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Sample records for coolant void reactivity

  1. Coolant Void Reactivity Analysis of CANDU Lattice

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  2. Burnup dependence of coolant void reactivity for ACR-1000 cell

    International Nuclear Information System (INIS)

    Le Tellier, R.; Marleau, G.; Hebert, A.; Roubstov, D.; Altiparmakov, D.; Irish, D.

    2008-01-01

    The Advanced Candu Reactor (ACR-1000) is light water cooled, fueled with enriched uranium and has a smaller lattice pitch than the Candu-6. As a result, the neutronic behavior of the ACR-1000 cell is expected to be somewhat different from that of the Candu-6 leading to a negative coolant void reactivity (CVR). Here we evaluate the CVR for the ACR-1000 cell using the lattice code DRAGON and compare our results with those obtained using the code WIMS-AECL. The differences observed between these two codes for the burnup dependence of the CVR is mainly explained in terms of the specific cell leakage model used by both codes. (authors)

  3. Physics study of Canada deuterium uranium lattice with coolant void reactivity analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin Su; Lee, Hyun Suk; Tak, Tae Woo; Lee, Deok Jung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Shin, Ho Cheol [Korea Hydro and Nuclear Power Central Research Institute (KHNP-CRI), Daejeon (Korea, Republic of)

    2017-02-15

    This study presents a coolant void reactivity analysis of Canada Deuterium Uranium (CANDU)-6 and Advanced Canada Deuterium Uranium Reactor-700 (ACR-700) fuel lattices using a Monte Carlo code. The reactivity changes when the coolant was voided were assessed in terms of the contributions of four factors and spectrum shifts. In the case of single bundle coolant voiding, the contribution of each of the four factors in the ACR-700 lattice is large in magnitude with opposite signs, and their summation becomes a negative reactivity effect in contrast to that of the CANDU-6 lattice. Unlike the coolant voiding in a single fuel bundle, the 2 x 2 checkerboard coolant voiding in the ACR-700 lattice shows a positive reactivity effect. The neutron current between the no-void and voided bundles, and the four factors of each bundle were analyzed to figure out the mechanism of the positive coolant void reactivity of the checkerboard voiding case. Through a sensitivity study of fuel enrichment, type of burnable absorber, and moderator to fuel volume ratio, a design strategy for the CANDU reactor was suggested in order to achieve a negative coolant void reactivity even for the checkerboard voiding case.

  4. Effects contributing to positive coolant void reactivity in CANDU

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  5. Models for coolant void reactivity evaluation in Candu Generation II and III+

    International Nuclear Information System (INIS)

    Popov, Alexi V.; Chambon, Richard P.; Le Tellier, Romain; Marleau, Guy; Hebert, Alain

    2008-01-01

    In the simulation of large-break loss-of-coolant accidents, homogenised cross-sections from trans- port calculations are used. These are usually computed in single cells or lattices representative for an infinite repeated pattern. Large coolant accidents in Candu, however, usually exhibit a checkerboard pattern of cooled and voided channels represented by lattices. It is reasonable, therefore, that homogenised cross-sections be produced in assemblies of lattices. This allows simulating the checkerboard voiding pat- tern and more realistically reproducing the lattice boundary conditions. The result is better simulation of the accident and more precise evaluation of coolant-void reactivity. For the present study, homogenised cross-sections are generated in a 2x2 heterogeneous assembly of four lattices for Generation II and III+ Candu designs. Results of reactivity calculations with the reactor code are compared to those using the traditional method. The difference is significant for Generation III+ Candu. (authors)

  6. On the difference between DRAGON and WIMS-AECL calculations of the coolant void reactivity

    International Nuclear Information System (INIS)

    Altiparmakov, D.; Roubtsov, D.; Irish, J.D.

    2009-01-01

    A difference in the shape of the burnup dependence of the coolant void reactivity (CVR) has been observed between DRAGON and WIMS-AECL calculations. This paper discusses the root cause of the difference and assesses the impact on burnup and full-core reactor calculations. A Fortran procedure has been developed to run WIMS-AECL as necessary in order to mimic DRAGON burnup calculations with leakage effects included. The comparison of standard WIMS-AECL results and simulated DRAGON results demonstrated that the difference is due to different definitions of CVR. If the same CVR definition is used, then the results of both WIMS-AECL and DRAGON analyses are essentially indistinguishable. The discrepancies in the fuel composition and cell-averaged two-group cross sections that are due to differences in WIMS-AECL and DRAGON leakage treatments are insignificant. (author)

  7. Coolant void reactivity adjustments in advanced CANDU lattices using adjoint sensitivity technique

    International Nuclear Information System (INIS)

    Assawaroongruengchot, M.; Marleau, G.

    2008-01-01

    Coolant void reactivity (CVR) is an important factor in reactor accident analysis. Here we study the adjustments of CVR at beginning of burnup cycle (BOC) and k eff at end of burnup cycle (EOC) for a 2D Advanced CANDU Reactor (ACR) lattice using the optimization and adjoint sensitivity techniques. The sensitivity coefficients are evaluated using the perturbation theory based on the integral neutron transport equations. The neutron and flux importance transport solutions are obtained by the method of cyclic characteristics (MOCC). Three sets of parameters for CVR-BOC and k eff -EOC adjustments are studied: (1) Dysprosium density in the central pin with Uranium enrichment in the outer fuel rings, (2) Dysprosium density and Uranium enrichment both in the central pin, and (3) the same parameters as in the first case but the objective is to obtain a negative checkerboard CVR-BOC (CBCVR-BOC). To approximate the EOC sensitivity coefficient, we perform constant-power burnup/depletion calculations using a slightly perturbed nuclear library and the unperturbed neutron fluxes to estimate the variation of nuclide densities at EOC. Our aim is to achieve a desired negative CVR-BOC of -2 mk and k eff -EOC of 0.900 for the first two cases, and a CBCVR-BOC of -2 mk and k eff -EOC of 0.900 for the last case. Sensitivity analyses of CVR and eigenvalue are also included in our study

  8. Coolant void reactivity adjustments in advanced CANDU lattices using adjoint sensitivity technique

    Energy Technology Data Exchange (ETDEWEB)

    Assawaroongruengchot, M. [Institut de Genie Nucleaire, Ecole Polytechnique de Montreal, P.O. Box 6079, stn. Centre-ville, Montreal, H3C3A7 (Canada)], E-mail: monchaia@gmail.com; Marleau, G. [Institut de Genie Nucleaire, Ecole Polytechnique de Montreal, P.O. Box 6079, stn. Centre-ville, Montreal, H3C3A7 (Canada)], E-mail: guy.marleau@polymtl.ca

    2008-03-15

    Coolant void reactivity (CVR) is an important factor in reactor accident analysis. Here we study the adjustments of CVR at beginning of burnup cycle (BOC) and k{sub eff} at end of burnup cycle (EOC) for a 2D Advanced CANDU Reactor (ACR) lattice using the optimization and adjoint sensitivity techniques. The sensitivity coefficients are evaluated using the perturbation theory based on the integral neutron transport equations. The neutron and flux importance transport solutions are obtained by the method of cyclic characteristics (MOCC). Three sets of parameters for CVR-BOC and k{sub eff}-EOC adjustments are studied: (1) Dysprosium density in the central pin with Uranium enrichment in the outer fuel rings, (2) Dysprosium density and Uranium enrichment both in the central pin, and (3) the same parameters as in the first case but the objective is to obtain a negative checkerboard CVR-BOC (CBCVR-BOC). To approximate the EOC sensitivity coefficient, we perform constant-power burnup/depletion calculations using a slightly perturbed nuclear library and the unperturbed neutron fluxes to estimate the variation of nuclide densities at EOC. Our aim is to achieve a desired negative CVR-BOC of -2 mk and k{sub eff}-EOC of 0.900 for the first two cases, and a CBCVR-BOC of -2 mk and k{sub eff}-EOC of 0.900 for the last case. Sensitivity analyses of CVR and eigenvalue are also included in our study.

  9. Sensitivity and Uncertainty Analysis for coolant void reactivity in a CANDU Fuel Lattice Cell Model

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Seung Yeol; Shim, Hyung Jin [Seoul National University, Seoul (Korea, Republic of)

    2016-10-15

    In this study, the EPBM is implemented in Seoul National university Monte Carlo (MC) code, McCARD which has the k uncertainty evaluation capability by the adjoint-weighted perturbation (AWP) method. The implementation is verified by comparing the sensitivities of the k-eigenvalue difference to the microscopic cross sections computed by the DPBM and the direct subtractions for the TMI-1 pin-cell problem. The uncertainty of the coolant void reactivity (CVR) in a CANDU fuel lattice model due to the ENDF/B-VII.1 covariance data is calculated by its sensitivities estimated by the EPBM. The method based on the eigenvalue perturbation theory (EPBM) utilizes the 1st order adjoint-weighted perturbation (AWP) technique to estimate the sensitivity of the eigenvalue difference. Furthermore this method can be easily applied in a S/U analysis code system equipped with the eigenvalue sensitivity calculation capability. The EPBM is implemented in McCARD code and verified by showing good agreement with reference solution. Then the McCARD S/U analysis have been performed with the EPBM module for the CVR in CANDU fuel lattice problem. It shows that the uncertainty contributions of nu of {sup 235}U and gamma reaction of {sup 238}U are dominant.

  10. Positive void reactivity

    International Nuclear Information System (INIS)

    Diamond, D.J.

    1992-09-01

    This report is a review of some of the important aspects of the analysis of large loss-of-coolant accidents (LOCAs). One important aspect is the calculation of positive void reactivity. To study this subject the lattice physics codes used for void worth calculations and the coupled neutronic and thermal-hydraulic codes used for the transient analysis are reviewed. Also reviewed are the measurements used to help validate the codes. The application of these codes to large LOCAs is studied with attention focused on the uncertainty factor for the void worth used to bias the results. Another aspect of the subject dealt with in the report is the acceptance criteria that are applied. This includes the criterion for peak fuel enthalpy and the question of whether prompt criticality should also be a criterion. To study the former, fuel behavior measurements and calculations are reviewed. (Author) (49 refs., 2 figs., tab.)

  11. Analysis on void reactivity of DCA lattice

    International Nuclear Information System (INIS)

    Min, B. J.; Noh, K. H.; Choi, H. B.; Yang, M. K.

    2001-01-01

    In case of loss of coolant accident, the void reactivity of CANDU fuel provides the positive reactivity and increases the reactor power rapidly. Therefore, it is required to secure credibility of the void reactivity for the design and analysis of reactor, which motivated a study to assess the measurement data of void reactivity. The assessment of lattice code was performed with the experimental data of void reactivity at 30, 70, 87 and 100% of void fractions. The infinite multiplication factors increased in four types of fuels as the void fractions of them grow. The infinite multiplication factors of uranium fuels are almost within 1%, but those of Pu fuels are over 10% by the results of WIMS-AECL and MCNP-4B codes. Moreover, coolant void reactivity of the core loaded with plutonium fuel is more negative compared with that with uranium fuel because of spectrum hardening resulting from large void fraction

  12. Estimation of coolant void reactivity for CANDU-NG lattice using DRAGON and validation using MCNP5 and TRIPOLI-4.3

    International Nuclear Information System (INIS)

    Karthikeyan, R.; Tellier, R. L.; Hebert, A.

    2006-01-01

    The Coolant Void Reactivity (CVR) is an important safety parameter that needs to be estimated at the design stage of a nuclear reactor. It helps to have an a priori knowledge of the behavior of the system during a transient initiated by the loss of coolant. In the present paper, we have attempted to estimate the CVR for a CANDU New Generation (CANDU-NG) lattice, as proposed at an early stage of the Advanced CANDU Reactor (ACR) development. We have attempted to estimate the CVR with development version of the code DRAGON, using the method of characteristics. DRAGON has several advanced self-shielding models incorporated in it, each of them compatible with the method of characteristics. This study will bring to focus the performance of these self-shielding models, especially when there is voiding of such a tight lattice. We have also performed assembly calculations in 2 x 2 pattern for the CANDU-NG fuel, with special emphasis on checkerboard voiding. The results obtained have been validated against Monte Carlo codes MCNP5 and TRIPOLI-4.3. (authors)

  13. Void effects on BWR Doppler and void reactivity feedback

    International Nuclear Information System (INIS)

    Hsiang-Shou Cheng; Diamond, D.J.

    1978-01-01

    The significance of steam voids and control rods on the Doppler feedback in a gadolinia shimmed BWR is demonstrated. The importance of bypass voids when determining void feedback is also shown. Calculations were done using a point model, i.e., feedback was expressed in terms of reactivity coefficients which were determined for individual four-bundle configurations and then appropriately combined to yield reactor results. For overpower transients the inclusion of the void effect of control rods is to reduce Doppler feedback. For overpressurization transients the inclusion of the effect of bypass void wil increase the reactivity due to void collapse. (author)

  14. Three-dimensional core analysis on a super fast reactor with negative local void reactivity

    International Nuclear Information System (INIS)

    Cao Liangzhi; Oka, Yoshiaki; Ishiwatari, Yuki; Ikejiri, Satoshi

    2009-01-01

    Keeping negative void reactivity throughout the cycle life is one of the most important requirements for the design of a supercritical water-cooled fast reactor (super fast reactor). Previous conceptual design has negative overall void reactivity. But the local void reactivity, which is defined as the reactivity change when the coolant of one fuel assembly disappears, also needs to be kept negative throughout the cycle life because the super fast reactor is designed with closed fuel assemblies. The mechanism of the local void reactivity is theoretically analyzed from the neutrons balance point of view. Three-dimensional neutronics/thermal-hydraulic coupling calculation is employed to analyze the characteristics of the super fast reactor including the local void reactivity. Some configurations of the core are optimized to decrease the local void reactivity. A reference core is successfully designed with keeping both overall and local void reactivity negative. The maximum local void reactivity is less than -30 pcm

  15. Coolant void effect investigation - case of a na-cooled fast reactor

    International Nuclear Information System (INIS)

    Glinatsis, G.; Gugiu, D.

    2013-01-01

    In the frame of the last EURATOM-FP7 Program, a large sized Sodium-cooled FR (SFR) has been studied. Mixed carbides fuel (U, Pu)C has been adopted for the backup core solution and important work has been also performed in order to obtain an ''optimised'' backup configuration ''close'' to the reference one, which is fueled by mixed oxides fuel (U, Pu)Ox. The peculiarity of both core designs (the reference configuration and the optimised backup configuration) is the adoption of a 60 cm Plenum zone in the upper part of each fuel assembly (FA), that is filled by coolant, in order to mitigate (when emptied) the core positive coolant void effect. This paper presents some results of a detailed study of the coolant void effect for the above SFR with mixed carbides core. Many aspects, like geometric heterogeneity, the burnup state, the operating conditions, etc., have been taken into consideration in order to obtain information about the ''propagation'' and the behaviour of the coolant void effect itself. The performed study investigates also the coolant void effect consequences on some reactivity coefficients, which are important for a safe behaviour of the reactor. The investigation consisted in the steady state simulations of the reactor on different operating conditions in Monte Carlo approach. (authors)

  16. A DRAGON-MCNP comparison of void reactivity calculations

    Energy Technology Data Exchange (ETDEWEB)

    Marleau, G [Ecole Polytechnique, Montreal, PQ (Canada). Inst. de Genie Nucleaire; Milgram, M S [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)

    1996-12-31

    The determination of the reactivity coefficients associated with coolant voiding in a CANDU reactor is a subject which has attracted a large amount of interest in the last few years both from the theoretical and experimental point of view. One expects that deterministic codes such as DRAGON and WIMS-AECL or the MCNP4 Monte Carlo code should be able to adequately simulate the cell behaviour upon coolant voiding. However, the absence of an experimental database at equilibrium and discharge burnups has not permitted the full validation of any of these lattice codes, although a partial validation through comparison of two different computer codes has been considered. Here we present a comparison between DRAGON and MCNP4 of the void reactivity evaluation for fresh fuel. (author). 16 refs., 5 tabs.

  17. A DRAGON-MCNP comparison of void reactivity calculations

    International Nuclear Information System (INIS)

    Marleau, G.

    1995-01-01

    The determination of the reactivity coefficients associated with coolant voiding in a CANDU reactor is a subject which has attracted a large amount of interest in the last few years both from the theoretical and experimental point of view. One expects that deterministic codes such as DRAGON and WIMS-AECL or the MCNP4 Monte Carlo code should be able to adequately simulate the cell behaviour upon coolant voiding. However, the absence of an experimental database at equilibrium and discharge burnups has not permitted the full validation of any of these lattice codes, although a partial validation through comparison of two different computer codes has been considered. Here we present a comparison between DRAGON and MCNP4 of the void reactivity evaluation for fresh fuel. (author). 16 refs., 5 tabs

  18. The use of graphite for the reduction of void reactivity in CANDU reactors

    International Nuclear Information System (INIS)

    Min, B.J.; Kim, B.G.; Sim, K-S.

    1995-01-01

    Coolant void reactivity can be reduced by using burnable poison in CANDU reactors. The use of graphite in the fuel bundle is introduced to reduce coolant void reactivity by adding an appropriate amount of burnable poison in the central rod. This study shows that sufficiently low void reactivity which in controllable by Reactor Regulating System (RRS) can be achieved by using graphite used fuel with slightly enriched uranium. Zero void reactivity can be also obtained by using graphite used fuel with a large central rod. A new fuel bundle with graphite rods can substantially reduce the void reactivity with less burnup penalty compared to previously proposed low void reactivity fuel with depleted uranium. (author)

  19. Coolant voiding analysis following SGTR for an HLMC reactor

    International Nuclear Information System (INIS)

    Farmer, M.T.; Spencer, B.W.; Sienicki, J.J.

    2000-01-01

    Concepts are under development at Argonne National Laboratory for a small, modular, proliferation-resistant nuclear power steam supply system. Of primary interest here is the simplified system design, featuring steam generators that are directly immersed in the lead-bismuth eutectic (LBE) coolant of the primary system. To support the safety case for this design approach, model development and analysis of transient coolant voiding during a postulated guillotine-type steam generator tube rupture event has been carried out. For the current design, the blowdown will occur from the steam generator shell into the ruptured 12.7-mm-inside-diameter tube through which the LBE coolant passes. The steam will expand biaxially in the tube, with a portion of the flow vented upward to eventually expand into the cover-gas region, while the balance of the flow is vented downward as a jet into the surrounding downward-flowing LBE. Coolant freezing is not an issue in this case because of high feedwater temperature in relation to the freezing point of the LBE. The specific objectives of the current work are to (a) determine the penetration behavior of the steam jet into the lower cold-leg region, (b) characterize the resultant void behavior in terms of coherent bubble versus breakup into a size distribution of small bubbles, and (c) characterize the motion of the bubbles with regard to rise to the cover-gas region (via the liner-to-coolant vessel gap) versus downward transport with the flowing LBE and subsequent upflow through the core to the cover-gas region

  20. Void fraction calculation in a channel containing boiling coolant

    International Nuclear Information System (INIS)

    Norelli, F.

    1978-01-01

    The problem of void fraction calculation was studied for a channel containing boiling coolant, when a slip ratio correlation is used. Use of fitting (e.g. polinomial or rational algebraic) for slip ratio correlation and the characteristic method are proposed in this work. In this way we are reduced to some elementary quadrature problem. Another problem discussed in the present work concerns what we must consider as ''initial condition'' in any initial value problem, in order to take into account different error distributions in steady state and in successive time-dependent calculations

  1. Detection of coolant void in lead-cooled fast reactors

    International Nuclear Information System (INIS)

    Wolniewicz, Peter; Håkansson, Ane; Jansson, Peter

    2015-01-01

    Highlights: • We model the ALFRED LFR using different Monte-Carlo codes. • We study the impact on coolant void on the fission cross section in fission chambers. • We develop a methodology to detect coolant void. • We study the impact of detector fissile coating burn-up. • We conclude that the developed methodology may be an attractive complement to LFR monitoring. - Abstract: Previous work (Wolniewicz et al., 2013) has indicated that using fission chambers coated with 242 Pu and 235 U, respectively, can provide the means of detecting changes in the neutron flux that are connected to coolant density changes in a small lead-cooled fast reactor. Such density changes may be due to leakages of gas into the coolant, which, over time, may coalesce to large bubbles implying a high risk of causing severe damage of the core. By using the ratio of the information provided by the two types of detectors a quantity is obtained that is sensitive to these density changes and, to the first order approximation, independent of the power level of the reactor. In this work we continue the investigation of this proposed methodology by applying it to the Advanced LFR European Demonstrator (ALFRED) and using realistic modelling of the neutron detectors. The results show that the methodology may be used to detect density changes indicating the initial stages of a coalescence process that may result in a large bubble. Also, it is shown that under certain circumstances, large bubbles passing through the core could be detected with this methodology

  2. Calculation of the void reactivity of CANDU lattices using the SCALE code system

    Energy Technology Data Exchange (ETDEWEB)

    Valko, J. [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.; Feher, S. [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.; Hoogenboom, J.E. [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.; Slobben, J. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands)

    1995-11-01

    The reactivity effect of coolant voiding in CANDU-type fuel lattices has been calculated with different methods using the SCALE code system. The known positive void reactivity coefficient of the original lattice was correctly obtained. A modified fuel bundle containing dysprosium and slightly enriched uranium to eliminate the positive reactivity effect was also calculated. Owing to the increased heterogeneity of this modified fuel the one-dimensional cylindrical calculation with XSDRN proved to be inadequate. Code options allowing bundle geometry were successfully used for the calculation of the strongly space dependent flux and spectrum changes which determine the void reactivity. (orig.).

  3. IAEA sodium void reactivity benchmark calculations

    International Nuclear Information System (INIS)

    Hill, R.N.; Finck, P.J.

    1992-01-01

    In this paper, the IAEA-1 992 ''Benchmark Calculation of Sodium Void Reactivity Effect in Fast Reactor Core'' problem is evaluated. The proposed design is a large axially heterogeneous oxide-fueled fast reactor as described in Section 2; the core utilizes a sodium plenum above the core to enhance leakage effects. The calculation methods used in this benchmark evaluation are described in Section 3. In Section 4, the calculated core performance results for the benchmark reactor model are presented; and in Section 5, the influence of steel and interstitial sodium heterogeneity effects is estimated

  4. Dependence of calculated void reactivity on film-boiling representation

    International Nuclear Information System (INIS)

    Whitlock, J.; Garland, W.

    1992-01-01

    Partial voiding of a fuel channel can lead to complicated neutronic analysis, because of highly nonuniform spatial distributions. An investigation of the distribution dependence of void reactivity in a Canada deuterium uranium (CANDU) lattice, specifically in the regime of film boiling, was done. Although the core is not expected to be critical at the time of sheath dryout, this study augments current knowledge of void reactivity in this type of lattice

  5. LIMBO computer code for analyzing coolant-voiding dynamics in LMFBR safety tests

    International Nuclear Information System (INIS)

    Bordner, G.L.

    1979-10-01

    The LIMBO (liquid metal boiling) code for the analysis of two-phase flow phenomena in an LMFBR reactor coolant channel is presented. The code uses a nonequilibrium, annular, two-phase flow model, which allows for slip between the phases. Furthermore, the model is intended to be valid for both quasi-steady boiling and rapid coolant voiding of the channel. The code was developed primarily for the prediction of, and the posttest analysis of, coolant-voiding behavior in the SLSF P-series in-pile safety test experiments. The program was conceived to be simple, efficient, and easy to use. It is particularly suited for parametric studies requiring many computer runs and for the evaluation of the effects of model or correlation changes that require modification of the computer program. The LIMBO code, of course, lacks the sophistication and model detail of the reactor safety codes, such as SAS, and is therefore intended to compliment these safety codes

  6. Effects of Void Uncertainties on Pin Power Distributions and the Void Reactivity Coefficient for a 10X10 BWR Assembly

    International Nuclear Information System (INIS)

    Jatuff, F.; Krouthen, J.; Helmersson, S.; Chawla, R.

    2004-01-01

    A significant source of uncertainty in Boiling Water Reactor physics is associated with the precise characterisation of the axially-dependent neutron moderation properties of the coolant inside the fuel assembly channel, and the corresponding effects on reactor physics parameters such as the lattice neutron multiplication, the neutron migration length, and the pin-by-pin power distribution. In this paper, the effects of particularly relevant void fraction uncertainties on reactor physics parameters have been studied for a BWR assembly of type Westinghouse SVEA-96 using the CASMO-4, HELIOS/PRESTO-2 and MCNP4C codes. The SVEA-96 geometry is characterised by the sub-division of the assembly into four different sub-bundles by means of an inner bypass with a cruciform shape. The study has covered the following issues: (a) the effects of different cross-section data libraries on the void coefficient of reactivity, for a wide range of void fractions; (b) the effects due to a heterogeneous vs. homogeneous void distribution inside the sub-bundles; and (c) the consequences of partly inserted absorber blades producing different void fractions in different sub-bundles. (author)

  7. Experiments in ZED-2 to study the physics of low-void reactivity fuel in CANDU

    International Nuclear Information System (INIS)

    Zeller, M.B.; Celli, A.; McPhee, G.P.

