.gov (United States)

    Ward, Benjamin G.

    2012-05-01

    An improved method and apparatus for passively conjugating the phases of a distorted wavefronts resulting from optical phase mismatch between elements of a fiber laser array are disclosed. A method for passively conjugating a distorted wavefront comprises the steps of: multiplexing a plurality of probe fibers and a bundle pump fiber in a fiber bundle array; passing the multiplexed output from the fiber bundle array through a collimating lens and into one portion of a non-linear medium; passing the output from a pump collection fiber through a focusing lens and into another portion of the non-linear medium so that the output from the pump collection fiber mixes with the multiplexed output from the fiber bundle; adjusting one or more degrees of freedom of one or more of the fiber bundle array, the collimating lens, the focusing lens, the non-linear medium, or the pump collection fiber to produce a standing wave in the non-linear medium.

  17. Twisted Vector Bundles on Pointed Nodal Curves

    Indian Academy of Sciences (India)

    Ivan Kausz

    2005-05-01

    Motivated by the quest for a good compactification of the moduli space of -bundles on a nodal curve we establish a striking relationship between Abramovich’s and Vistoli’s twisted bundles and Gieseker vector bundles.

  18. CANDU heat sinks improvements as a follow up to Fukushima Daiichi accident ''the regulator perspective''

    Energy Technology Data Exchange (ETDEWEB)

    Mesmous, Noreddine; Harwood, Chris [Canadian Nuclear Safety Commission, Ottawa, ON (Canada)

    2015-06-15

    The purpose of this paper is to provide a summary of the Canadian Nuclear Safety Commission (CNSC) recommendations related to improving the heat sink strategy as a follow up to the Fukushima Daiichi Accident (FDA). As a follow up to FDA, CNSC staff tasked the Nuclear Power Plant (NPP) licensees to review the lessons learned from the FDA and re-examine the NPP safety cases. The reviews have examined the CANDU defence-in-depth strategy and considered events more severe than those that have historically been regarded as credible, and evaluated their impact on the NPPs safety. Availability of emergency equipment was shown to be crucial during the FDA and its availability could have arrested the accident progression early enough to minimize any radioactive release to the environment. As a result, licensees presented appropriate evaluations of the means to provide coolant make-up to the primary Heat Transport System (HTS), boilers, moderator, calandria vault, and irradiated fuel pools.

  19. The Analysis of SBWR Critical Power Bundle Using Cobrag Code

    Directory of Open Access Journals (Sweden)

    Yohannes Sardjono

    2013-03-01

    Full Text Available The coolant mechanism of SBWR is similar with the Dodewaard Nuclear Power Plant (NPP in the Netherlands that first went critical in 1968. The similarity of both NPP is cooled by natural convection system. These coolant concept is very related with same parameters on fuel bundle design especially fuel bundle length, core pressure drop and core flow rate as well as critical power bundle. The analysis was carried out by using COBRAG computer code. COBRAG computer code is GE Company proprietary. Basically COBRAG computer code is a tool to solve compressible three-dimensional, two fluid, three field equations for two phase flow. The three fields are the vapor field, the continuous liquid field, and the liquid drop field. This code has been applied to analyses model flow and heat transfer within the reactor core. This volume describes the finitevolume equations and the numerical solution methods used to solve these equations. This analysis of same parameters has been done i.e.; inlet sub cooling 20 BTU/lbm and 40 BTU/lbm, 1000 psi pressure and R-factor is 1.038, mass flux are 0.5 Mlb/hr.ft2, 0.75 Mlb/hr.ft2, 1.00 Mlb/hr.ft2 and 1.25 Mlb/hr.ft2. Those conditions based on history operation of some type of the cell fuel bundle line at GE Nuclear Energy. According to the results, it can be concluded that SBWR critical power bundle is 10.5 % less than current BWR critical power bundle with length reduction of 12 ft to 9 ft.

  20. SAFIRE - a robotic inspection system for CANDU feeders

    Energy Technology Data Exchange (ETDEWEB)

    Buckingham, R. [OC Robotics, Bristol (United Kingdom)

    2011-07-01

    The condition of primary circuit feeder pipes in CANDU reactors is relevant to the commercial viability and plant life. One known wear mechanism is external fretting between feeder pipes and adjacent services or support structures, particularly within the Upper Feeder Cabinet (UFC). Fretting leads to wall thinning which must not exceed certain agreed limits. Chafe shields have been added to protect the feeder pipes. Regular inspections are required of the chafe shields, feeder pipes and other structures that may cause feeder damage. Historically, the dose received by inspectors conducting this work has been significant. For this reason Ontario Power Generation has invested in a remotely operated robot system to conduct visual inspections within the UFC. This system, called SAFIRE for 'Snake-Arm Feeder Inspection Robot Equipment' has been deployed at Pickering during 2010 and 2011 and has been used to inspect areas that are extremely difficult to inspect with existing manual techniques. The 2011 scope of work included inspection of a total of 660 feeder pipes in three UFC quadrants, in two reactors. The full scope was completed over a one-month period in Autumn 2011 in which SAFIRE was used during 23, twelve hour shifts. This included two periods each of 72 hours of continuous operation using multiple teams of operators. SAFIRE is remote controlled delivery system for multiple cameras to record still images and video. The main system elements include a snake-arm robot mounted on a mobile vehicle. It can be controlled from up to 500m away using a fibre/copper connection. The snake-arm is 2.2m long, 25mm wide and has 18 degrees of freedom. It is designed to snake between the rows of feeder pipes to inspect feeder/hanger interfaces, both above and below the feeder cabinet catwalks. Future upgrades offer the potential to add additional tools to increase functionality. This paper describes the SAFIRE development process from inception to operational experience

  1. Fission product inventory for CANFLEX-NU and CANFLEX-RU in the limiting channel of CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Jung, Jong Yeob

    2007-09-15

    The fission product inventory was evaluated for the 37-element fuel, CANFLEX-NU and CANFLEX-RU fuels which irradiated in the assumed limiting O6{sub m}od channel under a normal operating condition. Generally, the assessment of fission product release is divided into 3 steps which are a calculation of FGR during steady state, a calculation of a prompt release at the time of an accident, and a calculation of delayed release after the accident. Among them the first step was described in this report. The analysis was performed by using the ELESTRES code which is a computer code used in a performance evaluation of the fuel bundle during the normal irradiation. The calculation results of the fission product inventory for the 37-element fuel, CANFLEX-NU and CANFLEX-RU were compared to each other.

  2. Semiflexible Biopolymers in Bundled Arrangements

    Directory of Open Access Journals (Sweden)

    Jörg Schnauß

    2016-07-01

    Full Text Available Bundles and networks of semiflexible biopolymers are key elements in cells, lending them mechanical integrity while also enabling dynamic functions. Networks have been the subject of many studies, revealing a variety of fundamental characteristics often determined via bulk measurements. Although bundles are equally important in biological systems, they have garnered much less scientific attention since they have to be probed on the mesoscopic scale. Here, we review theoretical as well as experimental approaches, which mainly employ the naturally occurring biopolymer actin, to highlight the principles behind these structures on the single bundle level.

  3. Evaluating big deal journal bundles.

    Science.gov (United States)

    Bergstrom, Theodore C; Courant, Paul N; McAfee, R Preston; Williams, Michael A

    2014-07-01

    Large commercial publishers sell bundled online subscriptions to their entire list of academic journals at prices significantly lower than the sum of their á la carte prices. Bundle prices differ drastically between institutions, but they are not publicly posted. The data that we have collected enable us to compare the bundle prices charged by commercial publishers with those of nonprofit societies and to examine the types of price discrimination practiced by commercial and nonprofit journal publishers. This information is of interest to economists who study monopolist pricing, librarians interested in making efficient use of library budgets, and scholars who are interested in the availability of the work that they publish.

  4. Assessment of the thorium fuel cycle in power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Homan, F.J.; Allen, E.J.

    1977-01-01

    A study was conducted at Oak Ridge National Laboratory to evaluate the role of thorium fuel cycles in power reactors. Three thermal reactor systems were considered: Light Water Reactors (LWRs); High-Temperature Gas-Cooled Reactors (HTGRs); and Heavy Water Reactors (HWRs) of the Canadian Deuterium Uranium Reactor (CANDU) type; most of the effort was on these systems. A summary comparing thorium and uranium fuel cycles in Fast Breeder Reactors (FBRs) was also compiled.

  5. The Atiyah Bundle and Connections on a Principal Bundle

    Indian Academy of Sciences (India)

    Indranil Biswas

    2010-06-01

    Let be a ∞ manifold and a Lie a group. Let $E_G$ be a ∞ principal -bundle over . There is a fiber bundle $\\mathcal{C}(E_G)$ over whose smooth sections correspond to the connections on $E_G$. The pull back of $E_G$ to $\\mathcal{C}(E_G)$ has a tautological connection. We investigate the curvature of this tautological connection.

  6. Vector bundles on toric varieties

    CERN Document Server

    Gharib, Saman

    2011-01-01

    Following Sam Payne's work, we study the existence problem of nontrivial vector bundles on toric varieties. The first result we prove is that every complete fan admits a nontrivial conewise linear multivalued function. Such functions could potentially be the Chern classes of toric vector bundles. Then we use the results of Corti\\~nas, Haesemeyer, Walker and Weibel to show that the (non-equivariant) Grothendieck group of the toric 3-fold studied by Payne is large, so the variety has a nontrivial vector bundle. Using the same computation, we show that every toric 3-fold X either has a nontrivial line bundle, or there is a finite surjective toric morphism from Y to X, such that Y has a large Grothendieck group.

  7. Bundling ecosystem services in Denmark

    DEFF Research Database (Denmark)

    Turner, Katrine Grace; Odgaard, Mette Vestergaard; Bøcher, Peder Klith;

    2014-01-01

    We made a spatial analysis of 11 ecosystem services at a 10 km × 10 km grid scale covering most of Denmark. Our objective was to describe their spatial distribution and interactions and also to analyze whether they formed specific bundle types on a regional scale in the Danish cultural landscape....... We found clustered distribution patterns of ecosystem services across the country. There was a significant tendency for trade-offs between on the one hand cultural and regulating services and on the other provisioning services, and we also found the potential of regulating and cultural services...... to form synergies. We identified six distinct ecosystem service bundle types, indicating multiple interactions at a landscape level. The bundle types showed specialized areas of agricultural production, high provision of cultural services at the coasts, multifunctional mixed-use bundle types around urban...

  8. Left bundle-branch block

    DEFF Research Database (Denmark)

    Risum, Niels; Strauss, David; Sogaard, Peter;

    2013-01-01

    The relationship between myocardial electrical activation by electrocardiogram (ECG) and mechanical contraction by echocardiography in left bundle-branch block (LBBB) has never been clearly demonstrated. New strict criteria for LBBB based on a fundamental understanding of physiology have recently...

  9. Fabrication of electrospun nanofibers bundles

    Science.gov (United States)

    Ye, Junjun; Sun, Daoheng

    2007-12-01

    Aligned nanofibers, filament bundle composed of large number of nanofibers have potential applications such as bio-material, composite material etc. A series of electrospinning experiments have been conducted to investigate the electrospinning process,in which some parameters such as polymer solution concentration, bias voltage, distance between spinneret and collector, solution flow rate etc have been setup to do the experiment of nanofibers bundles construction. This work firstly reports electrospun nanofiber bundle through non-uniform electrical field, and nanofibers distributed in different density on electrodes from that between them. Thinner nanofibers bundle with a few numbers of nanofiber is collected for 3 seconds; therefore it's also possible that the addressable single nanofiber could be collected to bridge two electrodes.

  10. Simulation of bundle test Quench-12 with integral code MELCOR

    Energy Technology Data Exchange (ETDEWEB)

    Duspiva, J. [Nuclear Research Inst., Rez plc (Czech Republic)

    2011-07-01

    The past NRI analyses cover the Quench-01, Quench-03 and Quench-06 with version MELCOR 1.8.5 (including reflood model), and Quench-01 and Quench-11 tests with the latest version MELCOR 1.8.6. The Quench-12 test is specific, because it has different bundle configuration related to the VVER bundle configuration with hexagonal grid of pins and also used E110 cladding material. Specificity of Quench-12 test is also in the used material of fuel rod cladding - E110. The test specificities are a reason for the highest concern, because the VVER reactors are operated in the Czech Republic. The new input model was developed with the taking into account all experience from previous simulations of the Quench bundle tests. The recent version MELCOR 1.8.6 YU{sub 2}911 was used for the simulation with slightly modified ELHEAT package. Sensitivity studies on input parameters and oxidation kinetics were performed. (author)

  11. Reconnection of superfluid vortex bundles.

    Science.gov (United States)

    Alamri, Sultan Z; Youd, Anthony J; Barenghi, Carlo F

    2008-11-21

    Using the vortex filament model and the Gross-Pitaevskii nonlinear Schroedinger equation, we show that bundles of quantized vortex lines in He II are structurally robust and can reconnect with each other maintaining their identity. We discuss vortex stretching in superfluid turbulence and show that, during the bundle reconnection process, kelvin waves of large amplitude are generated, in agreement with the finding that helicity is produced by nearly singular vortex interactions in classical Euler flows.

  12. 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)

    1998-04-15

    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 first step of the whole project, thus focus to the establishment of improvement area. The study was performed by reconsideration of the previous code assessment works and investigation of AECL design analysis tools. In order to identify the thermal hydraulic phenomena for events, the whole system of CANDU plant was divided into main functional systems and subcomponents. Each phenomena was addressed to the each subcomponent. FinaIly improvement areas of model development for auditing tool were established based on the identified phenomena.

  13. 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, Taejeon (Korea)

    1998-04-01

    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 first step of the whole project, thus focus to the establishment of improvement area. The study was performed by reconsideration of the previous code assessment works and investigation of AECL design analysis tools. In order to identify the thermal hydraulic phenomena for events, the whole system of CANDU plant was divided into main functional systems and subcomponents. Each phenomena was addressed to the each subcomponent. Finally improvement areas of model development for auditing tool were established based on the identified phenomena. 8 refs., 21 figs., 19 tabs. (Author)

  14. A study on the interlink of CANDU safety analysis codes with development of GUI system

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. J.; Jeo, Y. J.; Park, Q. C. [Seoul National Univ., Seoul (Korea, Republic of); Kim, H. T.; Min, B. J. [KAERI, Taejon (Korea, Republic of)

    2003-10-01

    In order to improve the CANDU safety analysis code system, the interlink of containment analysis code, PRESCON2 to the system thermal hydraulics analysis code, CATHENA, has been implemented with development of the GUI system. Before the GUI development, we partly corrected two codes to optimize on the PC environment. The interlink of two codes could be executed by introducing three interlinking variables, mass flux, mixture enthalpy, and mixture specific volume. To guarantee the robustness of the codes, two codes are extremely linked by using the GUI system. The GUI system provides much of user-friendly functions and will be improved step by step. This study is expected to improve the safety assessment system and technology for CANDU NPPs.

  15. Evaluation of hydride blisters in zirconium pressure tube in CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cheong, Y. M.; Kim, Y. S.; Gong, U. S.; Kwon, S. C.; Kim, S. S.; Choo, K.N

    2000-09-01

    When the garter springs for maintaining the gap between the pressure tube and the calandria tube are displaced in the CANDU reactor, the sagging of pressure tube results in a contact to the calandria tube. This causes a temperature difference between the inner and outer surface of the pressure tube. The hydride can be formed at the cold spot of outer surface and the volume expansion by hydride dormation causes the blistering in the zirconium alloys. An incident of pressure tube rupture due to the hydride blisters had happened in the Canadian CANDU reactor. This report describes the theoretical development and models on the formation and growth of hydride blister and some experimental results. The evaluation methodology and non-destructive testing for hydride blister in operating reactors are also described.

  16. Implementation of an on-line monitoring system for transmitters in a CANDU nuclear power plant

    Science.gov (United States)

    Labbe, A.; Abdul-Nour, G.; Vaillancourt, R.; Komljenovic, D.

    2012-05-01

    Many transmitters (pressure, level and flow) are used in a nuclear power plant. It is necessary to calibrate them periodically to ensure that their measurements are accurate. These calibration tasks are time consuming and often contribute to worker radiation exposure. Human errors can also sometimes degrade their performance since the calibration involves intrusive techniques. More importantly, experience has shown that the majority of current calibration efforts are not necessary. These facts motivated the nuclear industry to develop new technologies for identifying drifting instruments. These technologies, well known as on-line monitoring (OLM) techniques, are non-intrusive and allow focusing the maintenance efforts on the instruments that really need a calibration. Although few OLM systems have been implemented in some PWR and BWR plants, these technologies are not commonly used and have not been permanently implemented in a CANDU plant. This paper presents the results of a research project that has been performed in a CANDU plant in order to validate the implementation of an OLM system. An application project, based on the ICMP algorithm developed by EPRI, has been carried out in order to evaluate the performance of an OLM system. The results demonstrated that the OLM system was able to detect the drift of an instrument in the majority of the studied cases. A feasibility study has also been completed and has demonstrated that the implementation of an OLM system at a CANDU nuclear power plant could be advantageous under certain conditions.

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

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Bub Dong; Lee, Won Jae; Hwang, Moon Kyu; Lim, Hong Sik [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-04-01

    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.The study was performed by reconsideration of the previous code assessment works and phenomena identification for essential accident scenario. Improvement areas of model development for auditing tool were identified based on the code comparison and PIRT results. Nine models have been improved significantly for the analysis of LOCA and Mon LOCA event. Conceptual problem or separate effect assessment have been performed to verify the model improvement. The linking calculation with CONTAIN 2.0 has been also enabled to establish the unified auditing code system. Analysis for the CANDU plant real transient and hypothetical LOCA bas been performed using the improved version. It has been concluded that the developed version can be utilized for the auditing analysis of LOCA and non-LOCA event for the CANDU reactor. 25 refs., 84 figs., 36 tabs. (Author)

  18. Mechanical engineering utilizing advanced engineering tools for the CANDU 9 project

    Energy Technology Data Exchange (ETDEWEB)

    Nuzzo, F.; Yu, S.K.W.; Hedges, K.R. [Atomic Energy of Canada Limited (AECL), Ontario (Canada)

    1998-05-01

    To meet the increasing challenging project requirements such as cost and schedule reduction, AECL has incorporated a comprehensive suite of integrated, advanced engineering tools for CANDU project engineering and delivery. This paper provides a description of the advanced engineering tools developed and used by AECL in the pre-project engineering of the CANDU 9 product and the construction projects such as the construction of two CANDU 6 units in Qinshan, China. The advanced mechanical engineering tools described include: the Process and Instrument Diagram ; the mechanical/piping 3D models; the CADDS/piping analysis interface (PAI) tool; the pipe support design system (SDS) tool; and the powerful equipment database tool - TeddyBase. A description of the enhanced work process will also be provided. The work process improvement is a direct result of the implementation of advanced information technology and the integration of AECL tools with commercial engineering and project tools available in the market. The use of these advanced tools results in better design quality; enhanced presentation of the engineering deliverables to construction and commissioning staff; and potential support to future plant operations (Ref.1). (author). 2 refs., 3 figs.

  19. Power distribution control of CANDU reactors based on modal representation of reactor kinetics

    Energy Technology Data Exchange (ETDEWEB)

    Xia, Lingzhi, E-mail: lxia4@uwo.ca [Department of Electrical and Computer Engineering, The University of Western Ontario, London, Ontario N6A 5B9 (Canada); Jiang, Jin, E-mail: jjiang@eng.uwo.ca [Department of Electrical and Computer Engineering, The University of Western Ontario, London, Ontario N6A 5B9 (Canada); Luxat, John C., E-mail: luxatj@mcmaster.ca [Department of Engineering Physics, McMaster University, Hamilton, Ontario L8S 4L7 (Canada)

    2014-10-15

    Highlights: • Linearization of the modal synthesis model of neutronic kinetic equations for CANDU reactors. • Validation of the linearized dynamic model through closed-loop simulations by using the reactor regulating system. • Design of a LQR state feedback controller for CANDU core power distribution control. • Comparison of the results of this new controller against those of the conventional reactor regulation system. - Abstract: Modal synthesis representation of a neutronic kinetic model for a CANDU reactor core has been utilized in the analysis and synthesis for reactor control systems. Among all the mode shapes, the fundamental mode of the power distribution, which also coincides with the desired reactor power distribution during operation, is used in the control system design. The nonlinear modal models are linearized around desired operating points. Based on the linearized model, linear quadratic regulator (LQR) control approach is used to synthesize a state feedback controller. The performance of this controller has been evaluated by using the original nonlinear models under load-following conditions. It has been demonstrated that the proposed reactor control system can produce more uniform power distribution than the traditional reactor regulation systems (RRS); in particular, it is more effective in compensating the Xenon induced transients.

  20. Bundle Formation in Biomimetic Hydrogels.

    Science.gov (United States)

    Jaspers, Maarten; Pape, A C H; Voets, Ilja K; Rowan, Alan E; Portale, Giuseppe; Kouwer, Paul H J

    2016-08-08

    Bundling of single polymer chains is a crucial process in the formation of biopolymer network gels that make up the extracellular matrix and the cytoskeleton. This bundled architecture leads to gels with distinctive properties, including a large-pore-size gel formation at very low concentrations and mechanical responsiveness through nonlinear mechanics, properties that are rarely observed in synthetic hydrogels. Using small-angle X-ray scattering (SAXS), we study the bundle formation and hydrogelation process of polyisocyanide gels, a synthetic material that uniquely mimics the structure and mechanics of biogels. We show how the structure of the material changes at the (thermally induced) gelation point and how factors such as concentration and polymer length determine the architecture, and with that, the mechanical properties. The correlation of the gel mechanics and the structural parameters obtained from SAXS experiments is essential in the design of future (synthetic) mimics of biopolymer networks.

  1. Principal bundles the classical case

    CERN Document Server

    Sontz, Stephen Bruce

    2015-01-01

    This introductory graduate level text provides a relatively quick path to a special topic in classical differential geometry: principal bundles.  While the topic of principal bundles in differential geometry has become classic, even standard, material in the modern graduate mathematics curriculum, the unique approach taken in this text presents the material in a way that is intuitive for both students of mathematics and of physics. The goal of this book is to present important, modern geometric ideas in a form readily accessible to students and researchers in both the physics and mathematics communities, providing each with an understanding and appreciation of the language and ideas of the other.

  2. Fabrication of CANFLEX bundle kit for irradiation test in NRU

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Moon Sung; Kwon, Hyuk Il; Ji, Chul Goo; Chang, Ho Il; Sim, Ki Seob; Suk, Ho Chun

    1997-10-01

    CANFLEX bundle kit was prepared at KAERI for the fabrication of complete bundle at AECL. Completed bundle will be used for irradiation test in NRU. Provisions in the `Quality Assurance Manual for HWR Fuel Projects,` `Manufacturing Plan` and `Quality Verification, Inspection and Test Plan` were implemented as appropriately for the preparation of CANFLEX kit. A set of CANFLEX kit consist of 43 fuel sheath of two different sizes with spacers, bearing pads and buttons attached, 2 pieces of end plates and 86 pieces of end caps with two different sizes. All the documents utilized as references for the fabrication such as drawings, specifications, operating instructions, QC instructions and supplier`s certificates are specified in this report. Especially, suppliers` certificates and inspection reports for the purchased material as well as KAERI`s inspection report are integrated as attachments to this report. Attached to this report are supplier`s certificates and KAERI inspection reports for the procured materials and KAERI QC inspection reports for tubes, pads, spacers, buttons, end caps, end plates and fuel sheath. (author). 37 refs.

  3. Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients

    Science.gov (United States)

    Todreas, N. E.; Cheng, S. K.; Basehore, K.

    1984-08-01

    The thermal hydraulic performance of wire wrapped fuel bundles of LMFBR configuration was investigated. Results obtained included phenomenological models for friction factors, flow split and mixing characteristics; correlations for predicting these characteristics suitable for insertion in design codes; numerical codes for analyzing bundle behavior both of the lumped subchannel and distributed parameter categories and experimental techniques for pressure velocity, flow split, salt conductivity and temperature measurement in water cooled mockups of bundles and subchannels. Flow regimes investigated included laminar, transition and turbulent flow under forced convection and mixed convection conditions. Forced convections conditions are emphasized. Outlet plenum behavior is also investigated.

  4. Recent Developments in the Management of Cameco Corporation's Fuel Services Division Waste - 13144

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Thomas P. [Cameco Corporation, Port Hope, Ontario (Canada)

    2013-07-01

    Cameco Corporation is a world leader in uranium production. Headquartered in Saskatoon, Saskatchewan our operations provide 16% of the world uranium mine production and we have approximately 435 million pounds of proven and probable uranium reserves. Cameco mining operations are located in Saskatchewan, Wyoming, Nebraska and Kazakhstan. Cameco is also a major supplier of uranium processing services required to produce fuel for the generation of clean energy. These operations are based in Blind River, Cobourg and Port Hope, Ontario and are collectively referred to as the Fuel Services Division. The Fuel Services Division produces uranium trioxide from uranium ore concentrate at the Blind River Refinery. Cameco produces uranium hexafluoride and uranium dioxide at the Port Hope Conversion Facility. Cameco operates a fuel manufacturing facility in Port Hope, Ontario and a metal fabrication facility located in Cobourg, Ontario. The company manufactures fuel bundles utilized in the Candu reactors. Cameco's Fuel Services Division produces several types of low-level radioactively contaminated wastes. Internal processing capabilities at both the Blind River Refinery and Port Hope Conversion Facility are extensive and allow for the recycling of several types of waste. Notwithstanding these capabilities there are certain wastes that are not amenable to the internal processing capabilities and must be disposed of appropriately. Disposal options for low-level radioactively contaminated wastes in Canada are limited primarily due to cost considerations. In recent years, Cameco has started to ship marginally contaminated wastes (<500 ppm uranium) to the United States for disposal in an appropriate landfill. The landfill is owned by US Ecology Incorporated and is located near Grand View, Idaho 70 miles southeast of Boise in the Owyhee Desert. The facility treats and disposes hazardous waste, non-hazardous industrial waste and low-activity radioactive material. The site

  5. Creep rupture of fiber bundles

    DEFF Research Database (Denmark)

    Linga, G.; Ballone, P.; Hansen, Alex

    2015-01-01

    The creep deformation and eventual breaking of polymeric samples under a constant tensile load F is investigated by molecular dynamics based on a particle representation of the fiber bundle model. The results of the virtual testing of fibrous samples consisting of 40000 particles arranged on Nc=4...

  6. Vector Bundles over Elliptic Fibrations

    CERN Document Server

    Friedman, R; Witten, Edward; Friedman, Robert; Morgan, John W.; Witten, Edward

    1997-01-01

    This paper gives various methods for constructing vector bundles over elliptic curves and more generally over families of elliptic curves. We construct universal families over generalized elliptic curves via spectral cover methods and also by extensions, and then give a relative version of the construction in families. We give various examples and make Chern class computations.

  7. Quantum principal bundles and corresponding gauge theories

    CERN Document Server

    Durdevic, M

    1995-01-01

    A generalization of classical gauge theory is presented, in the framework of a noncommutative-geometric formalism of quantum principal bundles over smooth manifolds. Quantum counterparts of classical gauge bundles, and classical gauge transformations, are introduced and investigated. A natural differential calculus on quantum gauge bundles is constructed and analyzed. Kinematical and dynamical properties of corresponding gauge theories are discussed.