    1994-01-01

    Prospective CANDU clients have indicated a desire for a zero or negative coolant void reactivity. In response to this market requirement AECL Research and AECL CANDU are jointly developing and testing a Low-Void Reactivity Fuel (LVRF) bundle, which will be retrofitable to the current generation of CANDU reactors. An important component of the LVRF program is the undertaking of reactor-physics experiments in the zero-energy ZED-2 lattice test facility at Chalk River Laboratories. Preliminary void-reactivity measurements have already been performed in ZED-2 using a limited amount of the prototype fuel. These experiments were to provide a proof-of-principle for the LVRF concept. A more comprehensive set of experiments are planned for later this year. Experiments to be performed include: measuring the critical buckling of CANDU-type lattices containing LVRF, with and without coolant in the channels; measuring the reactivity effect of heating the LVRF fuel and coolant in ZED-2 hot channels; and measuring detailed reaction rates and neutron density distributions across a LVRF bundle, in voided and D 2 O-cooled channels, by the foil activation method. This paper describes the experimental approach to be used for the study and presents calculations employing transport and diffusion theory to predict the results. The codes used for the simulations are the lattice code WIMS-AECL and the core code CONIFERS. Included in the paper are results from the preliminary measurement of void coefficient for LVRF in a ZED-2 lattice and a comparison of those results to predictions based on WIMS-AECL calculations. (author). 3 refs., 1 tab., 10 figs

  8. Core concept of fast power reactor with zero sodium void reactivity

    International Nuclear Information System (INIS)

    Matveev, V.I.; Chebeskov, A.N.; Krivitsky, I.Y.

    1991-01-01

    The paper presents a core concept of BN-800 - type fast power reactor with zero sodium void reactivity (SVR). Consideration is given to the layout-and some design features of such a core. Some considerations on the determination of the required SVR value as one of the fast reactor safety criteria in accidents with coolant boiling are presented. Some methodical considerations an the development of calculation models that give a correct description of the new core features are stated. The results of the integral SVR calculation studies are included. reactivity excursions under different scenarios of sodium boiling are estimated, some corrections into the calculated SVR value are discussed. (author)

  9. Influence of void effects on reactivity of coupled fast-thermal system HERBE

    International Nuclear Information System (INIS)

    Ljubenov, V.; Milovanovic, S.; Milovanovic, T.; Cuknic, O.

    1997-01-01

    Coupled fast-thermal system HERBE at the experimental zero power heavy water reactor RB is a system with the significant effects of the neutron leakage and neutron absorption. Presence of a coolant void introduces a new structure in an extremely heterogeneous core. In those conditions satisfactory results of the calculation are acquired only using specified space-energy homogenization procedure. In order to analyze transient appearances and accidental cases of the reactor systems, a procedure for modeling of influence of moderator and coolant loss on reactivity ('void effect') is developed. Reduction of the moderator volume fraction in some fuel channels due to air gaps or steam generation during the accidental moderator boiling, restricts validity of the diffusion approximation in the reactor calculations. In cases of high neutron flux gradients, which are consequence of high neutron absorption, application of diffusion approximation is questionable too. The problem may be solved using transport or Monte Carlo methods, but they are not acceptable in the routine applications. Applying new techniques based on space-energy core homogenization, such as the SPH method or the discontinuity factor method, diffusion calculations become acceptable. Calculations based on the described model show that loss of part of moderator medium introduce negative reactivity in the HERBE system. Calculated local void reactivity coefficients are used in safety analysis of hypothetical accidents

  10. Influence of the void fraction in the linear reactivity model

    International Nuclear Information System (INIS)

    Castillo, J.A.; Ramirez, J.R.; Alonso, G.

    2003-01-01

    The linear reactivity model allows the multicycle analysis in pressurized water reactors in a simple and quick way. In the case of the Boiling water reactors the void fraction it varies axially from 0% of voids in the inferior part of the fuel assemblies until approximately 70% of voids to the exit of the same ones. Due to this it is very important the determination of the average void fraction during different stages of the reactor operation to predict the burnt one appropriately of the same ones to inclination of the pattern of linear reactivity. In this work a pursuit is made of the profile of power for different steps of burnt of a typical operation cycle of a Boiling water reactor. Starting from these profiles it builds an algorithm that allows to determine the voids profile and this way to obtain the average value of the same one. The results are compared against those reported by the CM-PRESTO code that uses another method to carry out this calculation. Finally, the range in which is the average value of the void fraction during a typical cycle is determined and an estimate of the impact that it would have the use of this value in the prediction of the reactivity produced by the fuel assemblies is made. (Author)

  11. Possibilities of achieving non-positive void reactivity effect in fast sodium-cooled reactors with increased self-protection

    International Nuclear Information System (INIS)

    Alekseev, P.N.; Zverkov, Yu.A.; Morozov, A.G.; Orlov, V.V.; Slesarev, I.S.; Subbotin, S.A.

    1989-01-01

    The problems of self-protection inhancement for the liquid-metal cooled fast reactors with intra-assembly heterogeneity of the core are studied. Possible approaches to arrangement of such reactors with various powers characterized by high levels of coolant natural circulation, minimum reactivity changes during fuel burn-up and non-positive void effect of reactivity are found. 10 refs.; 11 figs

  12. Void coefficient of reactivity calculation for AP-600 core

    International Nuclear Information System (INIS)

    Suparlina, L.; Budiono, T.A.; Mardha, A.; Tukiran

    1998-01-01

    Void coefficient of reactivity as one of reactor kinetics parameters has been carried out. The calculation was done into two steps which is cell calculation using WIMSD/4 and core calculation using Batan-2DIFF code programs with the condition of beginning of cycle with all fresh fuels elements and all control rods withdrawn. The one dimension transport program in four neutron energy groups is used to calculate the cell generation of various core materials cell has been calculated in 1/4 fuel element with cluster model and square pitch arrange. Moderator density have been reduced until 20% for the void coefficient of reactivity calculation. Macroscopic cross-section as the out put of WIMSD/4 is being used as the input at the diffusion neutron program for core calculation. The void coefficient of reactivity of the AP-600 core can be determined with regular neutron flux and adjoint in four energy groups and X-Y geometry. The results is shown that the K eff calculation value is different 5.2% from the design data

  13. Breeding capability and void reactivity analysis of heavy-water-cooled thorium reactor

    International Nuclear Information System (INIS)

    Permana, Sidik; Takaki, Naoyuki; Sekimoto, Hiroshi

    2008-01-01

    The fuel breeding and void reactivity coefficient of thorium reactors have been investigated using heavy water as coolant for several parametric surveys on moderator-to-fuel ratio (MFR) and burnup. The equilibrium fuel cycle burnup calculation has been performed, which is coupled with the cell calculation for this evaluation. The η of 233 U shows its superiority over other fissile nuclides in the surveyed MFR ranges and always stays higher than 2.1, which indicates that the reactor has a breeding condition for a wide range of MFR. A breeding condition with a burnup comparable to that of a standard PWR or higher can be achieved by adopting a larger pin gap (1-6 mm), and a pin gap of about 2 mm can be used to achieve a breeding ratio (BR) of 1.1. A feasible design region of the reactors, which fulfills the breeding condition and negative void reactivity coefficient, has been found. A heavy-water-cooled PWR-type Th- 233 U fuel reactor can be designed as a breeder reactor with negative void coefficient. (author)

  14. Temporary solutions for a conservative estimation of void reactivity insertion in CANDU reactor

    International Nuclear Information System (INIS)

    Dumitrache, I.

    1997-01-01

    One of the most difficult task of the CANDU Reactor Physics Analysis is related to the correct treatment of the deviations from the reference coolant properties. The most significant problem is the reactivity inserted by a given coolant density variation. From the practical Nuclear Safety Analysis point of view, the solution must be not only conservative, but also adaptable to the current chain of codes utilized for accident simulation. The first set of experimental data was obtained by AECL many years ago. The fuel was fresh, clean and cold. Some of the currently used computer codes offer accurate predictions of the measured void reactivities. Unfortunately, the existing experimental data do not cover and are not significant for the burned CANDU fuel. A specific benchmark problem was suggested by the Institute for Nuclear Research (ICN) Pitesti. The problem was analysed and slightly modified during an IAEA Vienna RCM (Research Coordinating Meeting), Buenos Aires, 1990. Afterwards, the problem was independently solved in several countries, interested by the CANDU reactor. The results were presented and analysed at the Bombay RCM, 1992. It was clear that the interval defined by the code predictions is much too broad. New experimental data are necessary. They must cover the fuel isotopic composition specific for the burned CANDU fuel. The work is in progress at the Chalk River Laboratory. Temporary solutions have been analysed at the ICN Pitesti. The first aim was to identify the reactivity numerical values that are conservative, but not too inaccurate. The WIMS code predictions have been compared against other estimations, including the Monte-Carlo based ones. The second aim was to force the currently used code, PPV, to offer cell cross sections that are correct from the Reactor Physics point of view, and compatible with the imposed reactivity. Physical and mathematical procedures were proposed and evaluated. An additional solution was also taken into account: to

  15. Effect of nonlinear void reactivity on bifurcation characteristics of a lumped-parameter model of a BWR: A study relevant to RBMK

    Energy Technology Data Exchange (ETDEWEB)

    Verma, Dinkar, E-mail: dinkar@iitk.ac.in [Nuclear Engineering and Technology Program, Indian Institute of Technology Kanpur, Kanpur 208 016 (India); Kalra, Manjeet Singh, E-mail: drmanjeet.singh@dituniversity.edu.in [DIT University, Dehradun 248 009 (India); Wahi, Pankaj, E-mail: wahi@iitk.ac.in [Department of Mechanical Engineering, Indian Institute of Technology Kanpur, Kanpur 208 016 (India)

    2017-04-15

    Highlights: • A simplified model with nonlinear void reactivity feedback is studied. • Method of multiple scales for nonlinear analysis and oscillation characteristics. • Second order void reactivity dominates in determining system dynamics. • Opposing signs of linear and quadratic void reactivity enhances global safety. - Abstract: In the present work, the effect of nonlinear void reactivity on the dynamics of a simplified lumped-parameter model for a boiling water reactor (BWR) is investigated. A mathematical model of five differential equations comprising of neutronics and thermal-hydraulics encompassing the nonlinearities associated with both the reactivity feedbacks and the heat transfer process has been used. To this end, we have considered parameters relevant to RBMK for which the void reactivity is known to be nonlinear. A nonlinear analysis of the model exploiting the method of multiple time scales (MMTS) predicts the occurrence of the two types of Hopf bifurcation, namely subcritical and supercritical, leading to the evolution of limit cycles for a range of parameters. Numerical simulations have been performed to verify the analytical results obtained by MMTS. The study shows that the nonlinear reactivity has a significant influence on the system dynamics. A parametric study with varying nominal reactor power and operating conditions in coolant channel has also been performed which shows the effect of change in concerned parameter on the boundary between regions of sub- and super-critical Hopf bifurcations in the space constituted by the two coefficients of reactivities viz. the void and the Doppler coefficient of reactivities. In particular, we find that introduction of a negative quadratic term in the void reactivity feedback significantly increases the supercritical region and dominates in determining the system dynamics.

  16. Dependence of calculated void reactivity on film boiling representation in a CANDU lattice

    Energy Technology Data Exchange (ETDEWEB)

    Whitlock, J [McMaster Univ., Hamilton, ON (Canada). Dept. of Engineering Physics

    1994-12-31

    The distribution dependence of void reactivity in a CANDU (CANada Deuterium Uranium) lattice is studied, specifically in the regime of film boiling. A heterogeneous model of this phenomenon predicts a 4% increase in void reactivity over a homogeneous model for fresh fuel, and 11% at discharge. An explanation for this difference is offered, with regard to differing changes in neutron mean free path upon voiding. (author). 9 refs., 4 tabs., 6 figs.

  17. Spatial dependence of the void coefficient in the interstitial coolant channel positions of a stainless steel-clad TRIGA Mark I core

    International Nuclear Information System (INIS)

    Spriggs, Gregory D.; Nelson, George W.; Doane, Harry J.

    1982-01-01

    A new top grid plate was manufactured and installed in the U of A TRIGA. The new grid plate was identical to the old grid plate with respect to the fuel element array, but included two minor modifications; 1) 3/8'' holes were drilled in six interstitial positions between fuel element rings to allow for insertion of a small diameter void rod for void coefficient measurements in the coolant channels, and 2) flux wire holes were drilled in all remaining interstitial positions. The purpose of this report is to update the previously reported void coefficient measurements with data taken in one of the coolant channel positions

  18. Accurate reactivity void coefficient calculation for the fast spectrum reactor FBR-IME

    Energy Technology Data Exchange (ETDEWEB)

    Lima, Fabiano P.C.; Vellozo, Sergio de O.; Velozo, Marta J., E-mail: fabianopetruceli@outlook.com, E-mail: vellozo@cbpf.br, E-mail: martajann@gmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Secao de Engenharia Militar

    2017-07-01

    This paper aims to present an accurate calculation of the void reactivity coefficient for the FBR-IME, a fast spectrum reactor in development at the Engineering Military Institute (IME). The main design peculiarity lies in using mixed oxide [MOX - PuO{sub 2} + U(natural uranium)O{sub 2}] as fuel core. For this task, SCALE system was used to calculate the reactivity for several voids distributions generated by bubbles in the sodium beyond its boiling point. The results show that although the void reactivity coefficient is positive and location dependent, they are offset by other feedback effects, resulting in a negative overall coefficient. (author)

  19. Void reactivity decomposition for the Sodium-cooled Fast Reactor in equilibrium fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Sun Kaichao, E-mail: kaichao.sun@psi.ch [Paul Scherrer Institut (PSI), 5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne (EPFL), 1015 Lausanne (Switzerland); Krepel, Jiri; Mikityuk, Konstantin; Pelloni, Sandro [Paul Scherrer Institut (PSI), 5232 Villigen PSI (Switzerland); Chawla, Rakesh [Paul Scherrer Institut (PSI), 5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne (EPFL), 1015 Lausanne (Switzerland)

    2011-07-15

    Highlights: > We analyze the void reactivity effect for three ESFR core fuel cycle states. > The void reactivity effect is decomposed by neutron balance method. > Novelly, the normalization to the integral flux in the active core is applied. > The decomposition is compared with the perturbation theory based results. > The mechanism and the differences of the void reactivity effect are explained. - Abstract: The Sodium-cooled Fast Reactor (SFR) is one of the most promising Generation IV systems with many advantages, but has one dominating neutronic drawback - a positive sodium void reactivity. The aim of this study is to develop and apply a methodology, which should help better understand the causes and consequences of the sodium void effect. It focuses not only on the beginning-of-life (BOL) state of the core, but also on the beginning of open and closed equilibrium (BOC and BEC, respectively) fuel cycle conditions. The deeper understanding of the principal phenomena involved may subsequently lead to appropriate optimization studies. Various voiding scenarios, corresponding to different spatial zones, e.g. node or assembly, have been analyzed, and the most conservative case - the voiding of both inner and outer fuel zones - has been selected as the reference scenario. On the basis of the neutron balance method, the corresponding SFR void reactivity has been decomposed reaction-, isotope-, and energy-group-wise. Complementary results, based on generalized perturbation theory and sensitivity analysis, are also presented. The numerical analysis for both neutron balance and perturbation theory methods has been carried out using appropriate modules of the ERANOS code system. A strong correlation between the flux worth, i.e. the product of flux and adjoint flux, and the void reactivity importance distributions has been found for the node- and assembly-wise voiding scenarios. The neutron balance based decomposition has shown that the void effect is caused mainly by the

  20. Void reactivity decomposition for the Sodium-cooled Fast Reactor in equilibrium fuel cycle

    International Nuclear Information System (INIS)

    Sun Kaichao; Krepel, Jiri; Mikityuk, Konstantin; Pelloni, Sandro; Chawla, Rakesh

    2011-01-01

    Highlights: → We analyze the void reactivity effect for three ESFR core fuel cycle states. → The void reactivity effect is decomposed by neutron balance method. → Novelly, the normalization to the integral flux in the active core is applied. → The decomposition is compared with the perturbation theory based results. → The mechanism and the differences of the void reactivity effect are explained. - Abstract: The Sodium-cooled Fast Reactor (SFR) is one of the most promising Generation IV systems with many advantages, but has one dominating neutronic drawback - a positive sodium void reactivity. The aim of this study is to develop and apply a methodology, which should help better understand the causes and consequences of the sodium void effect. It focuses not only on the beginning-of-life (BOL) state of the core, but also on the beginning of open and closed equilibrium (BOC and BEC, respectively) fuel cycle conditions. The deeper understanding of the principal phenomena involved may subsequently lead to appropriate optimization studies. Various voiding scenarios, corresponding to different spatial zones, e.g. node or assembly, have been analyzed, and the most conservative case - the voiding of both inner and outer fuel zones - has been selected as the reference scenario. On the basis of the neutron balance method, the corresponding SFR void reactivity has been decomposed reaction-, isotope-, and energy-group-wise. Complementary results, based on generalized perturbation theory and sensitivity analysis, are also presented. The numerical analysis for both neutron balance and perturbation theory methods has been carried out using appropriate modules of the ERANOS code system. A strong correlation between the flux worth, i.e. the product of flux and adjoint flux, and the void reactivity importance distributions has been found for the node- and assembly-wise voiding scenarios. The neutron balance based decomposition has shown that the void effect is caused mainly

  1. Analysis of molten fuel-coolant interaction during a reactivity-initiated accident experiment

    International Nuclear Information System (INIS)

    El-Genk, M.S.; Hobbins, R.R.

    1981-01-01

    The results of a reactivity-initiated accident experiment, designated RIA-ST-4, are discussed and analyzed with regard to molten fuel-coolant interaction (MFCI). In this experiment, extensive amounts of molten UO 2 fuel and zircaloy cladding were produced and fragmented upon mixing with the coolant. Coolant pressurization up to 35 MPa and coolant overheating in excess of 940 K occurred after fuel rod failure. The initial coolant conditions were similar to those in boiling water reactors during a hot startup (that is, coolant pressure of 6.45 MPa, coolant temperature of 538 K, and coolant flow rate of 85 cm 3 /s). It is concluded that the high coolant pressure recorded in the RIA-ST-4 experiment was caused by an energetic MFCI and was not due to gas release from the test rod at failure, Zr/water reaction, or to UO 2 fuel vapor pressure. The high coolant temperature indicated the presence of superheated steam, which may have formed during the expansion of the working fluid back to the initial coolant pressure; yet, the thermal-to-mechanical energy conversion ratio is estimated to be only 0.3%

  2. An approach of SFR safety study through the most penalizing sodium void reactivity - 105

    International Nuclear Information System (INIS)

    Tiberi, V.; Ivanov, E.; Pignet, S.

    2010-01-01

    Sodium void reactivity effects can affect the plant safety significantly during accidental transients. Accordingly, they have to be accurately investigated for any new sodium cooled fast reactor concept, even if a fuel with a melting point lower than the sodium boiling temperature is adopted. Thus all new reactor concepts should be compared to each - others adopting the value of the maximal possible sodium void reactivity as a discrimination parameter. However, taking into account that the sodium void worth is spatially depended, it is not evident which volume could be voided in order to obtain the maximal reactivity increase. Typically the complete active core voiding (zones initially loaded with 235 U or 239 Pu) is taken into account. This paper summarizes the extensive work carried-out in the IRSN to investigate the sodium-void reactivity spatial profiles of a fast sodium-cooled reactor core in the aim of defining a methodology to search for the area where the void contribution to the reactivity is strictly positive. Perturbation theory design approach available in the ERANOS 2.1 code has been adopted to evaluate the 'area of positive void worth'. To do that, the code has been previously validated against experimental based benchmarks (IRPhEP) and reference calculations. The evaluation of the absolute values of reactivity profiles can be improved later-on adopting more sophisticated methodologies to perform more accurate calculations of the sample with the voided area determined adopting the rough procedure described here. It has been demonstrated that even the non-reference way of ERANOS calculations could be used to provide the basis for different core concepts inter-comparison. (authors)

  3. Analysis of void reactivity measurements in full MOX BWR physics experiments

    International Nuclear Information System (INIS)

    Ando, Yoshihira; Yamamoto, Toru; Umano, Takuya

    2008-01-01

    In the full MOX BWR physics experiments, FUBILA, four 9x9 test assemblies simulating BWR full MOX assemblies were located in the center of the core. Changing the in-channel moderator condition of the four assemblies from 0% void to 40% and 70% void mock-up, void reactivity was measured using Amplified Source Method (ASM) technique in the subcritical cores, in which three fission chambers were located. ASM correction factors necessary to express the consistency of the detector efficiency between measured core configurations were calculated using collision probability cell calculation and 3D-transport core calculation with the nuclear data library, JENDL-3.3. Measured reactivity worth with ASM correction factor was compared with the calculated results obtained through a diffusion, transport and continuous energy Monte Carlo calculation respectively. It was confirmed that the measured void reactivity worth was reproduced well by calculations. (author)

  4. Analysis of reactivity feedback effects of void and temperature in the MAPLE-X10 reactor

    International Nuclear Information System (INIS)

    Carlson, P.A.; Heeds, W.; Shim, S.Y.; King, S.G.

    1992-07-01

    The methods used for evaluating the void and temperature reactivity coefficients for the MAPLE-X10 Reactor are described and factors used in estimating their accuracy are discussed. The report presents representative transient analysis results using the CATHENA thermalhydraulics code. The role of the reactivity coefficients and their precision is discussed. The results are reviewed in terms of their safety implications

  5. A comparative neutronic analysis of KALIMER breeder core using Na or Pb-Bi coolant

    International Nuclear Information System (INIS)

    Yoo, J. W.; Kim, S. J.; Kim, Y. I.

    2000-01-01

    A comparative neutronic study has been conducted on KALIMER breeder core according to the replacement of sodium coolant by Pb-Bi coolant. Since the atomic weight of Pb and Bi is about 9 times heavier than that of Na, the energy loss by neutron colliding with Pb-Bi nucleus will be very small. Therefore, the reactor with Pb-Bi coolant will have a harder neutron spectrum than that with Na coolant. Consequently, the breeding ratio and burnup reactivity swing is expected to be enhanced. In addition, when Pb-Bi coolant is voided, a negative coolant void coefficient can be obtained by the net effects of smaller spectrum hardening and large neutron leakage. As a result, the breeding ratio was increased from 1.18 to 1.23 and burnup reactivity swing was reduced from 631 pcm to 150 pcm. When the coolant in the whole region of active core is voided, the coolant void coefficient was found to be -539 and -264 pcm at BOEC and EOEC, respectively. In the local voided case, the smaller coolant void coefficient was obtained than that of Na coolant. Accordingly, the use of Pb-Bi coolant in KALIMER gives an advantage of higher breeding ratio, smaller burnup reactivity swing and negative coolant void coefficient without any significant degradation of nuclear performance

  6. Reactivity effect of spent fuel due to spatial distributions for coolant temperature and burnup

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, T.; Yamane, Y. [Nagoya Univ., Dept. of Nuclear Engineering, Nagoya, Aichi (Japan); Suyama, K. [OECD/NEA, Paris (France); Mochizuki, H. [Japan Research Institute, Ltd., Tokyo (Japan)

    2002-03-01

    We investigated the reactivity effect of spent fuel caused by the spatial distributions of coolant temperature and burnup by using the integrated burnup calculation code system SWAT. The reactivity effect which arises from taking account of the spatial coolant temperature distribution increases as the average burnup increases, and reaches the maximum value of 0.69%{delta}k/k at 50 GWd/tU when the burnup distribution is concurrently considered. When the burnup distribution is ignored, the reactivity effect decreases by approximately one-third. (author)

  7. An analytical approach to the positive reactivity void coefficient of TRIGA Mark-II reactor

    International Nuclear Information System (INIS)

    Edgue, Erdinc; Yarman, Tolga

    1988-01-01

    Previous calculations of reactivity void coefficient of I.T.U. TRIGA Mark-II Reactor was done by the second author et al. The theoretical predictions were afterwards, checked in this reactor experimentally. In this work an analytical approach is developed to evaluate rather quickly the reactivity void coefficient of I.T.U. TRIGA Mark-II, versus the size of the void inserted into the reactor. It is thus assumed that the reactor is a cylindrical, bare nuclear system. Next a belt of water of 2πrΔrH is introduced axially at a distance r from the center line of the system. r here, is the thickness of the belt, and H is the height of the reactor. The void is described by decreasing the water density in the belt region. A two group diffusion theory is adopted to determine the criticality of our configuration. The space dependency of the group fluxes are, thereby, assumed to be J 0 (2.405 r / R) cos (π Z / H), the same as that associated with the original bare reactor uniformly loaded prior to the change. A perturbation type of approach, thence, furnishes the effect of introducing a void in the belt region. The reactivity void coefficient can, rather surprisingly, be indeed positive. To our knowledge, this fact had not been established, by the supplier. The agreement of our predictions with the experimental results is good. (author)

  8. Conclusive evidence of abrupt coagulation inside the void during cyclic nanoparticle formation in reactive plasma

    International Nuclear Information System (INIS)

    Wetering, F. M. J. H. van de; Nijdam, S.; Beckers, J.

    2016-01-01

    In this letter, we present scanning electron microscopy (SEM) results that confirm in a direct way our earlier explanation of an abrupt coagulation event as the cause for the void hiccup. In a recent paper, we reported on the fast and interrupted expansion of voids in a reactive dusty argon–acetylene plasma. The voids appeared one after the other, each showing a peculiar, though reproducible, behavior of successive periods of fast expansion, abrupt contraction, and continued expansion. The abrupt contraction was termed “hiccup” and was related to collective coagulation of a new generation of nanoparticles growing in the void using relatively indirect methods: electron density measurements and optical emission spectroscopy. In this letter, we present conclusive evidence using SEM of particles collected at different moments in time spanning several growth cycles, which enables us to follow the nanoparticle formation process in great detail.

  9. Optimization of binary breeder reactor. 1. Sodium void reactivity and Doppler effect in a new model

    International Nuclear Information System (INIS)

    Nascimento, J.A. do; Dias, A.F.; Ishiguro, Y.

    1985-01-01

    A model for the Binary Breeder Reactor (BBR) is examined for the inherent safety characteristics, sodium void reactivity and Doppler effect in the beginning of cycle and a hypothetical end of cycle. In addition to the standard fueling mode of the BBR, two others are considered: U 238 /U 233 -alternate fueling, and U 238 /PU-normal fueling of LMFBRs. (Author) [pt

  10. Shock loading and reactive flow modeling studies of void induced AP/AL/HTPB propellant

    Science.gov (United States)

    Miller, P. J.; Lindfors, A. J.

    1998-07-01

    The unreactive Hugoniot of a class 1.3 propellant has been investigated by shock compression experiments. The results are analyzed in terms of an ignition and growth reactive flow model using the DYNA2D hydrocode. The calculated shock ignition parameters of the model show a linear dependence on measured void volume which appears to reproduce the observed gauge records well. Shock waves were generated by impact in a 75 mm single stage powder gun. Manganin and PVDF pressure gauges provided pressure-time histories to 140 kbar. The propellants were of similar formulation differing only in AP particle size and the addition of a burn rate modifer (Fe2O3) from that of previous investigations. Results show neglible effect of AP particle size on shock response in contrast to the addition of Fe2O3 which appears to `stiffen' the unreactive Hugoniot and enhances significantly the reactive rates under shock. The unreactive Hugoniot, within experimental error, compares favorably to the solid AP Hugoniot. Shock experiments were performed on propellant samples strained to induce insitu voids. The material state was quantified by uniaxial tension dialatometry. The experimental records show a direct correlation between void volume (0 to 1.7%) and chemical reactivity behind the shock front. These results are discussed in terms of `hot spot' ignition resulting from the shock collapse of the voids.

  11. Experimental investigation of void distribution in Suppression Pool during the initial blowdown period of a Loss of Coolant Accident using air–water two-phase mixture

    International Nuclear Information System (INIS)

    Rassame, Somboon; Griffiths, Matthew; Yang, Jun; Lee, Doo Yong; Ju, Peng; Choi, Sung Won; Hibiki, Takashi; Ishii, Mamoru

    2014-01-01

    Highlights: • Basic understanding of the venting phenomena in the SP during a LOCA was obtained. • A series of experiment is carried out using the PUMA-E test facility. • Two phases of experiments, namely, an initial and a quasi-steady phase were observed. • The maximum void penetration depth was experienced during the initial phase. - Abstract: During the initial blowdown period of a Loss of Coolant Accident (LOCA), the non-condensable gas initially contained in the BWR containment is discharged to the pressure suppression chamber through the blowdown pipes. The performance of Emergency Core Cooling System (ECCS) can be degraded due to the released gas ingestion into the suction intakes of the ECCS pumps. The understanding of the relevant phenomena in the pressure suppression chamber is important in analyzing potential gas intrusion into the suction intakes of ECCS pumps. To obtain the basic understanding of the relevant phenomena and the generic data of void distribution in the pressure suppression chamber during the initial blowdown period of a LOCA, tests with various blowdown conditions were conducted using the existing Suppression Pool (SP) tank of the integral test facility, called Purdue University Multi-Dimensional Integral Test Assembly for ESBWR applications (PUMA-E) facility, a scaled downcomer pipe installed in the PUMA-E SP, and air discharge pipe system. Two different diameter sizes of air injection pipe (0.076 and 0.102 m), a range of air volumetric flux (7.9–24.7 m/s), initial void conditions in an air injection pipe (fully void, partially void, and fully filled with water) and different air velocity ramp rates (1.0, 1.5, and 2.0 s) are used to investigate the impact of the blowdown conditions to the void distribution in the SP. Two distinct phases of experiments, namely, an initial and a quasi-steady phase were observed. The maximum void penetration depth was experienced during the initial phase. The quasi-steady phase provided less void

  12. Temperature and void reactivity coefficient calculations for the high flux isotope reactor safety analysis report

    International Nuclear Information System (INIS)

    Engle, W.W. Jr.; Williams, L.R.

    1994-07-01

    This report provides documentation of a series of calculations performed in 1991 in order to provide input for the High Flux Isotope Reactor Safety Analysis Report. In particular, temperature and void reactivity coefficients were calculated for beginning-of-life, end-of-life, and xenon equilibrium (29 h) conditions. Much of the data used to prepare the computer models for these calculations was derived from the original HFIR nuclear design study

  13. Experimental investigation of void distribution in suppression pool over the duration of a loss of coolant accident using steam–water two-phase mixture

    International Nuclear Information System (INIS)

    Rassame, Somboon; Griffiths, Matthew; Yang, Jun; Ju, Peng; Sharma, Subash; Hibiki, Takashi; Ishii, Mamoru

    2015-01-01

    Highlights: • Experiments were conducted to study void fraction distribution in SP during blowdown. • 3 Experimental phases, namely, an initial and a quasi-steady phase, chugging were observed. • The maximum void penetration depth was experienced during the initial phase. • The quasi-steady phase provided less void penetration depth with oscillations. • The chugging phase was experienced at the end of experimental phase. - Abstract: Studies are underway to determine if a large amount gas discharged through the downcomer pipes in the pressure suppression chamber during the blowdown of Loss of Coolant Accident (LOCA) can potentially be entrained into the Emergency Core Cooling System (ECCS) suction piping of BWR. This may result in degraded ECCS pumps performance which could affect the ability to maintain or recover the water inventory level in the Reactor Pressure Vessel (RPV) during a LOCA. Therefore, it is very important to understand the void behavior in the pressure suppression chamber during the blowdown period of a LOCA. To address this issue, a set of experiments is conducted using the Purdue University Multi-Dimensional Integral Test Assembly for ESBWR applications (PUMA-E) facility. The geometry of the test apparatus is determined based on the basic geometrical scaling analysis from a prototypical BWR containment (MARK I) with a consideration of downcomer size, downcomer water submergence depth and Suppression Pool (SP) water level. Several instruments are installed in the test facility to measure the required experimental data such as the steam mass flow rate, void fraction, pressure and temperature. In the experiments, sequential flows of air, steam–air mixture and pure steam-each with the various flow rate conditions are injected from the Drywell (DW) through a downcomer pipe in the SP. Eight tests with two different downcomer sizes, various initial gas volumetric fluxes at the downcomer, and two different initial non-condensable gas

  14. Core dynamics analysis for reactivity insertion and loss of coolant flow tests using the HTTR

    International Nuclear Information System (INIS)

    Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Takeda, Tetsuaki

    2007-01-01

    The High Temperature engineering Test Reactor (HTTR) is a graphite-moderated and a gas-cooled reactor with a thermal power of 30 MW and a reactor outlet coolant temperature of 950degC (SAITO, 1994). Safety demonstration tests using the HTTR are in progress to verify its inherent safety features and improve the safety technology and design methodology for High-Temperature Gas-cooled Reactors (HTGRs) (TACHIBANA 2002) (NAKAGAWA 2004). The reactivity insertion test is one of the safety demonstration tests for the HTTR. This test simulates the rapid increase in the reactor power by withdrawing the control rod without operating the reactor power control system. In addition, the loss of coolant flow tests has been conducted to simulate the rapid decrease in the reactor power by tripping one, two or all out of three gas circulators. The experimental results have revealed the inherent safety features of HTGRs, such as the negative reactivity feedback effect. The numerical analysis code, which was named ACCORD (TAKAMATSU 2006), was developed to analyze the reactor dynamics including the flow behavior in the HTTR core. We used a conventional method, namely, a one-dimensional flow channel model and reactor kinetics model with a single temperature coefficient, taking into account the temperature changes in the core. However, a slight difference between the analytical and experimental results was observed. Therefore, we have modified this code to use a model with four parallel channels and twenty temperature coefficients in the core. Furthermore, we added another analytical model of the core for calculating the heat conduction between the fuel channels and the core in the case of the loss of coolant flow tests. This paper describes the validation results for the newly developed code using the experimental results of the reactivity insertion test as well as the loss of coolant flow tests by tripping one or two out of three gas circulators. Finally, the pre-analytical result of

  15. Void reactivity feedback analysis for U-based and Th-based LWR incineration cycles

    Energy Technology Data Exchange (ETDEWEB)

    Lindley, B.A.; Parks, G.T. [Cambridge University Engineering Department, Trumpington Street, Cambridge, CB2 1PZ (United Kingdom); Franceschini, F. [Westinghouse Electric Company LLC, Cranberry Township, PA (United States)

    2013-07-01

    In reduced-moderation LWRs, an external supply of transuranic (TRU) can be incinerated by mixing it with a fertile isotope ({sup 238}U or {sup 232}Th) and recycling all the actinides after each cycle. Performance is limited by coolant reactivity feedback - the moderator density coefficient (MDC) must be kept negative. The MDC is worse when more TRU is loaded, but TRU feed is also needed to maintain criticality. To assess the performance of this fuel cycle in different neutron spectra, three LWRs are considered: 'reference' PWRs and reduced-moderation PWRs and BWRs. The MDC of the equilibrium cycle is analysed by reactivity decomposition with perturbed coolant density by isotope and neutron energy. The results show that using {sup 232}Th as a fertile isotope yields superior performance to {sup 238}U. This is due essentially to the high resonance η of U bred from Th (U3), which increases the fissility of the U3-TRU isotope vector in the Th-fueled system relative to the U-fueled system, and also improves the MDC in a sufficiently hard spectrum. Spatial separation of TRU and U3 in the Th-fueled system renders further improvement by hardening the neutron spectrum in the TRU and softening it in the U3. This improves the TRU η and increases the negative MDC contribution from reduced thermal fission in U3. (authors)

  16. HELIOS/DRAGON/NESTLE codes' simulation of void reactivity in a CANDU core

    International Nuclear Information System (INIS)

    Sarsour, H.N.; Rahnema, F.; Mosher, S.; Turinsky, P.J.; Serghiuta, D.; Marleau, G.; Courau, T.

    2002-01-01

    This paper presents results of simulation of void reactivity in a CANDU core using the NESTLE core simulator, cross sections from the HELIOS lattice physics code in conjunction with incremental cross sections from the DRAGON lattice physics code. First, a sub-region of a CANDU6 core is modeled using the NESTLE core simulator and predictions are contrasted with predictions by the MCNP Monte Carlo simulation code utilizing a continuous energy model. In addition, whole core modeling results are presented using the NESTLE finite difference method (FDM), NESTLE nodal method (NM) without assembly discontinuity factors (ADF), and NESTLE NM with ADF. The work presented in this paper has been performed as part of a project sponsored by the Canadian Nuclear Safety Commission (CNSC). The purpose of the project was to gather information and assess the accuracy of best estimate methods using calculational methods and codes developed independently from the CANDU industry. (author)

  17. A comparative neutronic analysis of 150MWe TRU burner according to the coolant alteration

    International Nuclear Information System (INIS)

    Yoo, J. W.; Kim, S. J.; Kim, Y. I.

    2000-01-01

    A comparative neutronic analysis has been conducted for the small TRU burner according to their coolant material. The use of Pb-Bi coolant gave a low burnup reactivity swing and negative or less positive coolant void coefficient with harder neutron spectrum. By a lower burnup reactivity swing and higher conversion ratio of Pb-Bi cooled core, the total amount of TRU consumption was found to be small compared with Na cooled core despite of the higher MA consumption ratio of Pb-Bi cooled core. However, Pb-Bi cooled reactor have a lager margin in the coolant void coefficient, so that a variable MA composition can be loaded in the core. Accordingly, even though the Pb-Bi cooled TRU burner has not effectiveness on TRU burning in the same geometry and material condition, a flexible MA loading is envisaged to result in 10 times larger MA burning amount, still preserving a low coolant void worth

  18. CANDU with supercritical water coolant: conceptual design features

    International Nuclear Information System (INIS)

    Spinks, N.

    1997-01-01

    An advanced CANDU reactor, with supercritical water as coolant, has many attractive design features. The pressure exceeds 22 MPa but coolant temperatures in excess of 370 degrees C can be reached without encountering the two-phase region with its associated fuel-dry-out and flow-instability problems. Increased coolant temperature leads to increased plant thermodynamic efficiency reducing unit energy cost through reduced specific capital cost and reduced fueling cost. Increased coolant temperature leads to reduced void reactivity via reduced coolant in-core density. Light water becomes a coolant option. To preserve neutron economy, an advanced fuel channel is needed and is described below. A supercritical-water-cooled CANDU can evolve as fuel capabilities evolve to withstand increasing coolant temperatures. (author)

  19. Effects of Coolant Temperature Changes on Reactivity for Various Coolants in a Liquid Salt Cooled Very High Temperature Reactor (LS-VHTR)

    International Nuclear Information System (INIS)

    Casino, William A. Jr.

    2006-01-01

    The purpose of this study is to perform an investigation into the relative merit of various salts and salt compounds being considered for use as coolants in the liquid salt cooled very high temperature reactor platform (LS-VHTR). Most of the non-nuclear properties necessary to evaluate these salts are known, but the neutronic characteristics important to reactor core design are still in need of a more extensive examination. This report provides a two-fold approach to further this investigation. First, a list of qualifying salts is assembled based upon acceptable non-nuclear properties. Second, the effect on system reactivity for a secondary system transient or an off-normal or accident condition is examined for each of these salt choices. The specific incident to be investigated is an increase in primary coolant temperature beyond normal operating parameters. In order to perform the relative merit comparison of each candidate salt, the System Temperature Coefficient of Reactivity is calculated for each candidate salt at various state points throughout the core burn history. (author)

  20. Application of bias factor method with use of virtual experimental value to prediction uncertainty reduction in void reactivity worth of breeding light water reactor

    International Nuclear Information System (INIS)

    Kugo, Teruhiko; Mori, Takamasa; Kojima, Kensuke; Takeda, Toshikazu

    2007-01-01

    We have carried out the critical experiments for the MOX fueled tight lattice LWR cores using FCA facility and constructed the XXII-1 series cores. Utilizing the critical experiments carried out at FCA, we have evaluated the reduction of prediction uncertainty in the coolant void reactivity worth of the breeding LWR core based on the bias factor method with focusing on the prediction uncertainty due to cross section errors. In the present study, we have introduced a concept of a virtual experimental value into the conventional bias factor method to overcome a problem caused by the conventional bias factor method in which the prediction uncertainty increases in the case that the experimental core has the opposite reactivity worth and the consequent opposite sensitivity coefficients to the real core. To extend the applicability of the bias factor method, we have adopted an exponentiated experimental value as the virtual experimental value and formulated the prediction uncertainty reduction by the use of the bias factor method extended by the concept of the virtual experimental value. From the numerical evaluation, it has been shown that the prediction uncertainty due to cross section errors has been reduced by the use of the concept of the virtual experimental value. It is concluded that the introduction of virtual experimental value can effectively utilize experimental data and extend applicability of the bias factor method. (author)

  1. Assessment of Loss-of-Coolant Effect on Pressurized Heavy Water Reactors

    International Nuclear Information System (INIS)

    Kim, Won Young; Park, Joo Hwan; Kim, Bong Ghi

    2009-01-01

    A CANDU reactor is a heavy-water-moderated, natural uranium fuelled reactor with a pressure tube. The reactor contains a horizontal cylindrical vessel (calandria) and each pressure tube is isolated from the heavy-water moderator in a calandria. This allows the moderator system to be operated of a high-pressure and of a high-temperature coolant in pressure tube. This causes the pressurized liquid coolant in the channel to void and therefore give rise to a reactivity transient in the event of a break or fault in the coolant circuit. In particular, all CANDU reactors are well known to have a positive void reactivity coefficient and thus this phenomenon may lead to a positive feedback, which can cause a large power pulse. We assess the loss-of-coolant effect by coolant void reactivity versus fuel burnup, four factor parameters for fresh fuel and equilibrium fuel, reactivity change due to the change of coolant density and reactivity change in the case of half- and full-core coolant

  2. Coupling a system code with computational fluid dynamics for the simulation of complex coolant reactivity effects

    International Nuclear Information System (INIS)

    Bertolotto, D.

    2011-11-01

    The current doctoral research is focused on the development and validation of a coupled computational tool, to combine the advantages of computational fluid dynamics (CFD) in analyzing complex flow fields and of state-of-the-art system codes employed for nuclear power plant (NPP) simulations. Such a tool can considerably enhance the analysis of NPP transient behavior, e.g. in the case of pressurized water reactor (PWR) accident scenarios such as Main Steam Line Break (MSLB) and boron dilution, in which strong coolant flow asymmetries and multi-dimensional mixing effects strongly influence the reactivity of the reactor core, as described in Chap. 1. To start with, a literature review on code coupling is presented in Chap. 2, together with the corresponding ongoing projects in the international community. Special reference is made to the framework in which this research has been carried out, i.e. the Paul Scherrer Institute's (PSI) project STARS (Steady-state and Transient Analysis Research for the Swiss reactors). In particular, the codes chosen for the coupling, i.e. the CFD code ANSYS CFX V11.0 and the system code US-NRC TRACE V5.0, are part of the STARS codes system. Their main features are also described in Chap. 2. The development of the coupled tool, named CFX/TRACE from the names of the two constitutive codes, has proven to be a complex and broad-based task, and therefore constraints had to be put on the target requirements, while keeping in mind a certain modularity to allow future extensions to be made with minimal efforts. After careful consideration, the coupling was defined to be on-line, parallel and with non-overlapping domains connected by an interface, which was developed through the Parallel Virtual Machines (PVM) software, as described in Chap. 3. Moreover, two numerical coupling schemes were implemented and tested: a sequential explicit scheme and a sequential semi-implicit scheme. Finally, it was decided that the coupling would be single

  3. Sodium voiding analysis in Kalimer

    International Nuclear Information System (INIS)

    Chang, Won-Pyo; Jeong, Kwan-Seong; Hahn, Dohee

    2001-01-01

    A sodium boiling model has been developed for calculations of the void reactivity feedback as well as the fuel and cladding temperatures in the KALIMER core after onset of sodium boiling. The sodium boiling in liquid metal reactors using sodium as coolant should be modeled because of phenomenon difference observed from that in light water reactor systems. The developed model is a multiple -bubble slug ejection model. It allows a finite number of bubbles in a channel at any time. Voiding is assumed to result from formation of bubbles that fill the whole cross section of the coolant channel except for liquid film left on the cladding surface. The vapor pressure, currently, is assumed to be uniform within a bubble. The present study is focused on not only demonstration of the sodium voiding behavior predicted by the developed model, but also confirmation on qualitative acceptance for the model. In results, the model catches important phenomena for sodium boiling, while further effort should be made for the complete analysis. (author)

  4. Sodium-cooled fast reactor (SFR) fuel assembly design with graphite-moderating rods to reduce the sodium void reactivity coefficient

    Energy Technology Data Exchange (ETDEWEB)

    Won, Jong Hyuck; Cho, Nam Zin, E-mail: nzcho@kaist.ac.kr; Park, Hae Min; Jeong, Yong Hoon, E-mail: jeongyh@kaist.ac.kr

    2014-12-15

    Highlights: • The graphite rod-inserted SFR fuel assembly is proposed to achieve low sodium void reactivity. • The neutronics/thermal-hydraulics analyses are performed for the proposed SFR cores. • The sodium void reactivity is improved about 960–1030 pcm compared to reference design. - Abstract: The concept of a graphite-moderating rod-inserted sodium-cooled fast reactor (SFR) fuel assembly is proposed in this study to achieve a low sodium void reactivity coefficient. Using this concept, two types of SFR cores are analyzed; the proposed SFR type 1 core has new SFR fuel assemblies at the inner/mid core regions while the proposed SFR type 2 core has a B{sub 4}C absorber sandwich in the middle of the active core region as well as new SFR fuel assemblies at the inner/mid core regions. For the proposed SFR core designs, neutronics and thermal-hydraulic analyses are performed using the DIF3D, REBUS3, and the MATRA-LMR codes. In the neutronics analysis, the sodium void reactivity coefficient is obtained in various void situations. The two types of proposed core designs reduce the sodium void reactivity coefficient by about 960–1030 pcm compared to the reference design. However, the TRU enrichment for the proposed SFR core designs is increased. In the thermal hydraulic analysis, the temperature distributions are calculated for the two types of proposed core designs and the mass flow rate is optimized to satisfy the design constraints for the highest power generating assembly. The results of this study indicate that the proposed SFR assembly design concept, which adopts graphite-moderating rods which are inserted into the fuel assembly, can feasibly minimize the sodium void reactivity coefficient. Single TRU enrichment and an identical fuel slug diameter throughout the SFR core are also achieved because the radial power peak can be flattened by varying the number of moderating rods in each core region.

  5. Device for preventing coolant in a reactor from being lost

    International Nuclear Information System (INIS)

    Maruyama, Hiromi; Matsumoto, Tomoyuki.

    1975-01-01

    Object: To prevent all of coolant from being lost from the core at the time of failure in rupture of pipe in a recirculation system to cool the core with the coolant remained within the reactor. Structure: A valve, which will be closed when a water level of the coolant within the core is in a level less than a predetermined level, is provided on a recirculating water outlet nozzle in a pressure vessel to thereby prevent the coolant from being lost when the pipe is broken, thus cooling the core by means of reduced-pressure boiling of coolant remained within the core and boiling due to heat, and restraining core reactivity by means of void produced at that time. (Kamimura, M.)

  6. Local, zero-power void coefficient measurements in the ACPR

    Energy Technology Data Exchange (ETDEWEB)

    Rivard, J B; Thome, F V [Sandia Laboratories (United States)

    1974-07-01

    Changes in reactivity may be stimulated in the ACPR by the local introduction of voids into the reactor coolant. The local void coefficients of reactivity which describe this effect are of interest from a reactor safety point-of-view, and their determination is the subject of this presentation. Bottled nitrogen gas was used to produce the voids. The gas was forced out of a small diameter tube which was positioned vertically in the core lattice with its open end below the fuel. The gas was passed through a pressure regulator, a valve, and a flowmeter to establish a steady flow condition, following which a delayed-critical (zero-power) reactor state was established. Correlation of the average volume of core void created by the nitrogen flow with the reactivity worth of the delayed-critical control-rod bank position produced the values of the zero-power void coefficients of reactivity. The void coefficients were determined at various core positions from {approx}6 mm to 142 mm beyond the central irradiation space and for three different flow rates. For the range of void fractions investigated, these coefficients are negative, with values ranging between -$0.02 and -$0.12. Tabular and graphical results of the measurements are presented, and details of the coefficient determination are explained. (author)

  7. Local, zero-power void coefficient measurements in the ACPR

    International Nuclear Information System (INIS)

    Rivard, J.B.; Thome, F.V.

    1974-01-01

    Changes in reactivity may be stimulated in the ACPR by the local introduction of voids into the reactor coolant. The local void coefficients of reactivity which describe this effect are of interest from a reactor safety point-of-view, and their determination is the subject of this presentation. Bottled nitrogen gas was used to produce the voids. The gas was forced out of a small diameter tube which was positioned vertically in the core lattice with its open end below the fuel. The gas was passed through a pressure regulator, a valve, and a flowmeter to establish a steady flow condition, following which a delayed-critical (zero-power) reactor state was established. Correlation of the average volume of core void created by the nitrogen flow with the reactivity worth of the delayed-critical control-rod bank position produced the values of the zero-power void coefficients of reactivity. The void coefficients were determined at various core positions from ∼6 mm to 142 mm beyond the central irradiation space and for three different flow rates. For the range of void fractions investigated, these coefficients are negative, with values ranging between -$0.02 and -$0.12. Tabular and graphical results of the measurements are presented, and details of the coefficient determination are explained. (author)

  8. Influence of the void fraction in the linear reactivity model; Influencia de la fraccion de vacios en el modelo de reactividad lineal

    Energy Technology Data Exchange (ETDEWEB)

    Castillo, J.A.; Ramirez, J.R.; Alonso, G. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: jacm@nuclear.inin.mx

    2003-07-01

    The linear reactivity model allows the multicycle analysis in pressurized water reactors in a simple and quick way. In the case of the Boiling water reactors the void fraction it varies axially from 0% of voids in the inferior part of the fuel assemblies until approximately 70% of voids to the exit of the same ones. Due to this it is very important the determination of the average void fraction during different stages of the reactor operation to predict the burnt one appropriately of the same ones to inclination of the pattern of linear reactivity. In this work a pursuit is made of the profile of power for different steps of burnt of a typical operation cycle of a Boiling water reactor. Starting from these profiles it builds an algorithm that allows to determine the voids profile and this way to obtain the average value of the same one. The results are compared against those reported by the CM-PRESTO code that uses another method to carry out this calculation. Finally, the range in which is the average value of the void fraction during a typical cycle is determined and an estimate of the impact that it would have the use of this value in the prediction of the reactivity produced by the fuel assemblies is made. (Author)

  9. An assessment of methods of calculating sodium voiding reactivity in plutonium fuelled fast reactors

    International Nuclear Information System (INIS)

    Butland, A.T.D.; Simmons, W.N.; Stevenson, J.M.