  8. Strategic and welfare implications of bundling

    DEFF Research Database (Denmark)

    Martin, Stephen

    1999-01-01

    A standard oligopoly model of bundling shows that bundling by a firm with a monopoly over one product has a strategic effect because it changes the substitution relationships between the goods among which consumers choose. Bundling in appropriate proportions is privately profitable, reduces rival......' profits and overall welfare, and may drive rivals from the market...

  9. Logging residue bundling at the roadside in mountain operations

    Energy Technology Data Exchange (ETDEWEB)

    Spinelli, Raffaele; Magagnotti, Natascia (NR IVALSA, Sesto Fiorentino (Italy))

    2009-04-15

    The recovery of logging residue offers many benefits, and can be effectively performed with several techniques. This study presents the results of four distinct trials conducted on the Italian Alps with a truck-mounted bundler, and compares them with the figures obtained by the Austrian and German colleagues on the other side of the Alps. Average productivity varied between 14 and 22 bundles per productive machine hour (PMH), or between 10 and 15 bundles per scheduled machine hour (SMH). Mass production ranged from 3.1 to 4.3 oven-dry tonnes /SMH. If the machine is used for at least 800 SMH/year bundling costs will range between Euro 7 and 9/MWh, which is within the profitability limit indicated in Nordic studies. Although bundling may represent an additional cost in the slash-to-chip chain, it offers the important benefit of efficient outdoor winter storage, which is crucial in a region where the supply and demand of fuel biomass are often diachronic. The truck-mounted bundler used on the Alps adopts the same mechanism and achieves the same performance of the Nordic forwarder-mounted application. The main difference lies in its superior road mobility, which is crucial when covering scattered landings

  10. Development of thermal-hydraulic models for the safety evaluation of CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Young; Hwang, Gi Suk; Jung, Yun Sik [Handong Univ., Pohang (Korea, Republic of); No, Hee Cheon; Moon, Young Min [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2003-03-15

    The objective of the present research is to evaluate the safety analysis for CANDU and to improve the Horizontal Stratification Entrainment Model (HSEM) of RELAP5/MOD3.3. This report includes two items : the one is the development of experimental facility for the safety evaluation of CANDU, the other is the results of comparison with the existing correlations and data. The literature reviews are performed and the database for previous off-take experiments are built. By a survey of state-of-the-articles, the deficiencies of previous works and limitations of existing models are examined. The hydraulic behavior branching through the feeder pipes from the header pipe is analyzed and the test facility of off-take experiment is designed and manufactured as the prototype CANDU6, by a proper scaling methodologies. The test facility contains various branch pipes not only for three directions (top, side and bottom), but for arbitrary directions. The experiments about the onset of entrainment and branch quality only for three directions (top, side and bottom) are carried out by using air-water as working fluids. On the whole, the existing correlations predict the present experimental results well branch quality, entrainment, which validates the availability of experimental facility and methodology. Especially, for the branch quality with top and bottom branches, the different results are shown because of the unstable flow regimes in the horizontal pipe and the different branch diameters. The deficiencies of previous works and limitations of existing models are considered. The off-take experiment for arbitrary branch angles continues as the next year research.

  11. Higher order jet prolongations type gauge natural bundles over vector bundles

    Directory of Open Access Journals (Sweden)

    Jan Kurek

    2004-05-01

    Full Text Available Let $rgeq 3$ and $mgeq 2$ be natural numbers and $E$ be a vector bundle with $m$-dimensional basis. We find all gauge natural bundles ``similar" to the $r$-jet prolongation bundle $J^rE$ of $E$. We also find all gauge natural bundles ``similar" to the vector $r$-tangent bundle $(J^r_{fl}(E,R_0^*$ of $E$.

  12. An analytical model for the prediction of fluid-elastic forces in a rod bundle subjected to axial flow: theory, experimental validation and application to PWR fuel assemblies; Calcul des forces fluidelastiques dans les faisceaux de tubes sous ecoulement axial: theorie, validation, application aux assemblages combustibles des REP

    Energy Technology Data Exchange (ETDEWEB)

    Beaud, F. [Electricite de France (EDF), 78 - Chatou (France)

    1997-12-31

    A model predicting the fluid-elastic forces in a bundle of circular cylinders subjected to axial flow is presented in this paper. Whereas previously published models were limited to circular flow channel, the present one allows to take a rectangular flow external boundary into account. For that purpose, an original approach is derived from the standard method of images. This model will eventually be used to predict the fluid-structure coupling between the flow of primary coolant and a fuel assemblies in PWR nuclear reactors. It is indeed of major importance since the flow is shown to induce quite high damping and could therefore mitigate the incidence of an external load like a seismic excitation on the dynamics of the assemblies. The proposed model is validated on two cases from the literature but still needs further comparisons with the experiments being currently carried out on the EDF set-up. The flow has been shown to induce an approximate 12% damping on a PWR fuel assembly, at nominal reactor conditions. The possible grid effect on the fluid-structure coupling has been neglected so far but will soon be investigated at EDF. (author). 16 refs.

  13. CFD Analysis of Dry Storage System for CANDU Spent Fuel using Fluent 6.2

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Byung Soo; Jeong, Yong Hoon; Chang, Soon Heung Chang [KAIST, Taejon (Korea, Republic of)

    2006-07-01

    To obtain the licensing for MACSTOR/KN-400 developed by KHNP, 3-D CFD analysis was demanded to prove that the maximum temperature was not over the limited temperature (93 .deg. C locally and 66 .deg. C averagely for concrete). Though the thermal-hydraulic prediction by CATHENA-code show the reliable results, that could not the temperature distribution. That is, that could not predict the location of maximum temperature well. In this study, the analysis of the temperature distribution on the natural convection flow with thermal radiation shows the concrete temperature distribution. It was different from the predicted results by CATHENA-code. Therefore, to obtain the licensing for AMCSTOR/KN- 400, CFD analysis should be performed by 3-D CFD code like FLUENT at the same time.

  14. Multipath packet switch using packet bundling

    DEFF Research Database (Denmark)

    Berger, Michael Stubert

    2002-01-01

    The basic concept of packet bundling is to group smaller packets into larger packets based on, e.g., quality of service or destination within the packet switch. This paper presents novel applications of bundling in packet switching. The larger packets created by bundling are utilized to extend...... switching capacity by use of parallel switch planes. During the bundling operation, packets will experience a delay that depends on the actual implementation of the bundling and scheduling scheme. Analytical results for delay bounds and buffer size requirements are presented for a specific scheduling...

  15. Mathematical modelling for nanotube bundle oscillators

    Science.gov (United States)

    Thamwattana, Ngamta; Cox, Barry J.; Hill, James M.

    2009-07-01

    This paper investigates the mechanics of a gigahertz oscillator comprising a nanotube oscillating within the centre of a uniform concentric ring or bundle of nanotubes. The study is also extended to the oscillation of a fullerene inside a nanotube bundle. In particular, certain fullerene-nanotube bundle oscillators are studied, namely C60-carbon nanotube bundle, C60-boron nitride nanotube bundle, B36N36-carbon nanotube bundle and B36N36-boron nitride nanotube bundle. Using the Lennard-Jones potential and the continuum approach, we obtain a relation between the bundle radius and the radii of the nanotubes forming the bundle, as well as the optimum bundle size which gives rise to the maximum oscillatory frequency for both the fullerene and the nanotube bundle oscillators. While previous studies in this area have been undertaken through molecular dynamics simulations, this paper emphasizes the use of applied mathematical modelling techniques which provides considerable insight into the underlying mechanisms. The paper presents a synopsis of the major results derived in detail by the present authors in [1, 2].

  16. Fabrication of a CANFLEX-RU designed bundle for power ramp irradiation test in NRU

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Moon Sung

    2000-11-01

    The BDL-443 CANFLEX-RU bundle AKW was fabricated at Korea Atomic Energy Research Institute (KAERI) for power ramp irradiation testing in NRU reactor. The bundle was fabricated with IDR and ADU fuel pellets in adjacent elements and contains fuel pellets enriched to 1.65 wt% {sup 235}U in the outer and intermediate rings and also contains pellets enriched to 2.00 wt% {sup 235}U in the inner ring. This bundle does not have a center element to allow for insertion on a hanger bar. KAERI produced the IDR pellets with the IDR-source UO{sub 2} powder supplied by BNFL. ADU pellets were fabricated and supplied by AECL. Bundle kits (Zircaloy-4 end plates, end plugs, and sheaths with brazed appendages) manufactured at KAERI earlier in 1996 were used for the fabrication of the bundle. The CANFLEX bundle was fabricated successfully at KAERI according to the QA provisions specified in references and as per relevant KAERI drawings and technical specification. This report covers the fabrication activities performed at KAERI. Fabrication processes performed at AECL will be documented in a separate report.

  17. Development of scaling laws on thermal-hydraulic effect test facility for CANDU-6 moderator

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. Y.; Jung, Y. S.; Kim, N. S. [Handong University, Pohang (Korea, Republic of); Kim, M. W.; Kim, H. J. [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

    2004-07-01

    The scaling laws on thermal-hydraulic effect test facility for CANDU-6 moderator (HGU-KINS) have been investigated and manufactured. The basic laws are the satisfaction of energy conservation and dimensionless number, Ar and Re, for the similarities of thermal-hydraulic properties. And then the thermal-hydraulic scaling analyses of test facilities, SPEL(1/10 scale) and STERN(1/4 scale), have been identified by the present method. As a result, in the case of SPEL, the energy conservation is confirmed, but the similarities of Ar and the heat density are not considered. In the case of STERN, the energy conservation and the characteristics of Ar were well defined. But the similarity of the heat density is unsatisfied, either. Therefore the present method was applied with 1/8 length scale. For the performance test, CFD analysis has been accomplished by CFX5. The results of flow pattern certifications and variation of axial properties with CANDU show that the present scaling method is acceptable.

  18. The application of Plant Reliability Data Information System (PRINS) to CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, S. W.; Lim, Y. H.; Park, H. C. [Korea Hydro and Nuclear Power Co., Ltd., Naah-ri 260, Yangnam-myun, Gyeongju-si, Gyeong Buk (Korea, Republic of)

    2012-07-01

    As risk-informed applications (RIAs) are actively implanted in the nuclear industry, an issue associated with technical adequacy of Probabilistic Safety Assessment (PSA) arises in its modeling and data sourcing. In Korea, PSA for all Korean NPPs has been completed and KHNP(Korea Hydro and Nuclear Power Plant Company) developed the database called the Plant Reliability Data Information System (PRinS). It has several characteristics that distinguish it from other database system such as NPRDs (INPO,1994), PRIS (IAEA), and SRDF (EdF). This database has the function of systematic data management such as automatic data-gathering, periodic data deposition and updating, statistical analysis including Bayesian method, and trend analysis of failure rate or unavailability. In recent PSA for CANDU reactor, the component failure data of EPRI ALWR URD and Component Reliability Database were preferentially used as generic data set. The error factor for most component failure data was estimated by using the information NUREG/CR-4550 and NUREG/CR-4639. Also, annual trend analysis was performed for the functional losses of components using the statistical analysis and chart module of PRinS. Furthermore, the database has been updated regularly and maintained as a living program to reflect the current status. This paper presents the failure data analysis using PRinS which provides Bayesian analysis on main components in the CANDU reactor. (authors)

  19. Ludwig: A Training Simulator of the Safety Operation of a CANDU Reactor

    Directory of Open Access Journals (Sweden)

    Gustavo Boroni

    2011-01-01

    Full Text Available This paper presents the application Ludwig designed to train operators of a CANDU Nuclear Power Plant (NPP by means of a computer control panel that simulates the response of the evolution of the physical variables of the plant under normal transients. The model includes a close set of equations representing the principal components of a CANDU NPP plant, a nodalized primary circuit, core, pressurizer, and steam generators. The design of the application was performed using the object-oriented programming paradigm, incorporating an event-driven process to reflect the action of the human operators and the automatic control system. A comprehensive set of online graphical displays are provided giving an in-depth understanding of transient neutronic and thermal hydraulic response of the power plant. The model was validated against data from a real transient occurring in the Argentine NPP Embalse Río Tercero, showing good agreement. However, it should be stressed that the aim of the simulator is in the training of operators and engineering students.

  20. Safeguards and nonproliferation aspects of a dry fuel recycling technology

    Energy Technology Data Exchange (ETDEWEB)

    Pillay, K.K.S.

    1993-05-01

    Los Alamos National Laboratory undertook an independent assessment of the proliferation potentials and safeguardability of a dry fuel recycling technology, whereby spent pressurized-water reactor (PWR) fuels are used to fuel canadian deuterium uranium (CANDU) reactors. Objectives of this study included (1) the evaluation of presently available technologies that may be useful to safeguard technology options for dry fuel recycling (2) and identification of near-term and long-term research needs to develop process-specific safeguards requirements. The primary conclusion of this assessment is that like all other fuel cycle alternatives proposed in the past, the dry fuel recycle entails prolfferation risks and that there are no absolute technical fixes to eliminate such risks. This study further concludes that the proliferation risks of dry fuel recycling options are relatively minimal and presently known safeguards systems and technologies can be modified and/or adapted to meet the requirements of safeguarding such fuel recycle facilities.

  1. Reflood experiments in rod bundles with flow blockages due to clad ballooning

    Energy Technology Data Exchange (ETDEWEB)

    Moon, S.K.; Kim, J.; Kim, K.; Kim, B.J.; Park, J.K.; Youn, Y.J.; Choi, H.S.; Song, C.H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-07-15

    Clad ballooning and the resulting partial flow blockage are one of the major thermal-hydraulic concerns associated with the coolability of partially blocked cores during a loss-of-coolant accident (LOCA). Several in-pile tests have shown that fuel relocation causes a local power accumulation and a high thermal coupling between the clad and fuel debris in the ballooned regions. However, previous experiments in the 1980s did not take into account the fuel relocation phenomena and resulting local power increase in the ballooned regions. The present paper presents the results of systematic investigations on the coolability of rod bundles with flow blockages. The experiments were mainly performed in 5 x 5 rod bundles, 2 x 2 rod bundles and other test facilities. The experiments include a reflood heat transfer, single-phase convective heat transfer, flow redistributions phenomena, and droplet break-up behavior. The effects of the fuel relocation and resulting local power increase were investigated using a 5 x 5 rod bundle. The fuel relocation phenomena increase the peak cladding temperature.

  2. Macstor dry spent fuel storage system

    Energy Technology Data Exchange (ETDEWEB)

    Pare, F. E. [Atomic Energy of Canada Limited, Montreal (Canada)

    1996-04-15

    AECL, a Canadian Grown Corporation established since 1952, is unique among the world's nuclear organizations. It is both supplier of research reactors and heavy water moderated CANDU power reactors as well as operator of extensive nuclear research facilities. As part of its mandate, AECL has developed products and conceptual designs for the short, intermediate and long term storage and disposal of spent nuclear fuel. AECL has also assumed leadership in the area of dry storage of spent fuel. This Canadian Crown Corporation first started to look into dry storage for the management of its spent nuclear fuel in the early 1970's. After developing silo-like structures called concrete canisters for the storage of its research reactor enriched uranium fuel, AECL went on to perfect that technology for spent CANDU natural uranium fuel. In 1989 AECL teamed up with Trans nuclear, Inc.,(TN), a US based member of the international Trans nuclear Group, to extend its dry storage technology to LWR spent fuel. This association combines AECL's expertise and many years experience in the design of spent fuel storage facilities with TN's proven capabilities of processing, transportation, storage and handling of LWR spent fuel. From the early AECL-designed unventilated concrete canisters to the advanced MACSTOR concept - Modular Air-Cooled Canister Storage - now available also for LWR fuel - dry storage is proving to be safe, economical, practical and, most of all, well accepted by the general public. AECL's experience with different fuels and circumstances has been conclusive.

  3. General frame structures on quantum principal bundles

    CERN Document Server

    Durdevic, M

    1996-01-01

    A noncommutative-geometric generalization of the classical formalism of frame bundles is developed, incorporating into the theory of quantum principal bundles the concept of the Levi-Civita connection. The construction of a natural differential calculus on quantum principal frame bundles is presented, including the construction of the associated differential calculus on the structure group. General torsion operators are defined and analyzed. Illustrative examples are presented.

  4. ACM Bundles on Del Pezzo surfaces

    Directory of Open Access Journals (Sweden)

    Joan Pons-Llopis

    2009-11-01

    Full Text Available ACM rank 1 bundles on del Pezzo surfaces are classified in terms of the rational normal curves that they contain. A complete list of ACM line bundles is provided. Moreover, for any del Pezzo surface X of degree less or equal than six and for any n ≥ 2 we construct a family of dimension ≥ n − 1 of non-isomorphic simple ACM bundles of rank n on X.

  5. Large-scale Flow Pulsation in Tight Square Arrayed Rod Bundles of Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Hwan; Kim, Kyung Min; Cho, Hyung Hee [Yonsei University, Seoul (Korea, Republic of); Shin, Chang Hwan; In, Wang Kee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-05-15

    As a major component of modern nuclear reactor, the nuclear fuel rod bundles with liquid coolant have been studied by a lot of researchers to understand the flow structure between the fuel rods. Recently, rod arrays with much small pitch-to-diameter ratio have been being tried to increase performance of the nuclear reactor. The liquid coolant flowing axially through these small spaces between the rods is known to show some peculiar phenomena including large-scale, quasi-periodic flow pulsation. These flow pulsation phenomena dominate mixing process in the subchannels. Thus, precise understating of the flow structure is essential to predict thermal-hydraulic phenomena in nuclear rod bundles. In this present paper, the turbulent flow in tight square arrayed rod bundles is investigated with Hot-wire anemometry. Then, the measured velocity data are analyzed by using Fast Fourier Transform analysis to find characteristic frequency of the pulsation

  6. Entropy for frame bundle systems and Grassmann bundle systems induced by a diffeomorphism

    Institute of Scientific and Technical Information of China (English)

    SUN; Weniang(孙文祥)

    2002-01-01

    ALiao hyperbolic diffeomorphism has equal measure entropy and topological entropy to that ofits induced systems on frame bundles and Grassmann bundles. This solves a problem Liao posed in 1996 forLiao hyperbolic diffeomorphisms.

  7. Principal $G$-bundles over elliptic curves

    CERN Document Server

    Friedman, R; Witten, Edward; Friedman, Robert; Morgan, John W.; Witten, Edward

    1997-01-01

    Let $G$ be a simple and simply connected complex Lie group. We discuss the moduli space of holomorphic semistable principal $G$-bundles over an elliptic curve $E$. In particular, we give a new proof of a theorem of Looijenga and Bernshtein-Shvartsman, that the moduli space is a weighted projective space. The method of proof is to study the deformations of certain unstable bundles coming from special maximal parabolic subgroups of $G$. We also discuss the associated automorphism sheaves and universal bundles, as well as the relation between various universal bundles and spectral covers.

  8. Statistical Constitutive Equation of Aramid Fiber Bundles

    Institute of Scientific and Technical Information of China (English)

    熊杰; 顾伯洪; 王善元

    2003-01-01

    Tensile impact tests of aramid (Twaron) fiber bundles were carried om under high strain rates with a wide range of 0. 01/s~1000/s by using MTS and bar-bar tensile impact apparatus. Based on the statistical constitutive model of fiber bundles, statistical constitutive equations of aramid fiber bundles are derived from statistical analysis of test data at different strain rates. Comparison between the theoretical predictions and experimental data indicates statistical constitutive equations fit well with the experimental data, and statistical constitutive equations of fiber bundles at different strain rates are valid.

  9. Calculation of Quad-Cities Central Bundle Documented by the U.S. in FY98 Using Russian Computer Codes

    Energy Technology Data Exchange (ETDEWEB)

    Pavlovichev, A.M.

    2001-06-19

    The report presents calculation results of isotopic composition of irradiated fuel performed for the Quad Cities-1 reactor bundle with UO{sub 2} and MOX fuel. The MCU-REA code was used for calculations. The code is developed in Kurchatov Institute, Russia. The MCU-REA results are compared with the experimental data and HELIOS code results.

  10. Jacobi Structures on Affine Bundles

    Institute of Scientific and Technical Information of China (English)

    J. GRABOWSKI; D. IGLESIAS; J. C. MARRERO; E. PADR(O)N; P. URBA(N)SKI

    2007-01-01

    We study affine Jacobi structures (brackets) on an affine bundle π: A→M, i.e. Jacobi brackets that close on affine functions. We prove that if the rank of A is non-zero, there is a one-to- one correspondence between affine Jacobi structures on A and Lie algebroid structures on the vector bundle A+=∪p∈M Aff(Ap, R) of affine functionals. In the case rank A = 0, it is shown that there is a one-to-one correspondence between affins Jacobi structures on A and local Lie algebras on A+. Some examples and applications, also for the linear case, are discussed. For a special type of affine Jacobi structures which are canonically exhibited (strongly-affine or affine-homogeneous Jacobi structures) over a real vector space of finite dimension, we describe the leaves of its characteristic foliation as the orbits of an affine representation. These afline Jacobi structures can be viewed as an analog of the Kostant-Arnold-LiouviUe linear Poisson structure on the dual space of a real finite-dimensional Lie algebra.

  11. Post irradiation test report of irradiated DUPIC simulated fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Myung Seung; Jung, I. H.; Moon, J. S. and others

    2001-12-01

    The post-irradiation examination of irradiated DUPIC (Direct Use of Spent PWR Fuel in CANDU Reactors) simulated fuel in HANARO was performed at IMEF (Irradiated Material Examination Facility) in KAERI during 6 months from October 1999 to March 2000. The objectives of this post-irradiation test are i) the integrity of the capsule to be used for DUPIC fuel, ii) ensuring the irradiation requirements of DUPIC fuel at HANARO, iii) performance verification in-core behavior at HANARO of DUPIC simulated fuel, iv) establishing and improvement the data base for DUPIC fuel performance verification codes, and v) establishing the irradiation procedure in HANARO for DUPIC fuel. The post-irradiation examination performed are {gamma}-scanning, profilometry, density, hardness, observation the microstructure and fission product distribution by optical microscope and electron probe microanalyser (EPMA)

  12. Probabilistic integrity assessment of CANDU pressure tube for the consideration of flaw generation time

    Energy Technology Data Exchange (ETDEWEB)

    Kwak, Sang Log; Kim, Young Jin [Sungkyunkwan Univ., Seoul (Korea, Republic of); Lee, Joon Seong [Kyonggi Univ., Seoul (Korea, Republic of); Park, Youn Won [KINS, Taejon (Korea, Republic of)

    2001-07-01

    This paper describes a Probabilistic Fracture Mechanics (PFM) analysis based on Monte Carlo (MC) simulation. In the analysis of CANDU pressure tube, it is necessary to perform the PFM analyses based on statistical consideration of flaw generation time. A depth and an aspect ratio of initial semi-elliptical surface crack, a fracture toughness value, Delayed Hydride Cracking (DHC) velocity, and flaw generation time are assumed to be probabilistic variables. In all the analyses, degradation of fracture toughness due to neutron irradiation is considered. Also, the failure criteria considered are plastic collapse, unstable fracture and crack penetration. For the crack growth by DHC, the failure probability was evaluated in due consideration of flaw generation time.

  13. The Application of Advancements in Computer Technology to the Control and Safety System of CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chan, P. S. W. [AECL CANDU/Sheridan Park Research Community, Ontario (Canada)

    1992-04-15

    The present spatial control algorithm in CANDU reactors is based on flux synthesis from a set of parti-coloured harmonic flux modes. The design of the Rop system is also based on parti-coloured flux shapes, including both normal and abnormal reactor operating conditions. The dependency of the control and safety systems on parti-coloured data was necessitated by the slow CPU and by the scarcity of Ram which were available to the computer systems in the early seventies. Recent advancements in high speed microprocessors and high capacity Ram chips enable the development of the Pmfp computer code, which calculates reactor power distribution on-line, using diffusion theory and in-core self-powered flux detector readings as internal boundary conditions. The Pmfp based control and safety systems do not depend on parti-coloured flux shapes or preconceived reactor operating conditions.

  14. Fitness for service assessment of degraded CANDU feeder piping-Canadian regulatory expectations

    Energy Technology Data Exchange (ETDEWEB)

    Jin, John C., E-mail: john.jin@cnsc-ccsn.gc.c [Operational Engineering Assessment Div., Canadian Nuclear Safety Commission (Canada); Awad, Raoul [Operational Engineering Assessment Div., Canadian Nuclear Safety Commission (Canada)

    2011-03-15

    Allowance for the continued operation of feeder piping at some Canadian CANDU stations, which is experiencing active degradation mechanisms, has been based primarily on augmented inspection practices and conservative fitness for service assessments. The major degradation mechanisms identified to date are: pipe wall thinning due to Flow Accelerated Corrosion (FAC) and service induced cracking due to Intergranular Cracking due to Stress Corrosion Cracking (SCC) and potentially Low Temperature Creep Cracking (LTCC) mechanisms. Given that currently available industry codes and standards do not provide sufficient guidelines/criteria for assessing the degradation of feeder pipes, the Canadian Nuclear Safety Commission (CNSC) has asked the utilities to establish feeder pipe specific procedures to provide reasonable assurance that the risk associated with the feeder degradation is maintained at an acceptably low level. In response to this requirement, the Canadian CANDU industry has developed and continued to update feeder fitness for service guidelines to provide evaluation procedures and industry standard acceptance criteria for assessing the structural integrity of the feeder pipes. The scope and frequency of inspections are determined based on the results of the fitness for service assessments taking into account the relative susceptibility of feeder pipes to each specific degradation mechanism. While industry practices for the management of degraded feeder pipes have, in general, been complied with the regulatory expectations, outstanding issues still remain. Major regulatory concerns include uncertainties associated with limitations in both the inspection techniques and the mechanistic understanding of the degradation processes, which can impede inspection planning and fitness for service assessments. This paper presents the regulator's view of the current situation with respect to degradation of feeder piping, its implications for nuclear safety and the

  15. Anatomic Double-bundle ACL Reconstruction

    NARCIS (Netherlands)

    V.M. Schreiber; C.F. van Eck; F.H. Fu

    2010-01-01

    Rupture of the anterior cruciate ligament (ACL) is one of the most frequent forms of knee trauma. The traditional surgical treatment for ACL rupture is single-bundle reconstruction. However, during the past few years there has been a shift in interest toward double-bundle reconstruction to closely r

  16. The Verlinde formula for Higgs bundles

    CERN Document Server

    Andersen, Jørgen Ellegaard; Pei, Du

    2016-01-01

    We propose and prove the Verlinde formula for the quantization of the Higgs bundle moduli spaces and stacks for any simple and simply-connected group. This generalizes the equivariant Verlinde formula for the case of $SU(n)$ proposed previously by the second and third author. We further establish a Verlinde formula for the quantization of parabolic Higgs bundle moduli spaces and stacks.

  17. Principal Bundles on the Projective Line

    Indian Academy of Sciences (India)

    V B Mehta; S Subramanian

    2002-08-01

    We classify principal -bundles on the projective line over an arbitrary field of characteristic ≠ 2 or 3, where is a reductive group. If such a bundle is trivial at a -rational point, then the structure group can be reduced to a maximal torus.

  18. Line bundle embeddings for heterotic theories

    CERN Document Server

    Nibbelink, Stefan Groot

    2016-01-01

    In heterotic theories consistency requires the introduction of a non-trivial vector bundle. This bundle breaks the original ten-dimensional gauge groups E_8 x E_8 or SO(32) for the supersymmetric heterotic theories and SO(16) x SO(16) for the non-supersymmetric tachyon-free theory to smaller subgroups. A vast number of MSSM-like models have been constructed up to now, most of which describe the vector bundle as a sum of line bundles. However, there are several different ways of describing these line bundles and their embedding in the ten-dimensional gauge group. We recall and extend these different descriptions and explain how they can be translated into each other.

  19. Line bundle embeddings for heterotic theories

    Science.gov (United States)

    Nibbelin, Stefan Groot; Ruehle, Fabian

    2016-04-01

    In heterotic string theories consistency requires the introduction of a non-trivial vector bundle. This bundle breaks the original ten-dimensional gauge groups E8 × E8 or SO(32) for the supersymmetric heterotic string theories and SO(16) × SO(16) for the non-supersymmetric tachyon-free theory to smaller subgroups. A vast number of MSSM-like models have been constructed up to now, most of which describe the vector bundle as a sum of line bundles. However, there are several different ways of describing these line bundles and their embedding in the ten-dimensional gauge group. We recall and extend these different descriptions and explain how they can be translated into each other.

  20. Line bundle embeddings for heterotic theories

    Energy Technology Data Exchange (ETDEWEB)

    Groot Nibbelink, Stefan [Muenchen Univ. (Germany). Arnold Sommerfeld Center for Theoretical Physics; Ruehle, Fabian [Deutsches Elektronen-Synchrotron (DESY), Hamburg (Germany)

    2016-03-15

    In heterotic string theories consistency requires the introduction of a non-trivial vector bundle. This bundle breaks the original ten-dimensional gauge groups E{sub 8} x E{sub 8} or SO(32) for the supersymmetric heterotic string theories and SO(16) x SO(16) for the non-supersymmetric tachyon-free theory to smaller subgroups. A vast number of MSSM-like models have been constructed up to now, most of which describe the vector bundle as a sum of line bundles. However, there are several different ways of describing these line bundles and their embedding in the ten-dimensional gauge group. We recall and extend these different descriptions and explain how they can be translated into each other.

  1. Composite spinor bundles in gravitation theory

    CERN Document Server

    Sardanashvily, G

    1995-01-01

    In gravitation theory, the realistic fermion matter is described by spinor bundles associated with the cotangent bundle of a world manifold X. In this case, the Dirac operator can be introduced. There is the 1:1 correspondence between these spinor bundles and the tetrad gravitational fields represented by sections of the quotient \\Si of the linear frame bundle over X by the Lorentz group. The key point lies in the fact that different tetrad fields imply nonequivalent representations of cotangent vectors to X by the Dirac's matrices. It follows that a fermion field must be regarded only in a pair with a certain tetrad field. These pairs can be represented by sections of the composite spinor bundle S\\to\\Si\\to X where values of tetrad fields play the role of parameter coordinates, besides the familiar world coordinates.

  2. Requirements for disordered actomyosin bundle contractility

    CERN Document Server

    Lenz, Martin

    2011-01-01

    Actomyosin contractility is essential for biological force generation, and is well understood in highly ordered structures such as striated muscle. In vitro experiments have shown that non-sarcomeric bundles comprised only of F-actin and myosin thick filaments can also display contractile behavior, which cannot be described by standard muscle models. Here we investigate the microscopic symmetries underlying this process in large non-sarcomeric bundles with long actin filaments. We prove that contractile behavior requires non-identical motors that generate large enough forces to probe the nonlinear elastic behavior of F-actin. A simple disordered bundle model demonstrates a contraction mechanism based on these assumptions and predicts realistic bundle deformations. Recent experimental observations of F-actin buckling in in vitro contractile bundles support our model.

  3. Double Fell bundles and Spectral triples

    CERN Document Server

    Martins, Rachel A D

    2007-01-01

    As a natural and canonical extension of Kumjian's Fell bundles over groupoids \\cite{fbg}, we give a definition for a double Fell bundle (a double category) over a double groupoid. We show that finite dimensional double category Fell line bundles tensored with their dual with $S^o$-reality satisfy the finite real spectral triples axioms but not necessarily orientability. This means that these product bundles with noncommutative algebras can be regarded as noncommutative compact manifolds more general than real spectral triples as they are not necessarily orientable. By construction, they unify the noncommutative geometry axioms and hence provide an algebraic enveloping structure for finite spectral triples to give the Dirac operator $D$ new algebraic and geometric structures that are otherwise missing in the transition from Fredholm operator to Dirac operator. The Dirac operator in physical applications as a result becomes less ad hoc. The new noncommutative space we present is a complex line bundle over a dou...

  4. On Harder–Narasimhan Reductions for Higgs Principal Bundles

    Indian Academy of Sciences (India)

    Arijit Dey; R Parthasarathi

    2005-05-01

    The existence and uniqueness of – reduction for the Higgs principal bundles over nonsingular projective variety is shown. We also extend the notion of – reduction for (, )-bundles and ramified -bundles over a smooth curve.

  5. Functional bundles of the medial patellofemoral ligament.

    Science.gov (United States)

    Kang, Hui Jun; Wang, Fei; Chen, Bai Cheng; Su, Yan Ling; Zhang, Zhan Chi; Yan, Chang Bao

    2010-11-01

    The purpose of this study was to explore the anatomy and evaluate the function of the medial patellofemoral ligament (MPFL). Anatomical dissection was performed on 12 fresh-frozen knee specimens. The MPFL is a condensation of capsular fibers, which originates at the medial femoral condyle. It runs transversely and inserts to the medial edge of the patella. With the landmark of the medial femur epicondyle (MFE), the femoral origination was located: just 8.90 ± 3.27 mm proximally and 13.47 ± 3.68 mm posteriorly to the MFE. The most interesting finding in present study was functional bundles of its patellar insertion. Approximately from the femoral origination point, fibers of the MPFL form two relatively concentrated fiber bundles: the inferior-straight bundle and the superior-oblique bundle. The whole length of each was 71.78 ± 5.51 and 73.67 ± 5.40 mm, respectively. The included angle between bundles was 15.1° ± 2.1°. Although the superior-oblique bundle and the inferior-straight bundle run on the patellar MPFL inferiorly and superiorly, respectively, as their name indicates, the two bundles are not entirely separated, which make MPFL one intact structure. The inferior-straight bundle is the main static soft tissue restraints where the superior-oblique bundle associated with vastus medialis obliquus (VMO) is to serve as the main dynamic soft tissue restraints. So this finding may provide the theoretical foundation for the anatomical reconstruction of the MPFL and shed lights on the future researchers.

  6. Measurement of Velocity and Temperature Profiles in the 1/40 Scaled-Down CANDU-6 Moderator Tank

    Directory of Open Access Journals (Sweden)

    Hyoung Tae Kim

    2015-01-01

    Full Text Available In order to simulate the CANDU-6 moderator circulation phenomena during steady state operating and accident conditions, a scaled-down moderator test facility has been constructed at Korea Atomic Energy Institute (KAERI. In the present work an experiment using a 1/40 scaled-down moderator tank has been performed to identify the potential problems of the flow visualization and measurement in the scaled-down moderator test facility. With a transparent moderator tank model, a flow field is visualized with a particle image velocimetry (PIV technique under an isothermal state, and the temperature field is measured using a laser induced fluorescence (LIF technique. A preliminary CFD analysis is also performed to find out the flow, thermal, and heating boundary conditions with which the various flow patterns expected in the prototype CANDU-6 moderator tank can be reproduced in the experiment.

  7. Hydrodynamic Experiments for a Flow Distribution of a 61-pin Wire-wrapped Rod Bundle

    Energy Technology Data Exchange (ETDEWEB)

    Chang, S. K.; Euh, D. J.; Choi, H. S.; Kim, H. M.; Ko, Y. J.; Lee, D. W.; Lee, H. Y.; Choi, S. R. [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    Fuel assembly of the SFR (Sodium-cooled Fast breeder Reactor) type reactor generally has wire spacers which are wrapped around each fuel pin helically in axial direction. The configuration of a helical wire spacer guarantees the fuel rods integrity by providing the bundle rigidity, proper spacing between rods and promoting coolant mixing between subchannels. It is important to understand the flow characteristics in such a triangular array wire wrapped rod bundle in a hexagonal duct. The experimental work has been undertaken to quantify the friction and mixing parameters which characterize the flow distribution in subchannels for the KAERI's own bundle geometric configuration. This work presents the hydrodynamic experimental results for the flow distribution and the pressure drop in subchannels of a 61-pin wire wrapped rod bundle which has been fabricated considering the hydraulic similarity of the reference reactor. Hydrodynamic experiments for a 61-pin wire wrapped test assembly has been performed to provide the data of a flow distribution and pressure losses in subchannels for verifying the analysis capability of subchannel analysis codes for a KAERI's own prototype SFR reactor. Three type of sampling probes have been specially designed to conserve the shape of the flow area for each type of subchannels. All 126 subchannels have been measured to identify the characteristics of the flow distribution in a 37-pin rod assembly. Pressure drops at the interior and the edge subchannels have been also measured to recognize the friction losses of each type of subchannels.

  8. Higgs bundles and the real symplectic group

    CERN Document Server

    Gothen, Peter B

    2011-01-01

    We give an overview of the work of Corlette, Donaldson, Hitchin and Simpson leading to the non-abelian Hodge theory correspondence between representations of the fundamental group of a surface and the moduli space of Higgs bundles. We then explain how this can be generalized to a correspondence between character varieties for representations of surface groups in real Lie groups G and the moduli space of G-Higgs bundles. Finally we survey recent joint work with Bradlow, Garc\\'ia-Prada and Mundet i Riera on the moduli space of maximal Sp(2n,R)-Higgs bundles.

  9. Isothermal flow measurement using planar PIV in the 1/4 scaled model of CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Im, Sunghyuk; Sung, Hyung Jin [KAIST, Daejeon (Korea, Republic of); Seo, Han; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of); Kim, Hyoung Tae [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    The local temperature of the moderator is a key parameter in determining the available subcooling. To predict the flow field and local temperature distribution in the calandria, Korea Atomic Energy Research Institute (KAERI) started the experimental research on moderator circulation as one of a national R and D research programs from 2012. This research program includes the construction of the Moderator Circulation Test (MCT) facility, production of the validation data for self-reliant CFD tools, and development of optical measurement system using the Particle Image Velocimetry (PIV) and Laser Induced Fluorescence (LIF) techniques. Small-scale 1/40 and 1/8 small-scale model tests were performed prior to installation of the main MCT facility to identify the potential problems of the flow visualization and measurement expected in the 1/4 scale MCT facility. In the 1/40 scale test, a flow field was measured with a PIV measurement technique under an iso-thermal state, and the temperature field was visualized using a LIF technique. In this experiment, the key point was to illuminate the region of interest as uniformly as possible since the velocity and temperature fields in the shadow regions were distorted and unphysical. In the 1/8 scale test, the flow patterns from the inlet nozzles to the top region of the tank were investigated using PIV measurement at two different positions of the inlet nozzle. For each position of laser beam exposure the measurement sections were divided to 7 groups to overcome the limitation of the laser power to cover the relatively large test section. The MCT facility is the large-scale facility designed to reproduce the important characteristics of moderator circulation in a CANDU6 calandria under a range of operating conditions. It is reduced in a 1/4 scale and a moderator test vessel is built to the specifications of the CANDU6 reactor design, where a working fluid is sub-cooled water with atmospheric pressure. Previous studies were

  10. Effect of Testing Conditions on Fibre-Bundle Tensile Properties Part Ⅰ: Sample Preparation, Bundle Mass and Fibre Alignment of Wool Bundles

    Institute of Scientific and Technical Information of China (English)

    YU Wei-dong; YAN Hao-jing; Ron Postle; Yang Shouren

    2002-01-01

    Due to the effects of samples and testing conditions on fibre-bundle tensile behaviour, it is necessary to investigate the relationships between experimental factors and tensile properties for the fibre-bumdle tensile tester (TENSOR). The effects of bundle sample preparation, fibre bundle mass and fibre alignment have been tested. The experimental results indicated that (1) the low damage in combing and no free-end fibres in the cut bundle are most important for the sample preparation; (2) the reasonable bundle mass is 400- 700tex, but the tensile properties measured should bemodified with the bundle mass because a small amount of bundle mass causes the scatter results, while the larger is the bundle mass, the more difficult to comb fibres parallel and to clamp fibre evenly; and (3) the fibre irregular arrangement forms a slack bundle resulting in interaction between fibres, which will affect the reproducibility and accuracy of the tensile testing.

  11. Liquid Flow in Shaped Fiber Bundle

    Institute of Scientific and Technical Information of China (English)

    ZHANG Yan; WANG Hua-ping; CHEN Yue-hua

    2006-01-01

    By computation and comparison of the critical spreading coefficient parameter, it was found that shaped fiber bundle is better for wetting. Liquid-air interface tension of liquid arising the shaped fiber bundle body is considered as one critical factor besides liquid viscosity, inertia force and liquid-fiber interface tension. Experimental result simulation demonstrated that the liquid-air interface tension is correlated with the geometric size of the liquid arising in body, φ0 (x) and which is affected by the cross sectional shape of fiber and the radius of single fiber. The shaped fiber bundle model is introduced to investigate liquid flow in fiber assembly. The model is generated based on a random function for stochastic forming of fibers in bundle and it is necessary to combine this fundamental model with physical explanation for investigation of liquid flow in fiber assembly.

  12. Bundled Hybrid Offset Riser Global Strength Analysis

    Institute of Scientific and Technical Information of China (English)

    William C.Webster; Zhuang Kang; Wenzhou Liang; Youwei Kang; Liping Sun

    2011-01-01

    Bundled hybrid offset riser(BHOR)global strength analysis,which is more complex than single line offset riser global strength analysis,was carried out in this paper.At first,the equivalent theory is used to deal with BHOR,and then its global strength in manifold cases was analyzed,along with the use of a three-dimensional nonlinear time domain finite element program.So the max bending stress,max circumferential stress,and max axial stress in the BHOR bundle main section(BMS)were obtained,and the values of these three stresses in each riser were obtained through the "stress distribution method".Finally,the Max Von Mises stress in each riser was given and a check was made whether or not they met the demand.This paper provides a reference for strength analysis of the bundled hybrid offset riser and some other bundled pipelines.

  13. Noncommutative principal bundles through twist deformation

    CERN Document Server

    Aschieri, Paolo; Pagani, Chiara; Schenkel, Alexander

    2016-01-01

    We construct noncommutative principal bundles deforming principal bundles with a Drinfeld twist (2-cocycle). If the twist is associated with the structure group then we have a deformation of the fibers. If the twist is associated with the automorphism group of the principal bundle, then we obtain noncommutative deformations of the base space as well. Combining the two twist deformations we obtain noncommutative principal bundles with both noncommutative fibers and base space. More in general, the natural isomorphisms proving the equivalence of a closed monoidal category of modules and its twist related one are used to obtain new Hopf-Galois extensions as twists of Hopf-Galois extensions. A sheaf approach is also considered, and examples presented.

  14. Mobility of Taxol in Microtubule Bundles

    Science.gov (United States)

    Ross, J.

    2003-06-01

    Mobility of taxol inside microtubules was investigated using fluorescence recovery after photobleaching (FRAP) on flow-aligned bundles. Bundles were made of microtubules with either GMPCPP or GTP at the exchangeable site on the tubulin dimer. Recovery times were sensitive to bundle thickness and packing, indicating that taxol molecules are able to move laterally through the bundle. The density of open binding sites along a microtubule was varied by controlling the concentration of taxol in solution for GMPCPP samples. With > 63% sites occupied, recovery times were independent of taxol concentration and, therefore, inversely proportional to the microscopic dissociation rate, k_{off}. It was found that 10*k_{off} (GMPCPP) ~ k_{off} (GTP), consistent with, but not fully accounting for, the difference in equilibrium constants for taxol on GMPCPP and GTP microtubules. With taxol along the microtubule interior is hindered by rebinding events when open sites are within ~7 nm of each other.

  15. Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Todreas, N.E.; Cheng, S.K.; Basehore, K.

    1984-08-01

    This project principally undertook the investigation of the thermal hydraulic performance of wire wrapped fuel bundles of LMFBR configuration. Results obtained included phenomenological models for friction factors, flow split and mixing characteristics; correlations for predicting these characteristics suitable for insertion in design codes; numerical codes for analyzing bundle behavior both of the lumped subchannel and distributed parameter categories and experimental techniques for pressure velocity, flow split, salt conductivity and temperature measurement in water cooled mockups of bundles and subchannels. Flow regimes investigated included laminar, transition and turbulent flow under forced convection and mixed convection conditions. Forced convections conditions were emphasized. Continuing efforts are underway at MIT to complete the investigation of the mixed convection regime initiated here. A number of investigations on outlet plenum behavior were also made. The reports of these investigations are identified.

  16. Supporting the Secure Deployment of OSGi Bundles

    OpenAIRE

    Parrend, Pierre; Frénot, Stéphane

    2007-01-01

    International audience; The OSGi platform is a lightweight management layer over a Java virtual machine that makes runtime extensi- bility and multi-application support possible in mobile and constraint environments. This powerfull capability opens a particular attack vector against mobile platforms: the in- stallation of malicious OSGi bundles. The first countermea- sure is the digital signature of the bundles. We developed a tool suite that supports the signature, the publication and the va...

  17. A Geometric Approach to Noncommutative Principal Bundles

    CERN Document Server

    Wagner, Stefan

    2011-01-01

    From a geometrical point of view it is, so far, not sufficiently well understood what should be a "noncommutative principal bundle". Still, there is a well-developed abstract algebraic approach using the theory of Hopf algebras. An important handicap of this approach is the ignorance of topological and geometrical aspects. The aim of this thesis is to develop a geometrically oriented approach to the noncommutative geometry of principal bundles based on dynamical systems and the representation theory of the corresponding transformation group.

  18. Is It Complete Left Bundle Branch Block? Just Ablate the Right Bundle.

    Science.gov (United States)

    Ali, Hussam; Lupo, Pierpaolo; Foresti, Sara; De Ambroggi, Guido; Epicoco, Gianluca; Fundaliotis, Angelica; Cappato, Riccardo

    2017-03-01

    Complete left bundle branch block (LBBB) is established according to standard electrocardiographic criteria. However, functional LBBB may be rate-dependent or can perpetuate during tachycardia due to repetitive concealed retrograde penetration of impulses through the contralateral bundle "linking phenomenon." In this brief article, we present two patients with basal complete LBBB in whom ablating the right bundle unmasked the actual antegrade conduction capabilities of the left bundle. These cases highlight intriguing overlap between electrophysiological concepts of complete block, linking, extremely slow, and concealed conduction.

  19. Spent fuel characteristics & disposal considerations

    Energy Technology Data Exchange (ETDEWEB)

    Oversby, V.M.

    1996-06-01

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

  20. External cost assessment for nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Park, Byung Heung [Korea National University of Transportation, Chungju (Korea, Republic of); Ko, Won Il [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-12-15

    Nuclear power is currently the second largest power supply method in Korea and the number of nuclear power plants are planned to be increased as well. However, clear management policy for spent fuels generated from nuclear power plants has not yet been established. The back-end fuel cycle, associated with nuclear material flow after nuclear reactors is a collection of technologies designed for the spent fuel management and the spent fuel management policy is closely related with the selection of a nuclear fuel cycle. Cost is an important consideration in selection of a nuclear fuel cycle and should be determined by adding external cost to private cost. Unlike the private cost, which is a direct cost, studies on the external cost are focused on nuclear reactors and not at the nuclear fuel cycle. In this research, external cost indicators applicable to nuclear fuel cycle were derived and quantified. OT (once through), DUPIC (Direct Use of PWR SF in CANDU), PWR-MOX (PWR PUREX reprocessing), and Pyro-SFR (SFR recycling with pyroprocessing) were selected as nuclear fuel cycles which could be considered for estimating external cost in Korea. Energy supply security cost, accident risk cost, and acceptance cost were defined as external cost according to precedent and estimated after analyzing approaches which have been adopted for estimating external costs on nuclear power generation.

  1. Twisted Bundle on Noncommutative Space and U(1) Instanton

    CERN Document Server

    Ho, P M

    2000-01-01

    We study the notion of twisted bundles on noncommutative space. Due to theexistence of projective operators in the algebra of functions on thenoncommutative space, there are twisted bundles with non-constant dimension.The U(1) instanton solution of Nekrasov and Schwarz is such an example. As amathematical motivation for not excluding such bundles, we find gaugetransformations by which a bundle with constant dimension can be equivalent toa bundle with non-constant dimension.

  2. Simulation of In-Core Dose Rates for an Offline CANDU Reactor

    Science.gov (United States)

    Gilbert, Jordan

    This thesis describes the development of a Monte Carlo simulation to predict the dose rates that will be encountered by a novel robotic inspection system for the pressure tubes of an offline CANDU reactor. Simulations were performed using the Monte Carlo N-Particle (MCNP) radiation transport code, version 6.1. The radiation fields within the reactor, even when shut down, are very high, and can cause significant damage to certain structural components and the electronics of the inspection system. Given that the robotic system will rely heavily on electronics, it is important to know the dose rates that will be encountered, in order to estimate the component lifetimes. The MCNP simulation was developed and benchmarked against information obtained from Ontario Power Generation and the Canadian Nuclear Laboratories. The benchmarking showed a good match between the simulated values and the expected values. This simulation, coupled with the accompanying user interface, represent a tool in dose field prediction that is currently unavailable. Predicted dose rates for a postulated inspection at 7 days after shutdown, with 2:5 cm of tungsten shielding around the key components, would survive for approximately 7 hours in core. This is anticipated to be enough time to perform an inspection and shows that the use of this tool can aid in designing the new inspection system.

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

    Energy Technology Data Exchange (ETDEWEB)

    Choi, In Kil; Choun, Young Sun; Seo, Jeong Moon; Ahn, Seong Moon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Choi, In Kil; Ahn, Seong Moon; Choun, Young Sun; Seo, Jeong Moon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2004-07-01

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

  5. The histology of retinal nerve fiber layer bundles and bundle defects.

    Science.gov (United States)

    Radius, R L; Anderson, D R

    1979-05-01

    The fiber bundle striations recognized clinically in normal monkey eyes appear to be bundles of axons compartmentalized within glial tunnels formed by Müller's-cell processes, when viewed histologically. The dark boundaries that separate individual bundles are the broadened foot endings of these cells near the inner surface of the retina. Within one week after focal retinal photocoagulation, characteristic fundus changes could be seen in experimental eyes. In histologic sections of the involved retina, there was marked cystic degeneration of the retinal nerve fiber layer. Within one month, atrophy of distal axon segments was complete. With the drop-out of damaged axons and thinning of individual fiber bundles, retinal striations became less prominent. The resulting fundus picture in these experimental eyes is similar to fiber bundle defects that can be seen clinically in various neuro-ophthalmic disorders.

  6. Leeds-Lyon Symposium on Tribology ’The Tribology Design of Machine Elements’ (15th) 6-9 September 1988

    Science.gov (United States)

    1988-08-01

    In the Candu nuclear reactor, fuel bundles are moved into the channel by means of a ram connected to a ball screw. The screw is driven by a...lubricating film. Predominant mode of lubrication was probably mixed lubrication leading to high temperatures at the interface. Visual inspection of the...Velocity m/s Fig.4 ? -- SESSION XII (Paper (iv) - THURSDAY 8th SEPTEMBER WEAR PERFORMANCE OF MATERIALS FOR BALL SCREW AND SPLINE APPLICATIONS IN CANDU

  7. Comparison Study on Thermal-Hydraulic Analysis Depending on Liquid Relief Valve Response for an Station Blackout in CANDU-6

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. M.; Kho, D. W. [KHNP-CRI, Daejeon (Korea, Republic of); Choi, S. H.; Moon, B. J.; Kim, S. R. [Nuclear Engineering Service and Solution Co., Daejeon (Korea, Republic of)

    2014-10-15

    The purpose of this analysis is to compare the results of thermal-hydraulic analysis depending on liquid relief valve response during a station black out (SBO) events in CANDU-6. The primary heat transport system (PHTS) behavior following the postulated SBO is analyzed using CATHENA code. In the paper, analysis was performed to evaluate the effect on coolant system where LRVs are assumed to be opened or opened according to normal open characteristics in the condition of SBO. The result showed that the primary pressure boundary is extended from LRV to DCT and the effects on primary system behavior were neglectable.

  8. Hitachi's proposed DCS solution for new build CANDU EC6 using the G-HIACS unified platform

    Energy Technology Data Exchange (ETDEWEB)

    Tan, D.; Ishii, K.; Otsuka, Y.; Uemura, K., E-mail: daisuke.tan.ye@hitachi.com [Hitachi Ltd., Infrastructure Systems Co., Ibaraki (Japan); Marko, P.E. [Hitachi Power Systems Canada Ltd., Power and Industry Div., Ontario (Canada)

    2013-07-01

    Hitachi Ltd. has developed the safe and secure functional safety DCS controller for potential new build NPP projects in the global market. Hitachi has improved the availability, maintainability, and reliability for its latest DCS systems named G-HIACS. In this latest paper on its DCS product development program, Hitachi would like to report a proposed DCS solution for new build CANDU NSP and BOP based on the G-HIACS Unified Architecture (R800FS/HSC800FS vSAFE Functional Safety Controller and R900/HSC900 General Purpose Controller) hybrid control system. (author)

  9. Tangent bundle formulation of a charged gas

    CERN Document Server

    Sarbach, Olivier

    2013-01-01

    We discuss the relativistic kinetic theory for a simple, collisionless, charged gas propagating on an arbitrary curved spacetime geometry. Our general relativistic treatment is formulated on the tangent bundle of the spacetime manifold and takes advantage of its rich geometric structure. In particular, we point out the existence of a natural metric on the tangent bundle and illustrate its role for the development of the relativistic kinetic theory. This metric, combined with the electromagnetic field of the spacetime, yields an appropriate symplectic form on the tangent bundle. The Liouville vector field arises as the Hamiltonian vector field of a natural Hamiltonian. The latter also defines natural energy surfaces, called mass shells, which turn out to be smooth Lorentzian submanifolds. A simple, collisionless, charged gas is described by a distribution function which is defined on the mass shell and satisfies the Liouville equation. Suitable fibre integrals of the distribution function define observable fie...

  10. Twistor bundle theory and its application

    Institute of Scientific and Technical Information of China (English)

    2004-01-01

    Over an oriented even dimensional Riemannian manifold (M2m, ds2), in terms of the Levi-Civita connection form Ω and the canonical form Θ on the bundle of positive or→ J+(M, ds2) → M. The integrability on an almost complex structure J compatible with the metric and the orientation, is shown to be equivalent to the fact that the corresponding cross section of the twistor bundle is holomorphic with respect to J and the canonical almost complex structure J1 on J+(M, ds2), by using moving frame theory. Moreover, for various metrics and a fixed orientation on M, a canonical bundle isomorphism is established. As a consequence, we generalize a celebrated theorem of LeBrun.