    1979-01-01

    After a survey of the requirements an assessment of the accuracy of calculations of the sodium void effect using UK methods and data is made on the basis of the following work. First, the analysis of small and large sodium voids in the MOZART and Zebra 13 small (300 MW(E)) fast reactor mock-ups and the BIZET large fast reactor mock-ups, all of conventional design. The analysis was carried out using the UK FGL5 fine group nuclear data library, the MURAL cell code, whole reactor diffusion theory calculations of the neutron flux and perturbation theory methods. Exact perturbation theory was used in many cases, otherwise first order perturbation theory calculations were adjusted to give results equivalent to exact perturbation theory. Second, theoretical studies of some effects, including, the effects of extrapolating to fuel operating temperatures, fuel cycle and burn-up effects, and the heterogeneity effects of large fuelled subassemblies in pin geometry. Third, theoretical studies of approximations in the calculational methods including, the importance in the whole reactor calculation of the energy group structure and the spatial mesh, the importance of reactor material boundaries in the calculation of resonance shielding effects, and the use of neutron fluxes calculated using neutron diffusion theory rather than transport theory. (U.K.)

  10. An assessment of methods of calculating sodium-voiding reactivity in plutonium-fuelled fast reactors

    International Nuclear Information System (INIS)

    Butland, A.T.D.; Simmons, W.N.; Stevenson, J.M.

    1980-01-01

    After a survey of the requirements an assessment of the accuracy of calculations of the sodium-void effect using UK methods and data is made on the basis of the following work: (a) The analysis of small and large sodium voids in the MOZART and Zebra 13 small (300 MW(e)) fast reactor mock-ups and the BIZET large fast reactor mock-ups, all of conventional design. The analysis was carried out using the UK FGL5 fine group nuclear data library, the MURAL cell code, whole reactor diffusion theory calculations of the neutron flux and perturbation theory methods. Exact perturbation theory was used in many cases, otherwise first-order perturbation theory calculations were adjusted to give results equivalent to exact perturbation theory. (b) Theoretical studies of some effects, including the following: (i) The effects of extrapolating to fuel operating temperature; (ii) Fuel-cycle and burnup effects, including the gradual replacement through a fuel cycle of control-rod absorption by fission product absorption, the loss of fissile material and the change in fuel nuclide relative composition; (iii) The heterogeneity effects of large fuelled subassemblies in pin geometry. (c) Theoretical studies of approximations in the calculational methods, including the following: (i) The importance in the whole reactor calculation of the energy group structure and the spatial mesh, including comparisons of calculations in two (RZ) and three-dimensional geometry; (ii) The importance of reactor material boundaries in the calculation of resonance shielding effects; (iii) The use of neutron fluxes calculated using neutron diffusion theory rather than transport theory. (author)

  11. Prediction of the Sodium Void Reactivity in the Metal-fueled SFR Using the ENDF/B-VII.0 Library

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Sunghwan; Lim, Jae-Yong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The SVR (Sodium Void Reactivity) is one of the most important parameters in SFR (Sodium-cooled Fast Reactor) safety analysis. In this paper, to estimate the error of the SVR in metal-fueled SFR, three physics experiments named as BFS-75-1, BFS-109-2A, and BFS-84-1 were examined using recent cross-section library, ENDF/B-VII.0 and the MCNP code. In the MCNP6 calculation, two million histories/generation with 50 inactive/300 active generations are used with the continuous-energy ENDF/B-VII.0 library. We expect that accuracy of total cross-section of the sodium may play a dominant role in errors of SVRs at core peripheral and sodium plenum regions, whereas accuracy of capture cross-section of the sodium may play a dominant role for the results in errors of SVRs at core central region. In addition, capture cross-sections of the sodium in the ENDF/B-VII.0, the JEFF-3.2, and the JENDL-4.0 libraries show significant differences between each other, while total cross-sections of sodium in three libraries show good agreement.

  12. Computational benchmark on the void reactivity effect in MOX lattices. Contribution to a NEA-NSC benchmark study organized by the Working Party on Plutonium Recycling

    International Nuclear Information System (INIS)

    Freudenreich, W.E.; Aaldijk, J.K.

    1994-08-01

    The Working Party on Plutonium Recycling of the Nuclear Science Committee of the OECD Nuclear Energy Agency has initiated a benchmark study on the calculation of the void reactivity effect in MOX lattices. The results presented here were obtained with the continuous energy, generalized geometry Monte Carlo transport code MCNP. The cross-section libraries used were processed from the JEF-2.2 evaluation taking into account selfshielding in the unresolved resonance ranges (selfshielding in the resolved resonance ranges is treated by MCNP). For an infinite lattice of unit cells a positive void reactivity effect was found only for the MOX fuel with the largest Pu content. For an infinite lattice of macro cells (voidable inner zone with different fuel mixtures surrounded by an outer zone of UO 2 fuel with moderator) a positive void reactivity effect was obtained for the three MOX fuel types considered. The results are not representative for MOX-loaded power reactor lattices, but serve only to intercompare reactor physics codes and libraries. (orig.)

  13. Tradeoff of sodium void worth and burnup reactivity swing: Impacts on balance safety position in metallic-fueled cores

    International Nuclear Information System (INIS)

    Wigeland, R.A.; Turski, R.B.; Pizzica, P.A.

    1994-01-01

    A study has been conducted to investigate the effect of a lower sodium void worth on the consequences of severe accidents in metallic-fueled sodium-cooled reactors. Four 900 MWth designs were used for the study, where all of the reactor cores were designed based on the metallic fuel of the Integral Fast Reactor (IFR) concept. The four core designs each have different sodium void worth, in the range of -3$ to 5$. The purpose of the investigation was to determine the differences in severe accident response for the four core designs, in order to estimate the improvement in overall safety that could be achieved from a reduction in the sodium void worth for reactor cores which use a metallic fuel form

  14. Sensitivity and uncertainty analysis of reactivities for UO2 and MOX fueled PWR cells

    Energy Technology Data Exchange (ETDEWEB)

    Foad, Basma [Research Institute of Nuclear Engineering, University of Fukui, Kanawa-cho 1-2-4, Tsuruga-shi, Fukui-ken, 914-0055 (Japan); Egypt Nuclear and Radiological Regulatory Authority, 3 Ahmad El Zomar St., Nasr City, Cairo, 11787 (Egypt); Takeda, Toshikazu [Research Institute of Nuclear Engineering, University of Fukui, Kanawa-cho 1-2-4, Tsuruga-shi, Fukui-ken, 914-0055 (Japan)

    2015-12-31

    The purpose of this paper is to apply our improved method for calculating sensitivities and uncertainties of reactivity responses for UO{sub 2} and MOX fueled pressurized water reactor cells. The improved method has been used to calculate sensitivity coefficients relative to infinite dilution cross-sections, where the self-shielding effect is taken into account. Two types of reactivities are considered: Doppler reactivity and coolant void reactivity, for each type of reactivity, the sensitivities are calculated for small and large perturbations. The results have demonstrated that the reactivity responses have larger relative uncertainty than eigenvalue responses. In addition, the uncertainty of coolant void reactivity is much greater than Doppler reactivity especially for large perturbations. The sensitivity coefficients and uncertainties of both reactivities were verified by comparing with SCALE code results using ENDF/B-VII library and good agreements have been found.

  15. Improvement the value of sodium void reactivity effect of the fast neutron reactor by the instrumentality of the Monte Carlo code

    OpenAIRE

    P.A. Maslov; V.I. Matveev; I.V. Malysheva

    2015-01-01

    To fulfill safety of fast sodium reactors in a beyond design-basis accident (BDBA) like unprotected loss of flow accident (ULOF), the sodium void reactivity effect (SVRE) should be close to zero. Its value depends on the fuel burnup – the higher burnup the higher value of SVRE. We analyze limitation of the fuel burnup in the core of a large sodium reactor imposed by SVRE. The model of a large sodium-cooled reactor core is chosen for analysis. Two fuel types are considered – MOX and nitride...

  16. Void lattices

    International Nuclear Information System (INIS)

    Chadderton, L.T.; Johnson, E.; Wohlenberg, T.

    1976-01-01

    Void lattices in metals apparently owe their stability to elastically anisotropic interactions. An ordered array of voids on the anion sublattice in fluorite does not fit so neatly into this scheme of things. Crowdions may play a part in the formation of the void lattice, and stability may derive from other sources. (Auth.)

  17. Evaluation of the corrosion, reactivity and chemistry control aspects for the selection of an alternative coolant in the secondary circuit of sodium fast reactors

    International Nuclear Information System (INIS)

    Brissonneau, L.; Simon, N.; Balbaud-Celerier, F.; Courouau, J.L.; Martinelli, L.; Grabon, V.; Capitaine, A.; Conocar, O.; Blat, M.

    2009-01-01

    Full text of publication follows: Sodium Fast Reactors are promising fourth generation reactors as they can contribute to reduce resource demand in uranium and considerably reduce waste level due to their fast spectrum. However, progress can be obtained for these reactors on the investment cost and on safety improvement. To achieve these goals, one of the innovative solutions consists in eliminating the reaction of sodium with water in the steam generators, by replacing the sodium in the secondary circuit by another coolant. A work group composed of experts from CEA, Areva NP and EdF was in charge to evaluate several alternative coolants as Heavy Liquid Metals (HLM), nitrate salts and hydroxide mixtures, through a multi-criteria analysis. Three important criteria for the selection of one coolant are its 'Interactions with the structures', and its 'chemistry control', and 'Reactivity with fluids' which are strongly correlated. The assessment, mainly based on the state-of-art from published literature on these points, is detailed in this paper. The mechanisms of corrosion of steels by the HLM depend on the oxygen content. For Pb-Bi, it has been modelled for oxidation and release domains. The corrosion of steels by nitrate salts presents similarity with the oxidation induced by HLM. The highly corrosive hydroxide mixture requires the use of nickel base alloys, for which oxidation and mass transfer are nevertheless significant. The HLM requires a fine regulation of oxygen content, through measurements and control systems, both to prevent lead oxide precipitation at high level and release corrosion at low level. Nitrate salts decompose into nitrites at sufficiently high temperature, which might induce pressure build-up in the circuit. The hydroxides must be kept under reducing atmosphere to lower the corrosion rate. Though these coolants are relatively inert to air and water, one of the main drawbacks of HLM and nitrate salts are their reactivity with sodium. Bismuth

  18. Void worths in subcritical cores cooled by lead-bismuth

    International Nuclear Information System (INIS)

    Wallenius, Janne; Tucek, Kamil; Gudowski, Waclaw

    2001-01-01

    The introduction lead-bismuth coolant in accelerator driven transmutation systems (ADS) was: good neutron economy (higher source efficiency); natural circulation possible (decay heat removal); synergy with spallation target (simplified coolant management); high temperature of boiling (larger overpower margin); smaller void worths (operation at higher k-values). This paper deals with different aspects of the void worths in JAERI ADS

  19. Cure Cycle Design Methodology for Fabricating Reactive Resin Matrix Fiber Reinforced Composites: A Protocol for Producing Void-free Quality Laminates

    Science.gov (United States)

    Hou, Tan-Hung

    2014-01-01

    For the fabrication of resin matrix fiber reinforced composite laminates, a workable cure cycle (i.e., temperature and pressure profiles as a function of processing time) is needed and is critical for achieving void-free laminate consolidation. Design of such a cure cycle is not trivial, especially when dealing with reactive matrix resins. An empirical "trial and error" approach has been used as common practice in the composite industry. Such an approach is not only costly, but also ineffective at establishing the optimal processing conditions for a specific resin/fiber composite system. In this report, a rational "processing science" based approach is established, and a universal cure cycle design protocol is proposed. Following this protocol, a workable and optimal cure cycle can be readily and rationally designed for most reactive resin systems in a cost effective way. This design protocol has been validated through experimental studies of several reactive polyimide composites for a wide spectrum of usage that has been documented in the previous publications.

  20. Analysis of loss-of-coolant accidents in pressurized water reactors

    International Nuclear Information System (INIS)

    Moldaschl, H.

    1982-01-01

    Analysis of loss-of-coolant accidents in pressurized water reactors -Quantification of the influence of leak size, control assembly worth, boron concentration and initial power by a dynamic operations criterion. Neutronic and thermohydraulic behaviour of a pressurized water reactor during a loss-of-coolant accident (LOCA) is mainly influenced by -change of fuel temperature, -void in the primary coolant. They cause a local stabilization of power density, that means that also in the case of small leaks local void is the main stabilization effect. As a consequence the increase of fuel temperature remains very small even under extremely hypothetical assumptions: small leak, positive reactivity feedback (positive coolant temperature coefficient, negative density coefficient) at the beginning of the accident and all control assemblies getting stuck. Restrictions which have been valid up to now for permitted start-up conditions to fulfill inherent safety requirements can be lossened substantially by a dynamic operations criterion. Burnable poisons for compensation of reactivity theorefore can be omitted. (orig.)

  1. Development of a three dimensional homogeneous calculation model for the BFS-62 critical experiment. Preparation of adjusted equivalent measured values for sodium void reactivity values. Final report

    International Nuclear Information System (INIS)

    Manturov, G.; Semenov, M.; Seregin, A.; Lykova, L.

    2004-01-01

    The BFS-62 critical experiments are currently used as 'benchmark' for verification of IPPE codes and nuclear data, which have been used in the study of loading a significant amount of Pu in fast reactors. The BFS-62 experiments have been performed at BFS-2 critical facility of IPPE (Obninsk). The experimental program has been arranged in such a way that the effect of replacement of uranium dioxied blanket by the steel reflector as well as the effect of replacing UOX by MOX on the main characteristics of the reactor model was studied. Wide experimental program, including measurements of the criticality-keff, spectral indices, radial and axial fission rate distributions, control rod mock-up worth, sodium void reactivity effect SVRE and some other important nuclear physics parameters, was fulfilled in the core. Series of 4 BFS-62 critical assemblies have been designed for studying the changes in BN-600 reactor physics from existing state to hybrid core. All the assemblies are modeling the reactor state prior to refueling, i.e. with all control rod mock-ups withdrawn from the core. The following items are chosen for the analysis in this report: Description of the critical assembly BFS-62-3A as the 3rd assembly in a series of 4 BFS critical assemblies studying BN-600 reactor with MOX-UOX hybrid zone and steel reflector; Development of a 3D homogeneous calculation model for the BFS-62-3A critical experiment as the mock-up of BN-600 reactor with hybrid zone and steel reflector; Evaluation of measured nuclear physics parameters keff and SVRE (sodium void reactivity effect); Preparation of adjusted equivalent measured values for keff and SVRE. Main series of calculations are performed using 3D HEX-Z diffusion code TRIGEX in 26 groups, with the ABBN-93 cross-section set. In addition, precise calculations are made, in 299 groups and Ps-approximation in scattering, by Monte-Carlo code MMKKENO and discrete ordinate code TWODANT. All calculations are based on the common system

  2. Performance of Transuranic-Loaded Fully Ceramic Micro-Encapsulated Fuel in LWRs Final Report, Including Void Reactivity Evaluation

    International Nuclear Information System (INIS)

    Pope, Michael A.; Sen, R. Sonat; Boer, Brian; Ougouag, Abderrafi M.; Youinou, Gilles

    2011-01-01

    The current focus of the Deep Burn Project is on once-through burning of transuranics (TRU) in light-water reactors (LWRs). The fuel form is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the tri-isotropic (TRISO) fuel particle design from high-temperature reactor technology. In the Deep Burn LWR (DB-LWR) concept, these fuel particles are pressed into compacts using SiC matrix material and loaded into fuel pins for use in conventional LWRs. The TRU loading comes from the spent fuel of a conventional LWR after 5 years of cooling. Unit cell and assembly calculations have been performed using the DRAGON-4 code to assess the physics attributes of TRU-only FCM fuel in an LWR lattice. Depletion calculations assuming an infinite lattice condition were performed with calculations of various reactivity coefficients performed at each step. Unit cells and assemblies containing typical UO2 and mixed oxide (MOX) fuel were analyzed in the same way to provide a baseline against which to compare the TRU-only FCM fuel. Then, assembly calculations were performed evaluating the performance of heterogeneous arrangements of TRU-only FCM fuel pins along with UO2 pins.

  3. Analysis on the Multiplication Factor with the Change of Corium Mass and Void Fraction

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Hae Sun; Park, Chang Je; Song, Jin Ho; Ha, Kwang Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    The neutron absorbing materials and fuel rods would be separately arranged and relocated, since the control materials in metallic structures have lower melting points than that of the oxide fuel (UO{sub 2}) rod materials. In addition, core reflood for a BWR is normally accomplished by supplying unborated water unlikely for a PWR. Therefore, a potential for a recriticality event to occur may exist, if unborated coolant injection is initiated with this configuration in the reactor core. The re-criticality in this system, however, brings into question what the uranium mass is required to achieve a critical level. Furthermore, the additional decay heat from molten fuel (corium) will produce an increase of void and eventually results in under-moderation of neutrons. The prior verification of these consequential physical variations in criticality eigenvalue (effective multiplication factor, k{sub eff}) should be greatly contributed to control and termination of re-criticality. Therefore, this study addresses what uranium mass of corium could achieve re-criticality of an accident core, and how effect the coolant void fraction has on eigenvalue (k{sub eff}) and its reactivity. To analyze the critical mass and the effect on criticality upon changing coolant density, k{sub eff} values were calculated using the MCNPX 2.5.0 code, and the reactivity change was also investigated. As a result, a large change in corium mass leads to a little change in k{sub eff} value, nevertheless, only about 60 kg of uranium is necessary to achieve a critical level. Thus, the amounts to reach a re-criticality are not fairly large, considering the actual uranium quantities loaded in the reactor core. Based on the condition with k{sub eff} greater than unity, the absolute values of k{sub eff} decrease rate and the coolant density coefficient were gradually increased due to the steady increments of coolant void (i.e., decrease in coolant density). In addition, the k{sub eff} value approaches the

  4. Conceptual designing of reduced-moderation water reactor with heavy water coolant

    Energy Technology Data Exchange (ETDEWEB)

    Hibi, Kohki; Shimada, Shoichiro; Okubo, Tsutomu E-mail: okubo@hems.jaeri.go.jp; Iwamura, Takamichi; Wada, Shigeyuki

    2001-12-01

    The conceptual designing of reduced-moderation water reactors, i.e. advanced water-cooled reactors using plutonium mixed-oxide fuel with high conversion ratios more than 1.0 and negative void reactivity coefficients, has been carried out. The core is designed on the concept of a pressurized water reactor with a heavy water coolant and a triangular tight lattice fuel pin arrangement. The seed fuel assembly has an internal blanket region inside the seed fuel region as well as upper and lower blanket regions (i.e. an axial heterogeneous core). The radial blanket fuel assemblies are introduced in a checkerboard pattern among the seed fuel assemblies (i.e. a radial heterogeneous core). The radial blanket region is shorter than the seed fuel region. This study shows that the heavy water moderated core can achieve negative void reactivity coefficients and conversion ratios of 1.06-1.11.

  5. Void Reactivity Effects in the Second Charge of the Halden Boiling Water Reactor; Effets Cavitaires dans la Deuxieme Charge du Reacteur a Eau Lourde Bouillante de Halden (HBWR); Ehffekty pustotnoj reaktivnosti vo vtoroj zag HBWR; Effectos de Cavitacion en la Segunda Carga del Reactor de Agua Pesada Hirviente de Halden (HBWR)

    Energy Technology Data Exchange (ETDEWEB)

    Lunde, J. E. [OECD Halden Reactor Project (Norway)

    1964-02-15

    The reactivity effect of voids caused by boiling inside the coolant channels in the second fuel charge of the Halden Boiling Heavy Water Reactor has been measured both in void-simulated zero-power experiments and under actual power conditions. The void-simulated experiments consisted of measuring the reactivity effect of introducing void columns inside thin-walled tubes to various depths. The tubes were placed at different positions between die stringers in a single 7-rod cluster element practically identical with the normal second-charge fuel elements. This experiment enables an investigation of the reactivity dependence upon void fraction, and also the reactivity dependence of steam-bubble position in the coolant channel. The experiment was carried out in the Norwegian zero-power facility NORA, with a core consisting of 36 second-charge elements and with a lattice geometry identical to the one in HBWR. The temperature dependence of the void effect was investigated in a zero-power experiment with the 100 fuel-element core of HBWR. In a single fuel element the water level inside the coolant channel was depressed to various depths, and the reactivity effect of this perturbation was measured at different temperatures in the temperature interval 50 Degree-Sign C-220 Degree-Sign C. The power void reactivity has been measured in HBWR as a function of nuclear power at different moderator temperatures between 150 Degree-Sign C and 230 Degree-Sign C at powers up to about 16 MW at the highest temperature. The power-void reactivity coefficient is an important quantity in determining the dynamic behaviour of a boiling- water reactor. The theoretical determination of this quantity is, however, complicated by the fact that knowledge about the void distribution in the core is required. The detailed power-void distribution is not easily amenable to experimental determination, and accordingly the void-simulated experiments represent a better case for testing the reactor physics

  6. Criticality safety issues associated with the introduction of low void reactivity fuel in the Bruce reactors - a management and technical overview

    International Nuclear Information System (INIS)

    Thompson, J.W.; Austman, G.; Iglesias, F.; Schmeing, H.; Elliott, C.; Archinoff, G.

    2004-01-01

    The concept of criticality for operating reactor staff, particularly in a natural uranium-fuelled reactor, is relatively benign - the reactor is controlled at the critical condition by the regulating system. That is, issues related to criticality exist only within the reactor, in a set of carefully managed circumstances. With the introduction of enriched Low Void Reactivity Fuel (LVRF) into this operating environment comes a new 'concept of criticality', one which, although physically the same, cannot be treated in the same fashion. It may be the case that criticality can be achieved outside the reactor, albeit with a set of very pessimistic assumptions. Such 'inadvertent criticality' outside the reactor, should it occur, cannot be controlled. The consequences of such an inadvertent criticality could have far-reaching effects, not only in terms of severe health effects to those nearby, but also in terms of the negative impact on Bruce Power, and the Canadian nuclear industry in general. Thus the introduction of LVRF in the Bruce B reactors, and therefore the introduction of this new hazard, inadvertent criticality, warrants the development of a governance structure for its management. Such a program will consist of various elements, including the establishment of a framework to administer the criticality safety program, analytical assessment to support the process design, the development of operational procedures, the development of enhanced emergency procedures if necessary, and the implementation of a criticality safety training program. The entire package must be sufficient to demonstrate to station management, and the regulator, that the criticality safety risks associated with the implementation of enriched fuel have been properly evaluated, and that all necessary steps have been taken to effectively manage these risks. A well-founded Criticality Safety Program will offer such assurance. In this paper, we describe the establishment of a Criticality Safety

  7. Dynamics of core voiding during boiloff experiments

    International Nuclear Information System (INIS)

    Analytis, G.T.; Aksan, S.N.; Stierli, F.; Yadigaroglu, G.

    1987-01-01

    A series of boiloff experiments were conducted at the EIR NEPTUN test facility with a bundle consisting of 37 PWR fuel rod simulators. The test section was filled with subcooled coolant and the power was turned on. After an initial heatup phase, coolant was expelled from the test section due to rapid vapor generation near the mid-plane where the power generation was highest. Gradual boiloff of the remaining water followed. It was found that the ''collapsed liquid level'' (CLL) and the rod temperature histories could be predicted well, provided the initial expulsion of the coolant was calculated correctly. The axial void fraction and enthalpy profiles calculated with TRAC-BD/MOD1 provided the information needed for understanding the importance of heat transfer to the coolant before the inception of vapor generation, and the sensitivity of the results to the interfacial friction correlation used

  8. Measurements and calculation of reactivity in the IEA-R1 nuclear reactor

    International Nuclear Information System (INIS)

    Ferreira, P.S.B.

    1988-01-01

    Techniques and experimentals procedures utilized in the measurement of some nuclear parameters related to reactivity are presented. Measurements of reactivity coefficients, such as void, temperature and power, and control rod worth were made in the IEA-R1 Research Reactor. The techniques used to perform the measurements were: i) stable period (control rod calibration), ii) inverse kinetics (digital reactivity meter), iii) aluminium slab insertion in the fuel element coolant channels (void reactivity), iv) nuclear reactor core temperature changes by means of the changes in the coolant systems of reactor core (isothermal reactivity coefficient) and v) by making perturbation in the core through the control rod motions (power reactivity coefficient and control rod calibration). By using the computer codes HAMMER, HAMMER-TECHNION and CITATION, the experiments realized in the IEA-R1 reactor were simulated. From this simulation, the theoretical reactivity parameters were estimated and compared with the respective experimental results. Furthermore, in the second fuel load of Angra-1 Nuclear Power Station, the IPEN-CNEN/SP digital reactivity - meter were used in the lower power test with the aim to assess the equipment performance. Among several tests, the reacticity-meter were used in parallel with a Westinghouse analogic reativimeter-meter) to measure the heat additiona point, critical boron concentration, control rod calibration, isothermal and moderator reactivity coefficient. These tests, and the results obtained by the digital reactivity-meter are described. The results were compared with those obtained by Westinghouse analogic reactivity meter, showing excellent agreement. (author) [pt

  9. A core design study for 'zero-sodium-void-worth' cores

    International Nuclear Information System (INIS)

    Kawashima, Masatoshi; Suzuki, Masao; Hill, R.N.

    1992-01-01

    Recently, a number of low sodium-void-worth metal-fueled core design concepts have been proposed; to provide for flexibility in transuranic nuclide management strategy, core designs which exhibit a wide range of breeding characteristics have been developed. Two core concepts, a flat annular (transuranic burning) core and an absorber-type parfait (transuranic self-sufficient) core, are selected for this study. In this paper, the excess reactivity management schemes applied in the two designs are investigated in detail. In addition, the transient effect of reactivity insertions on the parfait core design is assessed. The upper and lower core regions in the parfait design are neutronically decoupled; however, the common coolant channel creates thermalhydraulic coupling. This combination of neutronic and thermalhydraulic characteristics leads to unique behavior in anticipated transient overpower events. (author)

  10. Partial discharges in spheroidal voids: Void orientation

    DEFF Research Database (Denmark)

    McAllister, Iain Wilson

    1997-01-01

    Partial discharge transients can be described in terms of the charge induced on the detecting electrode. The influence of the void parameters upon the induced charge is examined and discussed for spheroidal voids. It is shown that a quantitative interpretation of the induced charge requires...

  11. Pediatric Voiding Cystourethrogram

    Science.gov (United States)

    Scan for mobile link. Children's (Pediatric) Voiding Cystourethrogram A children’s (pediatric) voiding cystourethrogram uses fluoroscopy – a form of real-time x-ray – to examine a child’s bladder ...

  12. Cosmic void clumps

    Science.gov (United States)

    Lares, M.; Luparello, H. E.; Garcia Lambas, D.; Ruiz, A. N.; Ceccarelli, L.; Paz, D.

    2017-10-01

    Cosmic voids are of great interest given their relation to the large scale distribution of mass and the way they trace cosmic flows shaping the cosmic web. Here we show that the distribution of voids has, in consonance with the distribution of mass, a characteristic scale at which void pairs are preferentially located. We identify clumps of voids with similar environments and use them to define second order underdensities. Also, we characterize its properties and analyze its impact on the cosmic microwave background.

  13. Transient two-phase performance of LOFT reactor coolant pumps

    International Nuclear Information System (INIS)

    Chen, T.H.; Modro, S.M.

    1983-01-01

    Performance characteristics of Loss-of-Fluid Test (LOFT) reactor coolant pumps under transient two-phase flow conditions were obtained based on the analysis of two large and small break loss-of-coolant experiments conducted at the LOFT facility. Emphasis is placed on the evaluation of the transient two-phase flow effects on the LOFT reactor coolant pump performance during the first quadrant operation. The measured pump characteristics are presented as functions of pump void fraction which was determined based on the measured density. The calculated pump characteristics such as pump head, torque (or hydraulic torque), and efficiency are also determined as functions of pump void fractions. The importance of accurate modeling of the reactor coolant pump performance under two-phase conditions is addressed. The analytical pump model, currently used in most reactor analysis codes to predict transient two-phase pump behavior, is assessed

  14. Coolant leakage detection device

    International Nuclear Information System (INIS)

    Ito, Takao.

    1983-01-01

    Purpose: To surely detect the coolant leakage at a time when the leakage amount is still low in the intra-reactor inlet pipeway of FBR type reactor. Constitution: Outside of the intra-reactor inlet piping for introducing coolants at low temperature into a reactor core, an outer closure pipe is furnished. The upper end of the outer closure pipe opens above the liquid level of the coolants in the reactor, and a thermocouple is inserted to the opening of the upper end. In such a structure, if the coolants in the in-reactor piping should leak to the outer closure pipe, coolants over-flows from the opening thereof, at which the thermocouple detects the temperature of the coolants at a low temperature, thereby enabling to detect the leakage of the coolants at a time when it is still low. (Kamimura, M.)

  15. Calculation of Void in the Fort Saint Vrain Material

    Energy Technology Data Exchange (ETDEWEB)

    Potter, David Charles [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Taylor, Craig Michael [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Coons, James Elmer [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-05-11

    The percent void of the Fort Saint Vrain (FSV) material is estimated to be 21.1% based on the volume of the gap at the top of the drums, the volume of the coolant channels in the FSV fuel element, and the volume of the fuel handling channel in the FSV fuel element.

  16. On void nucleation

    International Nuclear Information System (INIS)

    Subbotin, A.V.

    1978-01-01

    Nucleation of viable voids in irradiated materials is considered. The mechanism of evaporation and absorption of interstitials and vacancies disregarding the possibility of void merging is laid down into the basis of the discussion. The effect of irradiated material structure on void nucleation is separated from the effect of the properties of supersaturated solutions of vacancies and interstitials. An analytical expression for the nucleation rate is obtained and analyzed in different cases. The interstitials are concluded to effect severely the nucleation rate of viable voids

  17. HANARO secondary coolant management

    International Nuclear Information System (INIS)

    Kim, Seon Duk.

    1998-02-01

    In this report, the basic theory for management of water quality, environmental factors influencing to the coolant, chemicals and its usage for quality control of coolant are mentioned, and water balance including the loss rate by evaporation (34.3 m 3 /hr), discharge rate (12.665 m 3 /hr), concentration ratio and feed rate (54.1 m 3 /hr) are calculated at 20 MW operation. Also, the analysis data of HANSU Limited for HANARO secondary coolant (feed water and circulating coolant) - turbidity, pH, conductivity, M-alkalinity, Ca-hardness, chloride ion, total iron ion, phosphoric ion and conversion rate are reviewed. It is confirmed that the feed water has good quality and the circulating coolant has been maintained within the control specification in general, but some items exceeded the control specification occasionally. Therefore it is judged that more regular discharge of coolant is needed. (author). 6 refs., 17 tabs., 18 figs

  18. Nuclear reactor coolant channels

    International Nuclear Information System (INIS)

    Macbeth, R.V.

    1978-01-01

    Reference is made to coolant channels for pressurised water and boiling water reactors and the arrangement described aims to improve heat transfer between the fuel rods and the coolant. Baffle means extending axially within the channel are provided and disposed relative to the fuel rods so as to restrict flow oscillations occurring within the coolant from being propagated transversely to the axis of the channel. (UK)

  19. Study on the effect of moderator density reactivity for Kartini reactor

    International Nuclear Information System (INIS)

    Budi Rohman; Widarto

    2009-01-01

    One of important characteristics of water-cooled reactors is the change of reactivity due to change in the density of coolant or moderator. This parameter generally has negative value and it has significant role in preventing the excursion of power during operation. Many thermal-hydraulic codes for nuclear reactors require this parameter as the input to account for reactivity feedback due to increase in moderator voids and the subsequent decrease in moderator density during operation. Kartini reactor is cooled and moderated by water, therefore, it is essential to study the effect of the change in moderator density as well as to determine the value of void or moderator density reactivity coefficient in order to characterize its behavior resulting from the presence of vapor or change of moderator density during operation. Analysis by MCNP code shows that the reactivity of core is decreasing with the decrease in moderator density. The analysis estimates the void or moderator density reactivity coefficient for Kartini Reactor to be -2.17×10-4 Δρ/ % void . (author)

  20. Temperature controlled 'void' formation

    International Nuclear Information System (INIS)

    Dasgupta, P.; Sharma, B.D.

    1975-01-01

    The nucleation and growth of voids in structural materials during high temperature deformation or irradiation is essentially dependent upon the existence of 'vacancy supersaturation'. The role of temperature dependent diffusion processes in 'void' formation under varying conditions, and the mechanical property changes associated with this microstructure are briefly reviewed. (author)

  1. Void nucleation at heterogeneities

    International Nuclear Information System (INIS)

    Seyyedi, S.A.; Hadji-Mirzai, M.; Russell, K.C.

    The energetics and kinetics of void nucleation at dislocations and interfaces are analyzed. These are potential void nucleation sites only when they are not point defect sinks. Both kinds of site are found to be excellent catalysts in the presence of inert gas

  2. Void hierarchy and cosmic structure

    International Nuclear Information System (INIS)

    Weygaert, Rien van de; Ravi Sheth

    2004-01-01

    Within the context of hierarchical scenarios of gravitational structure formation we describe how an evolving hierarchy of voids evolves on the basis of two processes, the void-in-void process and the void-in-cloud process. The related analytical formulation in terms of a two-barrier excursion problem leads to a self-similarly evolving peaked void size distribution

  3. Air void clustering.

    Science.gov (United States)

    2015-06-01

    Air void clustering around coarse aggregate in concrete has been identified as a potential source of : low strengths in concrete mixes by several Departments of Transportation around the country. Research was : carried out to (1) develop a quantitati...

  4. Calculation of Reactivity Build up in KANUPP core in Case of Large Break LOCA

    International Nuclear Information System (INIS)

    Arshad, M. W.

    2012-01-01

    Loss of Coolant Accident (LOCA) in a Pressurized Heavy Water Reactor (PHWR) leads to coolant expulsion in a primary heat transport system resulting in depressurization and possible core voiding. This results in deterioration of cooling conditions in reactor channels and increase in power before reactor shutdown, leading to higher fuel temperatures.The objective of this thesis is to couple Thermal Hydraulics Data for finding status of 2288 fuel bundles having unique coolant density along with continuous changing state of coolant. WIMCER and CITCER are used for the core calculation in case of LOCA and Thermal Hydraulic Data is obtained from the Thermal Hydraulic code TUF (two unequal flows). These codes are coupled with each other in C programming. Due to degradation of coolant in case of LOCA, the power and reactivity start increasing. Near to 5 mk of reactivity the moderator dump start and reactor goes shut down. The result obtained from these code is followed the same trend as shown in KFSAR. (author)

  5. The sodium coolant

    International Nuclear Information System (INIS)

    Rodriguez, G.

    2004-01-01

    The sodium is the best appropriate coolant for the fast neutrons reactors technology. Thus the fast neutrons reactors development is intimately bound to the sodium technology. This document presents the sodium as a coolant point of view: atomic structure and characteristics, sodium impacts on the fast neutron reactors technology, chemical properties of the sodium and the consequences, quality control in a nuclear reactor, sodium treatment. (A.L.B.)

  6. Nuclear reactor coolant channels

    International Nuclear Information System (INIS)

    Macbeth, R.V.

    1978-01-01

    A nuclear reactor coolant channel is described that is suitable for sub-cooled reactors as in pressurised water reactors as well as for bulk boiling, as in boiling water reactors and steam generating nuclear reactors. The arrangement aims to improve heat transfer between the fuel elements and the coolant. Full constructional details are given. See also other similar patents by the author. (U.K.)

  7. Extended Life Coolant Testing

    Science.gov (United States)

    2016-06-06

    number. PLEASE DO NOT RETURN YOUR FORM TO THE ABOVE ADDRESS. 1. REPORT DATE (DD-MM-YYYY) 06-06-2016 2. REPORT TYPE Interim Report 3. DATES COVERED ... Corrosion Testing of Traditional and Extended Life Coolants 5b. GRANT NUMBER 5c. PROGRAM ELEMENT NUMBER 6. AUTHOR(S) Hansen, Gregory A. T...providing vehicle specific coolants. Several laboratory corrosion tests were performed according to ASTM D1384 and D2570, but with a 2.5x extended time

  8. Research on coolant radiochemistry

    International Nuclear Information System (INIS)

    Yeon, Jei Won; Kim, W. H.; Park, Y. J.; Im, J. K.; Jung, Y. J.; Jee, K. Y.; Choi, K. C.

    2004-04-01

    The final objective of this study is to develop the technology on the reduction of radioactive material formed in reactor coolant circuit. The contents of this study are composed of the simulation of primary cooling system, chemistry measurement technology in the high-temperature high-pressure environments, and coolant chemistry control technology. The main results are as follows; High-temperature and high-pressure loop system was designed and fabricated, which is to inducing CRUD growth condition on the surface of cladding. The high-temperature pH measurement system was established with YSZ sensing electrode and Ag/AgCl reference electrode. The performance of pH electrode was confirmed in the temperature range 200∼280 .deg. C. Coolant chemistry control technologies such as the neutron irradiation technique of boric acid solution, the evaluation on high-temperature electrochemical behavior of coolant, and the measurement of physicochemical properties of micro-particles were developed. The results of this study can be useful for the understanding of chemical phenomena occurred in reactor coolant and for the study on the reduction of radioactive material in primary coolant, which will be carried out in the next research stage

  9. An experimental and theoretical analysis of void fraction dynamics in a boiling channel

    International Nuclear Information System (INIS)

    Romberg, T.M.

    1977-01-01

    This paper describes an experimental and theoretical investigation of the void fraction dynamics at the exit of a test boiling channel which is operated near the 'instability threshold power' (the power level at which coolant flow instabilities occur). Dynamic measurements of the perturbations in channel inlet flow-rate, power input and exit void fraction are analysed using multivariate spectral analysis. The resulting experimental cross-spectral density functions between flow-rate/exit void fraction and power input/exit void fraction agree favourably with those calculated by a linearised hydrodynamic model in the frequency domain. (Author)

  10. AIREKMOD-RR, Reactivity Transients in Nuclear Research Reactors

    International Nuclear Information System (INIS)

    Baggoura, B.; Mazrou, H.

    2001-01-01

    1 - Description of program or function: AIREMOD-RR is a point kinetics code which can simulate fast transients in nuclear research reactor cores. It can also be used for theoretical reactor dynamics studies. It is used for research reactor kinetic analysis and provides a point neutron kinetic capability. The thermal hydraulic behavior is governed by a one-dimensional heat balance equation. The calculations are restricted to a single equivalent unit cell which consists of fuel, clad and coolant. 2 - Method of solution: For transient reactor kinetic calculations a modified Runge Kutta numerical method is used. The external reactivity insertion, specified as a function of time, is converted in dollar ($) unit. The neutron density, energy release and feedback variables are given at each time step. The two types of reactivity feedback considered are: Doppler effect and moderator effect. A new expression for the reactivity dependence on the feedback variables has been introduced in the present version of the code. The feedback reactivities are fitted in power series expression. 3 - Restrictions on the complexity of the problem: The number of delayed neutron groups and the total number of equations are limited only by computer storage capabilities. - Coolant is always in liquid phase. - Void reactivity feedback is not considered

  11. SMART core power control method by coolant temperature variation

    International Nuclear Information System (INIS)

    Lee, Chung Chan; Cho, Byung Oh

    2001-08-01

    SMART is a soluble boron-free integral type pressurized water reactor. Its moderator temperature coefficient (MTC) is strongly negative throughout the cycle. The purpose of this report is how to utilize the primary coolant temperature as a second reactivity control system using the strong negative MTC. The reactivity components associated with reactor power change are Doppler reactivity due to fuel temperature change, moderator temperature reactivity and xenon reactivity. Doppler reactivity and moderator temperature reactivity take effects almost as soon as reactor power changes. On the other hand, xenon reactivity change takes more than several hours to reach an equilibrium state. Therefore, coolant temperature at equilibrium state is chosen as the reference temperature. The power dependent reference temperature line is limited above 50% power not to affect adversely in reactor safety. To compensate transient xenon reactivity, coolant temperature operating range is expanded. The suggested coolant temperature operation range requires minimum control rod motion for 50% power change. For smaller power changes such as 25% power change, it is not necessary to move control rods to assure that fuel design limits are not exceeded

  12. Coolant leakage detecting device

    International Nuclear Information System (INIS)

    Yamauchi, Kiyoshi; Kawai, Katsunori; Ishihara, Yoshinao.

    1995-01-01

    The device of the present invention judges an amount of leakage of primary coolants of a PWR power plant at high speed. Namely, a mass of coolants contained in a pressurizer, a volume controlling tank and loop regions is obtained based on a preset relational formula and signals of each of process amount, summed up to determine the total mass of coolants for every period of time. The amount of leakage for every period of time is calculated by a formula of Karman's filter based on the total mass of the primary coolants for every predetermined period of time, and displays it on CRT. The Karman's filter is formed on every formula for several kinds of states formed based on the preset amount of the leakage, to calculate forecasting values for every mass of coolants. An adaptable probability for every preset leakage amount is determined based on the difference between the forecast value and the observed value and the scattering thereof. The adaptable probability is compared with a predetermined threshold value, which is displayed on the CRT. This device enables earlier detection of leakage and identification of minute leakage amount as compared with the prior device. (I.S.)

  13. Reactor coolant cleanup device

    International Nuclear Information System (INIS)

    Igarashi, Noboru.

    1986-01-01

    Purpose: To enable to introduce reactor water at high temperature and high pressure as it is, as well as effectively adsorb to eliminate cobalt in reactor water. Constitution: The coolant cleanup device comprises a vessel main body inserted to coolant pipeway circuits in a water cooled reactor power plant and filters contained within the vessel main body. The filters are prepared by coating and baking powder of metal oxides such as manganese ferrite having a function capable of adsorbing cobalt in the coolants onto the surface of supports made of metals or ceramics resistant to strong acids and alkalies in the form of three-dimensional network structure, for example, zircaloy-2, SUS 303 and the zirconia (baking) to form a basic filter elements. The basic filter elements are charged in plurality to the vessel main body. (Kawaiami, Y.)

  14. Comparative sodium void effects for different advanced liquid metal reactor fuel and core designs

    International Nuclear Information System (INIS)

    Dobbin, K.D.; Kessler, S.F.; Nelson, J.V.; Gedeon, S.R.; Omberg, R.P.

    1991-01-01

    An analysis of metal-, oxide-, and nitride-fueled advanced liquid metal reactor cores was performed to investigate the calculated differences in sodium void reactivity, and to determine the relationship between sodium void reactivity and burnup reactivity swing using the three fuel types. The results of this analysis indicate that nitride fuel has the least positive sodium void reactivity for any given burnup reactivity swing. Thus, it appears that a good design compromise between transient overpower and loss of flow response is obtained using nitride fuel. Additional studies were made to understand these and other nitride advantages. (author)

  15. 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.

  16. Determination of the void nucleation rate from void size distributions

    International Nuclear Information System (INIS)

    Brailsford, A.D.

    1977-01-01

    A method of estimating the void nucleation rate from one void size distribution and from observation of the maximum void radius at prior times is proposed. Implicit in the method are the assumptions that both variations in the critical radius with dose and vacancy thermal emission processes during post-nucleation quasi-steady-state growth may be neglected. (Auth.)

  17. Void shape effects and voids starting from cracked inclusion

    DEFF Research Database (Denmark)

    Tvergaard, Viggo

    2011-01-01

    Numerical, axisymmetric cell model analyses are used to study the growth of voids in ductile metals, until the mechanism of coalescence with neighbouring voids sets in. A special feature of the present analyses is that extremely small values of the initial void volume fraction are considered, dow...

  18. Measurement of local void fraction in a ribbed annulus

    International Nuclear Information System (INIS)

    Steimke, J.L.

    1992-01-01

    The computer code FLOWTRAN-TF is used to analyze hypothetical hydraulic accidents for the nuclear reactor at the Savannah River Site. During a hypothetical Large Break Loss-of-Coolant Accident (LOCA), reactor assemblies would contain a two-phase mixture of air and water which flows downward. Reactor assemblies consist of nested, ribbed annuli. Longitudinal ribs divide each annulus into four subchannels. For accident conditions, air and water can flow past ribs from one subchannel to another. For FLOWTRAN-TF to compute the size of those flows, it is necessary to know the local void fraction in the region of the rib. Measurements have previously been made of length-average void fraction in a ribbed annulus. However, no direct measurements were available of local void fraction. Due to the lack of data, a test was designed to measure local void fraction at the rib. One question addressed by the test was whether void fraction at the rib is solely a function of azimuthal-average void fraction or a function of additional variables such as pressure boundary conditions. This report provides a discussion of this test

  19. Coolant system decontamination

    International Nuclear Information System (INIS)

    Anstine, L.D.; James, D.B.; Melaika, E.A.; Peterson, J.P.

    1981-01-01

    An improved method for decontaminating the coolant system of water cooled nuclear power reactors and for regenerating the decontamination solution is described. A small amount of one or more weak-acid organic complexing agents is added to the reactor coolant, and the pH is adjusted to form a decontamination solution which is circulated throughout the coolant system to dissolve metal oxides from the interior surfaces and complex the resulting metal ions and radionuclide ions. The coolant containing the complexed metal ions and radionuclide ions is passed through a strong-base anion exchange resin bed which has been presaturated with a solution containing the complexing agents in the same ratio and having the same pH as the decontamination solution. As the decontamination solution passes through the resin bed, metal-complexed anions are exchanged for the metal-ion-free anions on the bed, while metal-ion-free anions in the solution pass through the bed, thus removing the metal ions and regenerating the decontamination solution. (author)

  20. Research on Coolant Radiochemistry

    International Nuclear Information System (INIS)

    Ha, Yeong Keong; Kim, W. H.; Yeon, J. W.; Jung, Y. J.; Choi, K. C.; Choi, K. S.; Park, Y. J.; Cho, Y. H.

    2007-06-01

    The final objective of this study is to develop a method for reducing radioactive materials formed in the reactor coolant circuit. This second stage research was categorized into the following three subgroups: the development of the estimation technique of microscopic chemical variation at high temperatures and pressures, the fundamental study on the thermodynamics at high temperatures and pressures, and the study on the deposition of metal oxides and the determination of the main factors responsible for the growth of CRUD. First, in the development of the estimation technique of microscopic chemical change at high temperatures and pressures, the technique for measuring coolant chemistry such as pH, conductivity and Eh was developed to be appropriate for the high temperature and pressure condition. The coolant chemistry measuring system including the self-devised high temperature pH sensor can be applied to the field of nuclear reactor and contribute on a large scale in the automation of the coolant chemistry control and the establishment of the real-time on-line measuring technique. Secondly, the dissociation constant of water and the solubility of metal oxides were measured in the fundamental study on the thermodynamics at high temperatures and pressures. Finally, in the study on the deposition of metal oxides and the determination of the main factors responsible for the growth of CRUD, the careful investigation of the deposition phenomena of micro particles on the cladding surface showed that subcooled boiling and the dissolved hydrogen are the main factors responsible for the growth of CRUD. In addition, the basis was provided for the construction of a new particle behavior model in the reactor coolant circuit

  1. Archaeology of Void Spaces

    Science.gov (United States)

    Look, Cory

    The overall goal of this research is to evaluate the efficacy of pXRF for the identification of ancient activity areas at Pre-Columbian sites in Antigua that range across time periods, geographic regions, site types with a variety of features, and various states of preservation. These findings have important implications for identifying and reconstructing places full of human activity but void of material remains. A synthesis for an archaeology of void spaces requires the construction of new ways of testing anthrosols, and identifying elemental patterns that can be used to connect people with their places and objects. This research begins with an exploration of rich middens in order to study void spaces. Midden archaeology has been a central focus in Caribbean research, and consists of an accumulation of discarded remnants from past human activities that can be tested against anthrosols. The archaeological collections visited for this research project involved creating new databases to generate a comprehensive inventory of sites, materials excavated, and assemblages available for study. Of the more than 129 Pre-Columbian sites documented in Antigua, few sites have been thoroughly surveyed or excavated. Twelve Pre-Columbian sites, consisting of thirty-six excavated units were selected for study; all of which contained complete assemblages for comparison and soil samples for testing. These excavations consisted almost entirely of midden excavations, requiring new archaeological investigations to be carried out in spaces primarily void of material remains but within the village context. Over the course of three seasons excavations, shovel test pits, and soil augers were used to obtain a variety of anthrosols and archaeological assemblages in order to generate new datasets to study Pre-Columbian activity areas. The selection of two primary case study sites were used for comparison: Indian Creek and Doigs. Findings from this research indicate that accounting for the

  2. Three-dimensional simulations of void collapse in energetic materials

    Science.gov (United States)

    Rai, Nirmal Kumar; Udaykumar, H. S.