  11. Abelian conformal field theory and determinant bundles

    DEFF Research Database (Denmark)

    Andersen, Jørgen Ellegaard; Ueno, K.

    2007-01-01

    Following [10], we study a so-called bc-ghost system of zero conformal dimension from the viewpoint of [14, 16]. We show that the ghost vacua construction results in holomorphic line bundles with connections over holomorphic families of curves. We prove that the curvature of these connections...... are up to a scale the same as the curvature of the connections constructed in [14, 16]. We study the sewing construction for nodal curves and its explicit relation to the constructed connections. Finally we construct preferred holomorphic sections of these line bundles and analyze their behaviour near...

  12. Classical Higgs fields on gauge gluon bundles

    Directory of Open Access Journals (Sweden)

    Palese Marcella

    2016-01-01

    Full Text Available Classical Higgs fields and related canonical conserved quantities are defined by invariant variational problems on suitably defined gauge gluon bundles. We consider Lagrangian field theories which are assumed to be invariant with respect to the action of a gauge-natural group. As an illustrative example we exploit the ‘gluon Lagrangian’, i.e. a Yang-Mills Lagrangian on the (1, 1-order gauge-natural bundle of SU(3-principal connections. The kernel of the gauge-natural Jacobi morphism for such a Lagrangian, by inducing a reductive split structure, canonically defines a ‘gluon classical Higgs field’.

  13. Assessment of System Behavior and Actions Under Loss of Electric Power For CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Kang, San Ha; Moon, Bok Ja; Kim, Seoung Rae [Nuclear Engineering Service and Solution Co., Ltd., Daejeon (Korea, Republic of)

    2014-05-15

    For the analysis, the CANDU-6 plant in Korea is considered and only the passive components are operable. The other systems are assumed to be at failed condition due to the loss of electric power. At this accident, only the inventories remained in the primary heat transport system (PHTS) and steam generator can be used for the decay heat removal. Due to the transfer of decay heat, the inventory of steam generator secondary side is discharged to the air through passive operation of main steam safety valves (MSSVs). After the steam generators are dried, the PHTS is over-pressurized and the coolant is discharged to fuelling machine vault through passive operation of degasser condenser tank relief valves (DCRVs). Under this situation, the maintenance of the integrity of PHTS is important for the protection of radionuclides release to the environment. Thus, deterministic analysis using CATHENA code is carried out for the simulation of the accident and the appropriate operator action is considered. The loss of electric power results in the depletion of steam generator inventory which is necessary for the decay heat removal. If only the passive system is credited, the PT can be failed after the steam generator is depleted. For the prevention of the PT failure, the feedwater should be supplied to the steam generator before 4,800s after the accident. The feedwater can be supplied using water in dousing tank if the steam generators are depressurized. The decay heat from the core is removed through natural circulation if the feedwater can be supplied continuously.

  14. Characteristics of U-tube assembly design for CANDU 6 type steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jun Su; Jeong, Seung Ha [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-06-01

    Since the first operation of nuclear steam generator early 1960s, its performance requirements have been met but the steam generator problems have been met but the steam generator problems have been major cause of reducing the operational reliability, plant safety and availability. U-tube assembly of steam generator forms the primary system pressure boundary of the plant and have experienced several types of tube degradation problems. Tube failure and leakage resulting from the degradation will cause radioactive contamination of secondary system by the primary coolant, and this may lead to unplanned plant outages and costly repair operations such as tube plugging or steam generator replacement. For the case of steam generators for heavy water reactors, e.g. Wolsong 2, 3, and 4 NPP, a high cost of heavy water will be imposed additionally. During the plant operation, steam generator tubes can potentially be subject to adverse environmental conditions which will cause damages to U-tube assembly. Types of the damage depend upon the combined effects of design factors, materials and chemical environment of steam generator, and they are the pure water stress corrosion cracking, intergranular attack, pitting, wastage, denting, fretting and fatigue, etc. In this report, a comprehensive review of major design factors of recirculating steam generators has been performed against the potential tube damages. Then the design characteristics of CANDU-type Wolsong steam generator were investigated in detail, including tube material, thermalhydraulic aspects, tube-to-tubesheet joint, tube supports, water chemistry and sludge management. 9 tabs., 18 figs., 38 refs. (Author) .new.

  15. Lazarsfeld-Mukai bundles and applications

    CERN Document Server

    Aprodu, Marian

    2012-01-01

    We survey the development of the notion of Lazarsfeld-Mukai bundles together with various applications, from the classification of Mukai manifolds to Brill-Noether theory and syzygies of $K3$ sections. To see these techniques at work, we present a short proof of a result of M. Reid on the existence of elliptic pencils.

  16. The Hodge bundle on Hurwitz spaces

    NARCIS (Netherlands)

    van der Geer, G.; Kouvidakis, A.

    2011-01-01

    In 2009 Kokotov, Korotkin and Zograf gave in [7] a formula for the class of the Hodge bundle on the Hurwitz space of admissible covers of genus g and degree d of the projective line. They gave an analytic proof of it. In this note we give an algebraic proof and an extension of the result.

  17. η-Invariant and Flat Vector Bundles

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    We present an alternate definition of the mod Z component of the AtiyahPatodi-Singer η invariant associated to (not necessary unitary) fiat vector bundles, which identifies explicitly its real and imaginary parts. This is done by combining a deformation of flat connections introduced in a previous paper with the analytic continuation procedure appearing in the original article of Atiyah, Parodi and Singer.

  18. Meromorphic Higgs bundles And Related Geometries

    OpenAIRE

    Dalakov, Peter

    2016-01-01

    The present note is mostly a survey on the generalised Hitchin integrable system and moduli spaces of meromorphic Higgs bundles. We also fill minor gaps in the existing literature, outline a calculation of the infinitesimal period map and review briefly some related geometries.

  19. Meromorphic Higgs bundles and related geometries

    Science.gov (United States)

    Dalakov, Peter

    2016-11-01

    The present note is mostly a survey on the generalised Hitchin integrable system and moduli spaces of meromorphic G-Higgs bundles. We also fill minor gaps in the existing literature, outline a calculation of the infinitesimal period map and review some related geometries.

  20. Computations in intersection rings of flag bundles

    CERN Document Server

    Grayson, Daniel R; Stillman, Michael E

    2012-01-01

    Intersection rings of flag varieties and of isotropic flag varieties are generated by Chern classes of the tautological bundles modulo the relations coming from multiplicativity of total Chern classes. In this paper we describe the Groebner bases of the ideals of relations and give applications to computation of intersections, as implemented in Macaulay2.

  1. Capacity efficiency of recovery request bundling

    DEFF Research Database (Denmark)

    Ruepp, Sarah Renée; Dittmann, Lars; Berger, Michael Stübert

    2010-01-01

    This paper presents a comparison of recovery methods in terms of capacity efficiency. In particular, a method where recovery requests are bundled towards the destination (Shortcut Span Protection) is evaluated against traditional recovery methods. Our simulation results show that Shortcut Span...

  2. Active Hair-Bundle Motility by the Vertebrate Hair Cell

    Science.gov (United States)

    Tinevez, J.-Y.; Martin, P.; Jülicher, F.

    2009-02-01

    The hair bundle is both a mechano-sensory antenna and a force generator that might help the vertebrate hair cell from the inner ear to amplify its responsiveness to small stimuli. To study active hair-bundle motility, we combined calcium iontophoresis with mechanical stimulation of single hair bundles from the bullfrog's sacculus. A hair bundle could oscillate spontaneously, or be quiescent but display non-monotonic movements in response to abrupt force steps. Extracellular calcium changes or static biases to the bundle's position at rest could affect the kinetics of bundle motion and evoke transitions between the different classes of motility. The calcium-dependent location of a bundle's operating point within its nonlinear force-displacement relation controlled the type of movements observed. A unified theoretical description, in which mechanical activity stems from myosin-based adaptation and electro-mechanical feedback by Ca2+, could account for the fast and slow manifestations of active hair-bundle motility.

  3. LUSTERNIK-S CHNIRELMANN CATEGORY AND EMBEDDING FINITE COVERING MAPS, PRINCIPAL G-BUNDLES INTO BUNDLES

    Institute of Scientific and Technical Information of China (English)

    LIULUOFEI

    1996-01-01

    The author proves several embedding theorems for finite covering maps,principal G-bundies into bundles.The main results are 1. Let π:E→X be a finite covering map, and X a connected locally path-connected paracompact space. If cat X≤k, then the finite covering space π:E→X can be embedded into the trivial real k-plane bundle. 2. Let π:E→X be a principal G-bundle over a paracompact space. If there exists a linera action of Gon F(F=R or C)and cat X≤k ,then π:E→X can be embedded into ξ1 … ξn for any F-vector bundles ξi,i=1,…k.

  4. Holomorphic Vector Bundle on Hopf Manifolds with Abelian Fundamental Groups

    Institute of Scientific and Technical Information of China (English)

    Xiang Yu ZHOU; Wei Ming LIU

    2004-01-01

    Let X be a Hopf manifolds with an Abelian fundamental group. E is a holomorphic vector bundle of rank r with trivial pull-back to W = Cn - {0}. We prove the existence of a non-vanishing section of L(×) E for some line bundle on X and study the vector bundles filtration structure of E. These generalize the results of D. Mall about structure theorem of such a vector bundle E.

  5. Post irradiation examination of thermal reactor fuels

    Science.gov (United States)

    Sah, D. N.; Viswanathan, U. K.; Ramadasan, E.; Unnikrishnan, K.; Anantharaman, S.

    2008-12-01

    The post irradiation examination (PIE) facility at the Bhabha Atomic Research Centre (BARC) has been in operation for more than three decades. Over these years this facility has been utilized for examination of experimental fuel pins and fuels from commercial power reactors operating in India. In a program to assess the performance of (U,Pu)O 2 MOX fuel prior to its introduction in commercial reactors, three experimental MOX fuel clusters irradiated in the pressurized water loop (PWL) of CIRUS up to burnup of 16 000 MWd/tU were examined. Fission gas release from these pins was measured by puncture test. Some of these fuel pins in the cluster contained controlled porosity pellets, low temperature sintered (LTS) pellets, large grain size pellets and annular pellets. PIE has also been carried out on natural UO 2 fuel bundles from Indian PHWRs, which included two high burnup (˜15 000 MWd/tU) bundles. Salient investigations carried out consisted of visual examination, leak testing, axial gamma scanning, fission gas analysis, microstructural examination of fuel and cladding, β, γ autoradiography of the fuel cross-section and fuel central temperature estimation from restructuring. A ThO 2 fuel bundle irradiated in Kakrapar Atomic Power Station (KAPS) up to a nominal fuel burnup of ˜11 000 MWd/tTh was also examined to evaluate its in-pile performance. The performance of the BWR fuel pins of Tarapur Atomic Power Stations (TAPS) was earlier assessed by carrying out PIE on 18 fuel elements selected from eight fuel assemblies irradiated in the two reactors. The burnup of these fuel elements varied from 5000 to 29 000 MWd/tU. This paper provides a brief review of some of the fuels examined and the results obtained on the performance of natural UO 2, enriched UO 2, MOX, and ThO 2 fuels.

  6. Review of design criteria and safety analysis of safety class electric building for fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.

    1998-02-01

    Steady state fuel test loop will be equipped in HANARO to obtain the development and betterment of advanced fuel and materials through the irradiation tests. HANARO fuel test loop was designed for CANDU and PWR fuel testing. Safety related system of Fuel Test Loop such as emergency cooling water system, component cooling water system, safety ventilation system, high energy line break mitigation system and remote control room was required 1E class electric supply to meet the safety operation in accordance with related code. Therefore, FTL electric building was designed to construction and install the related equipment based on seismic category I. The objective of this study is to review the design criteria and analysis the safety function of safety class electric building for fuel test loop, and this results will become guidance for the irradiation testing in future. (author). 10 refs., 6 tabs., 30 figs.

  7. A Comparison between Clinical Results of Selective Bundle and Double Bundle Anterior Cruciate Ligament Reconstruction

    Science.gov (United States)

    Yoo, Yon-Sik; Song, Si Young; Yang, Cheol Jung; Ha, Jong Mun; Kim, Yoon Sang

    2016-01-01

    Purpose The purpose of this study was to compare the clinical outcomes of arthroscopic anatomical double bundle (DB) anterior cruciate ligament (ACL) reconstruction with either selective anteromedial (AM) or posterolateral (PL) bundle reconstruction while preserving a relatively healthy ACL bundle. Materials and Methods The authors evaluated 98 patients with a mean follow-up of 30.8±4.0 months who had undergone DB or selective bundle ACL reconstructions. Of these, 34 cases underwent DB ACL reconstruction (group A), 34 underwent selective AM bundle reconstruction (group B), and 30 underwent selective PL bundle reconstructions (group C). These groups were compared with respect to Lysholm and International Knee Documentation Committee (IKDC) score, side-to-side differences of anterior laxity measured by KT-2000 arthrometer at 30 lbs, and stress radiography and Lachman and pivot shift test results. Pre- and post-operative data were objectively evaluated using a statistical approach. Results The preoperative anterior instability measured by manual stress radiography at 90° of knee flexion in group A was significantly greater than that in groups B and C (all pACL tears offers comparable clinical results to DB reconstruction in complete ACL tears. PMID:27401652

  8. QTLs analysis of rice peduncle vascular bundle and panicle traits

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    @@The vascular bundle in plants plays an important role in transportation of photosynthetic products, mineral nutrients, water, and so on. Significant positive correlations were found between grain yield, panicle traits and the No. Of peduncle vascular bundles. So, it is very important to study the inheritance of peduncle vascular bundle, which is a quantitative trait.

  9. VECTOR BUNDLE, KILLING VECTOR FIELD AND PONTRYAGIN NUMBERS

    Institute of Scientific and Technical Information of China (English)

    周建伟

    1991-01-01

    Let E be a vector bundle over a compact Riemannian manifold M. We construct a natural metric on the bundle space E and discuss the relationship between the killing vector fields of E and M. Then we give a proof of the Bott-Baum-Cheeger Theorem for vector bundle E.

  10. Heat exchanger with helical bundles of finned tubes

    Energy Technology Data Exchange (ETDEWEB)

    Eyking, H.J.

    1975-01-23

    The invention applies to a heat exchanger with helical bundles of tubes consisting of finned tubes separated by spacers. The spacers are designed as closed holding cylinders with holding devices for the tube bundles, each ot which surrounds a bundle of tubes. This construction serves to simplify the production process and to enable the use of the heat exchanger at higher loads.

  11. Stability of Picard Bundle Over Moduli Space of Stable Vector Bundles of Rank Two Over a Curve

    Indian Academy of Sciences (India)

    Indranil Biswas; Tomás L Gómez

    2001-08-01

    Answering a question of [BV] it is proved that the Picard bundle on the moduli space of stable vector bundles of rank two, on a Riemann surface of genus at least three, with fixed determinant of odd degree is stable.

  12. Thermal Analysis of CANDU6-Moderator System for Loss of Cooling%CANDU6慢化剂系统丧失冷却情况下的温度分析

    Institute of Scientific and Technical Information of China (English)

    徐珍

    2012-01-01

    The coolant system and moderator system of CANDU6 are independent. The prompt neutrons are moderated as thermal neutrons by the moderator and the continuous nuclear fission in the reactor is maintained. At the same time the moderator system supplies the heat sink for the heat produced by the neutrons moderation. During the in-service maintenance of plant, the standby RCW which will only cool down reactor coolant system operates instead of RCW and can not supply heat sink for moderator system heat exchanger. As the result, the moderator system will lose heat sink during the operation of standby RCW. To estimate the moderator temperature, was compared with the experiment data and evaluated in this paper. the thermal analysis of moderator system for loss of cooling the system failure caused by the temperature raising wasevaluated in this paper.%压力管卧式重水反应堆(CANDU6)具有相互独立的冷却剂系统和慢化剂系统。慢化剂系统将堆芯高能裂变中子慢化到能维持持续裂变所需的热中子水平,并将慢化中子过程中产生的热量带出。在反应堆大修期间,需要对再循环冷却水系统(RCW)进行检修,则需要并投入其备用系统,但是RCW备用系统仅对反应堆冷却剂系统进行冷却,不提供慢化剂系统热交换器冷却水。所以在RCW备用系统投入的情况下,慢化剂系统丧失冷却。为判断在此情况下慢化剂的温度变化情况,本文对CANDU6大修期间慢化剂系统丧失冷却情况下的温度变化进行分析并与试验结果进行比较,评估是否会由于温度过高而导致系统失效。

  13. Bundling of harvesting residues and whole-trees and the treatment of bundles; Hakkuutaehteiden ja kokopuiden niputus ja nippujen kaesittely

    Energy Technology Data Exchange (ETDEWEB)

    Kaipainen, H.; Seppaenen, V.; Rinne, S.

    1996-12-31

    The conditions on which the bundling of the harvesting residues from spruce regeneration fellings would become profitable were studied. The calculations showed that one of the most important features was sufficient compaction of the bundle, so that the portion of the wood in the unit volume of the bundle has to be more than 40 %. The tests showed that the timber grab loader of farm tractor was insufficient for production of dense bundles. The feeding and compression device of the prototype bundler was constructed in the research and with this device the required density was obtained.The rate of compaction of the dry spruce felling residues was about 40 % and that of the fresh residues was more than 50 %. The comparison between the bundles showed that the calorific value of the fresh bundle per unit volume was nearly 30 % higher than that of the dry bundle. This means that the treatment of the bundles should be done of fresh felling residues. Drying of the bundles succeeded well, and the crushing and chipping tests showed that the processing of the bundles at the plant is possible. The treatability of the bundles was also excellent. By using the prototype, developed in the research, it was possible to produce a bundle of the fresh spruce harvesting residues, the diameter of which was about 50 cm and the length about 3 m, and the rate of compaction over 50 %. By these values the reduction target of the costs is obtainable

  14. A study on the environmental friendliness of nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. J.; Lee, B. H.; Lee, S. Y.; Lim, C. Y.; Choi, Y. S.; Lee, Y. E.; Hong, D. S.; Cheong, J. H; Park, J. B.; Kim, K. K.; Cheong, H. Y; Song, M. C; Lee, H. J. [Korea Advanced Inst. of Science and Technology, Taejon (Korea, Republic of)

    1998-01-01

    The purpose of this study is to develop methodologies for quantifying environmental and socio-political factors involved with nuclear fuel cycle and finally to evaluate nuclear fuel cycle options with special emphasis given to the factors. Moreover, methodologies for developing practical radiological health risk assessment code system will be developed by which the assessment could be achieved for the recycling and reuse of scrap materials containing residual radioactive contamination. Selected scenarios are direct disposal, DUPIC(Direct use of PWR spent fuel in CANDU), and MOX recycle, land use, radiological effect, and non-radiological effect were chosen for environmental criteria and public acceptance and non-proliferation of nuclear material for socio-political ones. As a result of this study, potential scenarios to be chosen in Korea were selected and methodologies were developed to quantify the environmental and socio-political criteria. 24 refs., 27 tabs., 29 figs. (author)

  15. Emitters of N-photon bundles.

    Science.gov (United States)

    Muñoz, C Sánchez; Del Valle, E; Tudela, A González; Müller, K; Lichtmannecker, S; Kaniber, M; Tejedor, C; Finley, J J; Laussy, F P

    2014-07-01

    Controlling the ouput of a light emitter is one of the basic tasks of photonics, with landmarks such as the laser and single-photon sources. The development of quantum applications makes it increasingly important to diversify the available quantum sources. Here, we propose a cavity QED scheme to realize emitters that release their energy in groups, or "bundles" of N photons, for integer N. Close to 100% of two-photon emission and 90% of three-photon emission is shown to be within reach of state of the art samples. The emission can be tuned with system parameters so that the device behaves as a laser or as a N-photon gun. The theoretical formalism to characterize such emitters is developed, with the bundle statistics arising as an extension of the fundamental correlation functions of quantum optics. These emitters will be useful for quantum information processing and for medical applications.

  16. Care bundles reduce readmissions for COPD.

    Science.gov (United States)

    Matthews, Healther; Tooley, Cathy; Nicholls, Carol; Lindsey-Halls, Anna

    In 2011, the respiratory nursing team at the James Paget University Hospital Foundation Trust were considering introducing a discharge care bundle for patients admitted with an acute exacerbation of chronic obstructive pulmonary disease. At the same time, the trust was asking for applications for Commissioning for Quality and Innovation schemes (CQUINs). These are locally agreed packages of quality improvement goals and indicators, which, if achieved in total, enable the provider to earn its full CQUIN payment. A CQUIN scheme should address the three domains of quality, safety and effectiveness, patient experience and also show innovation. This article discusses how the care bundle was introduced and how, over a 12-month period, it showed tangible results in improving the care pathway for COPD patients as well as reducing readmissions and saving a significant amount of money.

  17. Phase Slips in Oscillatory Hair Bundles

    Science.gov (United States)

    Roongthumskul, Yuttana; Shlomovitz, Roie; Bruinsma, Robijn; Bozovic, Dolores

    2013-01-01

    Hair cells of the inner ear contain an active amplifier that allows them to detect extremely weak signals. As one of the manifestations of an active process, spontaneous oscillations arise in fluid immersed hair bundles of in vitro preparations of selected auditory and vestibular organs. We measure the phase-locking dynamics of oscillatory bundles exposed to low-amplitude sinusoidal signals, a transition that can be described by a saddle-node bifurcation on an invariant circle. The transition is characterized by the occurrence of phase slips, at a rate that is dependent on the amplitude and detuning of the applied drive. The resultant staircase structure in the phase of the oscillation can be described by the stochastic Adler equation, which reproduces the statistics of phase slip production. PMID:25167040

  18. Client Provider Collaboration for Service Bundling

    Directory of Open Access Journals (Sweden)

    LETIA, I. A.

    2008-04-01

    Full Text Available The key requirement for a service industry organization to reach competitive advantages through product diversification is the existence of a well defined method for building service bundles. Based on the idea that the quality of a service or its value is given by the difference between expectations and perceptions, we draw the main components of a frame that aims to support the client and the provider agent in an active collaboration meant to co-create service bundles. Following e3-value model, we structure the supporting knowledge around the relation between needs and satisfying services. We deal with different perspectives about quality through an ontological extension of Value Based Argumentation. The dialog between the client and the provider takes the form of a persuasion whose dynamic object is the current best configuration. Our approach for building service packages is a demand driven approach, allowing progressive disclosure of private knowledge.

  19. Deformations of Fell bundles and twisted graph algebras

    Science.gov (United States)

    Raeburn, Iain

    2016-11-01

    We consider Fell bundles over discrete groups, and the C*-algebra which is universal for representations of the bundle. We define deformations of Fell bundles, which are new Fell bundles with the same underlying Banach bundle but with the multiplication deformed by a two-cocycle on the group. Every graph algebra can be viewed as the C*-algebra of a Fell bundle, and there are are many cocycles of interest with which to deform them. We thus obtain many of the twisted graph algebras of Kumjian, Pask and Sims. We demonstate the utility of our approach to these twisted graph algebras by proving that the deformations associated to different cocycles can be assembled as the fibres of a C*-bundle.

  20. Quantum principal bundles and their characteristic classes

    CERN Document Server

    Durdevic, M

    1996-01-01

    A brief exposition of the general theory of characteristic classes of quantum principal bundles is given. The theory of quantum characteristic classes incorporates ideas of classical Weil theory into the conceptual framework of non-commutative differential geometry. A purely cohomological interpretation of the Weil homomorphism is given, together with a standard geometrical interpretation via quantum invariant polynomials. A natural spectral sequence is described. Some quantum phenomena appearing in the formalism are discussed.

  1. On Complex Supermanifolds with Trivial Canonical Bundle

    CERN Document Server

    Groeger, Josua

    2016-01-01

    We give an algebraic characterisation for the triviality of the canonical bundle of a complex supermanifold in terms of a certain Batalin-Vilkovisky superalgebra structure. As an application, we study the Calabi-Yau case, in which an explicit formula in terms of the Levi-Civita connection is achieved. Our methods include the use of complex integral forms and the recently developed theory of superholonomy.

  2. Uncontrolled inexact information within bundle methods

    OpenAIRE

    Malick, Jérôme; Welington De Oliveira, ·; Zaourar-Michel, Sofia

    2016-01-01

    International audience; We consider convex nonsmooth optimization problems where additional information with uncontrolled accuracy is readily available. It is often the case when the objective function is itself the output of an optimization solver, as for large-scale energy optimization problems tackled by decomposition. In this paper, we study how to incorporate the uncontrolled linearizations into (proximal and level) bundle algorithms in view of generating better iterates and possibly acc...

  3. Uncovering ecosystem service bundles through social preferences.

    Directory of Open Access Journals (Sweden)

    Berta Martín-López

    Full Text Available Ecosystem service assessments have increasingly been used to support environmental management policies, mainly based on biophysical and economic indicators. However, few studies have coped with the social-cultural dimension of ecosystem services, despite being considered a research priority. We examined how ecosystem service bundles and trade-offs emerge from diverging social preferences toward ecosystem services delivered by various types of ecosystems in Spain. We conducted 3,379 direct face-to-face questionnaires in eight different case study sites from 2007 to 2011. Overall, 90.5% of the sampled population recognized the ecosystem's capacity to deliver services. Formal studies, environmental behavior, and gender variables influenced the probability of people recognizing the ecosystem's capacity to provide services. The ecosystem services most frequently perceived by people were regulating services; of those, air purification held the greatest importance. However, statistical analysis showed that socio-cultural factors and the conservation management strategy of ecosystems (i.e., National Park, Natural Park, or a non-protected area have an effect on social preferences toward ecosystem services. Ecosystem service trade-offs and bundles were identified by analyzing social preferences through multivariate analysis (redundancy analysis and hierarchical cluster analysis. We found a clear trade-off among provisioning services (and recreational hunting versus regulating services and almost all cultural services. We identified three ecosystem service bundles associated with the conservation management strategy and the rural-urban gradient. We conclude that socio-cultural preferences toward ecosystem services can serve as a tool to identify relevant services for people, the factors underlying these social preferences, and emerging ecosystem service bundles and trade-offs.

  4. Remote fabrication and irradiation test of recycled nuclear fuel prepared by the oxidation and reduction of spent oxide fuel

    Science.gov (United States)

    Jin Ryu, Ho; Chan Song, Kee; Il Park, Geun; Won Lee, Jung; Seung Yang, Myung

    2005-02-01

    A direct dry recycling process was developed in order to reuse spent pressurized light water reactor (LWR) nuclear fuel in CANDU reactors without the separation of sensitive nuclear materials such as plutonium. The benefits of the dry recycling process are the saving of uranium resources and the reduction of spent fuel accumulation as well as a higher proliferation resistance. In the process of direct dry recycling, fuel pellets separated from spent LWR fuel rods are oxidized from UO2 to U3O8 at 500 °C in an air atmosphere and reduced into UO2 at 700 °C in a hydrogen atmosphere, which is called OREOX (oxidation and reduction of oxide fuel). The pellets are pulverized during the oxidation and reduction processes due to the phase transformation between cubic UO2 and orthorhombic U3O8. Using the oxide powder prepared from the OREOX process, the compaction and sintering processes are performed in a remote manner in a shielded hot cell due to the high radioactivity of the spent fuel. Most of the fission gas and volatile fission products are removed during the OREOX and sintering processes. The mini-elements fabricated by the direct dry recycling process are irradiated in the HANARO research reactor for the performance evaluation of the recycled fuel pellets. Post-irradiation examination of the irradiated fuel showed that microstructural evolution and fission gas release behavior of the dry-recycled fuel were similar to high burnup UO2 fuel.

  5. Development of Strengthened Bundle High Temperature Superconductors

    Energy Technology Data Exchange (ETDEWEB)

    Lue, J.W.; Lubell, M.S. [Oak Ridge National Lab., TN (United States); Demko, J.A. [Oak Ridge Inst. for Science and Education, TN (United States); Tomsic, M. [Plastronic, Inc., Troy, OH (United States); Sinha, U. [Southwire Company, Carollton, GA (United States)

    1997-12-31

    In the process of developing high temperature superconducting (HTS) transmission cables, it was found that mechanical strength of the superconducting tape is the most crucial property that needs to be improved. It is also desirable to increase the current carrying capacity of the conductor so that fewer layers are needed to make the kilo-amp class cables required for electric utility usage. A process has been developed by encapsulating a stack of Bi-2223/Ag tapes with a silver or non-silver sheath to form a strengthened bundle superconductor. This process was applied to HTS tapes made by the Continuous Tube Forming and Filling (CTFF) technique pursued by Plastronic Inc. and HTS tapes obtained from other manufacturers. Conductors with a bundle of 2 to 6 HTS tapes have been made. The bundled conductor is greatly strengthened by the non-silver sheath. No superconductor degradation as compared to the sum of the original critical currents of the individual tapes was seen on the finished conductors.

  6. Bundling harvester; Harvennuspuun automaattisen nippukorjausharvesterin kehittaeminen

    Energy Technology Data Exchange (ETDEWEB)

    Koponen, K. [Eko-Log Oy, Kuopio (Finland)

    1997-12-01

    The starting point of the project was to design and construct, by taking the silvicultural point of view into account, a harvesting and processing system especially for energy-wood, containing manually driven bundling harvester, automating of the harvester, and automated loading. The equipment forms an ideal method for entrepreneur`s-line harvesting. The target is to apply the system also for owner`s-line harvesting. The profitability of the system promotes the utilisation of the system in both cases. The objectives of the project were: to construct a test equipment and prototypes for all the project stages, to carry out terrain and strain tests in order to examine the usability and durability, as well as the capacity of the machine, to test the applicability of the Eko-Log system in simultaneous harvesting of energy and pulp woods, and to start the marketing and manufacturing of the products. The basic problems of the construction of the bundling harvester have been solved using terrain-tests. The prototype machine has been shown to be operable. Loading of the bundles to form sufficiently economically transportable loads has been studied, and simultaneously, the branch-biomass has been tried to be utilised without loosing the profitability of transportation. The results have been promising, and will promote the profitable utilisation of wood-energy. (orig.)

  7. Subchannel Code Benchmarking to Columbia University 4x4 and Pacific Northwest Laboratory 2x6 Bundle Test Data

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Kang Hoon; Oezdemir, Erdal; Oh, Seung Jong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2014-10-15

    The subchannel code is used to assess the safety of a reactor core at the steady-state and transient conditions. KEPCO Nuclear Fuel (KNF) has been developed new subchannel code, THALES, for PWR core design application. In this study, we are comparing the THALES result with VIPRE-01 code result utilizing bundle test data. VIPRE-01 was developed under EPRI sponsorship and has been used by U.S. PWR commercial nuclear utilities, historically. THALES and VIPRE-01 codes were benchmarked to two kind of bundle test data which were at the steady-state and transient conditions. THALES predicted fluid velocity and temperature profile of bundle test data well and the error rate between THALES and VIPRE-01 was very small.

  8. Velocity and turbulence distributions in wall subchannels of a road bundle in three axial planes downstream of a spacer grid

    Science.gov (United States)

    Rehme, K.

    1987-03-01

    The velocity, turbulence, and temperature distributions in nuclear fuel element bundles of nuclear reactors were investigated. The mean velocity, the wall shear stresses, and the turbulence were measured in two wall subchannels of a rod bundle of four parallel rods, arranged in a rectangular channel, for three axial planes. A spacer grid was inserted in the rod bundle, for ratios between the distance spacer grid/measuring plane and the hydraulic diameter (LIDh) of 40.4, 32.8 and 16.9. The Reynolds number was 145,000. The results show that the distributions of the velocity and the turbulence are affected by the spacer grid, already for LIDh = 40.4. The effects of the spacer grid increase with decreasing distance to the spacer grid.

  9. An analytical fiber bundle model for pullout mechanics of root bundles

    Science.gov (United States)

    Cohen, D.; Schwarz, M.; Or, D.

    2011-09-01

    Roots in soil contribute to the mechanical stability of slopes. Estimation of root reinforcement is challenging because roots form complex biological networks whose geometrical and mechanical characteristics are difficult to characterize. Here we describe an analytical model that builds on simple root descriptors to estimate root reinforcement. Root bundles are modeled as bundles of heterogeneous fibers pulled along their long axes neglecting root-soil friction. Analytical expressions for the pullout force as a function of displacement are derived. The maximum pullout force and corresponding critical displacement are either derived analytically or computed numerically. Key model inputs are a root diameter distribution (uniform, Weibull, or lognormal) and three empirical power law relations describing tensile strength, elastic modulus, and length of roots as functions of root diameter. When a root bundle with root tips anchored in the soil matrix is pulled by a rigid plate, a unique parameter, ?, that depends only on the exponents of the power law relations, dictates the order in which roots of different diameters break. If ? 1, large roots break first. When ? = 1, all fibers break simultaneously, and the maximum tensile force is simply the roots' mean force times the number of roots in the bundle. Based on measurements of root geometry and mechanical properties, the value of ? is less than 1, usually ranging between 0 and 0.7. Thus, small roots always fail first. The model shows how geometrical and mechanical characteristics of roots and root diameter distribution affect the pullout force, its maximum and corresponding displacement. Comparing bundles of roots that have similar mean diameters, a bundle with a narrow variance in root diameter will result in a larger maximum force and a smaller displacement at maximum force than a bundle with a wide diameter distribution. Increasing the mean root diameter of a bundle without changing the distribution's shape increases

  10. Modelling packing interactions in parallel helix bundles: pentameric bundles of nicotinic receptor M2 helices.

    Science.gov (United States)

    Sankararamakrishnan, R; Sansom, M S

    1995-11-01

    The transbilayer pore of the nicotinic acetylcholine receptor (nAChR) is formed by a pentameric bundle of M2 helices. Models of pentameric bundles of M2 helices have been generated using simulated annealing via restrained molecular dynamics. The influence of: (a) the initial C alpha template; and (b) screening of sidechain electrostatic interactions on the geometry of the resultant M2 helix bundles is explored. Parallel M2 helices, in the absence of sidechain electrostatic interactions, pack in accordance with simple ridges-in-grooves considerations. This results in a helix crossing angle of ca. +12 degrees, corresponding to a left-handed coiled coil structure for the bundle as a whole. Tilting of M2 helices away from the central pore axis at their C-termini and/or inclusion of sidechain electrostatic interactions may perturb such ridges-in-grooves packing. In the most extreme cases right-handed coiled coils are formed. An interplay between inter-helix H-bonding and helix bundle geometry is revealed. The effects of changes in electrostatic screening on the dimensions of the pore mouth are described and the significance of these changes in the context of models for the nAChR pore domain is discussed.

  11. Influence on rewetting temperature and wetting delay during rewetting rod bundle by various radial jet models

    Energy Technology Data Exchange (ETDEWEB)

    Debbarma, Ajoy; Pandey, Krishna Murari [National Institute of Technology, Assam (India). Dept. of Mechanical Engineering

    2016-03-15

    Numerical investigation of the rewetting of single sector fuel assembly of Advanced Heavy Water Reactor (AHWR) has been carried out to exhibit the effect of coolant jet diameters (2, 3 and 4 mm) and jet directions (Model: M, X and X2). The rewetting phenomena with various jet models are compared on the basis of rewetting temperature and wetting delay. Temperature-time curve have been evaluated from rods surfaces at different circumference, radial and axial locations of rod bundle. The cooling curve indicated the presence of vapor in respected location, where it prevents the contact between the firm and fluid phases. The peak wall temperature represents as rewetting temperature. The time period observed between initial to rewetting temperature point is wetting delay. It was noted that as improved in various jet models, rewetting temperature and wetting delay reduced, which referred the coolant stipulation in the rod bundle dominant vapor formation.

  12. Heat transfer in a vertical 7-element bundle cooled with supercritical Freon-12

    Energy Technology Data Exchange (ETDEWEB)

    Richards, G.; Milner, A.; Pascoe, C.; Patel, H.; Peiman, W.; Saltanov, Eu. [Univ. of Ontario Inst. of Tech., Faculty of Energy Systems and Nuclear Science, Oshawa, Ontario (Canada); Pometko, R.S.; Opanasenko, A.N. [Inst. of Physics and Power Engineering, Obninsk (Russian Federation); Shelegov, A.S. [IATE, Obninsk (Russian Federation); Kirillov, P.L. [Inst. of Physics and Power Engineering, Obninsk (Russian Federation); Pioro, I. [Univ. of Ontario Inst. of Tech., Faculty of Energy Systems and Nuclear Science, Oshawa, Ontario (Canada)

    2010-07-01

    Currently, SuperCritical Water-cooled nuclear Reactor (SCWR) concepts are being developed worldwide with an objective to increase thermal efficiencies of future Nuclear Power Plants (NPPs) on 10 -15% compared to those of current water-cooled NPPs. With such an increase in the thermal efficiencies, SCW NPPs will be at the current level of the most advanced thermal power plants: coal-fired SCW NPPs and combined-cycle gas-fired NPPs. However, to be able to develop SCWRs at least several key technical problems should be resolved. One of these problems is limited amount of experimental data on heat transfer in fuel bundles and based on that SCW heat-transfer correlations. Experiments in SCW are very complicated and expensive due to high critical parameters of water (pressure 22.064 MPa and temperature 374.95°C). Moreover, there are only few SCW test rigs, which capable to perform experiments in full-scale bundles. As a preliminary approach supercritical-pressure heat-transfer experiments in bundles can be performed in modeling fluids such as Freons or carbon dioxide. Therefore, a set of experimental data was obtained in Freon-12-cooled bundle simulator at the Institute of Physics and Power Engineering (IPPE, Obninsk, Russia). A vertical 7-element bundle was installed in a hexagonal flow channel. The test section consisted of elements that were 9.5 mm in diameter with the total heated length of 1 m. Bulk-fluid and wall temperature profiles were recorded using thermocouples. Several heat-transfer regimes were tested. Also, this paper references thermophysical properties of supercritical Freon-12 at the critical pressure (4.1361 MPa) and test pressure of 4.65 MPa. (author)

  13. Experimental investigation of heat transfer from a 2 × 2 rod bundle to supercritical pressure water

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Han [State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, Xi’an 710049 (China); Bi, Qincheng, E-mail: qcbi@mail.xjtu.edu.cn [State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, Xi’an 710049 (China); Wang, Linchuan; Lv, Haicai [State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, Xi’an 710049 (China); Leung, Laurence K.H. [Atomic Energy of Canada Limited, Chalk River, Ont., Canada K0J 1J0 (Canada)

    2014-08-15

    Highlights: • Heat transfer of supercritical water through a 2 × 2 rod bundles is investigated. • Circumferential wall temperature distribution is obtained. • Effects of system parameters on heat transfer characteristics are analyzed. • Heat transfer correlations are compared against the rod bundle test data. - Abstract: Heat transfer experiments with supercritical pressure water flowing vertically upward through a 2 × 2 rod bundle have been performed at Xi’an Jiaotong University. A fuel-assembly simulator with four heated rods was installed inside a square channel with rounded corner. The outer diameter of each heated rod is 8 mm with an effective heated length of 600 mm. The experiments covered the pressure range of 23–28 MPa, mass-flux range of 350–1000 kg/(m{sup 2} s) and heat-flux range on the rod surface of 200–1000 kW/m{sup 2}. Heat transfer characteristics of supercritical pressure water through the bundle were examined with respect to variations of heat flux, system pressure, and mass flux. These characteristics were shown to be similar to those previously observed in tubes or annuli. The experimental data indicate a non-uniform circumferential wall-temperature distribution around the heated rod. A maximum wall temperature was observed at the surface facing the corner gap between the heated rod and the ceramic tube, while the minimum wall temperature was observed at the surface facing the center subchannel. The difference between maximum and minimum wall temperatures varies with heat flux and/or mass flux. Eight heat transfer correlations developed for supercritical water were assessed against the current set of test data. Prediction of the Jackson correlation agrees closely with the experimental Nusselt number. A new correlation has been derived based on Jackson correlation to improve the prediction accuracy of supercritical heat transfer coefficient in a 2 × 2 rod bundle.

  14. Modelling material effects on flow-accelerated corrosion in primary CANDU coolant and secondary reactor feed-water

    Energy Technology Data Exchange (ETDEWEB)

    Phromwong, P.; Lister, D., E-mail: c7r13@unb.ca [Univ. of New Brunswick, Dept. of Chemical Engineering, Fredericton, New Brunswick (Canada); Uchida, S. [Japan Atomic Energy Agency, Tokai-mura, Ibaraki (Japan)

    2012-07-01

    The effects of chromium content on flow-accelerated corrosion (FAC) of carbon steel have been predicted very well by including a passivating layer, which is a chromium-dependent diffusion barrier at the metal-oxide interface. By adjusting the properties of the chromium-dependent layer, described with a Passivation Parameter (PP), we can predict the FAC of carbon steel of different chromium contents in typical reactor feed-water environments (140{sup o}C and neutral or ammoniated chemistry). The model and an appropriate PP are also applied to the environment typical of carbon-steel feeders in the primary coolant of a CANDU reactor (310{sup o}C and lithiated chemistry). The model predicts FAC rate very well (with a deviation of 10% or less) in both situations. (author)

  15. Probabilistic seismic safety assessment of a CANDU 6 nuclear power plant including ambient vibration tests: Case study

    Energy Technology Data Exchange (ETDEWEB)

    Nour, Ali [Hydro Québec, Montréal, Québec H2L4P5 (Canada); École Polytechnique de Montréal, Montréal, Québec H3C3A7 (Canada); Cherfaoui, Abdelhalim; Gocevski, Vladimir [Hydro Québec, Montréal, Québec H2L4P5 (Canada); Léger, Pierre [École Polytechnique de Montréal, Montréal, Québec H3C3A7 (Canada)

    2016-08-01

    Highlights: • In this case study, the seismic PSA methodology adopted for a CANDU 6 is presented. • Ambient vibrations testing to calibrate a 3D FEM and to reduce uncertainties is performed. • Procedure for the development of FRS for the RB considering wave incoherency effect is proposed. • Seismic fragility analysis for the RB is presented. - Abstract: Following the 2011 Fukushima Daiichi nuclear accident in Japan there is a worldwide interest in reducing uncertainties in seismic safety assessment of existing nuclear power plant (NPP). Within the scope of a Canadian refurbishment project of a CANDU 6 (NPP) put in service in 1983, structures and equipment must sustain a new seismic demand characterised by the uniform hazard spectrum (UHS) obtained from a site specific study defined for a return period of 1/10,000 years. This UHS exhibits larger spectral ordinates in the high-frequency range than those used in design. To reduce modeling uncertainties as part of a seismic probabilistic safety assessment (PSA), Hydro-Québec developed a procedure using ambient vibrations testing to calibrate a detailed 3D finite element model (FEM) of the containment and reactor building (RB). This calibrated FE model is then used for generating floor response spectra (FRS) based on ground motion time histories compatible with the UHS. Seismic fragility analyses of the reactor building (RB) and structural components are also performed in the context of a case study. Because the RB is founded on a large circular raft, it is possible to consider the effect of the seismic wave incoherency to filter out the high-frequency content, mainly above 10 Hz, using the incoherency transfer function (ITF) method. This allows reducing significantly the non-necessary conservatism in resulting FRS, an important issue for an existing NPP. The proposed case study, and related methodology using ambient vibration testing, is particularly useful to engineers involved in seismic re-evaluation of

  16. Hybrid Monte Carlo deterministic and probabilistic core assessment for flaws and leak-before break for CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shi, P.; Mok, D.H.B. [AMEC NSS, Toronto, Ontario (Canada)

    2011-07-01

    Even though pressure tubes are major components of a CANDU reactor, only small proportions of pressure tubes are sampled for in-service inspections due to execution cost, outage duration, and site cumulative radiation exposure limits. In general, a realistic core assessment was not carried out based on all known information related to in-service degradation mechanisms. Recently, a hybrid deterministic and probabilistic core assessment (HDPCA) has been introduced to address the uncertainties associated with uninspected pressure tubes and diverse degradation mechanisms. In the present paper, the HDPCA was carried out for a CANDU unit based on cumulative operating experience and history in order to satisfy the requirements of Clause 7 of CSA Standard N285.8 by considering the uncertainties associated with the estimated distribution parameters, the limited inspected data, and pressure tube properties. The HDPCA is composed of two parts: a simulation part and a deterministic evaluation part. The outcome of the core assessment is the expected pressure tube failure frequency due to pressure tube flaws. In the simulations, pressure tube material properties were sampled from distributions derived from material surveillance and testing programs. The flaw dimensions and intensities were sampled from distributions fitted to in-service inspection data. The pressure tubes were then populated with flaws. Each simulated flaw was evaluated for DHC initiation under constant loading conditions. When Delayed Hydride Cracking initiation from a flaw was predicted, the pressure tube was evaluated for rupture in the Leak-Before-Break evaluation. Based on all the predicted pressure tube ruptures from simulations, the failure frequency was calculated on an annual basis. The largest expected mean and the 95% upper bound of the mean failure frequencies for any evaluation subinterval to the end of pressure tube design life of 210,000 EFPH are significantly below the allowable failure frequency

  17. A Tannakian approach to dimensional reduction of principal bundles

    CERN Document Server

    Álvarez-Cónsul, Luis; García-Prada, Oscar

    2016-01-01

    Let $P$ be a parabolic subgroup of a connected simply connected complex semisimple Lie group $G$. Given a compact K\\"ahler manifold $X$, the dimensional reduction of $G$-equivariant holomorphic vector bundles over $X\\times G/P$ was carried out by the first and third authors. This raises the question of dimensional reduction of holomorphic principal bundles over $X\\times G/P$. The method used for equivariant vector bundles does not generalize to principal bundles. In this paper, we adapt to equivariant principal bundles the Tannakian approach of Nori, to describe the dimensional reduction of $G$-equivariant principal bundles over $X\\times G/P$, and to establish a Hitchin--Kobayashi type correspondence. In order to be able to apply the Tannakian theory, we need to assume that $X$ is a complex projective manifold.

  18. Monopoles and Modifications of Bundles over Elliptic Curves

    Directory of Open Access Journals (Sweden)

    Andrey M. Levin

    2009-06-01

    Full Text Available Modifications of bundles over complex curves is an operation that allows one to construct a new bundle from a given one. Modifications can change a topological type of bundle. We describe the topological type in terms of the characteristic classes of the bundle. Being applied to the Higgs bundles modifications establish an equivalence between different classical integrable systems. Following Kapustin and Witten we define the modifications in terms of monopole solutions of the Bogomolny equation. We find the Dirac monopole solution in the case R × (elliptic curve. This solution is a three-dimensional generalization of the Kronecker series. We give two representations for this solution and derive a functional equation for it generalizing the Kronecker results. We use it to define Abelian modifications for bundles of arbitrary rank. We also describe non-Abelian modifications in terms of theta-functions with characteristic.

  19. Amplitude death of coupled hair bundles with stochastic channel noise

    CERN Document Server

    Kim, Kyung-Joong

    2014-01-01

    Hair cells conduct auditory transduction in vertebrates. In lower vertebrates such as frogs and turtles, due to the active mechanism in hair cells, hair bundles(stereocilia) can be spontaneously oscillating or quiescent. Recently, the amplitude death phenomenon has been proposed [K.-H. Ahn, J. R. Soc. Interface, {\\bf 10}, 20130525 (2013)] as a mechanism for auditory transduction in frog hair-cell bundles, where sudden cessation of the oscillations arises due to the coupling between non-identical hair bundles. The gating of the ion channel is intrinsically stochastic due to the stochastic nature of the configuration change of the channel. The strength of the noise due to the channel gating can be comparable to the thermal Brownian noise of hair bundles. Thus, we perform stochastic simulations of the elastically coupled hair bundles. In spite of stray noisy fluctuations due to its stochastic dynamics, our simulation shows the transition from collective oscillation to amplitude death as inter-bundle coupling str...

  20. The avalanche process of the fiber bundle model with defect

    Science.gov (United States)

    Hao, Da-Peng; Tang, Gang; Xia, Hui; Xun, Zhi-Peng; Han, Kui

    2017-04-01

    In order to explore the impacts of defect on the tensile fracture process of materials, the fiber bundle model with defect is constructed based on the classical fiber bundle model. In the fiber bundle model with defect, the two key parameters are the mean size and the density of defects. In both uniform and Weibull threshold distributions, the mean size and density all bring impacts on the threshold distribution of fibers. By means of analytical approximation and numerical simulation, we show that the two key parameters of the model have substantial effects on the failure process of the bundle. From macroscopic view, the defect described by the altering of threshold distribution of fibers will has a significant impact on the mechanical properties of the bundle. While in microscopic scale, the statistical properties of the model are still harmonious with the classical fiber bundle model.

  1. Vector bundles on complex projective spaces

    CERN Document Server

    Okonek, Christian; Spindler, Heinz

    1980-01-01

    This expository treatment is based on a survey given by one of the authors at the Séminaire Bourbaki in November 1978 and on a subsequent course held at the University of Göttingen. It is intended to serve as an introduction to the topical question of classification of holomorphic vector bundles on complex projective spaces, and can easily be read by students with a basic knowledge of analytic or algebraic geometry. Short supplementary sections describe more advanced topics, further results, and unsolved problems.

  2. Bundling Products and Services Through Modularization Strategies

    DEFF Research Database (Denmark)

    Bask, Anu; Hsuan, Juliana; Rajahonka, Mervi;

    2012-01-01

    Modularity has been recognized as a powerful tool in improving the efficiency and management of product design and manufacturing. However, the integrated view on covering both, product and service modularity for product-service systems (PSS), is under researched. Therefore, in this paper our...... objective is to contribute to the PSS modularity. Thus, we describe configurations of PSSs and the bundling of products and services through modularization strategies. So far there have not been tools to analyze and determine the correct combinations of degrees of product and service modularities....

  3. Higher order mechanics on graded bundles

    Science.gov (United States)

    Bruce, Andrew James; Grabowska, Katarzyna; Grabowski, Janusz

    2015-05-01

    In this paper we develop a geometric approach to higher order mechanics on graded bundles in both, the Lagrangian and Hamiltonian formalism, via the recently discovered weighted algebroids. We present the corresponding Tulczyjew triple for this higher order situation and derive in this framework the phase equations from an arbitrary (also singular) Lagrangian or Hamiltonian, as well as the Euler-Lagrange equations. As important examples, we geometrically derive the classical higher order Euler-Lagrange equations and analogous reduced equations for invariant higher order Lagrangians on Lie groupoids.

  4. Compression of a bundle of light rays.

    Science.gov (United States)

    Marcuse, D

    1971-03-01

    The performance of ray compression devices is discussed on the basis of a phase space treatment using Liouville's theorem. It is concluded that the area in phase space of the input bundle of rays is determined solely by the required compression ratio and possible limitations on the maximum ray angle at the output of the device. The efficiency of tapers and lenses as ray compressors is approximately equal. For linear tapers and lenses the input angle of the useful rays must not exceed the compression ratio. The performance of linear tapers and lenses is compared to a particular ray compressor using a graded refractive index distribution.

  5. Differential geometry of complex vector bundles

    CERN Document Server

    Kobayashi, Shoshichi

    2014-01-01

    Holomorphic vector bundles have become objects of interest not only to algebraic and differential geometers and complex analysts but also to low dimensional topologists and mathematical physicists working on gauge theory. This book, which grew out of the author's lectures and seminars in Berkeley and Japan, is written for researchers and graduate students in these various fields of mathematics. Originally published in 1987. The Princeton Legacy Library uses the latest print-on-demand technology to again make available previously out-of-print books from the distinguished backlist of Princeto

  6. Cosmic multimuon bundles detected by DELPHI

    CERN Document Server

    Rídky, J

    2004-01-01

    The DELPHI detector located at LEP accelerator has been used also to measure multimuon bundles originated from cosmic ray interactions. Two subdetectors-hadron calorimeter and time projection chamber, are used for this purpose. The 1999 and 2000 data are analyzed over wide range of multiplicities. The multiplicity distribution is compared with prediction of Monte Carlo simulation based on CORSIKA/QGSJET. The Monte-Carlo does not describe the large multiplicity part of data. Even the extreme assumption on the cosmic ray composition (pure iron nuclei) hardly predicts comparable number of high-multiplicity events.

  7. Improving the neutronic characteristics of a boiling water reactor by using uranium zirconium hydride fuel instead of uranium dioxide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Galahom, Ahmed Abdelghafar [Higher Technological Institute, Ramadan (Egypt)

    2016-06-15

    The present work discusses two different models of boiling water reactor (BWR) bundle to compare the neutronic characteristics of uranium dioxide (UO{sub 2}) and uranium zirconium hydride (UZrH{sub 1.6}) fuel. Each bundle consists of four assemblies. The BWR assembly fueled with UO{sub 2} contains 8 × 8 fuel rods while that fueled with UZrH{sub 1.6} contains 9 × 9 fuel rods. The Monte Carlo N-Particle Transport code, based on the Mont Carlo method, is used to design three dimensional models for BWR fuel bundles at typical operating temperatures and pressure conditions. These models are used to determine the multiplication factor, pin-by-pin power distribution, axial power distribution, thermal neutron flux distribution, and axial thermal neutron flux. The moderator and coolant (water) are permitted to boil within the BWR core forming steam bubbles, so it is important to calculate the reactivity effect of voiding at different values. It is found that the hydride fuel bundle design can be simplified by eliminating water rods and replacing the control blade with control rods. UZrH{sub 1.6} fuel improves the performance of the BWR in different ways such as increasing the energy extracted per fuel assembly, reducing the uranium ore, and reducing the plutonium accumulated in the BWR through burnup.

  8. Heat Transfer Analysis in Wire Bundles for Aerospace Vehicles

    Science.gov (United States)

    Rickman, S. L.; Iamello, C. J.

    2016-01-01

    Design of wiring for aerospace vehicles relies on an understanding of "ampacity" which refers to the current carrying capacity of wires, either, individually or in wire bundles. Designers rely on standards to derate allowable current flow to prevent exceedance of wire temperature limits due to resistive heat dissipation within the wires or wire bundles. These standards often add considerable margin and are based on empirical data. Commercial providers are taking an aggressive approach to wire sizing which challenges the conventional wisdom of the established standards. Thermal modelling of wire bundles may offer significant mass reduction in a system if the technique can be generalized to produce reliable temperature predictions for arbitrary bundle configurations. Thermal analysis has been applied to the problem of wire bundles wherein any or all of the wires within the bundle may carry current. Wire bundles present analytical challenges because the heat transfer path from conductors internal to the bundle is tortuous, relying on internal radiation and thermal interface conductance to move the heat from within the bundle to the external jacket where it can be carried away by convective and radiative heat transfer. The problem is further complicated by the dependence of wire electrical resistivity on temperature. Reduced heat transfer out of the bundle leads to higher conductor temperatures and, hence, increased resistive heat dissipation. Development of a generalized wire bundle thermal model is presented and compared with test data. The steady state heat balance for a single wire is derived and extended to the bundle configuration. The generalized model includes the effects of temperature varying resistance, internal radiation and thermal interface conductance, external radiation and temperature varying convective relief from the free surface. The sensitivity of the response to uncertainties in key model parameters is explored using Monte Carlo analysis.