    2018-03-01

    The collapse of voids in porous energetic materials leads to hot-spot formation and reaction initiation. This work advances the current knowledge of the dynamics of void collapse and hot-spot formation using 3D reactive void collapse simulations in HMX. Four different void shapes, i.e., sphere, cylinder, plate, and ellipsoid, are studied. For all four shapes, collapse generates complex three-dimensional (3D) baroclinic vortical structures. The hot spots are collocated with regions of intense vorticity. The differences in the vortical structures for the different void shapes are shown to significantly impact the relative sensitivity of the voids. Voids of high surface area generate hot spots of greater intensity; intricate, highly contorted vortical structures lead to hot spots of corresponding tortuosity and therefore enhanced growth rates of reaction fronts. In addition, all 3D voids are shown to be more sensitive than their two-dimensional (2D) counterparts. The results provide physical insights into hot-spot formation and growth and point to the limitations of 2D analyses of hot-spot formation.

  3. Compartmentalized safety coolant injection system

    International Nuclear Information System (INIS)

    Johnson, F.T.

    1983-01-01

    A safety coolant injection system for nuclear reactors wherein a core reflood tank is provided to afford more reliable reflooding of the reactor core in the event of a break in one of the reactor coolant supply loops. Each reactor coolant supply loop is arranged in a separate compartment in the containment structure to contain and control the flow of spilled coolant so as to permit its use during emergency core cooling procedures. A spillway allows spilled coolant in the compartment to pass into the emergency water storage tank from where it can be pumped back to the reactor vessel. (author)

  4. Assessment of Candidate Molten Salt Coolants for the Advanced High Temperature Reactor (AHTR)

    Energy Technology Data Exchange (ETDEWEB)

    Williams, D.F.

    2006-03-24

    exhibit better heat transfer and nuclear performance metrics. Lighter salts also tend to have more favorable (larger) moderating ratios, and thus should have a more favorable coolant-voiding behavior in-core. Heavy (high-Z) salts tend to have lower heat capacities and thermal conductivities and more significant activation and transmutation products. However, all of the salts are relatively good heat-transfer agents. A detailed discussion of each property and the combination of properties that served as a heat-transfer metric is presented in the body of this report. In addition to neutronic metrics, such as moderating ratio and neutron absorption, the activation properties of the salts were investigated (Table C). Again, lighter salts tend to have more favorable activation properties compared to salts with high atomic-number constituents. A simple model for estimating the reactivity coefficients associated with a reduction of salt content in the core (voiding or thermal expansion) was also developed, and the primary parameters were investigated. It appears that reasonable design flexibility exists to select a safe combination of fuel-element design and salt coolant for most of the candidate salts. Materials compatibility is an overriding consideration for high-temperature reactors; therefore the question was posed whether any one of the candidate salts was inherently, or significantly, more corrosive than another. This is a very complex subject, and it was not possible to exclude any fluoride salts based on the corrosion database. The corrosion database clearly indicates superior container alloys, but the effect of salt identity is masked by many factors which are likely more important (impurities, redox condition) in the testing evidence than salt identity. Despite this uncertainty, some reasonable preferences can be recommended, and these are indicated in the conclusions. The reasoning to support these conclusions is established in the body of this report.

  5. Revised Mark 22 coolant temperature coefficients

    International Nuclear Information System (INIS)

    Graves, W.E.

    1987-01-01

    Coolant temperature coefficients for the Mark 22 charge published previously are non-conservative because of the neglect of a significant mechanism which has a positive contribution to reactivity. Even after correcting for this effect, dynamic tests made on a Mark VIB charge in the early 60's suggest the results are still non-conservative. This memorandum takes both of these sources of information into account in making a best estimate of the prompt (coolant plus metal) temperature coefficient. Although no safety issues arise from this work (the overall temperature coefficient still strongly contributes to reactor stability), it is obviously desirable to use best estimates for prompt coefficients in limits and other calculations

  6. Displacive stability of a void in a void lattice

    International Nuclear Information System (INIS)

    Brailsford, A.D.

    1977-01-01

    It has recently been suggested that the stability of the void-lattice structure in irradiated metals may be attributed to the effect of the overlapping of the point-defect diffusion fields associated with each void. It is shown here, however, that the effect is much too weak. When one void is displaced from its lattice site, the displacement is shown to relax to zero as proposed, but a conservative estimate indicates that the characteristic time is equivalent to an irradiation dose of the order of 300 displacements per atom which is generally much greater than the dose necessary for void-lattice formation

  7. CTF Void Drift Validation Study

    Energy Technology Data Exchange (ETDEWEB)

    Salko, Robert K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gosdin, Chris [Pennsylvania State Univ., University Park, PA (United States); Avramova, Maria N. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gergar, Marcus [Pennsylvania State Univ., University Park, PA (United States)

    2015-10-26

    This milestone report is a summary of work performed in support of expansion of the validation and verification (V&V) matrix for the thermal-hydraulic subchannel code, CTF. The focus of this study is on validating the void drift modeling capabilities of CTF and verifying the supporting models that impact the void drift phenomenon. CTF uses a simple turbulent-diffusion approximation to model lateral cross-flow due to turbulent mixing and void drift. The void drift component of the model is based on the Lahey and Moody model. The models are a function of two-phase mass, momentum, and energy distribution in the system; therefore, it is necessary to correctly model the ow distribution in rod bundle geometry as a first step to correctly calculating the void distribution due to void drift.

  8. Fuel-Coolant Interactions: Visualization and Mixing Measurements

    International Nuclear Information System (INIS)

    Loewen, Eric P.; Bonazza, Riccardo; Corradini, Michael L.; Johannesen, Robert E.

    2002-01-01

    Dynamic X-ray imaging of fuel-coolant interactions (FCI), including quantitative measurement of fuel-coolant volume fractions and length scales, has been accomplished with a novel imaging system at the Nuclear Safety Research Center at the University of Wisconsin, Madison. The imaging system consists of visible-light high-speed digital video, low-energy X-ray digital imaging, and high-energy X-ray digital imaging subsystems. The data provide information concerning the melt jet velocity, melt jet configuration, melt volume fractions, void fractions, and spatial and temporal quantification of premixing length scales for a model fuel-coolant system of molten lead poured into a water pool (fuel temperatures 500 to 1000 K; jet diameters 10 to 30 mm; coolant temperatures 20 to 90 deg. C). Overall results indicate that the FCI has three general regions of behavior, with the high fuel-coolant temperature region similar to what might be expected under severe accident conditions. It was observed that the melt jet leading edge has the highest void fraction and readily fragments into discrete masses, which then subsequently subdivide into smaller masses of length scales <10 mm. The intact jet penetrates <3 to 5 jet length/jet diameter before this breakup occurs into discrete masses, which continue to subdivide. Hydrodynamic instabilities can be visually identified at the leading edge and along the jet column with an interfacial region that consists of melt, vapor, and water. This interface region was observed to grow in size as the water pool temperature was increased, indicating mixing enhancement by boiling processes

  9. Shock-induced hotspot formation and chemical reaction initiation in PETN containing a spherical void

    International Nuclear Information System (INIS)

    Shan, Tzu-Ray; Thompson, Aidan P

    2014-01-01

    We present results of reactive molecular dynamics simulations of hotspot formation and chemical reaction initiation in shock-induced compression of pentaerythritol tetranitrate (PETN) with the ReaxFF reactive force field. A supported shockwave is driven through a PETN crystal containing a 20 nm spherical void at a sub-threshold impact velocity of 2 km/s. Formation of a hotspot due to shock-induced void collapse is observed. During void collapse, NO 2 is the dominant species ejected from the upstream void surface. Once the ejecta collide with the downstream void surface and the hotspot develops, formation of final products such as N 2 and H 2 O is observed. The simulation provides a detailed picture of how void collapse and hotspot formation leads to initiation at sub-threshold impact velocities.

  10. Simulation of steam explosion in stratified melt-coolant configuration

    International Nuclear Information System (INIS)

    Leskovar, Matjaž; Centrih, Vasilij; Uršič, Mitja

    2016-01-01

    Highlights: • Strong steam explosions may develop spontaneously in stratified configurations. • Considerable melt-coolant premixed layer formed in subcooled water with hot melts. • Analysis with MC3D code provided insight into stratified steam explosion phenomenon. • Up to 25% of poured melt was mixed with water and available for steam explosion. • Better instrumented experiments needed to determine dominant mixing process. - Abstract: A steam explosion is an energetic fuel coolant interaction process, which may occur during a severe reactor accident when the molten core comes into contact with the coolant water. In nuclear reactor safety analyses steam explosions are primarily considered in melt jet-coolant pool configurations where sufficiently deep coolant pool conditions provide complete jet breakup and efficient premixture formation. Stratified melt-coolant configurations, i.e. a molten melt layer below a coolant layer, were up to now believed as being unable to generate strong explosive interactions. Based on the hypothesis that there are no interfacial instabilities in a stratified configuration it was assumed that the amount of melt in the premixture is insufficient to produce strong explosions. However, the recently performed experiments in the PULiMS and SES (KTH, Sweden) facilities with oxidic corium simulants revealed that strong steam explosions may develop spontaneously also in stratified melt-coolant configurations, where with high temperature melts and subcooled water conditions a considerable melt-coolant premixed layer is formed. In the article, the performed study of steam explosions in a stratified melt-coolant configuration in PULiMS like conditions is presented. The goal of this analytical work is to supplement the experimental activities within the PULiMS research program by addressing the key questions, especially regarding the explosivity of the formed premixed layer and the mechanisms responsible for the melt-water mixing. To

  11. Dual coolant blanket concept

    International Nuclear Information System (INIS)

    Malang, S.; Schleisiek, K.

    1994-11-01

    A self-cooled liquid metal breeder blanket with helium-cooled first wall ('Dual Coolant Blanket Concept') for a fusion DEMO reactor is described. This is one of the four blanket concepts under development in the frame of the European fusion technology program with the aim to select in 1995 the two most promising ones for further development. Described are the design of the blankets including the ancillary loop system and the results of the theoretical and experimental work in the fields of neutronics, magnetohydrodynamics, thermohydraulics, mechanical stresses, compatibility and purification of lead-lithium, tritium control, safety, reliability, and electrically insulating coatings. The remaining open questions and the required R and D programme are identified. (orig.) [de

  12. Effect of main stream void distribution on cavitating hydrofoil

    International Nuclear Information System (INIS)

    Ito, J.

    1993-01-01

    For the safety analysis of a loss of coolant accident in a pressurized water reactor, it is important to establish an analytical method which predicts the pump performance under gas-liquid two-phase flow condition. J.H. Kim briefly reviewed several major two-phase flow pump models, and discussed the parameters that could significantly affect two-phase pump behavior. The parameter pointed out to be of the most importance is void distribution at the pump inlet. This says that the pipe bend near the pump inlet makes the void distribution at the pump inlet nonuniform, and this matter can have a significant effect on the impeller blade performance. This paper proposes an analytical method of solution for a partially cavitating hydrofoil placed in the main stream of incompressible homogeneous bubbly two-phase flow conditions whose void fraction is exponentially distributed normal to chordline. The paper clarifies the effect of main stream void distribution parameter on the partially cavitating hydrofoil characteristics

  13. PRECISION COSMOGRAPHY WITH STACKED VOIDS

    International Nuclear Information System (INIS)

    Lavaux, Guilhem; Wandelt, Benjamin D.

    2012-01-01

    We present a purely geometrical method for probing the expansion history of the universe from the observation of the shape of stacked voids in spectroscopic redshift surveys. Our method is an Alcock-Paczyński (AP) test based on the average sphericity of voids posited on the local isotropy of the universe. It works by comparing the temporal extent of cosmic voids along the line of sight with their angular, spatial extent. We describe the algorithm that we use to detect and stack voids in redshift shells on the light cone and test it on mock light cones produced from N-body simulations. We establish a robust statistical model for estimating the average stretching of voids in redshift space and quantify the contamination by peculiar velocities. Finally, assuming that the void statistics that we derive from N-body simulations is preserved when considering galaxy surveys, we assess the capability of this approach to constrain dark energy parameters. We report this assessment in terms of the figure of merit (FoM) of the dark energy task force and in particular of the proposed Euclid mission which is particularly suited for this technique since it is a spectroscopic survey. The FoM due to stacked voids from the Euclid wide survey may double that of all other dark energy probes derived from Euclid data alone (combined with Planck priors). In particular, voids seem to outperform baryon acoustic oscillations by an order of magnitude. This result is consistent with simple estimates based on mode counting. The AP test based on stacked voids may be a significant addition to the portfolio of major dark energy probes and its potentialities must be studied in detail.

  14. Coolant channel module CCM

    International Nuclear Information System (INIS)

    Hoeld, Alois

    2007-01-01

    A complete and detailed description of the theoretical background of an '(1D) thermal-hydraulic drift-flux based mixture-fluid' coolant channel model and its resulting module CCM will be presented. The objective of this module is to simulate as universally as possible the steady state and transient behaviour of the key characteristic parameters of a single- or two-phase fluid flowing within any type of heated or non-heated coolant channel. Due to the possibility that different flow regimes can appear along any channel, such a 'basic (BC)' 1D channel is assumed to be subdivided into a number of corresponding sub-channels (SC-s). Each SC can belong to only two types of flow regime, an SC with just a single-phase fluid, containing exclusively either sub-cooled water or superheated steam, or an SC with a two-phase mixture flow. After an appropriate nodalisation of such a BC (and therefore also its SC-s) a 'modified finite volume method' has been applied for the spatial discretisation of the partial differential equations (PDE-s) which represent the basic conservation equations of thermal-hydraulics. Special attention had to be given to the possibility of variable SC entrance or outlet positions (which describe boiling boundaries or mixture levels) and thus the fact that an SC can even disappear or be created anew. The procedure yields for each SC type (and thus the entire BC), a set of non-linear ordinary 1st order differential equations (ODE-s). To link the resulting mean nodal with the nodal boundary function values, both of which are present in the discretised differential equations, a special quadratic polygon approximation procedure (PAX) had to be constructed. Together with the very thoroughly tested packages for drift-flux, heat transfer and single- and two-phase friction factors this procedure represents the central part of the here presented 'Separate-Region' approach, a theoretical model which provides the basis to the very effective working code package CCM

  15. Thorium Fuel Utilization Analysis on Small Long Life Reactor for Different Coolant Types

    Science.gov (United States)

    Permana, Sidik

    2017-07-01

    A small power reactor and long operation which can be deployed for less population and remote area has been proposed by the IAEA as a small and medium reactor (SMR) program. Beside uranium utilization, it can be used also thorium fuel resources for SMR as a part of optimalization of nuclear fuel as a “partner” fuel with uranium fuel. A small long-life reactor based on thorium fuel cycle for several reactor coolant types and several power output has been evaluated in the present study for 10 years period of reactor operation. Several key parameters are used to evaluate its effect to the reactor performances such as reactor criticality, excess reactivity, reactor burnup achievement and power density profile. Water-cooled types give higher criticality than liquid metal coolants. Liquid metal coolant for fast reactor system gives less criticality especially at beginning of cycle (BOC), which shows liquid metal coolant system obtains almost stable criticality condition. Liquid metal coolants are relatively less excess reactivity to maintain longer reactor operation than water coolants. In addition, liquid metal coolant gives higher achievable burnup than water coolant types as well as higher power density for liquid metal coolants.

  16. Reactivity initiated accidents and loss of shutdown - 20 years later

    International Nuclear Information System (INIS)

    Luxat, J.C.

    2007-01-01

    A review of the safety of Ontario's nuclear power reactors was conducted in 1987 after the Chernobyl accident. As part of this review an analysis was performed of a Loss of Coolant Accident in a Pickering A unit with coincident failure to shutdown. This analysis showed that the power excursion was halted by channel and calandria vessel failures leading to moderator fluid displacement. The containment structure did not fail and, at worst might suffer minor cracking at the top of the dome of the reactor building. Overall the dose consequences of such an accident were no worse than the limiting design basis dual failure event. In the intervening twenty years following this analysis, Significant experimental information has been obtained that relates to power pulse behaviour. This information, together with conservatisms in he original analysis, are reviewed and assessed in this paper. In addition, the issue of reactivity initiated events in other reactor types is reviewed to identify the reactor design characteristics that are of importance in these events. Contrary to popular belief the existence of positive coolant void reactivity is not as significant a factor as it is sometimes stated to be. On balance, with appropriate design measures, no one reactor type can be claimed to be 'more safe' than another. The underlying basis for this statement is articulated in this paper. (author)

  17. Fuel-Coolant Interactions - some Basic Studies at the UKAEA Culham Laboratory

    International Nuclear Information System (INIS)

    Reynolds, J.A.; Dullforce, T.A.; Peckover, R.S.; Vaughan, G.J.

    1976-01-01

    In a hypothetical fault sequence important effects of fuel-coolant interactions include voiding and dispersion of core debris as well as the pressure damage usually discussed. The development of the fuel-coolant interaction probably depends on any pre-mixing Weber break-up that may occur, and is therefore a function of the way the fuel and coolant come together. Four contact modes are identified: jetting, shock tube, drops and static, and Culham's experiments have been mainly concerned with simulating the falling drop mode by using molten tin in water. It was observed that the fuel-coolant interaction is a short series of violent coolant oscillations centred at a localized position on the drop, generating a spray of submillimeter sized debris. The interaction started spontaneously at a specific time after the drop first contacted the water. There was a definite limited fuel-coolant interaction zone on a plot of initial coolant temperature versus initial fuel temperature outside which interactions never occurred. The. interaction time was a function of the initial temperatures. Theoretical scaling formulae are given which describe the fuel-coolant interaction zone and dwell time. Bounds of fuel and coolant temperature below which fuel-coolant interactions do not occur are explained by freezing. Upper bounds of fuel and coolant temperatures above which there were no fuel-coolant interactions are interpreted in terms of heat transfer through vapour films of various thicknesses. In conclusion: We have considered the effects of fuel-coolant interactions in a hypothetical fault sequence, emphasising that debris and vapour production as well as the pressure pulse can be important factors. The fuel-coolant interaction has been classified into types, according to possible modes of mixing in the fault sequence. Culham has been studying one type, the self-triggering of falling drops, by simulant experiments. It is found that there is a definite zone of interaction on a plot

  18. Air void clustering : [technical summary].

    Science.gov (United States)

    2015-06-01

    Air void clustering around coarse aggregate in concrete has been : identified as a potential source of low strengths in concrete mixes by : several Departments of Transportation around the country. Research : was carried out to (1) develop a quantita...

  19. Secondary coolant purification system

    International Nuclear Information System (INIS)

    Stiteler, F.Z.; Donohue, J.P.

    1978-01-01

    The present invention combines the attributes of volatile chemical addition, continuous blowdown, and full flow condensate demineralization. During normal plant operation (defined as no primary to secondary leakage) condensate from the condenser is pumped through a full flow condensate demineralizer system by the condensate pumps. Volatile chemical additions are made. Dissolved and suspended solids are removed in the condensate polishers by ion exchange and/or filtration. At the same time a continuous blowdown of approximately 1 percent of the main steaming rate of the steam generators is maintained. Radiation detectors monitor the secondary coolant. If these monitors indicate no primary to secondary leakage, the blowdown is cooled and returned directly to the condensate pump discharge. If one of the radiation monitors should indicate a primary to secondary leak, when the temperature of the effluent exiting from the blowdown heat exchanger is compatible with the resin specifications of the ion exchangers, the bypass valve causes the blowdown flow to pass through the blowdown ion exchangers

  20. Determination of void fraction from source range monitor and mass flow rate data

    International Nuclear Information System (INIS)

    McCormick, R.D.

    1986-09-01

    This is a report on the calculation of the TMI-2 primary coolant system local void fraction from source range neutron flux monitor data and from hot leg mass flowrate meter data during the first 100 minutes of the accident. The methods of calculation of void fraction from the two data sources is explained and the results are compared. It is indicated that the void fraction determined using the mass flowrate data contained an error of unknown magnitude due to the assumption of constant homogeneous volumetric flowrate used in the calculation and required further work. Void fraction determined from the source range monitor data is felt to be usable although an uncertainty analysis has not been performed

  1. Analysis of the reactivity coefficients of the advanced high-temperature reactor for plutonium and uranium fuels

    International Nuclear Information System (INIS)

    Zakova, Jitka; Talamo, Alberto

    2008-01-01

    The conceptual design of the advanced high-temperature reactor (AHTR) has recently been proposed by the Oak Ridge National Laboratory, with the intention to provide and alternative energy source for very high temperature applications. In the present study, we focused on the analyses of the reactivity coefficients of the AHTR core fueled with two types of fuel: enriched uranium and plutonium from the reprocessing of light water reactors irradiated fuel. More precisely, we investigated the influence of the outer graphite reflectors on the multiplication factor of the core, the fuel and moderator temperature reactivity coefficients and the void reactivity coefficient for five different molten salts: NaF, BeF 2 , LiF, ZrF 4 and Li 2 BeF 4 eutectic. In order to better illustrate the behavior of the previous parameters for different core configurations, we evaluated the moderating ratio of the molten salts and the absorption rate of the key fuel nuclides, which, of course, are driven by the neutron spectrum. The results show that the fuel and moderator temperature reactivity coefficients are always negative, whereas the void reactivity coefficient can be set negative provided that the fuel to moderator ratio is optimized (the core is undermoderated) and the moderating ratio of the coolant is large

  2. Analysis of the reactivity coefficients of the advanced high-temperature reactor for plutonium and uranium fuels

    Energy Technology Data Exchange (ETDEWEB)

    Zakova, Jitka [Department of Nuclear and Reactor Physics, Royal Institute of Technology, KTH, Roslagstullsbacken 21, S-10691, Stockholm (Sweden)], E-mail: jitka.zakova@neutron.kth.se; Talamo, Alberto [Nuclear Engineering Division, Argonne National Laboratory, ANL, 9700 South Cass Avenue, Argonne, IL 60439 (United States)], E-mail: alby@anl.gov

    2008-05-15

    The conceptual design of the advanced high-temperature reactor (AHTR) has recently been proposed by the Oak Ridge National Laboratory, with the intention to provide and alternative energy source for very high temperature applications. In the present study, we focused on the analyses of the reactivity coefficients of the AHTR core fueled with two types of fuel: enriched uranium and plutonium from the reprocessing of light water reactors irradiated fuel. More precisely, we investigated the influence of the outer graphite reflectors on the multiplication factor of the core, the fuel and moderator temperature reactivity coefficients and the void reactivity coefficient for five different molten salts: NaF, BeF{sub 2}, LiF, ZrF{sub 4} and Li{sub 2}BeF{sub 4} eutectic. In order to better illustrate the behavior of the previous parameters for different core configurations, we evaluated the moderating ratio of the molten salts and the absorption rate of the key fuel nuclides, which, of course, are driven by the neutron spectrum. The results show that the fuel and moderator temperature reactivity coefficients are always negative, whereas the void reactivity coefficient can be set negative provided that the fuel to moderator ratio is optimized (the core is undermoderated) and the moderating ratio of the coolant is large.

  3. Analysis of sodium-void experiments in ZPPR-3 modified Phase 3 core

    Energy Technology Data Exchange (ETDEWEB)

    Yoshida, T.

    1978-08-01

    An analysis is presented of a series of sodium-void reactivity measurements performed in assembly 3 of Zero Power Plutonium Reactor (ZPPR-3), a mockup of the US Demoplant. In this series, large-zone sodium-void effects were studied in detail in the presence of many singularities, namely, control rods (CRs) and control rod positions (CRPs). The Karlsruhe data-and-method have been applied to an analysis of these experiments, and the results are presented. The work is aimed at complementing the sodium-void reactivity analysis based on the SNEAK experiments, where it was difficult to simulate a large plutonium-core of a prototype fast breeder reactor.

  4. Decontamination of main coolant pumps

    International Nuclear Information System (INIS)

    Roofthooft, R.

    1988-01-01

    Last year a number of main coolant pumps in Belgian nuclear power plants were decontaminated. A new method has been developed to reduce the time taken for decontamination and the volume of waste to be treated. The method comprises two phases: Oxidation with permanganate in nitric acid and dissolution in oxalic acid. The decontamination of main coolant pumps can now be achieved in less than one day. The decontamination factors attained range between 15 and 150. (orig.) [de

  5. Alignment of voids in the cosmic web

    NARCIS (Netherlands)

    Platen, Erwin; van de Weygaert, Rien; Jones, Bernard J. T.

    2008-01-01

    We investigate the shapes and mutual alignment of voids in the large-scale matter distribution of a Lambda cold dark matter (Lambda CDM) cosmology simulation. The voids are identified using the novel watershed void finder (WVF) technique. The identified voids are quite non-spherical and slightly

  6. On cavitation instabilities with interacting voids

    DEFF Research Database (Denmark)

    Tvergaard, Viggo

    2012-01-01

    voids so far apart that the radius of the plastic zone around each void is less than 1% of the current spacing between the voids, can still affect each others at the occurrence of a cavitation instability such that one void stops growing while the other grows in an unstable manner. On the other hand...

  7. Coolant circuit water chemistry of the Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    Tilky, Peter; Doma, Arpad

    1985-01-01

    The numerous advantages of the proper selection of water chemistry parameters including low corrosion rate of the structural materials, hence the low-level activity build-up, depositions, radiation doses were emphasized. Major characteristics of water chemistry applied to the primary coolant of pressurized water reactors including neutral, slightly basic and strong basic ones are discussed. Boric acid is widely used to control reactivity. Primary coolant water chemistry of WWER type reactors which is based on the addition of ammonia and potassium hydroxide to boric acid is compared with that of other reactors. The demineralization of the total condensate of the steam turbines became a general trend in the water chemistry of the secondary coolant circuits. (V.N.)

  8. Fuel cycles with high fuel burn-up: analysis of reactivity coefficients

    International Nuclear Information System (INIS)

    Kryuchkov, E.F.; Shmelev, A.N.; Ternovykh, M.J.; Tikhomirov, G.V.; Jinhong, L.; Saito, M.

    2003-01-01

    Fuel cycles of light-water reactors (LWR) with high fuel burn-up (above 100 MWd/kg), as a rule, involve large amounts of fissionable materials. It leads to forming the neutron spectrum harder than that in traditional LWR. Change of neutron spectrum and significant amount of non-traditional isotopes (for example, 237 Np, 238 Pu, 231 Pa, 232 U) in such fuel compositions can alter substantially reactivity coefficients as compared with traditional uranium-based fuel. The present work addresses the fuel cycles with high fuel burn-up which are based on Th-Pa-U and U-Np-Pu fuel compositions. Numerical analyses are carried out to determine effective neutron multiplication factor and void reactivity coefficient (VRC) for different values of fuel burn-up and different lattice parameters. The algorithm is proposed for analysis of isotopes contribution to these coefficients. Various ways are considered to upgrade safety of nuclear fuel cycles with high fuel burn-up. So, the results obtained in this study have demonstrated that: -1) Non-traditional fuel compositions developed for achievement of high fuel burn-up in LWR can possess positive values of reactivity coefficients that is unacceptable from the reactor operation safety point of view; -2) The lattice pitch of traditional LWR is not optimal for non-traditional fuel compositions, the increased value of the lattice pitch leads to larger value of initial reactivity margin and provides negative VRC within sufficiently broad range of coolant density; -3) Fuel burn-up has an insignificant effect on VRC dependence on coolant density, so, the measures undertaken to suppress positive VRC of fresh fuel will be effective for partially burnt-up fuel compositions also and; -4) Increase of LWR core height and introduction of additional moderators into the fuel lattice can be used as the ways to reach negative VRC values for full range of possible coolant density variations

  9. Fuel cycles with high fuel burn-up: analysis of reactivity coefficients

    Energy Technology Data Exchange (ETDEWEB)

    Kryuchkov, E.F.; Shmelev, A.N.; Ternovykh, M.J.; Tikhomirov, G.V.; Jinhong, L. [Moscow Engineering Physics Institute (State University) (Russian Federation); Saito, M. [Tokyo Institute of Technology (Japan)

    2003-07-01

    Fuel cycles of light-water reactors (LWR) with high fuel burn-up (above 100 MWd/kg), as a rule, involve large amounts of fissionable materials. It leads to forming the neutron spectrum harder than that in traditional LWR. Change of neutron spectrum and significant amount of non-traditional isotopes (for example, {sup 237}Np, {sup 238}Pu, {sup 231}Pa, {sup 232}U) in such fuel compositions can alter substantially reactivity coefficients as compared with traditional uranium-based fuel. The present work addresses the fuel cycles with high fuel burn-up which are based on Th-Pa-U and U-Np-Pu fuel compositions. Numerical analyses are carried out to determine effective neutron multiplication factor and void reactivity coefficient (VRC) for different values of fuel burn-up and different lattice parameters. The algorithm is proposed for analysis of isotopes contribution to these coefficients. Various ways are considered to upgrade safety of nuclear fuel cycles with high fuel burn-up. So, the results obtained in this study have demonstrated that: -1) Non-traditional fuel compositions developed for achievement of high fuel burn-up in LWR can possess positive values of reactivity coefficients that is unacceptable from the reactor operation safety point of view; -2) The lattice pitch of traditional LWR is not optimal for non-traditional fuel compositions, the increased value of the lattice pitch leads to larger value of initial reactivity margin and provides negative VRC within sufficiently broad range of coolant density; -3) Fuel burn-up has an insignificant effect on VRC dependence on coolant density, so, the measures undertaken to suppress positive VRC of fresh fuel will be effective for partially burnt-up fuel compositions also and; -4) Increase of LWR core height and introduction of additional moderators into the fuel lattice can be used as the ways to reach negative VRC values for full range of possible coolant density variations.

  10. Air void structure and frost resistance

    DEFF Research Database (Denmark)

    Hasholt, Marianne Tange

    2014-01-01

    ). This observation is interesting as the parameter of total surface area of air voids normally is not included in air void analysis. The following reason for the finding is suggested: In the air voids conditions are favourable for ice nucleation. When a capillary pore is connected to an air void, ice formation...... on that capillary pores are connected to air voids. The chance that a capillary pore is connected to an air void depends on the total surface area of air voids in the system, not the spacing factor.......This article compiles results from 4 independent laboratory studies. In each study, the same type of concrete is tested at least 10 times, the air void structure being the only variable. For each concrete mix both air void analysis of the hardened concrete and a salt frost scaling test...

  11. Labelling Of Coolant Flow Anomaly Using Fractal Structure

    International Nuclear Information System (INIS)

    Djainal, Djen Djen

    1996-01-01

    This research deals with the instrumentation of the detection and characterization of vertical two-phase flow coolant. This type of work is particularly intended to find alternative method for the detection and identification of noise in vertical two-phase flow in a nuclear reactor environment. Various new methods have been introduced in the past few years, an attempt to developed an objective indicator off low patterns. One of new method is Fractal analysis which can complement conventional methods in the description of highly irregular fluctuations. In the present work, Fractal analysis was applied to analyze simulated boiling coolant signal. This simulated signals were built by sum random elements in small subchannels of the coolant channel. Two modes are defined and both are characterized by their void fractions. In the case of uni modal -PDF signals, the difference between these modes is relatively small. On other hand, bimodal -PDF signals have relative large range. In this research, Fractal dimension can indicate the characters of that signals simulation

  12. Reactor coolant pump transportation incident

    International Nuclear Information System (INIS)

    Noce, D.

    1992-01-01

    This paper reports on an incident, which occurred on August 27, 1991, in which a Reactor Coolant Pump motor en route from Surry Power Station to Westinghouse repair facilities struck the overpass at the junction of Interstate 64 and Jefferson Avenue in Newport News, Virginia. The transport container that housed the reactor coolant pump motor failed to clear the overpass. The force of the impact dislodged the container and motor from the truck bed, and it landed on the acceleration land and road shoulder. Upon impact, the container broke open and exposed the reactor coolant pump motor. Incidental radioactively contaminated water that remained in the motor coolers drained onto the road, contaminating the aggregate as well as the underlying gravel

  13. Finding Brazing Voids by Holography

    Science.gov (United States)

    Galluccio, R.

    1986-01-01

    Vibration-induced interference fringes reveal locations of defects. Holographic apparatus used to view object while vibrated ultrasonically. Interference fringes in hologram reveal brazing defects. Holographic technique locates small voids in large brazed joints. Identifies unbrazed regions 1 in. to second power (6 cm to the second power) or less in area.

  14. Fast instrumentation for loss of coolant accident (LOCA) experimental studies pertaining to nuclear reactors

    International Nuclear Information System (INIS)

    Venkat Raj, V.; Sreenivas Rao, G.; Belokar, D.G.; Dolas, P.K.

    1989-01-01

    The loss of coolant accident (LOCA) which involves a breach in the pressure boundary of the primary coolant system (PCS) is one of the postulated accident conditions against which the safety of the reactor system is to be ensured. Mathematical models have been developed to analyse this kind of transients. However, because of the extremely complicated nature of the phenomena involved, it is necessary to validate the analytical models with appropriate experimental data. Many parameters are to be measured during the experiments, out of which temperature, pressure, void fraction and two-phase mass flow rate are the most important parameters. Since the phenomenon is very fast, special fast response instruments are required. This paper deals with the considerations that govern the selection of appropriate instruments and the development of suitable instruments for transient two-phase flow and void fraction measurements. The requirements of the associated fast data acquisition system are also discussed. (author). 4 figs

  15. Full sized tests on a french coolant pump under two-phase flow

    International Nuclear Information System (INIS)

    Huchard, J.C.; Bore, C.; Dueymes, E.

    1997-01-01

    The French Safety Authorities required EDF to demonstrate the ability of the new N4 main coolant pump to withstand two-phase flow conditions without damage. Therefore three full sized tests, simulating a bleeding flow on the primary system, were performed on a laboratory test loop under real operating conditions (temperature = 290 deg. C, pressure = 155 b, flowrate = 7 m 3 /s; electrical power = 7 MW). The maximum value of the mean void fraction reached 75 %. The outcome of the tests is very positive: the mechanical behaviour of the main coolant pump is good, even at high void fraction. The maximum vibration levels were below the limits fixed by the manufacturer. Correlations between the mechanical behaviour of the pump and the pressure pulsation in the test loop have been found. (authors)

  16. Analysis Of Primary Coolant Suction Side Pressure In The Delay Chamber Of The RSG-GAS

    International Nuclear Information System (INIS)

    Dibyo, Sukmanto

    2000-01-01

    Delay chamber is a tank to delay flow that located in the primary cooling suction side of RSG-GAS. A void occurred when operation reactor caused by too high the delta P at inlet suction pump. The condition may be avoided by using one line mode of the cooling flow. The analysis show that void volume in the delay chamber is occurred because the coolant negative pressure lowers the saturation pressure should be avoided though decreasing the delta P until about 0.1 bar at about 45 exp 0 C. Solution suggested are to use bypass flow from the spent fuel to the delay chamber. Coolant temperature can be also decreased by decreasing the power level of the reactor as well as improving the heat exchanger and cooling tower performances

  17. Sodium as a reactor coolant

    International Nuclear Information System (INIS)

    Cesar, S.B.G.

    1989-01-01

    This work is related to the use of sodium as a reactor coolant, to the advantages and problems related to its use, its mechanical, thermophysics, eletronical, magnetic and nuclear properties. It is mainly a bibliographic review, with the aim of gathering the necessary information to persons initiating in the study of sodium and also as reference source. (author) [pt

  18. Vertical reactor coolant pump instabilities

    International Nuclear Information System (INIS)

    Jones, R.M.

    1985-01-01

    The investigation conducted at the Tennessee Valley Authority's Sequoyah Nuclear Power Plant to determine and correct increasing vibrations in the vertical reactor coolant pumps is described. Diagnostic procedures to determine the vibration causes and evaluate the corractive measures taken are also described

  19. Organic coolants and their applications to fusion reactors

    International Nuclear Information System (INIS)

    Gierszewski, P.; Hollies, B.

    1986-08-01

    Organic coolants offer a unique set of characteristics for fusion applications. Their advantages include high-temperature (670 K or 400 degrees C) but low-pressure (2 MPa) operation, limited reactivity with lithium and lithium-lead, reduced corrosion and activation, good heat-transfer capabilities, no magnetohydrodynamic (MHD) effects, and an operating temperature range that extends to room temperature. The major disadvantages are decomposition and flammability. However, organic coolants have been extensively studied in Canada, including nineteen years with an operating 60-MW organic-cooled reactor. Proper attention to design and coolant chemistry controlled these potential problems to acceptable levels. This experience provides an extensive data base for design under fusion conditions. The organic fluid characteristics are described in sufficient detail to allow fusion system designers to evaluate organic coolants for specific applications. To illustrate and assess the potential applications, analyses are presented for organic-cooled blankets, first walls, high heat flux components and thermal power cycles. Designs are identified that take advantage of organic coolant features, yet have fluid decomposition related costs that are a small fraction of the overall cost of electricity. For example, organic-cooled first walls make lithium/ferritic steel blankets possible in high-field, high-surface-heat-flux tokamaks, and organic-cooled limiters (up to about 8 MW/m 2 surface heating) are a safer alternative to water cooling for liquid metal blanket concept. Organics can also be used in intermediate heat exchanger loops to provide efficient heat transfer with low reactivity and a large tritium barrier. 55 refs

  20. CT measurements of SAP voids in concrete

    DEFF Research Database (Denmark)

    Laustsen, Sara; Bentz, Dale P.; Hasholt, Marianne Tange

    2010-01-01

    X-ray computed tomography (CT) scanning is used to determine the SAP void distribution in hardened concrete. Three different approaches are used to analyse a binary data set created from CT measurement. One approach classifies a cluster of connected, empty voxels (volumetric pixel of a 3D image......) as one void, whereas the other two approaches are able to classify a cluster of connected, empty voxels as a number of individual voids. Superabsorbent polymers (SAP) have been used to incorporate air into concrete. An advantage of using SAP is that it enables control of the amount and size...... of the created air voids. The results indicate the presence of void clusters. To identify the individual voids, special computational approaches are needed. The addition of SAP results in a dominant peak in two of the three air void distributions. Based on the position (void diameter) of the peak, it is possible...

  1. Nuclear reactor coolant and cover gas system

    International Nuclear Information System (INIS)

    George, J.A.; Redding, A.H.; Tower, S.N.

    1976-01-01

    A core cooling system is disclosed for a nuclear reactor of the type utilizing a liquid coolant with a cover gas above free surfaces of the coolant. The disclosed system provides for a large inventory of reactor coolant and a balanced low pressure cover gas arrangement. A flow restricting device disposed within a reactor vessel achieves a pressure of the cover gas in the reactor vessel lower than the pressure of the reactor coolant in the vessel. The low gas pressure is maintained over all free surfaces of the coolant in the cooling system including a coolant reservoir tank. Reactor coolant stored in the reservoir tank allows for the large reactor coolant inventory provided by the invention

  2. Void migration in fusion materials

    International Nuclear Information System (INIS)

    Cottrell, G.A.

    2002-01-01

    Neutron irradiation in a fusion power plant will cause helium bubbles and voids to form in the armour and blanket structural materials. If sufficiently large densities of such defects accumulate on the grain boundaries of the materials, the strength and the lifetimes of the metals will be reduced by helium embrittlement and grain boundary failure. This Letter discusses void migration in metals, both by random Brownian motion and by biassed flow in temperature gradients. In the assumed five-year blanket replacement time of a fusion power plant, approximate calculations show that the metals most resilient to failure are tungsten and molybdenum, and marginally vanadium. Helium embrittlement and grain boundary failure is expected to be more severe in steel and beryllium

  3. Void migration in fusion materials

    Science.gov (United States)

    Cottrell, G. A.

    2002-04-01

    Neutron irradiation in a fusion power plant will cause helium bubbles and voids to form in the armour and blanket structural materials. If sufficiently large densities of such defects accumulate on the grain boundaries of the materials, the strength and the lifetimes of the metals will be reduced by helium embrittlement and grain boundary failure. This Letter discusses void migration in metals, both by random Brownian motion and by biassed flow in temperature gradients. In the assumed five-year blanket replacement time of a fusion power plant, approximate calculations show that the metals most resilient to failure are tungsten and molybdenum, and marginally vanadium. Helium embrittlement and grain boundary failure is expected to be more severe in steel and beryllium.

  4. Spatial dependence of void coefficient in the University of Arizona TRIGA research reactor

    International Nuclear Information System (INIS)

    Spriggs, Gregory D.; Doane, Harry; Wells, Robert

    1980-01-01

    The spatial dependence of the moderator void coefficient of reactivity in the axial direction was experimentally measured in the A-ring using a hollow, air-filled aluminum cylinder. It was found that the void coefficient was positive in the central region of the fuel section reaching a maximum value of approximately + .045 cents/cm 3 and was negative towards the outer edges of the fuel section reaching a maximum of - .09 cents/cm 3 . (author)

  5. Coolant monitoring systems for PWR reactors

    International Nuclear Information System (INIS)

    Luzhnov, A.M.; Morozov, V.V.; Tsypin, S.G.

    1987-01-01

    The ways of improving information capacity of existing monitoring systems and the necessity of designing new ones for coolant monitoring are reviewed. A wide research program on development of coolant monitoring systems in PWR reactors is analyzed. The possible applications of in-core and out-of-core detectors for coolant monitoring are demonstrated

  6. Specificities of reactor coolant pumps units with lead and lead-bismuth coolant

    International Nuclear Information System (INIS)

    Beznosov, A.V.; Anotonenkov, M.A.; Bokov, P.A.; Baranova, V.S.; Kustov, M.S.

    2009-01-01

    The analysis results of impact of lead and lead-bismuth coolants specific properties on the coolants flow features in flow channels of the main and auxiliary circulating pumps are presented. Impossibility of cavitation initiation in flow channels of vane pumps pumping lead and lead-bismuth coolants was demonstrated. The experimental research results of discontinuity of heavy liquid metal coolant column were presented and conditions of gas cavitation initiation in coolant flow were discussed. Invalidity of traditional calculation methods of water and sodium coolants circulation pumps calculations for lead and lead-bismuth coolants circulation pumps was substantiated [ru

  7. Trace organics in AGR coolants

    International Nuclear Information System (INIS)

    Smith, R.; Green, L.O.; Johnson, P.A.V.

    1980-01-01

    Several analytical techniques have been employed in previous studies of the stable organic compounds arising from the radiolysis of methane/carbon monoxide/carbon dioxide coolants. The majority of this early information was collected from the Windscale AGR prototype. Analyses were also carried out on the liquors obtained from the WAGR humidryers. Three classes of compound were found in the liquors; aliphatic acids in the aqueous phase and methyl ketones and aromatic hydrocarbons in the oily phase. Acetic acid was found to be the predominant carboxylic acid. This paper outlines the major findings from a recent analytical survey of coolants taken over a wide range of dose rate, pressure, temperature and composition, from materials testing reactor facilities, WAGR and CAGR. (author)

  8. Size-Effects in Void Growth

    DEFF Research Database (Denmark)

    Niordson, Christian Frithiof

    2005-01-01

    The size-effect on ductile void growth in metals is investigated. The analysis is based on unit cell models both of arrays of cylindrical voids under plane strain deformation, as well as arrays of spherical voids using an axisymmetric model. A recent finite strain generalization of two higher order...... strain gradient plasticity models is implemented in a finite element program, which is used to study void growth numerically. The results based on the two models are compared. It is shown how gradient effects suppress void growth on the micron scale when compared to predictions based on conventional...... models. This increased resistance to void growth, due to gradient hardening, is accompanied by an increase in the overall strength for the material. Furthermore, for increasing initial void volume fraction, it is shown that the effect of gradients becomes more important to the overall response but less...

  9. Radiolytic reactions in the coolant of helium cooled reactors

    International Nuclear Information System (INIS)

    Tingey, G.L.; Morgan, W.C.

    1975-01-01

    The success of helium cooled reactors is dependent upon the ability to prevent significant reaction between the coolant and the other components in the reactor primary circuit. Since the thermal reaction of graphite with oxidizing gases is rapid at temperatures of interest, the thermal reactions are limited primarily by the concentration of impurity gases in the helium coolant. On the other hand, the rates of radiolytic reactions in helium are shown to be independent of reactive gas concentration until that concentration reaches a very low level. Calculated steady-state concentrations of reactive species in the reactor coolant and core burnoff rates are presented for current U. S. designed, helium cooled reactors. Since precise base data are not currently available for radiolytic rates of some reactions and thermal reaction rate data are often variable, the accuracy of the predicted gas composition is being compared with the actual gas compositions measured during startup tests of the Fort Saint Vrain high temperature gas-cooled reactor. The current status of these confirmatory tests is discussed. 12 references

  10. Coolant clean-up and recycle systems

    International Nuclear Information System (INIS)

    Ito, Takao.

    1979-01-01

    Purpose: To increase the service life of mechanical seals in a shaft sealing device, eliminate leakages and improve the safety by providing a recycle pump for feeding coolants to a coolant clean-up device upon reactor shut-down and adapting the pump treat only low temperature and low pressure coolants. Constitution: The system is adapted to partially take out coolants from the pipeways of a recycling pump upon normal operation and feed them to a clean-up device. Upon reactor shut-down, the recycle pump is stopped and coolants are extracted by the recycle pump for shut-down into the clean-up device. Since the coolants are not fed to the clean-up device by the recycle pump during normal operation as conducted so far, high temperature and high pressure coolants are not directly fed to the recycle pump, thereby enabling to avoid mechanical problems in the pump. (Kamimura, M.)

  11. Design comparisons of TRU burner cores with similar sodium void worth

    International Nuclear Information System (INIS)

    Sang Ji, Kim; Young Il, Kim; Young Jin, Kim; Nam Zin, Cho

    2001-01-01

    This study summarizes the neutronic performance and fuel cycle behavior of five geometrically-different transuranic (TRU) burner cores with similar low sodium void reactivity. The conceptual cores encompass core geometries for annular, two-region homogeneous, dual pin type, pan-shaped and H-shaped cores. They have been designed with the same assembly specifications and managed to have similar end-of-cycle sodium void reactivities and beginning-of-cycle peak power densities through the changes in the core size and configuration. The requirement of low sodium void reactivity is shown to lead each design concept to characteristic neutronics performance and fuel cycle behavior. The H-/pan-shaped cores allow the core compaction as well as higher rate of TRU burning. (author)

  12. Nocturia: The circadian voiding disorder

    Directory of Open Access Journals (Sweden)

    Jin Wook Kim

    2016-05-01

    Full Text Available Nocturia is a prevalent condition of waking to void during the night. The concept of nocturia has evolved from being a symptomatic aspect of disease associated with the prostate or bladder to a form of lower urinary tract disorder. However, recent advances in circadian biology and sleep science suggest that it might be important to consider nocturia as a form of circadian dysfunction. In the current review, nocturia is reexamined with an introduction to sleep disorders and recent findings in circadian biology in an attempt to highlight the importance of rediscovering nocturia as a problem of chronobiology.

  13. The dark matter of galaxy voids

    Science.gov (United States)

    Sutter, P. M.; Lavaux, Guilhem; Wandelt, Benjamin D.; Weinberg, David H.; Warren, Michael S.

    2014-03-01

    How do observed voids relate to the underlying dark matter distribution? To examine the spatial distribution of dark matter contained within voids identified in galaxy surveys, we apply Halo Occupation Distribution models representing sparsely and densely sampled galaxy surveys to a high-resolution N-body simulation. We compare these galaxy voids to voids found in the halo distribution, low-resolution dark matter and high-resolution dark matter. We find that voids at all scales in densely sampled surveys - and medium- to large-scale voids in sparse surveys - trace the same underdensities as dark matter, but they are larger in radius by ˜20 per cent, they have somewhat shallower density profiles and they have centres offset by ˜ 0.4Rv rms. However, in void-to-void comparison we find that shape estimators are less robust to sampling, and the largest voids in sparsely sampled surveys suffer fragmentation at their edges. We find that voids in galaxy surveys always correspond to underdensities in the dark matter, though the centres may be offset. When this offset is taken into account, we recover almost identical radial density profiles between galaxies and dark matter. All mock catalogues used in this work are available at http://www.cosmicvoids.net.

  14. The Beckoning Void in Moravagine

    Directory of Open Access Journals (Sweden)

    Stephen K. Bellstrom

    1979-01-01

    Full Text Available The Chapter «Mascha,» lying at the heart of Cendrars's Moravagine , contains within it a variety of images and themes suggestive of emptiness. The philosophy of nihilism is exemplified in the motivations and actions of the group of terrorists seeking to plunge Russia into revolutionary chaos. Mascha's anatomical orifice, symbolizing both a biological and a psychological fault, and the abortion of her child, paralleled by the abortion of the revolutionary ideal among her comrades, are also emblematic of the chapter's central void. Moreover, Cendrars builds the theme of hollowness by describing Moravagine with images of omission, such as «empan» (space or span, «absent,» and «étranger.» Moravagine's presence, in fact, characteristically causes an undercurrent of doubt and uncertainty about the nature of reality to become overt. It is this parodoxical presence which seems to cause the narrator (and consequently the narrative to «lose» a day at the most critical moment of the story. By plunging the reader into the narrator's lapsus memoriae , Cendrars aims at creating a feeling of the kind of mental and cosmic disorder for which Moravagine is the strategist and apologist. This technique of insufficiency is an active technique, even though it relies on the passive idea of removing explanation and connecting details. The reader is invited, or lured, into the central void of the novel and, faced with unresolvable dilemmas, becomes involved in the same disorder that was initially produced.

  15. In-core failure of the instrumented BWR rod by locally induced high coolant temperature

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki

    1985-12-01

    In the BWR type light water loop instrumented in HBWR, a current BWR type fuel rod pre-irradiated up to 5.6 MWd/kgU was power ramped to 50 kW/m. During the ramp, the diameter of the rod was expanded significantly at the bottom end. The behaviour was different from which caused by pellet-cladding interaction (PCI). In the post-irradiation examination, the rod was found to be failed. In this paper, the cause of the failure was studied and obtained the followings. (1) The significant expansion of the rod diameter was attributed to marked oxidation of cladding outer diameter, appeared in the direction of 0 0 -180 0 degree with a shape of nodular. (2) The cladding in the place was softened by high coolant temperature. Coolant pressure, 7MPa intruded the cladding into inside chamfer void at pellet interface. (3) At the place of the significant oxidation, an instrumented transformer was existed and the coolant flow area was very little. The reduction of the coolant flow was enhanced by the bending of the cladding which was caused in pre-irradiation stage. They are considered to be a principal cause of local closure of coolant flow and resultant high temperature in the place. (author)

  16. Elastic wave scattering from multiple voids (porosity)

    International Nuclear Information System (INIS)

    Thompson, D.O.; Rose, J.H.; Thompson, R.B.; Wormley, S.J.