  9. Heat transfer in bundles of finned tubes in crossflow

    Energy Technology Data Exchange (ETDEWEB)

    Stasiulevicius, J.; Skrinska, A.; Zukauskas, A.; Hewitt, G.F.

    1986-01-01

    This book provides correlations of heat transfer and hydraulic data for bundles of finned tubes in crossflow at high Reynolds numbers. Results of studies of the effectiveness of the fin, local, and mean heat transfer coefficients are presented. The effect of geometric parameters of the fins and of the location of tubes in the bundle on heat transfer and hydraulic drag are described. The resistance of the finned tube bundles under study and other factors are examined.

  10. Severe accident analysis of a station blackout accident using MAAP-CANDU for the Point Lepreau station refurbishment project level 2 PSA

    Energy Technology Data Exchange (ETDEWEB)

    Brown, M.J.; Petoukhov, S.M. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2012-07-01

    A Level 2 Probabilistic Safety Assessment was performed for the Point Lepreau Generating Station, using the MAAP-CANDU code to simulate the progression of severe core damage accidents and fission product releases. Five representative severe accidents were selected: Station Blackout, Small Loss-of-Coolant, Stagnation Feeder Break, Steam Generator Tube Rupture, and Shutdown State. Analysis results for the reference station blackout accident are discussed in this paper. (author)

  11. Isothermal microcalorimetry, a tool for probing SWNT bundles.

    Science.gov (United States)

    Marquis, Renaud; Greco, Carla; Schultz, Patrick; Meunier, Stéphane; Mioskowski, Charles

    2009-11-01

    The bundling state of several dry single-walled carbon nanotube (SWNT) samples is compared using isothermal microcalorimetry (IMC). So as to get different dry samples with various bundling states, the pristine SWNTs were pretreated with a solution of an aromatic amphiphile with or without sonication, washed and dried before being studied by IMC. The bundling state of the different SWNT samples, which was first analyzed by TEM, was then correlated to the obtained IMC data thanks to the interpretation of the observed energy transfer phenomena. From our results, IMC appears to be an interesting technique for the surface probing of dry SWNT samples, and herein for the evaluation of the bundling state.

  12. Restriction Theorem for Principal bundles in Arbitrary Characteristic

    DEFF Research Database (Denmark)

    Gurjar, Sudarshan

    2015-01-01

    The aim of this paper is to prove two basic restriction theorem for principal bundles on smooth projective varieties in arbitrary characteristic generalizing the analogues theorems of Mehta-Ramanathan for vector bundles. More precisely, let G be a reductive algebraic group over an algebraically...... closed field k and let X be a smooth, projective variety over k together with a very ample line bundle O(1). The main result of the paper is that if E is a semistable (resp. stable) principal G-bundle on X w.r.t O(1), then the restriction of E to a general, high multi-degree, complete-intersection curve...

  13. Development boiling to sprinkled tube bundle

    Directory of Open Access Journals (Sweden)

    Kracík Petr

    2016-01-01

    Full Text Available This paper presents results of a studied heat transfer coefficient at the surface of a sprinkled tube bundle where boiling occurs. Research in the area of sprinkled exchangers can be divided into two major parts. The first part is research on heat transfer and determination of the heat transfer coefficient at sprinkled tube bundles for various liquids, whether boiling or not. The second part is testing of sprinkle modes for various tube diameters, tube pitches and tube materials and determination of individual modes’ interface. All results published so far for water as the falling film liquid apply to one to three tubes for which the mentioned relations studied are determined in rigid laboratory conditions defined strictly in advance. The sprinkled tubes were not viewed from the operational perspective where there are more tubes and various modes may occur in different parts with various heat transfer values. The article focuses on these processes. The tube is located in a low-pressure chamber where vacuum is generated using an exhauster via ejector. The tube consists of smooth copper tubes of 12 mm diameter placed horizontally one above another.

  14. Development boiling to sprinkled tube bundle

    Science.gov (United States)

    Kracík, Petr; Pospíšil, Jiří

    2016-03-01

    This paper presents results of a studied heat transfer coefficient at the surface of a sprinkled tube bundle where boiling occurs. Research in the area of sprinkled exchangers can be divided into two major parts. The first part is research on heat transfer and determination of the heat transfer coefficient at sprinkled tube bundles for various liquids, whether boiling or not. The second part is testing of sprinkle modes for various tube diameters, tube pitches and tube materials and determination of individual modes' interface. All results published so far for water as the falling film liquid apply to one to three tubes for which the mentioned relations studied are determined in rigid laboratory conditions defined strictly in advance. The sprinkled tubes were not viewed from the operational perspective where there are more tubes and various modes may occur in different parts with various heat transfer values. The article focuses on these processes. The tube is located in a low-pressure chamber where vacuum is generated using an exhauster via ejector. The tube consists of smooth copper tubes of 12 mm diameter placed horizontally one above another.

  15. Bundled capillary electrophoresis using microstructured fibres.

    Science.gov (United States)

    Rogers, Benjamin; Gibson, Graham T T; Oleschuk, Richard D

    2011-01-01

    Joule heating, arising from the electric current passing through the capillary, causes many undesired effects in CE that ultimately result in band broadening. The use of narrow-bore capillaries helps to solve this problem as smaller cross-sectional area results in decreased Joule heating and the rate of heat dissipation is increased by the larger surface-to-volume ratio. Issues arising from such small capillaries, such as poor detection sensitivity, low loading capacity and high flow-induced backpressure (complicating capillary loading) can be avoided by using a bundle of small capillaries operating simultaneously that share buffer reservoirs. Microstructured fibres, originally designed as waveguides in the telecommunication industry, are essentially a bundle of parallel ∼5 μm id channels that extend the length of a fibre having otherwise similar dimensions to conventional CE capillaries. This work presents the use of microstructured fibres for CZE, taking advantage of their relatively high surface-to-volume ratio and the small individual size of each channel to effect highly efficient separations, particularly for dye-labelled peptides.

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

    CERN Document Server

    Jonkmans, G; Jewett, C; Thompson, M

    2012-01-01

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

  17. A thermal-hydraulic code for transient analysis in a channel with a rod bundle

    Energy Technology Data Exchange (ETDEWEB)

    Khodjaev, I.D. [Research & Engineering Centre of Nuclear Plants Safety, Electrogorsk (Russian Federation)

    1995-09-01

    The paper contains the model of transient vapor-liquid flow in a channel with a rod bundle of core of a nuclear power plant. The computer code has been developed to predict dryout and post-dryout heat transfer in rod bundles of nuclear reactor core under loss-of-coolant accidents. Economizer, bubble, dispersed-annular and dispersed regimes are taken into account. The computer code provides a three-field representation of two-phase flow in the dispersed-annular regime. Continuous vapor, continuous liquid film and entrained liquid drops are three fields. For the description of dispersed flow regime two-temperatures and single-velocity model is used. Relative droplet motion is taken into account for the droplet-to-vapor heat transfer. The conservation equations for each of regimes are solved using an effective numerical technique. This technique makes it possible to determine distribution of the parameters of flows along the perimeter of fuel elements. Comparison of the calculated results with the experimental data shows that the computer code adequately describes complex processes in a channel with a rod bundle during accident.

  18. Evaluation of Single-Bundle versus Double-Bundle PCL Reconstructions with More Than 10-Year Follow-Up

    Directory of Open Access Journals (Sweden)

    Masataka Deie

    2015-01-01

    Full Text Available Background. Posterior cruciate ligament (PCL injuries are not rare in acute knee injuries, and several recent anatomical studies of the PCL and reconstructive surgical techniques have generated improved patient results. Now, we have evaluated PCL reconstructions performed by either the single-bundle or double-bundle technique in a patient group followed up retrospectively for more than 10 years. Methods. PCL reconstructions were conducted using the single-bundle (27 cases or double-bundle (13 cases method from 1999 to 2002. The mean age at surgery was 34 years in the single-bundle group and 32 years in the double-bundle group. The mean follow-up period was 12.5 years. Patients were evaluated by Lysholm scoring, the gravity sag view, and knee arthrometry. Results. The Lysholm score after surgery was 89.1±5.6 points for the single-bundle group and 91.9±4.5 points for the double-bundle group. There was no significant difference between the methods in the side-to-side differences by gravity sag view or knee arthrometer evaluation, although several cases in both groups showed a side-to-side difference exceeding 5 mm by the latter evaluation method. Conclusions. We found no significant difference between single- and double-bundle PCL reconstructions during more than 10 years of follow-up.

  19. Evolution of the MACSTOR{trademark} dry spent fuel storage system

    Energy Technology Data Exchange (ETDEWEB)

    Pare, F.E.; Joubert, W.M. [Atomic Energy of Canada Ltd., Montreal, Quebec (Canada)

    1995-12-31

    The MACSTOR{trademark} (Modular Air-Cooled Canister Storage) system was developed by Atomic Energy of Canada Limited (AECL) for the interim storage of spent fuel discharged by light water reactors. It is a hybrid system which combines the operational economies of metal cask technology with the capital economies of concrete technology. The system includes all the necessary equipment to transfer spent fuel from a storage pool to an independent interim dry spent fuel storage site. After presenting a description of the system and a brief history of its development, the paper addresses its thermal performance and modeling for various design configurations. Finally, a brief summary of the experience being gained during the implementation of a MACSTOR{trademark} system modified for CANDU spent fuel at the Gentilly-2 NPP in Quebec is presented. It includes progress made in licensing activities and in public hearings pertinent to the initiation of the project.

  20. On the nuclear oxide fuel densification, swelling and thermal re-sintering

    Energy Technology Data Exchange (ETDEWEB)

    Paraschiv, M.C. E-mail: mariusparaschiv@mail.rtns.ro; Paraschiv, A.; Grecu, V.V

    2002-04-01

    A method of calculating the fuel densification and swelling as a result of only the initial fuel porosity has been developed. Well-characterized UO{sub 2} fuel pellets, from the Romanian CANDU type fuel fabrication route, were used to fit the specific pore volume distribution resulted from the sintering process to a lognormal distribution function and thermal re-sintering tests up to 280 h at 1700 deg. C were done to fit the coefficients of vacancies diffusion and the self-diffusion of uranium, required by the model. Careful analyses proved that the irradiation-induced resolution is strongly dependent of the hydrostatic pressure and a corrected formula has been proposed. Analyses of the time evolution of the pore size distribution during re-sintering tests and annealing tests similar with some as described in the open literature were also done.

  1. Coherent hollow-core waveguide bundles for thermal imaging.

    Science.gov (United States)

    Gal, Udi; Harrington, James; Ben-David, Moshe; Bledt, Carlos; Syzonenko, Nicholas; Gannot, Israel

    2010-09-01

    There has been very little work done in the past to extend the wavelength range of fiber image bundles to the IR range. This is due, in part, to the lack of IR transmissive fibers with optical and mechanical properties analogous to the oxide glass fibers currently employed in the visible fiber bundles. Our research is aimed at developing high-resolution hollow-core coherent IR fiber bundles for transendoscopic infrared imaging. We employ the hollow glass waveguide (HGW) technology that was used successfully to make single-HGWs with Ag/AgI thin film coatings to form coherent bundles for IR imaging. We examine the possibility of developing endoscopic systems to capture thermal images using hollow waveguide fiber bundles adjusted to the 8-10?mum spectral range and investigate the applicability of such systems. We carried out a series of measurements in order to characterize the optical properties of the fiber bundles. These included the attenuation, resolution, and temperature response. We developed theoretical models and simulation tools that calculate the light propagation through HGW bundles, and which can be used to calculate the optical properties of the fiber bundles. Finally, the HGW fiber bundles were used to transmit thermal images of various heated objects; the results were compared with simulation results. The experimental results are encouraging, show an improvement in the resolution and thermal response of the HGW fiber bundles, and are consistent with the theoretical results. Nonetheless, additional improvements in the attenuation of the bundles are required in order to be able to use this technology for medical applications.

  2. CFD Verification of 5x5 Rod Bundle with Mixing Vane Spacer Grids

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sungkew; Jang, Hyungwook; Lim, Jongseon; Park, Eungjun; Nahm, Keeyil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Results of the CHF test are used for determining the CHF correlation, which is used to evaluate the thermal margin in the reactor core. Computational fluid dynamics (CFD) has been used to save the time and cost for experimental tests, components design and complicated phenomena in all industries including the reactor coolant system. L. D. Smith et al. applied the CFD methodology in a 5x5 rod bundle with the mixing vane spacer grid using the renormalization group (RNG) k-epsilon model. This CFD model agreed reasonably well with the test data. M. E. Conner et al. conducted experiments to validate the CFD methodology for the single-phase flow conditions in PWR fuel assemblies. In this validation case, the CFD code predicted very similar flow field structures as the test data. In this study, a CFD simulation under single-phase flow condition was conducted for one specific condition in a thermal mixing flow test of 5x5 rod bundle with some mixing vane spacer grids. In this study, a CFD simulation under a single-phase flow condition was conducted for one specific condition in a thermal mixing flow test of 5x5 rod bundle with the mixing vane spacer grids to verify the applicability of the CFD model for predicting the outlet temperature distribution. FLUENT 14.5 Version was used in this CFD analysis. For the successful prediction of the wall bounded turbulent flows, the y+ with 3 prism layers was determined within 5. At this time, k-epsilon standard turbulence model was used. The temperature distribution of CFD for each sub-channel at the outlet region of test bundle showed the difference approximately within 1.1% and 0.2% while comparing to that of test and sub-channel analysis code, respectively.

  3. On the general elephant conjecture for Mori conic bundles

    CERN Document Server

    Prokhorov, Yu G

    1996-01-01

    Let $f:X\\to S$ be an extremal contraction from a threefolds with terminal singularities onto a surface (so called Mori conic bundle). We study some particular cases of such contractions: quotients of usual conic bundles and index two contractions. Assuming Reid's general elephants conjecture we also obtain a rough classification. We present many examples.

  4. Lexical Bundles in L1 and L2 Academic Writing

    Science.gov (United States)

    Chen, Yu-Hua; Baker, Paul

    2010-01-01

    This paper adopts an automated frequency-driven approach to identify frequently-used word combinations (i.e., "lexical bundles") in academic writing. Lexical bundles retrieved from one corpus of published academic texts and two corpora of student academic writing (one L1, the other L2), were investigated both quantitatively and qualitatively.…

  5. An integral Riemann-Roch theorem for surface bundles

    DEFF Research Database (Denmark)

    Madsen, Ib Henning

    2010-01-01

    This paper is a response to a conjecture by T. Akita about an integral Riemann–Roch theorem for surface bundles.......This paper is a response to a conjecture by T. Akita about an integral Riemann–Roch theorem for surface bundles....

  6. Parabolic stable Higgs bundles over complete noncompact Riemann surfaces

    Institute of Scientific and Technical Information of China (English)

    李嘉禹; 王友德

    1999-01-01

    Let M be an open Riemann surface with a finite set of punctures, a complete Poincar(?)-like metric is introduced near the punctures and the equivalence between the stability of an indecomposable parabolic Higgs bundle, and the existence of a Hermitian-Einstein metric on the bundle is established.

  7. Moduli of Parabolic Higgs Bundles and Atiyah Algebroids

    DEFF Research Database (Denmark)

    Logares, Marina; Martens, Johan

    2010-01-01

    In this paper we study the geometry of the moduli space of (non-strongly) parabolic Higgs bundles over a Riemann surface with marked points. We show that this space possesses a Poisson structure, extending the one on the dual of an Atiyah algebroid over the moduli space of parabolic vector bundles...

  8. On the Classification of Complex Vector Bundles of Stable Rank

    Indian Academy of Sciences (India)

    Constantin Bǎnicǎ; Mihai Putinar

    2006-08-01

    One describes, using a detailed analysis of Atiyah–Hirzebruch spectral sequence, the tuples of cohomology classes on a compact, complex manifold, corresponding to the Chern classes of a complex vector bundle of stable rank. This classification becomes more effective on generalized flag manifolds, where the Lie algebra formalism and concrete integrability conditions describe in constructive terms the Chern classes of a vector bundle.

  9. Presenting Lexical Bundles for Explicit Noticing with Schematic Linguistic Representation

    Science.gov (United States)

    Thomson, Haidee Elizabeth

    2016-01-01

    Lexical bundles are essential for fluency, but their incompleteness is a stumbling block for learners. In this study, two presentation methods to increase awareness of lexical bundles through explicit noticing are explored and compared with incidental exposure. The three conditions in this study were as follows: noticing with schematic linguistic…

  10. Smooth Bundling of Large Streaming and Sequence Graphs

    NARCIS (Netherlands)

    Hurter, C.; Ersoy, O.; Telea, A.

    2013-01-01

    Dynamic graphs are increasingly pervasive in modern information systems. However, understanding how a graph changes in time is difficult. We present here two techniques for simplified visualization of dynamic graphs using edge bundles. The first technique uses a recent image-based graph bundling met

  11. Helical twist controls the thickness of F-actin bundles

    NARCIS (Netherlands)

    Claessens, M.M.A.E.; Semmrich, C.; Ramos, L.; Bausch, A.R.

    2008-01-01

    In the presence of condensing agents such as nonadsorbing polymer, multivalent counter ions, and specific bundling proteins, chiral biopolymers typically form bundles with a finite thickness, rather than phase-separating into a polymer-rich phase. Although short-range repulsive interactions or geome

  12. Subanalytic Bundles and Tubular Neighbourhoods of Zero-Loci

    Indian Academy of Sciences (India)

    Vishwambhar Pati

    2003-08-01

    We introduce the natural and fairly general notion of a subanalytic bundle (with a finite dimensional vector space of sections) on a subanalytic subset of a real analytic manifold , and prove that when is compact, there is a Baire subset of sections in whose zero-loci in have tubular neighbourhoods, homeomorphic to the restriction of the given bundle to these zero-loci.

  13. Non-abelian higher gauge theory and categorical bundle

    CERN Document Server

    Viennot, David

    2012-01-01

    A gauge theory is associated with a principal bundle endowed with a connection permitting to define horizontal lifts of paths. The horizontal lifts of surfaces cannot be defined into a principal bundle structure. An higher gauge theory is an attempt to generalize the bundle structure in order to describe horizontal lifts of surfaces. A such attempt is particularly difficult for the non-abelian case. Some structures have been proposed to realize this goal (twisted bundle, gerbes with connection, bundle gerbe, 2-bundle). Each of them uses a category in place of the total space manifold of the usual principal bundle structure. Some of them replace also the structure group by a category (more precisely a Lie crossed module viewed as a category). But the base space remains still a simple manifold (possibly viewed as a trivial category with only identity arrows). We propose a new principal categorical bundle structure, with a Lie crossed module as structure groupoid, but with a base space belonging to a bigger clas...

  14. Effects of sleeve blockages on axial velocity and intensity of turbulence in an unheated 7 x 7 rod bundle. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Creer, J.M.; Rowe, D.S.; Bates, J.M.; Sutey, A.M.

    1976-01-01

    An experimental study is described which was performed to investigate the turbulent flow phenomena near postulated sleeve blockages in a model nuclear fuel rod bundle. The sleeve blockages were characteristic of fuel clad ''swelling'' or ''ballooning'' which could occur during loss-of-coolant accidents (LOCA) in pressurized water reactors. The study was conducted to provide information relative to the flow phenomena near postulated blockages to support detailed safety analyses of LOCAs. The results of the study are especially useful for verification of the hydraulic treatment of reactor core computer programs such as COBRA.

  15. On The Motive of G-bundles

    CERN Document Server

    Habibi, Somayeh

    2011-01-01

    Let $G$ be a reductive algebraic group over a perfect field $k$ and $\\mathcal{G}$ a $G$-bundle over a scheme $X/k$. The main aim of this article is to study the motive associated with $\\mathcal{G}$, inside the Veovodsky Motivic categories. We consider the case that $\\charakt k=0$ (resp. $\\charakt k\\geq 0$), the motive associated to $X$ is geometrically mixed Tate (resp. geometrically cellular) and $\\mathcal{G}$ is locally trivial for the Zariski (resp. \\'etale) topology on $X$ and show that the motive of $\\mathcal{G}$ is a geometrically mixed Tate motive. Moreover for a general $X$ we construct a filtration on the motive associated to $\\mathcal{G}$ in terms of weight polytopes. Along the way we give some applications and examples.

  16. Bundled automobile insurance coverage and accidents.

    Science.gov (United States)

    Li, Chu-Shiu; Liu, Chwen-Chi; Peng, Sheng-Chang

    2013-01-01

    This paper investigates the characteristics of automobile accidents by taking into account two types of automobile insurance coverage: comprehensive vehicle physical damage insurance and voluntary third-party liability insurance. By using a unique data set in the Taiwanese automobile insurance market, we explore the bundled automobile insurance coverage and the occurrence of claims. It is shown that vehicle physical damage insurance is the major automobile coverage and affects the decision to purchase voluntary liability insurance coverage as a complement. Moreover, policyholders with high vehicle physical damage insurance coverage have a significantly higher probability of filing vehicle damage claims, and if they additionally purchase low voluntary liability insurance coverage, their accident claims probability is higher than those who purchase high voluntary liability insurance coverage. Our empirical results reveal that additional automobile insurance coverage information can capture more driver characteristics and driving behaviors to provide useful information for insurers' underwriting policies and to help analyze the occurrence of automobile accidents.

  17. Bundles over Quantum RealWeighted Projective Spaces

    Directory of Open Access Journals (Sweden)

    Tomasz Brzeziński

    2012-09-01

    Full Text Available The algebraic approach to bundles in non-commutative geometry and the definition of quantum real weighted projective spaces are reviewed. Principal U(1-bundles over quantum real weighted projective spaces are constructed. As the spaces in question fall into two separate classes, the negative or odd class that generalises quantum real projective planes and the positive or even class that generalises the quantum disc, so do the constructed principal bundles. In the negative case the principal bundle is proven to be non-trivial and associated projective modules are described. In the positive case the principal bundles turn out to be trivial, and so all the associated modules are free. It is also shown that the circle (coactions on the quantum Seifert manifold that define quantum real weighted projective spaces are almost free.

  18. The 2-Hilbert Space of a Prequantum Bundle Gerbe

    CERN Document Server

    Bunk, Severin; Szabo, Richard J

    2016-01-01

    We construct a prequantum 2-Hilbert space for any line bundle gerbe whose Dixmier-Douady class is torsion. Analogously to usual prequantisation, this 2-Hilbert space has the category of sections of the line bundle gerbe as its underlying 2-vector space. These sections are obtained as certain morphism categories in Waldorf's version of the 2-category of line bundle gerbes. We show that these morphism categories carry a monoidal structure under which they are semisimple and abelian. We introduce a dual functor on the sections, which yields a closed structure on the morphisms between bundle gerbes and turns the category of sections into a 2-Hilbert space. We discuss how these 2-Hilbert spaces fit various expectations from higher prequantisation. We then extend the transgression functor to the full 2-category of bundle gerbes and demonstrate its compatibility with the additional structures introduced. We discuss various aspects of Kostant-Souriau prequantisation in this setting, including its dimensional reductio...

  19. [Bundle-branch block depending on the heart rate].

    Science.gov (United States)

    Apostolov, L

    1975-01-01

    Five patients are reported, admitted to the hospital, with diseases predominantly of the cardio-vascular system. During the electrocardiographic examinations bundle branch block was established, depending on heart rate. It fluctuated within the physiological limits from 50 to 90/min. In three of the patients, the bundle branch block appeared with the quickening of the heart rate (tachycardia-depending bundle branch block) and in two of the patients--the bundle branch block appeared during the slowing down of the heart action and disappeared with its quickening (bradicardia-depending bundle branch block). A brief literature review is presented and attention is paid to the possible diagnostic errors and the treatment mode of those patients with cardiac tonic and antiarrhythmic medicaments.

  20. Voltage- and calcium-dependent motility of saccular hair bundles

    Science.gov (United States)

    Quiñones, Patricia M.; Meenderink, Sebastiaan W. F.; Bozovic, Dolores

    2015-12-01

    Active bundle motility, which is hypothesized to supply feedback for mechanical amplification of signals, is thought to enhance sensitivity and sharpen tuning in vestibular and auditory organs. To study active hair bundle motility, we combined high-speed camera recordings of bullfrog sacculi, which were mounted in a two-compartment chamber, and voltage-clamp of the hair cell membrane potential. Using this paradigm, we measured three types of bundle motions: 1) spontaneous oscillations which can be analyzed to measure the physiological operating range of the transduction channel; 2) a sustained quasi-static movement of the bundle that depends on membrane potential; and 3) a fast, transient and asymmetric movement that resets the bundle position and depends on changes in the membrane potential. These data support a role for both calcium and voltage in the transduction-channel function.

  1. A geometric approach to noncommutative principal torus bundles

    DEFF Research Database (Denmark)

    Wagner, Stefan

    2013-01-01

    for noncommutative algebras and say that a dynamical system (A, 핋n,α) is called a noncommutative principal 핋n-bundle, if localization leads to a trivial noncommutative principal 핋n-bundle. We prove that this approach extends the classical theory of principal torus bundles and present a bunch of (nontrivial......A (smooth) dynamical system with transformation group 핋n is a triple (A, 핋n,α), consisting of a unital locally convex algebra A, the n-torus 핋n and a group homomorphism α:핋n→Aut(A), which induces a (smooth) continuous action of 핋n on A. In this paper, we present a new, geometrically oriented...... approach to the noncommutative geometry of principal torus bundles based on such dynamical systems. Our approach is inspired by the classical setting: In fact, after recalling the definition of a trivial noncommutative principal torus bundle, we introduce a convenient (smooth) localization method...

  2. Morse theory for the space of Higgs G-bundles

    CERN Document Server

    Biswas, Indranil

    2010-01-01

    Fix a $C^\\infty$ principal $G$--bundle $E^0_G$ on a compact connected Riemann surface $X$, where $G$ is a connected complex reductive linear algebraic group. We consider the gradient flow of the Yang--Mills--Higgs functional on the cotangent bundle of the space of all smooth connections on $E^0_G$. We prove that this flow preserves the subset of Higgs $G$--bundles, and, furthermore, the flow emanating from any point of this subset has a limit. Given a Higgs $G$--bundle, we identify the limit point of the integral curve passing through it. These generalize the results of the second named author on Higgs vector bundles.

  3. HORIZONTAL LAPLACE OPERATOR IN REAL FINSLER VECTOR BUNDLES

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    A vector bundle F over the tangent bundle TM of a manifold M is said to be a Finsler vector bundle if it is isomorphic to the pull-back π*E of a vector bundle E over M([1]). In this article the authors study the h-Laplace operator in Finsler vector bundles.An h-Laplace operator is defined, first for functions and then for horizontal Finsler forms on E. Using the h-Laplace operator, the authors define the h-harmonic function and h-harmonic horizontal Finsler vector fields, and furthermore prove some integral formulas for the h-Laplace operator, horizontal Finsler vector fields, and scalar fields on E.

  4. Artificial ciliary bundles with nano fiber tip links

    CERN Document Server

    Asadnia, Mohsen; Miao, Jianmin; Triantafyllou, Michael

    2015-01-01

    Mechanosensory ciliary bundles in fishes are the inspiration for carefully engineered artificial flow sensors. We report the development of a new class of ultrasensitive MEMS flow sensors that mimic the intricate morphology of the ciliary bundles, including the stereocilia, tip links, and the cupula, and thereby achieve threshold detection limits that match the biological example. An artificial ciliary bundle is achieved by fabricating closely-spaced arrays of polymer micro-pillars with gradiating heights. Tip links that form the fundamental sensing elements are realized through electrospinning aligned PVDF piezoelectric nano-fibers that link the distal tips of the polymer cilia. An optimized synthesis of hyaluronic acid-methacrylic anhydride hydrogel that results in properties close to the biological cupula, together with drop-casting method are used to form the artificial cupula that encapsulates the ciliary bundle. In testing, fluid drag force causes the ciliary bundle to slide, stretching the flexible nan...

  5. Superconductivity in an Inhomogeneous Bundle of Metallic and Semiconducting Nanotubes

    Directory of Open Access Journals (Sweden)

    Ilya Grigorenko

    2013-01-01

    Full Text Available Using Bogoliubov-de Gennes formalism for inhomogeneous systems, we have studied superconducting properties of a bundle of packed carbon nanotubes, making a triangular lattice in the bundle's transverse cross-section. The bundle consists of a mixture of metallic and doped semiconducting nanotubes, which have different critical transition temperatures. We investigate how a spatially averaged superconducting order parameter and the critical transition temperature depend on the fraction of the doped semiconducting carbon nanotubes in the bundle. Our simulations suggest that the superconductivity in the bundle will be suppressed when the fraction of the doped semiconducting carbon nanotubes will be less than 0.5, which is the percolation threshold for a two-dimensional triangular lattice.

  6. Fiber bundle model under fluid pressure

    Science.gov (United States)

    Amitrano, David; Girard, Lucas

    2016-03-01

    Internal fluid pressure often plays an important role in the rupture of brittle materials. This is a major concern for many engineering applications and for natural hazards. More specifically, the mechanisms through which fluid pressure, applied at a microscale, can enhance the failure at a macroscale and accelerate damage dynamics leading to failure remains unclear. Here we revisit the fiber bundle model by accounting for the effect of fluid under pressure that contributes to the global load supported by the fiber bundle. Fluid pressure is applied on the broken fibers, following Biot's theory. The statistical properties of damage avalanches and their evolution toward macrofailure are analyzed for a wide range of fluid pressures. The macroscopic strength of the new model appears to be strongly controlled by the action of the fluid, particularly when the fluid pressure becomes comparable with the fiber strength. The behavior remains consistent with continuous transition, i.e., second order, including for large pressure. The main change concerns the damage acceleration toward the failure that is well modeled by the concept of sweeping of an instability. When pressure is increased, the exponent β characterizing the power-law distribution avalanche sizes significantly decreases and the exponent γ characterizing the cutoff divergence when failure is approached significantly increases. This proves that fluid pressure plays a key role in failure process acting as destabilization factor. This indicates that macrofailure occurs more readily under fluid pressure, with a behavior that becomes progressively unstable as fluid pressure increases. This may have considerable consequences on our ability to forecast failure when fluid pressure is acting.

  7. Fuel enrichment and temperature distribution in nuclear fuel rod in (D-T) driven hybrid reactor system

    Energy Technology Data Exchange (ETDEWEB)

    Osman, Ypek [Suleyman Demirel Universitesi Muhendislik-Mimarlyk Fakultesi, Isparta (Turkey)

    2001-07-01

    In this study, melting point of the fuel rod and temperature distribution in nuclear fuel rod are investigated for different coolants under various first wall loads (P{sub w}, =5, 6, 7, 8, 9, and 10 MWm{sup -2}) in Fusion-Fission reactor fueled with 50%LWR +50%CANDU. The fusion source of neutrons of 14.1 MeV is simulated by a movable target along the main axis of cylindrical geometry as a line source. In addition, the fusion chamber was thought as a cylindrical cavity with a diameter of 300 cm that is comparatively small value. The fissile fuel zone is considered to be cooled with four different coolants, gas, flibe (Li{sub 2}BeF{sub 4}), natural lithium (Li), and eutectic lithium (Li{sub 17}Pb{sub 83}). Investigations are observed during 4 years for discrete time intervals of{delta}t= 0.5 month and by a plant factor (PF) of 75%. Volumetric ratio of coolant-to fuel is 1:1, 45.515% coolant, 45.515% fuel, 8.971% clad, in fuel zone. (author)

  8. Literature search on Light Water Reactor (LWR) fuel and absorber rod fabrication, 1960--1976

    Energy Technology Data Exchange (ETDEWEB)

    Sample, C R [comp.

    1977-02-01

    A literature search was conducted to provide information supporting the design of a conceptual Light Water Reactor (LWR) Fuel Fabrication plant. Emphasis was placed on fuel processing and pin bundle fabrication, effects of fuel impurities and microstructure on performance and densification, quality assurance, absorber and poison rod fabrication, and fuel pin welding. All data have been taken from publicly available documents, journals, and books. This work was sponsored by the Finishing Processes-Mixed Oxide (MOX) Fuel Fabrication Studies program at HEDL.

  9. Restriction of Preferences to the Set of Consumption Bundles, In a Model with Production and Consumption Bundles

    NARCIS (Netherlands)

    Schalk, S.

    1999-01-01

    In contrast to the neo-classical theory of Arrow and Debreu, a model of a private ownership economy is presented, in which production and consumption bundles are treated separately. Each of the two types of bundles is assumed to establish a con- vex cone. Production technologies can convert producti

  10. Effect of Flow Blockage on the Coolability during Reflood in a 2 × 2 Rod Bundle

    Directory of Open Access Journals (Sweden)

    Kihwan Kim

    2014-01-01

    Full Text Available During the reflood phase of a large-break loss-of-coolant accident (LBLOCA in a pressurized-water reactor (PWR, the fuel rods can be ballooned or rearranged owing to an increase in the temperature and internal pressure of the fuel rods. In this study, an experimental study was performed to understand the thermal behavior and effect of the ballooned region on the coolability using a 2 × 2 rod bundle test facility. The electrically heated rod bundle was used and the ballooning shape of the rods was simulated by superimposing hollow sleeves, which have a 90% blockage ratio. Forced reflood tests were performed to examine the transient two-phase heat transfer behavior for different reflood rates and rod powers. The droplet behaviors were also investigated by measuring the velocity and size of droplets near the blockage region. The results showed that the heat transfer was enhanced in the downstream of the blockage region, owing to the reduced flow area of the subchannel, intensification of turbulence, and deposition of the droplet.

  11. Development of finite element models for the study of ageing effects in CANDU 6 concrete containment buildings

    Energy Technology Data Exchange (ETDEWEB)

    Ding, Y.; Jaffer, S., E-mail: Yuqing.Ding@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2016-06-15

    In nuclear power plants (NPPs), concrete containment buildings (CCBs) provide the final physical barrier against the release of radioactive materials into the environment and protect the nuclear structures housed within the containment building. CCBs have to be maintained to ensure leak tightness and sound structural integrity for the safe operation throughout the life of NPPs. However, the integrity of CCBs may be affected by the ageing of its concrete, post-tensioning cables and reinforcing bars (rebars). Finite element models (FEMs) of CANDU 6 CCBs have been developed using 2 independent finite element programs for the study of the effect of ageing of CCBs. These FEMs have been validated using multiple-source data and have been used for preliminary analyses of the effect of thermal load and ageing degradation on the concrete structure. The modelling assumptions and simplifications, approach, and validation are discussed in this paper. The preliminary analyses for temperature effects and potential applications to the study of ageing degradation in CCBs using the FEMs are briefly introduced. (author)

  12. CONSIDERATIONS FOR THE DEVELOPMENT OF A DEVICE FOR THE DECOMMISSIONING OF THE HORIZONTAL FUEL CHANNELS IN THE CANDU 6 NUCLEAR REACTOR PART 4 - FUEL CHANNEL ASSEMBLY

    Directory of Open Access Journals (Sweden)

    Gabi ROSCA FARTAT

    2014-05-01

    Full Text Available As many nuclear power plants are reaching their end of lifecycle, the decommissioning of these installations has become one of the 21st century’s great challenges. Each project may be managed differently, depending on the country, development policies, financial considerations, and the availability of qualified engineers or specialized companies to handle such projects. The principle objective of decommissioning is to place a facility into such a condition that there is no unacceptable risk from the decommissioned facility to public health and safety of the environment. In order to ensure that at the end of its life the risk from a facility is within acceptable bounds, action is normally required. The overall decommissioning strategy is to deliver a timely, costeffective program while maintaining high standards of safety, security and environmental protection. If facilities were not decommissioned, they could degrade and potentially present an environmental radiological hazard in the future. Simply abandoning or leaving a facility after ceasing operations is not considered to be an acceptable alternative to decommissioning. The final aim of decommissioning is to recover the geographic site to its original condition.

  13. CONSIDERATIONS FOR THE DEVELOPMENT OF A DEVICE FOR THE DECOMMISSIONING OF THE HORIZONTAL FUEL CHANNELS IN THE CANDU 6 NUCLEAR REACTOR PART 5 - FUEL CHANEL DECOMMISSIONING

    Directory of Open Access Journals (Sweden)

    Gabi ROSCA FARTAT

    2014-05-01

    Full Text Available As many nuclear power plants are reaching their end of lifecycle, the decommissioning of these installations has become one of the 21st century’s great challenges. Each project may be managed differently, depending on the country, development policies, financial considerations, and the availability of qualified engineers or specialized companies to handle such projects. The principle objective of decommissioning is to place a facility into such a condition that there is no unacceptable risk from the decommissioned facility to public health and safety of the environment. In order to ensure that at the end of its life the risk from a facility is within acceptable bounds, action is normally required. The overall decommissioning strategy is to deliver a timely, cost-effective program while maintaining high standards of safety, security and environmental protection. If facilities were not decommissioned, they could degrade and potentially present an environmental radiological hazard in the future. Simply abandoning or leaving a facility after ceasing operations is not considered to be an acceptable alternative to decommissioning. The final aim of decommissioning is to recover the geographic site to its original condition.

  14. CONSIDERATIONS FOR THE DEVELOPMENT OF A DEVICE FOR THE DECOMMISSIONING OF THE HORIZONTAL FUEL CHANNELS IN THE CANDU 6 NUCLEAR REACTOR PART 3 - FUEL CHANNEL REFERENCES

    Directory of Open Access Journals (Sweden)

    Gabi ROSCA FARTAT

    2014-05-01

    Full Text Available As many nuclear power plants are reaching their end of lifecycle, the decommissioning of these installations has become one of the 21st century’s great challenges. Each project may be managed differently, depending on the country, development policies, financial considerations, and the availability of qualified engineers or specialized companies to handle such projects. The principle objective of decommissioning is to place a facility into such a condition that there is no unacceptable risk from the decommissioned facility to public health and safety of the environment. In order to ensure that at the end of its life the risk from a facility is within acceptable bounds, action is normally required. The overall decommissioning strategy is to deliver a timely, cost-effective program while maintaining high standards of safety, security and environmental protection. If facilities were not decommissioned, they could degrade and potentially present an environmental radiological hazard in the future. Simply abandoning or leaving a facility after ceasing operations is not considered to be an acceptable alternative to decommissioning. The final aim of decommissioning is to recover the geographic site to its original condition.

  15. CONSIDERATIONS FOR THE DEVELOPMENT OF A DEVICE FOR THE DECOMMISSIONING OF THE HORIZONTAL FUEL CHANNELS IN THE CANDU 6 NUCLEAR REACTOR PART 2 - FUEL CHANNEL PRESENTATION

    Directory of Open Access Journals (Sweden)

    Gabi ROSCA FARTAT

    2014-05-01

    Full Text Available As many nuclear power plants are reaching their end of lifecycle, the decommissioning of these installations has become one of the 21st century’s great challenges. Each project may be managed differently, depending on the country, development policies, financial considerations, and the availability of qualified engineers or specialized companies to handle such projects. The principle objective of decommissioning is to place a facility into such a condition that there is no unacceptable risk from the decommissioned facility to public health and safety of the environment. In order to ensure that at the end of its life the risk from a facility is within acceptable bounds, action is normally required. The overall decommissioning strategy is to deliver a timely, costeffective program while maintaining high standards of safety, security and environmental protection. If facilities were not decommissioned, they could degrade and potentially present an environmental radiological hazard in the future. Simply abandoning or leaving a facility after ceasing operations is not considered to be an acceptable alternative to decommissioning.The final aim of decommissioning is to recover the geographic site to its original condition.

  16. PWR FLECHT SEASET 163-Rod Bundle Flow Blockage Task data report. NRC/EPRI/Westinghouse report No. 13, August-October 1982

    Energy Technology Data Exchange (ETDEWEB)

    Loftus, M J; Hochreiter, L E; McGuire, M F; Valkovic, M M

    1983-10-01

    This report presents data from the 163-Rod Bundle Blow Blockage Task of the Full-Length Emergency Cooling Heat Transfer Systems Effects and Separate Effects Test Program (FLECHT SEASET). The task consisted of forced and gravity reflooding tests utilizing electrical heater rods with a cosine axial power profile to simulate PWR nuclear core fuel rod arrays. These tests were designed to determine effects of flow blockage and flow bypass on reflooding behavior and to aid in the assessment of computational models in predicting the reflooding behavior of flow blockage in rod bundle arrays.

  17. CANDU堆应用RU的PWR/CANDU联合核燃料循环的研究%Study of RU Utilization in CANDU Reactor-an Advanced Nuclear Fuel Cycle of PWR/CANDU Synergism

    Institute of Scientific and Technical Information of China (English)

    霍小东; 谢仲生

    2003-01-01

    对压水堆乏燃料后处理回收铀(RU)在秦山三期CANDU堆中应用的可行性和经济性进行分析.使用ORIGEN2程序,对后处理回收铀在生产后放置不同时间后核素的成份和放射性活度进行了计算.证明RU燃料元件生产的放射性水平是可以接受的.使用DRAGON/DONJON程序对应用RU的秦山三期CANDU堆的时均堆芯和瞬时堆芯校验分析表明:采用简单的2燃耗区,2、4棒束的换料方案能满足最大通道功率、最大棒束功率限制.通过放射性分析和堆芯物理分析可以看出,秦山三期CANDU堆在不改变堆芯结构及运行模式的条件下,从天然铀(NU)燃料过渡到RU燃料是可行的.通过对秦山三期CANDU堆应用RU的经济性分析,可以看出PWR/CANDU联合核燃料循环的策略既可节约铀资源(23%),提高燃料的能量输出(41%),又减少了废燃料的处置量(66%),可大大降低核电成本.

  18. Bundling Actin Filaments From Membranes: Some Novel Players

    Directory of Open Access Journals (Sweden)

    Clément eThomas

    2012-08-01

    Full Text Available Progress in live-cell imaging of the cytoskeleton has significantly extended our knowledge about the organization and dynamics of actin filaments near the plasma membrane of plant cells. Noticeably, two populations of filamentous structures can be distinguished. On the one hand, fine actin filaments which exhibit an extremely dynamic behavior basically characterized by fast polymerization and prolific severing events, a process referred to as actin stochastic dynamics. On the other hand, thick actin bundles which are composed of several filaments and which are comparatively more stable although they constantly remodel as well. There is evidence that the actin cytoskeleton plays critical roles in trafficking and signaling at both the cell cortex and organelle periphery but the exact contribution of actin bundles remains unclear. A common view is that actin bundles provide the long-distance tracks used by myosin motors to deliver their cargo to growing regions and accordingly play a particularly important role in cell polarization. However, several studies support that actin bundles are more than simple passive highways and display multiple and dynamic roles in the regulation of many processes, such as cell elongation, polar auxin transport, stomatal and chloroplast movement, and defense against pathogens. The list of identified plant actin-bundling proteins is ever expanding, supporting that plant cells shape structurally and functionally different actin bundles. Here I review the most recently characterized actin-bundling proteins, with a particular focus on those potentially relevant to membrane trafficking and/or signaling.

  19. Two-categorical bundles and their classifying spaces

    DEFF Research Database (Denmark)

    Baas, Nils A.; Bökstedt, M.; Kro, T.A.

    2012-01-01

    For a 2-category 2C we associate a notion of a principal 2C-bundle. In case of the 2-category of 2-vector spaces in the sense of M.M. Kapranov and V.A. Voevodsky this gives the the 2-vector bundles of N.A. Baas, B.I. Dundas and J. Rognes. Our main result says that the geometric nerve of a good 2......-category is a classifying space for the associated principal 2-bundles. In the process of proving this we develop a lot of powerful machinery which may be useful in further studies of 2-categorical topology. As a corollary we get a new proof of the classification of principal bundles. A calculation based...... on the main theorem shows that the principal 2-bundles associated to the 2-category of 2-vector spaces in the sense of J.C. Baez and A.S. Crans split, up to concordance, as two copies of ordinary vector bundles. When 2C is a cobordism type 2-category we get a new notion of cobordism-bundles which turns out...

  20. A secondary fuel removal process: plasma processing

    Energy Technology Data Exchange (ETDEWEB)

    Min, J. Y.; Kim, Y. S. [Hanyang Univ., Seoul (Korea, Republic of); Bae, K. K.; Yang, M. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-07-01

    Plasma etching process of UO{sub 2} by using fluorine containing gas plasma is studied as a secondary fuel removal process for DUPIC (Direct Use of PWR spent fuel Into Candu) process which is taken into consideration for potential future fuel cycle in Korea. CF{sub 4}/O{sub 2} gas mixture is chosen for reactant gas and the etching rates of UO{sub 2} by the gas plasma are investigated as functions of CF{sub 4}/O{sub 2} ratio, plasma power, substrate temperature, and plasma gas pressure. It is found that the optimum CF{sub 4}/O{sub 2} ratio is around 4:1 at all temperatures up to 400 deg C and the etching rate increases with increasing r.f. power and substrate temperature. Under 150W r.f. power the etching rate reaches 1100 monolayers/min at 400 deg C, which is equivalent to about 0.5mm/min. (author).

  1. Development of equipment for fabricating DUPIC fuel powder

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ki Ho; Yang, M. S.; Park, J. J.; Lee, J. W.; Kim, J. H.; Cho, K. H.; Lee, D. Y.; Lee, Y. S.; Na, S. H

    1999-06-01

    The powder fabrication processes, as the first stage of manufacturing DUPIC (Direct Use of PWR spent fuel In CANDU) fuel, consist of the slitting of spent PWR fuel rods, REOX (Oxidation and REduction of Oxide Fuels) processing to produce the powder feedstock, the milling of the produced powder, the granulation of the milled powder, and the mixing of the granulated powder with pressing lubricants. All these processes should be conducted by remote means in a hot-cell environment where the direct human access is limited to the strictest minimum due to the high radioactivity. This report describe the development of the equipment for fabricating DUPIC fuel powder. These equipment are Slitting Machine, Oxidation and Reduction (OREOX) Furnace, Mill, Roll Compactor, and Mixer. Remote design concept was applied to all the equipment for use in the M6 hot-cell of the IMEF. Mechanical design considerations and capabilities of the equipment for remote operation and maintenance are presented. First prototypes were developed and installed in the DUPIC full scale mock-up and tested using a master-slave manipulator. Redesign and reconstruction were made on each equipment based on mock-up test results. The remote technology acquired through this research was utilized in developing other equipment for DUPIC fuel fabrication, thereby improving safety and increasing productivity. This technology could also be extended to the area of remote handling equipment development for use in hazardous environments. (author). 14 refs., 9 tabs., 21 figs.

  2. —Impact of Customer Knowledge Heterogeneity on Bundling Strategy

    OpenAIRE

    Amiya Basu; Padmal Vitharana

    2009-01-01

    We consider a marketer of components who can select one of three alternative pricing strategies: (1) a pure component strategy (i.e., the customer can only buy the components individually), (2) a pure bundling strategy (i.e., the components must be purchased together), or (3) a mixed bundling strategy (i.e., the customer may buy a component individually, or buy the bundle). We consider a market where customer knowledge of components varies and propose that a high-knowledge customer can determ...

  3. Systematic evaluation of bundled SPC water for biomolecular simulations.

    Science.gov (United States)

    Gopal, Srinivasa M; Kuhn, Alexander B; Schäfer, Lars V

    2015-04-07

    In bundled SPC water models, the relative motion of groups of four water molecules is restrained by distance-dependent potentials. Bundled SPC models have been used in hybrid all-atom/coarse-grained (AA/CG) multiscale simulations, since they enable to couple atomistic SPC water with supra-molecular CG water models that effectively represent more than a single water molecule. In the present work, we systematically validated and critically tested bundled SPC water models as solvent for biomolecular simulations. To that aim, we investigated both thermodynamic and structural properties of various biomolecular systems through molecular dynamics (MD) simulations. Potentials of mean force of dimerization of pairs of amino acid side chains as well as hydration free energies of single side chains obtained with bundled SPC and standard (unrestrained) SPC water agree closely with each other and with experimental data. Decomposition of the hydration free energies into enthalpic and entropic contributions reveals that in bundled SPC, this favorable agreement of the free energies is due to a larger degree of error compensation between hydration enthalpy and entropy. The Ramachandran maps of Ala3, Ala5, and Ala7 peptides are similar in bundled and unrestrained SPC, whereas for the (GS)2 peptide, bundled water leads to a slight overpopulation of extended conformations. Analysis of the end-to-end distance autocorrelation times of the Ala5 and (GS)2 peptides shows that sampling in more viscous bundled SPC water is about two times slower. Pronounced differences between the water models were found for the structure of a coiled-coil dimer, which is instable in bundled SPC but not in standard SPC. In addition, the hydration of the active site of the serine protease α-chymotrypsin depends on the water model. Bundled SPC leads to an increased hydration of the active site region, more hydrogen bonds between water and catalytic triad residues, and a significantly slower exchange of water

  4. The Born rule as structure of spectral bundles (extended abstract

    Directory of Open Access Journals (Sweden)

    Bertfried Fauser

    2012-10-01

    Full Text Available Topos approaches to quantum foundations are described in a unified way by means of spectral bundles, where the base space is a space of contexts and each fibre is its spectrum. Differences in variance are due to the bundle being a fibration or opfibration. Relative to this structure, the probabilistic predictions of the Born rule in finite dimensional settings are then described as a section of a bundle of valuations. The construction uses in an essential way the geometric nature of the valuation locale monad.

  5. Bondage Numbers of C4 Bundles over a Cycle Cn

    Directory of Open Access Journals (Sweden)

    Moo Young Sohn

    2013-01-01

    Full Text Available Graph bundles generalize the notion of covering graphs and graph products. Graph bundles have been applied in computer architecture and communication networks. The bondage number is an important parameter for measuring the vulnerability and stability of the network domination under link failure. The bondage number b(G of a graph G is the minimum number of edges whose removal enlarges the domination number. In this paper, we show that the bondage number of every C4 bundles over a cycle Cn  (n≥4 is equal to 4.

  6. The effect of hair bundle shape on hair bundle hydrodynamics of inner ear hair cells at low and high frequencies.

    Science.gov (United States)

    Shatz, L F

    2000-03-01

    The relationship between size and shape of the hair bundle of a hair cell in the inner ear and its sensitivity at asymptotically high and low frequencies was determined, thereby extending the results of an analysis of hair bundle hydrodynamics in two dimensions (Freeman and Weiss, 1990. Hydrodynamic analysis of a two-dimensional model for micromechanical resonance of free-standing hair bundles. Hear. Res. 48, 37-68) to three dimensions. A hemispheroid was used to represent the hair bundle. The hemispheroid had a number of advantages: it could represent shapes that range from thin, pencil-like shapes, to wide, flat, disk-like shapes. Also analytic methods could be used in the high frequency range to obtain an exact solution to the equations of motion. In the low frequency range, where an approximate solution was found using boundary element methods, the sensitivity of the responses of hair cells was mainly proportional to the cube of the heights of their hair bundles, and at high frequencies, the sensitivity of the hair cells was mainly proportional to the inverse of their heights. An excellent match was obtained between measurements of sensitivity curves in the basillar papilla of the alligator and bobtail lizards and the model's predictions. These results also suggest why hair bundles of hair cells in vestibular organs which are sensitive to low frequencies have ranges of heights that are an order of magnitude larger than the range of heights of hair bundles of hair cells found in auditory organs.

  7. Nonproliferation and safeguards aspects of the DUPIC fuel cycle concept

    Energy Technology Data Exchange (ETDEWEB)

    Persiani, P. K. [Argonne National Lab., IL (United States)

    1997-07-01

    The purpose of the study is to comment on the proliferation characteristic profiles of some of the proposed fuel cycle alternatives to help ensure that nonproliferation concerns are introduced into the early stages of a fuel cycle concept development program, and to perhaps aid in the more effective implementation of the international nonproliferation regime initiative and safeguards systems. Alternative recycle concepts proposed by several countries involve the recycle of spent fuel without the separation of plutonium from uranium and fission products. The concepts are alternatives to either the direct long-term storage deposition of or the purex reprocessing of the spent fuels. The alternate fuel cycle concepts reviewed include: the dry-recycle processes such as the direct use of reconfigured PWR spent fuel assemblies into CANDU reactors(DUPIC); low-decontamination, single-cycle co-extraction of fast reactor fuels in a wet-purex type of reprocessing; and on a limited scale the thorium-uranium fuel cycle. The nonproliferation advantages usually associated with the above non-separation processes are: the highly radioactive spent fuel presents a barrier to the physical diversion of the nuclear material; avoid the need to dissolve and chemically separate the plutonium from the uranium and fission products; and that the spent fuel isotopic quality of the plutonium vector is further degraded. Although the radiation levels and the need for reprocessing may be perceived as barriers to the terrorist or the subnational level of safeguards, the international level of nonproliferation concerns is addressed primarily by material accountancy and verification activities. On the international level of nonproliferation concerns, the non-separation fuel cycle concepts involved have to be evaluated on the bases of the impact the processes may have on nuclear materials accountancy. (author).

  8. A study of bacterial flagellar bundling.

    Science.gov (United States)

    Flores, Heather; Lobaton, Edgar; Méndez-Diez, Stefan; Tlupova, Svetlana; Cortez, Ricardo

    2005-01-01

    Certain bacteria, such as Escherichia coli (E. coli) and Salmonella typhimurium (S. typhimurium), use multiple flagella often concentrated at one end of their bodies to induce locomotion. Each flagellum is formed in a left-handed helix and has a motor at the base that rotates the flagellum in a corkscrew motion. We present a computational model of the flagellar motion and their hydrodynamic interaction. The model is based on the equations of Stokes flow to describe the fluid motion. The elasticity of the flagella is modeled with a network of elastic springs while the motor is represented by a torque at the base of each flagellum. The fluid velocity due to the forces is described by regularized Stokeslets and the velocity due to the torques by the associated regularized rotlets. Their expressions are derived. The model is used to analyze the swimming motion of a single flagellum and of a group of three flagella in close proximity to one another. When all flagellar motors rotate counterclockwise, the hydrodynamic interaction can lead to bundling. We present an analysis of the flow surrounding the flagella. When at least one of the motors changes its direction of rotation, the same initial conditions lead to a tumbling behavior characterized by the separation of the flagella, changes in their orientation, and no net swimming motion. The analysis of the flow provides some intuition for these processes.

  9. Semi-annual status report of the Canadian Nuclear Fuel Waste Management Program, April 1--September 30, 1991

    Energy Technology Data Exchange (ETDEWEB)

    Wright, E.D. [comp.

    1992-02-01

    This report is the eleventh in a series of semi-annual status reports on the research and development program for the safe management and disposal of Canada's nuclear fuel waste. it describes progress achieved in the three major subprograms, engineered systems, natural systems and performance assessment, from 1991 April 1 to September 30. It also gives a brief description of the activities being carried out in preparation for the public and governmental review of the disposal concept. Since 1987, this program has been jointly funded by AECL and Ontario Hydro under the auspices of the CANDU Owners Group (COG).

  10. Computer-Based Procedure Systems: Technical Basis and Human Factors Review Guidance

    Science.gov (United States)

    2000-03-01

    computer-assisted system for fuel reloading while at power was designed for CANDU NPPs (Gertman et al., 1994). AECL has several aids under...for visualizing the interior of a reactor fuel channel to support the removal of stuck fuel bundles. This system is envisioned as a training aid...and vendor documents and correspondence; NRC correspondence and internal memoranda; bulletins and information no- tices; inspection and

  11. Severe accident analysis of a small LOCA accident using MAAP-CANDU support level 2 PSA for the Point Lepreau station refurbishment project

    Energy Technology Data Exchange (ETDEWEB)

    Petoukhov, S.M.; Brown, M.J.; Mathew, P.M. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2012-07-01

    A Level 2 Probabilistic Safety Assessment was performed for the Point Lepreau Generating Station. The MAAP4-CANDU code was used to calculate the progression of postulated severe core damage accidents and fission product releases. Five representative severe core damage accidents were selected: Station Blackout, Small Loss-of-Coolant Accident, Stagnation Feeder Break, Steam Generator Tube Rupture, and Shutdown State Accident. Analysis results for only the reference Small LOCA Accident scenario (which is a very low probability event) are discussed in this paper. (author)

  12. Radionuclide Release after LBLOCA with Loss of Class IV Power Accident in CANDU-6 Plant

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hoon [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2011-10-15

    A large break in a pipe train of a primary heat transport system discharges coolant, which has high energy and large mass, into the containment building. Reactor shutdown and emergency core cooling water will limit the fuel cladding failure, but cannot prevent it entirely. The containment building is the last barrier of radionuclide release to the environment. Containment isolation and pressure suppression by dousing and local air cooler reduce the amount of radionuclide release to the environment. The objective of containment behavior analysis for large break loss of coolant with loss of class IV power accident is to assess the amount of radionuclide release to the ambient atmosphere. Radionuclide release rates in this event, with all safety system available, that is, the containment building is intact, as well as with containment system impairment, are analyzed with GOTHIC and SMART code

  13. Introductory lectures on fibre bundles and topology for physicists

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, G.H.

    1978-05-01

    These lectures may provide useful background material for understanding gauge theories, particularly the nonperturbative effects such as instantons and monopoles. The mathematical language of topology and fibre bundles is introduced.

  14. National Partnership for Maternal Safety: Consensus Bundle on Obstetric Hemorrhage.

    Science.gov (United States)

    Main, Elliott K; Goffman, Dena; Scavone, Barbara M; Low, Lisa Kane; Bingham, Debra; Fontaine, Patricia L; Gorlin, Jed B; Lagrew, David C; Levy, Barbara S

    2015-07-01

    Hemorrhage is the most frequent cause of severe maternal morbidity and preventable maternal mortality and therefore is an ideal topic for the initial national maternity patient safety bundle. These safety bundles outline critical clinical practices that should be implemented in every maternity unit. They are developed by multidisciplinary work groups of the National Partnership for Maternal Safety under the guidance of the Council on Patient Safety in Women's Health Care. The safety bundle is organized into four domains: Readiness, Recognition and Prevention, Response, and Reporting and System Learning. Although the bundle components may be adapted to meet the resources available in individual facilities, standardization within an institution is strongly encouraged. References contain sample resources and "Potential Best Practices" to assist with implementation.

  15. National Partnership for Maternal Safety Consensus Bundle on Obstetric Hemorrhage.

    Science.gov (United States)

    Main, Elliott K; Goffman, Dena; Scavone, Barbara M; Low, Lisa Kane; Bingham, Debra; Fontaine, Patricia L; Gorlin, Jed B; Lagrew, David C; Levy, Barbara S

    2015-01-01

    Hemorrhage is the most frequent cause of severe maternal morbidity and preventable maternal mortality and therefore is an ideal topic for the initial national maternity patient safety bundle. These safety bundles outline critical clinical practices that should be implemented in every maternity unit. They are developed by multidisciplinary work groups of the National Partnership for Maternal Safety under the guidance of the Council on Patient Safety in Women's Health Care. The safety bundle is organized into 4 domains: Readiness, Recognition and Prevention, Response, and Reporting and Systems Learning. Although the bundle components may be adapted to meet the resources available in individual facilities, standardization within an institution is strongly encouraged. References contain sample resources and "Potential Best Practices" to assist with implementation.

  16. Vertical, Bubbly, Cross-Flow Characteristics over Tube Bundles

    Science.gov (United States)

    Iwaki, C.; Cheong, K. H.; Monji, H.; Matsui, G.

    2005-12-01

    Two-phase flow over tube bundles is commonly observed in shell and tube-type heat exchangers. However, only limited amount of data concerning flow pattern and void fraction exists due to the flow complexity and the difficulties in measurement. The detailed flow structure in tube bundles needs to be understood for reliable and effective design. Therefore, the objective of this study was to clarify the two-phase structure of cross-flow in tube bundles by PIV. Experiments were conducted using two types of models, namely in-line and staggered arrays with a pitch-to-diameter ratio of 1.5. Each test section contains 20 rows of five 15 mm O.D. tubes in each row. The experiment’s data were obtained under very low void fraction (αtube bundles were described in terms of the velocity vector field, turbulence intensity and void fraction.

  17. Mechanical Models of Microtubule Bundle Collapse in Alzheimer's Disease

    Science.gov (United States)

    Sendek, Austin; Singh, Rajiv; Cox, Daniel

    2013-03-01

    Amyloid-beta aggregates initiate Alzheimer's disease, and downstream trigger degradation of tau proteins that act as microtubule bundle stabilizers and mechanical spacers. Currently it is unclear which of tau cutting by proteases, tau phosphorylation, or tau aggregation are responsible for cytoskeleton degradation., We construct a percolation simulation of the microtubule bundle using a molecular spring model for the taus and including depletion force attraction between microtubules and membrane/actin cytoskeletal surface tension. The simulation uses a fictive molecular dynamics to model the motion of the individual microtubules within the bundle as a result of random tau removal, and calculates the elastic modulus of the bundle as the tau concentration falls. We link the tau removal steps to kinetic tau steps in various models of tau degradation. Supported by US NSF Grant DMR 1207624

  18. Zeta Functions for Elliptic Curves I. Counting Bundles

    CERN Document Server

    Weng, Lin

    2012-01-01

    To count bundles on curves, we study zetas of elliptic curves and their zeros. There are two types, i.e., the pure non-abelian zetas defined using moduli spaces of semi-stable bundles, and the group zetas defined for special linear groups. In lower ranks, we show that these two types of zetas coincide and satisfy the Riemann Hypothesis. For general cases, exposed is an intrinsic relation on automorphism groups of semi-stable bundles over elliptic curves, the so-called counting miracle. All this, together with Harder-Narasimhan, Desale-Ramanan and Zagier's result, gives an effective way to count semi-stable bundles on elliptic curves not only in terms of automorphism groups but more essentially in terms of their $h^0$'s. Distributions of zeros of high rank zetas are also discussed.

  19. On exact triangles consisting of stable vector bundles on tori

    CERN Document Server

    Kobayashi, Kazushi

    2016-01-01

    In this paper, we consider the exact triangles consisting of stable holomorphic vector bundles on one-dimensional complex tori, and discuss their relations with the corresponding Fukaya category via the homological mirror symmetry.

  20. Steric effects induce geometric remodeling of actin bundles in filopodia

    CERN Document Server

    Dobramysl, Ulrich; Erban, Radek

    2016-01-01

    Filopodia are ubiquitous fingerlike protrusions, spawned by many eukaryotic cells, to probe and interact with their environments. Polymerization dynamics of actin filaments, comprising the structural core of filopodia, largely determine their instantaneous lengths and overall lifetimes. The polymerization reactions at the filopodial tip require transport of G-actin, which enter the filopodial tube from the filopodial base and diffuse toward the filament barbed ends near the tip. Actin filaments are mechanically coupled into a tight bundle by cross-linker proteins. Interestingly, many of these proteins are relatively short, restricting the free diffusion of cytosolic G-actin throughout the bundle and, in particular, its penetration into the bundle core. To investigate the effect of steric restrictions on G-actin diffusion by the porous structure of filopodial actin filament bundle, we used a particle-based stochastic simulation approach. We discovered that excluded volume interactions result in partial and the...

  1. Bundles of Norms About Teen Sex and Pregnancy.

    Science.gov (United States)

    Mollborn, Stefanie; Sennott, Christie

    2015-09-01

    Teen pregnancy is a cultural battleground in struggles over morality, education, and family. At its heart are norms about teen sex, contraception, pregnancy, and abortion. Analyzing 57 interviews with college students, we found that "bundles" of related norms shaped the messages teens hear. Teens did not think their communities encouraged teen sex or pregnancy, but normative messages differed greatly, with either moral or practical rationalizations. Teens readily identified multiple norms intended to regulate teen sex, contraception, abortion, childbearing, and the sanctioning of teen parents. Beyond influencing teens' behavior, norms shaped teenagers' public portrayals and post hoc justifications of their behavior. Although norm bundles are complex to measure, participants could summarize them succinctly. These bundles and their conflicting behavioral prescriptions create space for human agency in negotiating normative pressures. The norm bundles concept has implications for teen pregnancy prevention policies and can help revitalize social norms for understanding health behaviors.

  2. Design and synthesis of DNA four-helix bundles

    Energy Technology Data Exchange (ETDEWEB)

    Rangnekar, Abhijit; Gothelf, Kurt V [Department of Chemistry, Centre for DNA Nanotechnology (CDNA) and Interdisciplinary Nanoscience Center (iNANO), Aarhus University, DK-8000 Aarhus C (Denmark); LaBean, Thomas H, E-mail: kvg@chem.au.dk, E-mail: thl@cs.duke.edu [Department of Chemistry, Duke University, Durham, NC 27708 (United States)

    2011-06-10

    The field of DNA nanotechnology has evolved significantly in the past decade. Researchers have succeeded in synthesizing tile-based structures and using them to form periodic lattices in one, two and three dimensions. Origami-based structures have also been used to create nanoscale structures in two and three dimensions. Design and construction of DNA bundles with fixed circumference has added a new dimension to the field. Here we report the design and synthesis of a DNA four-helix bundle. It was found to be extremely rigid and stable. When several such bundles were assembled using appropriate sticky-ends, they formed micrometre-long filaments. However, when creation of two-dimensional sheet-like arrays of the four-helix bundles was attempted, nanoscale rings were observed instead. The exact reason behind the nanoring formation is yet to be ascertained, but it provides an exciting prospect for making programmable circular nanostructures using DNA.

  3. Performance-based bundled payments: potential benefits and burdens.

    Science.gov (United States)

    Satin, David J; Miles, Justin

    2009-10-01

    Performance-based bundled payments have emerged as the most recent iteration of pay for performance. These are programs in which providers are paid a single fee for a set of evidenced-based services related to a diagnosis. The payments are typically linked to outcomes as well as other quality measures. This paper reviews two prominent bundled payment programs--PROMETHEUS and ProvenCare--and discusses the potential pitfalls of these approaches.

  4. Identity-Based Cryptosystems for Enhanced Deployment of OSGi Bundles

    OpenAIRE

    Parrend, Pierre; Galice, Samuel; Frénot, Stéphane; Ubéda, Stéphane

    2007-01-01

    International audience; The OSGi platform is designed to make Java soft- ware extensible at runtime. This undeniably presents a great interest in several domains like embedded plat- forms or enterprise application servers. However, se- curing the deployment of the OSGi components, or bundles, proves to be a major challenge. The current approach consists in digitally signing the bundles and certifying the signature through a Public Key Infras- tructure. We propose to replace this technology wi...

  5. Dunkl Operators as Covariant Derivatives in a Quantum Principal Bundle

    OpenAIRE

    Micho Đurđevich; Stephen Bruce Sontz

    2011-01-01

    A quantum principal bundle is constructed for every Coxeter group acting on a finite-dimensional Euclidean space $E$, and then a connection is also defined on this bundle. The covariant derivatives associated to this connection are the Dunkl operators, originally introduced as part of a program to generalize harmonic analysis in Euclidean spaces. This gives us a new, geometric way of viewing the Dunkl operators. In particular, we present a new proof of the commutativity of these operators amo...

  6. Simplified modeling of EM field coupling to complex cable bundles

    OpenAIRE

    Schetelig, B.; J. Keghie; Kanyou Nana, R.; Fichte, L.-O.; S. Potthast; Dickmann, S.

    2010-01-01

    In this contribution, the procedure "Equivalent Cable Bundle Method" is used for the simplification of large cable bundles, and it is extended to the application on differential signal lines. The main focus is on the reduction of twisted-pair cables. Furthermore, the process presented here allows to take into account cables with wires that are situated quite close to each other. The procedure is based on a new approach to calculate the geometry of the simplified cable and us...

  7. Generalized holomorphic bundles and the B-field action

    CERN Document Server

    Hitchin, Nigel

    2010-01-01

    On a generalized complex manifold there is an associated definition of a generalized holomorphic bundle, introduced by Gualtieri. This notion in the case of an ordinary complex structure yields an object which we call a co-Higgs bundle and we consider the B-field action of a closed form of type (1,1), both local and global. The effect makes contact with both Nahm's equations and holomorphic gerbes.

  8. Generalized holomorphic bundles and the B-field action

    Science.gov (United States)

    Hitchin, Nigel

    2011-01-01

    On a generalized complex manifold, there is an associated definition of a generalized holomorphic bundle, introduced by Gualtieri. In the case of an ordinary complex structure, this notion yields an object which we call a co-Higgs bundle, and we consider the B-field action of a closed form of type (1,1), both local and global. The effect makes contact with both Nahm's equations and holomorphic gerbes.

  9. A study on framework for assessing limiting condition for operation(LCOs) quantitatively in a CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kyung Min

    2003-02-15

    Limiting conditions for operations define the allowed outage times and the actions to be taken if the repair cannot be completed within the AOT. Typically the actions required are plant shutdown. In situations where the risk associated with the action, i.e., the risk of plant shutdown given a failure in the safety system, may be substantial, a strategy is needed to control the risk implications. In this study the strategy is evaluated quantitatively using a tool of system dynamics. The strategic actions associated with LCO needs to be assessed dynamically. System dynamics technique is easy enough to be applied for quantitative assessment of LCO, where the system dynamics is an important factor in evaluating operational modes of nuclear power plants. The review on the improvements of overly conservative technical specification has been performed in this study. The VENSIM tool has been applied to evaluate quantitatively the LCO of the auxiliary feed water systems in the Wol-Sung nuclear power plant. A value of core damage frequency in PSA is used as risk measure. The analysis of both full power operation and shutdown operation has been compared for the value of the CDF. This study shows a plot of LCO full power operating and shutdown risks in term of core damage probability for failure of AFWS (auxiliary feedwater system). Obtaining a lower risk level in a stable mode, compared to the LCO operation alternative, is the principal motivation of going to full power operational mode or shutdown operational mode. A time dependent framework developed in this study has been applied to assess the LCO of the example problem and it is shown that it is very flexible in that it can be applied to assess LCO quantitatively under any operational context of the technical specifications. This study contributes risk informed regulation to enhance and optimize the LCO of previous technical specifications for CANDU through developing framework in application proposed in this methodology.

  10. BiSet: Semantic Edge Bundling with Biclusters for Sensemaking.

    Science.gov (United States)

    Sun, Maoyuan; Mi, Peng; North, Chris; Ramakrishnan, Naren

    2016-01-01

    Identifying coordinated relationships is an important task in data analytics. For example, an intelligence analyst might want to discover three suspicious people who all visited the same four cities. Existing techniques that display individual relationships, such as between lists of entities, require repetitious manual selection and significant mental aggregation in cluttered visualizations to find coordinated relationships. In this paper, we present BiSet, a visual analytics technique to support interactive exploration of coordinated relationships. In BiSet, we model coordinated relationships as biclusters and algorithmically mine them from a dataset. Then, we visualize the biclusters in context as bundled edges between sets of related entities. Thus, bundles enable analysts to infer task-oriented semantic insights about potentially coordinated activities. We make bundles as first class objects and add a new layer, "in-between", to contain these bundle objects. Based on this, bundles serve to organize entities represented in lists and visually reveal their membership. Users can interact with edge bundles to organize related entities, and vice versa, for sensemaking purposes. With a usage scenario, we demonstrate how BiSet supports the exploration of coordinated relationships in text analytics.

  11. Oscillation of carbon molecules inside carbon nanotube bundles

    Science.gov (United States)

    Thamwattana, Ngamta; Cox, Barry J.; Hill, James M.

    2009-04-01

    In this paper, we investigate the mechanics of a nanoscaled gigahertz oscillator comprising a carbon molecule oscillating within the centre of a uniform concentric ring or bundle of carbon nanotubes. Two kinds of oscillating molecules are considered, which are a carbon nanotube and a C60 fullerene. Using the Lennard-Jones potential and the continuum approach, we obtain a relation between the bundle radius and the radii of the nanotubes forming the bundle, as well as the optimum bundle size which gives rise to the maximum oscillatory frequency for both the nanotube-bundle and the C60-bundle oscillators. While previous studies in this area have been undertaken through molecular dynamics simulations, this paper emphasizes the use of applied mathematical modelling techniques, which provides considerable insight into the underlying mechanisms of the nanoscaled oscillators. The paper presents a synopsis of the major results derived in detail by the present authors (Cox et al 2007 Proc. R. Soc. A 464 691-710 and Cox et al 2007 J. Phys. A: Math. Theor. 40 13197-208).

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

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Dong Keun

    2005-12-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-04-01

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

  15. Preliminary study of CANDU moderator thermal hydraulics using the CUPID code

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Gi; Jeong Jae Jun [Pusan National Univ., Busan (Korea, Republic of); Lee, Jae Ryong; Kim, Hyoung Tae [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    When the moderator cooling system fails, moderator may act as to remove decay heat which occurs in fuel. During loss of coolant accident (LOCA), the film boiling occurs in the Calandria tube (CT) because the hot pressure tube would deform into contacting with the calandria tube. And lower subcooling would decrease the margin of the CT to dryout. So, it is important to estimate a local subcooling of the moderator inside the Calandria vessel. However, in order to predict the internal temperature the study of empirical experiments and calculations are needed because only the inlet/outlet temperature can be measured in real reactor. In this study, the internal flow of the moderator was predicted by using the CUPID code, which has been developed in KAERI. The CUPID adopts three dimensional, transient, two phase and three field model, and includes various physical models and correlations of the interfacial mass, momentum and energy transfer for the closure relations of the two fluid model. The CUPID code shows single phase and two phase flow through two phase flow calculations of virtual can be applied.

  16. Thermal Aspects of Using Alternative Nuclear Fuels in Supercritical Water-Cooled Reactors

    Science.gov (United States)

    Grande, Lisa Christine

    A SuperCritical Water-cooled Nuclear Reactor (SCWR) is a Generation IV concept currently being developed worldwide. Unique to this reactor type is the use of light-water coolant above its critical point. The current research presents a thermal-hydraulic analysis of a single fuel channel within a Pressure Tube (PT)-type SCWR with a single-reheat cycle. Since this reactor is in its early design phase many fuel-channel components are being investigated in various combinations. Analysis inputs are: steam cycle, Axial Heat Flux Profile (AHFP), fuel-bundle geometry, and thermophysical properties of reactor coolant, fuel sheath and fuel. Uniform and non-uniform AHFPs for average channel power were applied to a variety of alternative fuels (mixed oxide, thorium dioxide, uranium dicarbide, uranium nitride and uranium carbide) enclosed in an Inconel-600 43-element bundle. The results depict bulk-fluid, outer-sheath and fuel-centreline temperature profiles together with the Heat Transfer Coefficient (HTC) profiles along the heated length of fuel channel. The objective is to identify the best options in terms of fuel, sheath material and AHFPS in which the outer-sheath and fuel-centreline temperatures will be below the accepted temperature limits of 850°C and 1850°C respectively. The 43-element Inconel-600 fuel bundle is suitable for SCWR use as the sheath-temperature design limit of 850°C was maintained for all analyzed cases at average channel power. Thoria, UC2, UN and UC fuels for all AHFPs are acceptable since the maximum fuel-centreline temperature does not exceed the industry accepted limit of 1850°C. Conversely, the fuel-centreline temperature limit was exceeded for MOX at all AHFPs, and UO2 for both cosine and downstream-skewed cosine AHFPs. Therefore, fuel-bundle modifications are required for UO2 and MOX to be feasible nuclear fuels for SCWRs.

  17. Turbulence Model Evaluation Study for a Secondary Flow and a Flow Pulsation in the Sub-Channels of an 18-Finned Rod Bundle by Using Computational Fluid Dynamics

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Hark; Chae, Hee Taek; Park, Cheol; Kim, Heon Il

    2008-09-15

    Since the heat flux of the rod type fuel used in the HANARO, a research reactor being operated in the KAERI, is substantially higher than the heat flux of power reactors, the HANARO fuel has 8 longitudinal fins for enhancing the heat release from the fuel rod surface. This unique shape of a nuclear fuel led us to study the flows and thermal hydraulic characteristics of it. Especially because the flows through the narrow channels built up by these finned rod fuels would be different from the flow characteristics in the coolant channels formed by bare rod fuels, some experimental studies to investigate the flow behaviors and structures in a finned rod bundle were done by other researchers. But because of the very complex geometries of the flow channels in the finned rod bundle only allowed us to obtain limited information about the flow characteristics, a numerical study by a computational fluid dynamics technique has been adopted to elucidate more about such a complicated flow in a finned rod bundle. In this study, for the development of an adequate computational model to simulate such a complex geometry, a mesh sensitivity study and the effects of various turbulence models were examined. The CFD analysis results were compared with the experimental results. Some of them have a good agreement with the experimental results. All linear eddy viscosity turbulence models could hardly predict the secondary flows near the fuel surfaces and in the sub-channel, but the RSM (Reynolds Stress Model) revealed very different results from the eddy viscosity turbulence models. In the transient analysis all turbulence model predicted flow pulsation at the center of a subchannel as well as at the gap between rods in spite of large P/D. The flow pulsation showed different results with turbulence models and the location in the sub-channels.

  18. External Verification of the Bundle Adjustment in Photogrammetric Software Using the Damped Bundle Adjustment Toolbox

    Science.gov (United States)

    Börlin, Niclas; Grussenmeyer, Pierre

    2016-06-01

    The aim of this paper is to investigate whether the Matlab-based Damped Bundle Adjustment Toolbox (DBAT) can be used to provide independent verification of the BA computation of two popular software—PhotoModeler (PM) and PhotoScan (PS). For frame camera data sets with lens distortion, DBAT is able to reprocess and replicate subsets of PM results with high accuracy. For lens-distortion-free data sets, DBAT can furthermore provide comparative results between PM and PS. Data sets for the discussed projects are available from the authors. The use of an external verification tool such as DBAT will enable users to get an independent verification of the computations of their software. In addition, DBAT can provide computation of quality parameters such as estimated standard deviations, correlation between parameters, etc., something that should be part of best practice for any photogrammetric software. Finally, as the code is free and open-source, users can add computations of their own.

  19. An in vitro biomechanical comparison of anterior cruciate ligament reconstruction: single bundle versus anatomical double bundle techniques

    Directory of Open Access Journals (Sweden)

    Sandra Umeda Sasaki

    2008-01-01

    Full Text Available INTRODUCTION: Anterior cruciate ligament ruptures are frequent, especially in sports. Surgical reconstruction with autologous grafts is widely employed in the international literature. Controversies remain with respect to technique variations as continuous research for improvement takes place. One of these variations is the anatomical double bundle technique, which is performed instead of the conventional single bundle technique. More recently, there has been a tendency towards positioning the two bundles through double bone tunnels in the femur and tibia (anatomical reconstruction. OBJECTIVES: To compare, through biomechanical tests, the practice of anatomical double bundle anterior cruciate ligament reconstruction with a patellar graft to conventional single bundle reconstruction with the same amount of patellar graft in a paired experimental cadaver study. METHODS: Nine pairs of male cadaver knees ranging in age from 44 to 63 years were randomized into two groups: group A (single bundle and group B (anatomical reconstruction. Each knee was biomechanically tested under three conditions: intact anterior cruciate ligament, reconstructed anterior cruciate ligament, and injured anterior cruciate ligament. Maximum anterior dislocation, rigidity, and passive internal tibia rotation were recorded with knees submitted to a 100 N horizontal anterior dislocation force applied to the tibia with the knees at 30, 60 and 90 degrees of flexion. RESULTS: There were no differences between the two techniques for any of the measurements by ANOVA tests. CONCLUSION: The technique of anatomical double bundle reconstruction of the anterior cruciate ligament with bone-patellar tendon-bone graft has a similar biomechanical behavior with regard to anterior tibial dislocation, rigidity, and passive internal tibial rotation.

  20. Pressure Loss across Tube Bundles in Two-phase Flow

    Energy Technology Data Exchange (ETDEWEB)

    Sim, Woo Gun; Banzragch, Dagdan [Hannam Univ., Daejon (Korea, Republic of)

    2016-03-15

    An analytical model was developed by Sim to estimate the two-phase damping ratio for upward two-phase flow perpendicular to horizontal tube bundles. The parameters of two-phase flow, such as void fraction and pressure loss evaluated in the model, were calculated based on existing experimental formulations. However, it is necessary to implement a few improvements in the formulations for the case of tube bundles. For the purpose of the improved formulation, we need more information about the two-phase parameters, which can be found through experimental test. An experiment is performed with a typical normal square array of cylinders subjected to the two-phase flow of air-water in the tube bundles, to calculate the two-phase Euler number and the two-phase friction multiplier. The pitch-to-diameter ratio is 1.35 and the diameter of cylinder is 18mm. Pressure loss along the flow direction in the tube bundles is measured with a pressure transducer and data acquisition system to calculate the two-phase Euler number and the two-phase friction multiplier. The void fraction model by Feenstra et al. is used to estimate the void fraction of the two-phase flow in tube bundles. The experimental results of the two phase friction multiplier and two-phase Euler number for homogeneous and non-homogeneous two-phase flows are compared and evaluated against the analytical results given by Sim's model.