    1983-01-01

    This paper describes the development of an ultrasonic backscatter measurement technique which provides a convenient way to determine certain characteristics of a distribution of voids (porosity) in materials. A typical ultrasonic sample prepared by placing the ''frit'' in a crucible in an RF induction heater is shown. The results of the measurements were Fourier transformed into an amplitude-frequency description, and were then deconvolved with the transducer response function. Several properties needed to characterize a void distribution are obtained from the experimental results, including average void size, the spatial extent of the voids region, the average void separation, and the volume fraction of material contained in the void distribution. A detailed comparison of values obtained from the ultrasonic measurements with visually determined results is also given

  17. Cosmology with void-galaxy correlations.

    Science.gov (United States)

    Hamaus, Nico; Wandelt, Benjamin D; Sutter, P M; Lavaux, Guilhem; Warren, Michael S

    2014-01-31

    Galaxy bias, the unknown relationship between the clustering of galaxies and the underlying dark matter density field is a major hurdle for cosmological inference from large-scale structure. While traditional analyses focus on the absolute clustering amplitude of high-density regions mapped out by galaxy surveys, we propose a relative measurement that compares those to the underdense regions, cosmic voids. On the basis of realistic mock catalogs we demonstrate that cross correlating galaxies and voids opens up the possibility to calibrate galaxy bias and to define a static ruler thanks to the observable geometric nature of voids. We illustrate how the clustering of voids is related to mass compensation and show that volume-exclusion significantly reduces the degree of stochasticity in their spatial distribution. Extracting the spherically averaged distribution of galaxies inside voids from their cross correlations reveals a remarkable concordance with the mass-density profile of voids.

  18. Development of the impedance void meter

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Moon Ki; Song, Chul Hwa; Won, Soon Yeon; Kim, Bok Deuk [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-06-01

    An impedance void meter is developed to measure the area-averaged void fraction. Its basic principle is based on the difference in the electrical conductivity between phases. Several methods of measuring void fraction are briefly reviewed and the reason why this type of void meter is chosen to develop is discussed. Basic principle of the measurement is thoroughly described and several design parameters to affect the overall function are discussed in detail. As example of applications is given for vertical air-water flow. It is shown that the current design has good dynamic response as well as very fine spatial resolution. (Author) 47 refs., 37 figs.

  19. Coolant inlet device for nuclear reactors

    International Nuclear Information System (INIS)

    Ando, Hiroshi; Abe, Yasuhiro; Iwabuchi, Toshihiko; Yamamoto, Kenji.

    1969-01-01

    Herein disclosed is a coolant inlet device for liquid-metal cooled reactors which employs a coolant distributor serving also as a supporting means for the reactor core. The distributor is mounted within the reactor vessel so as to slide horizontally on supporting lugs, and is further slidably connected via a junction pipe to a coolant inlet conduit protruding through the floor of the vessel. The distributor is adapted to uniformly disperse the highly pressured coolant over the reactor core so as to reduce the stresses sustained by the reactor vessel as well as the supporting lugs. Moreover, the slidable nature of the distributor allows thermal shock and excessive coolant pressures to be prevented or alleviated, factors which posed major difficulties in conventional coolant inlet devices. (Owens, K. J.)

  20. Speed control device for coolant recycling pump

    International Nuclear Information System (INIS)

    Kageyama, Takao.

    1992-01-01

    The present invention intends to increase a margin relative of the oscillations of neutron fluxes when the temperature of feedwater is lowered in a compulsory recycling type BWR reactor. That is, when the operation point represented by a reactor thermal power and a reactor core inlet flow rate is in a state approximate to an oscillation limit of the reactor power, the device of the present invention controls the recycling pump speed in the increasing direction depending on the lowering range of the feedwater temperature from a stationary state. With such a constitution, even if the reactor power is in the operation region near the oscillation limit in the BWR type reactor and a feedwater heating loss is caused, the speed of the coolant recycling pump is increased by 10% at the maximum depending on the extent of the reduction of the feedwater temperature, so that the oscillation of the reactor power can be prevented from lasting for a long period of time even if a reactivity external disturbance should occur in the reactor. (I.S.)

  1. Organic coolant for ARIES-III

    International Nuclear Information System (INIS)

    Sze, D.K.; Sviatoslavsky, I.; Sawan, M.; Gierszewski, P.; Hollies, R.; Sharafat, S.; Herring, S.

    1991-04-01

    ARIES-III is a D-He 3 reactor design study. It is found that the organic coolant is well suited for the D-He 3 reactor. This paper discusses the unique features of the D-He 3 reactor, and the reason that the organic coolant is compatible with those features. The problems associated with the organic coolant are also discussed. 8 refs., 2 figs., 6 tabs

  2. Physical properties of organic coolants

    International Nuclear Information System (INIS)

    Debbage, A.G.; Garton, D.A.; Kinneir, J.H.

    1963-03-01

    Density, viscosity, specific heat, vapour pressure and calorific value were measured within the temperature range 100 - 400 deg C for mixtures of Santowax R with pyrolytic high boiler and Santowax R with O.M.R.E. radiolytic high boiler; in addition measurements were made on Santowax OM, X-7 standard, X-7 loop coolant and O.M.R.E. coolant supplied by Atomic Energy of Canada Ltd. The accuracy of the measurements made were density (± 1/4%), viscosity (± 2%), specific heat (± 2%), vapour pressure (± 2%) and calorific value (± 1/2%). Thermal conductivity was calculated from an improved form of the Smiths equation with an accuracy within ± 6%. Equations fitted to the vapour pressure results were used to provide data outside the experimental range for burnout correlation purposes. The general effect of high boiler content on the specific heat and calorific values was small. The differences in physical property values for corresponding values of either pyrolytic or radiolytic high boiler were small for density (0.3%) and specific heat (2%), but quite large for viscosity (70%) with the pyrolytic high boiler mixture giving the higher value. The chemical analysis of all materials was based on gas chromatography and the relationship between this and an earlier distillation method established. (author)

  3. Effects of void anisotropy on the ignition and growth rates of energetic materials

    Science.gov (United States)

    Rai, Nirmal Kumar; Sen, Oishik; Udaykumar, H. S.

    2017-06-01

    Initiation of heterogeneous energetic materials is thought to occur at hot spots; reaction fronts propagate from sites of such hot spots into the surrounding material resulting in complete consumption of the material. Heterogeneous materials, such as plastic bonded explosives (PBXs) and pressed materials contain numerous voids, defects and interfaces at which hot spots can occur. Amongst the various mechanisms of hot spot formation, void collapse is considered to be the predominant one in the high strain rate loading conditions. It is established in the past the shape of the voids has a significant effect on the initiation behavior of energetic materials. In particular, void aspect ratio and orientations play an important role in this regard. This work aims to quantify the effects of void aspect ratio and orientation on the ignition and growth rates of chemical reaction from the hot spot. A wide range of aspect ratio and orientations is considered to establish a correlation between the ignition and growth rates and the void morphology. The ignition and growth rates are obtained from high fidelity reactive meso-scale simulations. The energetic material considered in this work is HMX and Tarver McGuire HMX decomposition model is considered to capture the reaction mechanism of HMX. The meso-scale simulations are performed using a Cartesian grid based Eulerian solver SCIMITAR3D. The void morphology is shown to have a significant effect on the ignition and growth rates of HMX.

  4. Analysis of sodium-void experiments in ZPPR-3 modified phase 3 core

    International Nuclear Information System (INIS)

    Yoshida, T.

    1978-08-01

    In this work, large-zone sodium-void effects are studied in detail in the presence of many singularities, namely, control rods (CRs) and control rod positions (CRPs). The results of measurements and calculations are compared by CIE (calculation/experiment) values, which are 1.07 when the voided core region is free of singularities. When the void region includes CPRs, which are concurrently voided, the CIE value deteriorates and varies from 0.35 to 1.58. The agreement can be improved considerably by correcting the reactivity worth of the sodium contained in the CRPs with the aid of experimental data (CIE = 1.00 +- 0.15). The heterogeneity correction for the fuel elements was performed by the plate-cell vollision probability code KAPPER. (GL) [de

  5. Cleaning of aluminum after machining with coolants

    International Nuclear Information System (INIS)

    Roop, B.

    1992-01-01

    An x-ray photoemission spectroscopic study was undertaken to compare the cleaning of the Advanced Photon Source (APS) aluminum extrusion storage ring vacuum chambers after machining with and without water soluble coolants. While there was significant contamination left by the coolants, the cleaning process was capable of removing the residue. The variation of the surface and near surface composition of samples machined either dry or with coolants was negligible after cleaning. The use of such coolants in the machining process is therefore recommended

  6. Coolant clean-up system in the primary coolant circuit for nuclear reactor

    International Nuclear Information System (INIS)

    Saito, Michio.

    1981-01-01

    Purpose: To maintain the quality of coolants at a prescribed level by distillating coolants in the primary coolant circuit for a BWR type reactor to remove impurities therefrom, taking out the condensates from the top of the distillation column and extracting impurities in a concentrated state from the bottom. Constitution: Coolant water for cooling the core is recycled by a recycling pump by way of a recycling pipeway in a reactor. The coolants extracted from an extraction pipeway connected to the recycling pipeway are fed into a distillation column, where distillation is taken place. Impurities in the coolants, that is, in-core corrosion products, fission products generated in the reactor core, etc. are separated by the distillation, concentrated and solidified in the bottom of the distillation column. While on the other hand, condensates removed with the impurities, that is, coolants cleaned-up are recycled to the coolant water for cooling the reactor core. (Moriyama, K.)

  7. Prediction of void fraction in subcooled flow boiling

    International Nuclear Information System (INIS)

    Petelin, S.; Koncar, B.

    1998-01-01

    The information on heat transfer and especially on the void fraction in the reactor core under subcooled conditions is very important for the water-cooled nuclear reactors, because of its influence upon the reactivity of the systems. This paper gives a short overview of subcooled boiling phenomenon and indicates the simplifications made by the RELAP5 model of subcooled boiling. RELAP5/MOD3.2 calculations were compared with simple one-dimensional models and with high-pressure Bartolomey experiments.(author)

  8. Selection of an Alternate Biocide for the ISS Internal Thermal Control System Coolant, Phase 2

    Science.gov (United States)

    Wilson, Mark E.; Cole, Harold; Weir, Natalee; Oehler, Bill; Steele, John; Varsik, Jerry; Lukens, Clark

    2004-01-01

    The ISS (International Space Station) ITCS (Internal Thermal Control System) includes two internal coolant loops that utilize an aqueous based coolant for heat transfer. A silver salt biocide had previously been utilized as an additive in the coolant formulation to control the growth and proliferation of microorganisms within the coolant loops. Ground-based and in-flight testing demonstrated that the silver salt was rapidly depleted, and did not act as an effective long-term biocide. Efforts to select an optimal alternate biocide for the ITCS coolant application have been underway and are now in the final stages. An extensive evaluation of biocides was conducted to down-select to several candidates for test trials and was reported on previously. Criteria for that down-select included: the need for safe, non-intrusive implementation and operation in a functioning system; the ability to control existing planktonic and biofilm residing microorganisms; a negligible impact on system-wetted materials of construction; and a negligible reactivity with existing coolant additives. Candidate testing to provide data for the selection of an optimal alternate biocide is now in the final stages. That testing has included rapid biocide effectiveness screening using Biolog MT2 plates to determine minimum inhibitory concentration (amount that will inhibit visible growth of microorganisms), time kill studies to determine the exposure time required to completely eliminate organism growth, materials compatibility exposure evaluations, coolant compatibility studies, and bench-top simulated coolant testing. This paper reports the current status of the effort to select an alternate biocide for the ISS ITCS coolant. The results of various test results to select the optimal candidate are presented.

  9. Design and fabrication of magnetic coolant filter

    Science.gov (United States)

    Prashanth, B. N.

    2017-07-01

    Now a day's use of coolants in industry has become dominant because of high production demands. Coolants not only help in speeding up the production but also provide many advantages in the metal working operation. As the consumption of coolants is very high a system is badly in need, so as to recirculate the used coolant. Also the amount of hazardous waste generated by industrial plants has become an increasingly costly problem for the manufactures and an additional stress on the environment. Since the purchase and disposal of the spent cutting fluids is becoming increasingly expensive, fluid recycling is a viable option for minimizing the cost. Separation of metallic chips from the coolants by using magnetic coolant separation has proven a good management and maintenance of the cutting fluid. By removing the metallic chips, the coolant life is greatly extended, increases the machining quality and reduces downtime. Above being the case, a magnetic coolant filter is developed which utilizes high energy permanent magnets to develop a dense magnetic field along a narrow flow path into which the contaminated coolant is directed. The ferromagnetic particles captured and aligned by the dense magnetic field, from the efficient filter medium. This enables the unit to remove ferromagnetic particles from the coolant. Magnetic coolant filters use the principle of magnetic separation to purify the used coolant. The developed magnetic coolant separation has the capability of purifying 40 litres per minute of coolant with the size of the contaminants ranging from 1 µm to 30 µm. The filter will be helpful in saving the production cost as the cost associated with the proposed design is well justified by the cost savings in production. The magnetic field produced by permanent magnets will be throughout the area underneath the reservoir. This produces magnetic field 30mm above the coolant reservoir. Very fine particles are arrested without slip. The magnetic material used will not

  10. Coolant mixing in pressurized water reactors. Proceedings

    International Nuclear Information System (INIS)

    Hoehne, T.; Grunwald, G.; Rohde, U.

    1998-10-01

    For the analysis of boron dilution transients and main steam like break scenarios the modelling of the coolant mixing inside the reactor vessel is important. The reactivity insertion due to overcooling or deboration depends strongly on the coolant temperature and boron concentration. The three-dimensional flow distribution in the downcomer and the lower plenum of PWR's was calculated with a computational fluid dynamics (CFD) code (CFX-4). Calculations were performed for the PWR's of SIEMENS KWU, Westinghouse and VVER-440 / V-230 type. The following important factors were identified: exact representation of the cold leg inlet region (bend radii etc.), extension of the downcomer below the inlet region at the PWR Konvoi, obstruction of the flow by the outlet nozzles penetrating the downcomer, etc. The k-ε turbulence model was used. Construction elements like perforated plates in the lower plenum have large influence on the velocity field. It is impossible to model all the orifices in the perforated plates. A porous region model was used to simulate perforated plates and the core. The porous medium is added with additional body forces to simulate the pressure drop through perforated plates in the VVER-440. For the PWR Konvoi the whole core was modelled with porous media parameters. The velocity fields of the PWR Konvoi calculated for the case of operation of all four main circulation pumps show a good agreement with experimental results. The CFD-calculation especially confirms the back flow areas below the inlet nozzles. The downcomer flow of the Russian VVER-440 has no recirculation areas under normal operation conditions. By CFD calculations for the downcomer and the lower plenum an analytical mixing model used in the reactor dynamic code DYN3D was verified. The measurements, the analytical model and the CFD-calculations provided very well agreeing results particularly for the inlet region. The difficulties of analytical solutions and the uncertainties of turbulence

  11. Studies of void formation in pure metals

    International Nuclear Information System (INIS)

    Lanore, J.M.; Glowinski, L.; Risbet, A.; Regnier, P.; Flament, J.L.; Levy, V.; Adda, Y.

    1975-01-01

    Recent experiments on the effect of gases on the final configuration of vacancy clustering (void or loop), and on the local effects at dislocations are described. The contribution of this data to a general knowledge of void formation will be discussed, and Monte Carlo calculations of swelling induced by irradiation with different particles presented [fr

  12. Void Fraction Instrument operation and maintenance manual

    International Nuclear Information System (INIS)

    Borgonovi, G.; Stokes, T.I.; Pearce, K.L.; Martin, J.D.; Gimera, M.; Graves, D.B.

    1994-09-01

    This Operations and Maintenance Manual (O ampersand MM) addresses riser installation, equipment and personnel hazards, operating instructions, calibration, maintenance, removal, and other pertinent information necessary to safely operate and store the Void Fraction Instrument. Final decontamination and decommissioning of the Void Fraction Instrument are not covered in this document

  13. Studies of void formation in pure metals

    International Nuclear Information System (INIS)

    Lanore, J.M.; Glowinski, L.; Risbet, A.; Regnier, P.; Flament, J.L.

    1975-01-01

    Recent experiments on the effect of gases on the final configuration of vacancy clustering (void or loop), and on the local effects at dislocations are described. The contribution of this data to our general knowledge of void formation will be discussed, and Monte Carlo calculations of swelling induced by irradiation with different particles presented

  14. Statistics and geometry of cosmic voids

    International Nuclear Information System (INIS)

    Gaite, José

    2009-01-01

    We introduce new statistical methods for the study of cosmic voids, focusing on the statistics of largest size voids. We distinguish three different types of distributions of voids, namely, Poisson-like, lognormal-like and Pareto-like distributions. The last two distributions are connected with two types of fractal geometry of the matter distribution. Scaling voids with Pareto distribution appear in fractal distributions with box-counting dimension smaller than three (its maximum value), whereas the lognormal void distribution corresponds to multifractals with box-counting dimension equal to three. Moreover, voids of the former type persist in the continuum limit, namely, as the number density of observable objects grows, giving rise to lacunar fractals, whereas voids of the latter type disappear in the continuum limit, giving rise to non-lacunar (multi)fractals. We propose both lacunar and non-lacunar multifractal models of the cosmic web structure of the Universe. A non-lacunar multifractal model is supported by current galaxy surveys as well as cosmological N-body simulations. This model suggests, in particular, that small dark matter halos and, arguably, faint galaxies are present in cosmic voids

  15. Void formation in irradiated binary nickel alloys

    International Nuclear Information System (INIS)

    Shaikh, M.A.; Ahmed, M.; Akhter, J.I.

    1994-01-01

    In this work a computer program has been used to compute void radius, void density and swelling parameter for nickel and binary nickel-carbon alloys irradiated with nickel ions of 100 keV. The aim is to compare the computed results with experimental results already reported

  16. Void Measurement by the ({gamma}, n) Reaction

    Energy Technology Data Exchange (ETDEWEB)

    Rouhani, S Zia

    1962-09-15

    It is proposed to use the ({gamma}, n) reaction for the measurement of the integrated void volume fraction in two phase flow of D{sub 2}O inside a duct. This method is applicable to different channel geometries, and it is shown to be insensitive to the pattern of void distribution over the cross-sectional area of the channels The method has been tested on mock-ups of voids in a round duct of 6 mm inside diameter. About 40 m.c. {sup 24}Na was used as a source of gamma-rays. The test results show that the maximum measured error in this arrangement is less than 2.5 % (net void) for a range of 2.7 % to 44.44 % actual void volume fractions.

  17. Void Measurement by the (γ, n) Reaction

    International Nuclear Information System (INIS)

    Rouhani, S. Zia

    1962-09-01

    It is proposed to use the (γ, n) reaction for the measurement of the integrated void volume fraction in two phase flow of D 2 O inside a duct. This method is applicable to different channel geometries, and it is shown to be insensitive to the pattern of void distribution over the cross-sectional area of the channels The method has been tested on mock-ups of voids in a round duct of 6 mm inside diameter. About 40 m.c. 24 Na was used as a source of gamma-rays. The test results show that the maximum measured error in this arrangement is less than 2.5 % (net void) for a range of 2.7 % to 44.44 % actual void volume fractions

  18. Pores and Void in Asclepiades’ Physical Theory

    Science.gov (United States)

    Leith, David

    2012-01-01

    This paper examines a fundamental, though relatively understudied, aspect of the physical theory of the physician Asclepiades of Bithynia, namely his doctrine of pores. My principal thesis is that this doctrine is dependent on a conception of void taken directly from Epicurean physics. The paper falls into two parts: the first half addresses the evidence for the presence of void in Asclepiades’ theory, and concludes that his conception of void was basically that of Epicurus; the second half focuses on the precise nature of Asclepiadean pores, and seeks to show that they represent void interstices between the primary particles of matter which are the constituents of the human body, and are thus exactly analogous to the void interstices between atoms within solid objects in Epicurus’ theory. PMID:22984299

  19. Reactor having coolant recycling pump

    International Nuclear Information System (INIS)

    Goto, Tadashi; Karatsuka, Shigeki; Yamamoto, Hajime.

    1991-01-01

    In a coolant recycling pump for an LMFBR type reactor, vertical grooves are formed to a static portion which surrounds a pump shaft as far as the lower end thereof. Sodium mists present in an annular gap of the pump shaft form a rotational flow, lose its centrifugal force at the grooved portion and are collected positively to the grooved portion. Further, since the rotational flow in the grooved channel is in a state of a cavity flow, the pressure is released in the grooved portion and a secondary eddy current is formed thereby providing a depressurized state. Accordingly, by a synergestic effect of the centrifugal force and the cavity flow, sodium mists can be recovered completely. (T.M.)

  20. On the abundance of extreme voids II: a survey of void mass functions

    International Nuclear Information System (INIS)

    Chongchitnan, Siri; Hunt, Matthew

    2017-01-01

    The abundance of cosmic voids can be described by an analogue of halo mass functions for galaxy clusters. In this work, we explore a number of void mass functions: from those based on excursion-set theory to new mass functions obtained by modifying halo mass functions. We show how different void mass functions vary in their predictions for the largest void expected in an observational volume, and compare those predictions to observational data. Our extreme-value formalism is shown to be a new practical tool for testing void theories against simulation and observation.

  1. Towards the reanalysis of void coefficients measurements at proteus for high conversion light water reactor lattices

    Energy Technology Data Exchange (ETDEWEB)

    Hursin, M.; Koeberl, O.; Perret, G. [Paul Scherrer Institut PSI, 5232 Villigen (Switzerland)

    2012-07-01

    High Conversion Light Water Reactors (HCLWR) allows a better usage of fuel resources thanks to a higher breeding ratio than standard LWR. Their uses together with the current fleet of LWR constitute a fuel cycle thoroughly studied in Japan and the US today. However, one of the issues related to HCLWR is their void reactivity coefficient (VRC), which can be positive. Accurate predictions of void reactivity coefficient in HCLWR conditions and their comparisons with representative experiments are therefore required. In this paper an inter comparison of modern codes and cross-section libraries is performed for a former Benchmark on Void Reactivity Effect in PWRs conducted by the OECD/NEA. It shows an overview of the k-inf values and their associated VRC obtained for infinite lattice calculations with UO{sub 2} and highly enriched MOX fuel cells. The codes MCNPX2.5, TRIPOLI4.4 and CASMO-5 in conjunction with the libraries ENDF/B-VI.8, -VII.0, JEF-2.2 and JEFF-3.1 are used. A non-negligible spread of results for voided conditions is found for the high content MOX fuel. The spread of eigenvalues for the moderated and voided UO{sub 2} fuel are about 200 pcm and 700 pcm, respectively. The standard deviation for the VRCs for the UO{sub 2} fuel is about 0.7% while the one for the MOX fuel is about 13%. This work shows that an appropriate treatment of the unresolved resonance energy range is an important issue for the accurate determination of the void reactivity effect for HCLWR. A comparison to experimental results is needed to resolve the presented discrepancies. (authors)

  2. Mathematical model of the reactor coolant pump

    International Nuclear Information System (INIS)

    Kozuh, M.

    1989-01-01

    The mathematical model of reactor coolant pump is described in this paper. It is based on correlations for centrifugal reactor coolant pumps. This code is one of the elements needed for the simulation of the whole NPP primary system. In subroutine developed according to this model we tried in every possible detail to incorporate plant specific data for Krsko NPP. (author)

  3. Lecture background notes on transient sodium boiling and voiding in fast reactors

    International Nuclear Information System (INIS)

    Okrent, D.; Fauske, H.K.

    1972-01-01

    This set of lecture background notes includes the following: (1) Introductory remarks on fast reactor safety, which are intended to provide some perspective on the role played by sodium boiling. (2) A discussion of superheat which reviews the experimental data and nucleation models with emphasis on the pressure-temperature history effect on radius of active cavity sites, including the role played by inert gas. (3) A discussion of the growth and collapse of spherical bubbles. (4) A historical description of the development of computer codes to describe voiding and a detailed description of the analytical formulation of typical models for calculating voiding due to boiling, fission gas release, and molten fuel-coolant interaction. (U.S.)

  4. Organic coolant in Winnipeg riverbed sediments

    International Nuclear Information System (INIS)

    Guthrie, J.E.; Acres, O.E.

    1979-03-01

    Between January and May 1977 a prolonged leak of organic coolant occurred from the Whiteshell Nuclear Research Establishment's nuclear reactor, and a minimum of 1450 kg of coolant entered the Winnipeg River and was deposited on the riverbed. The level of radioactivity associated with this coolant was low, contributing less than 0.2 μGy (0.02 mrad) a year to the natural background gamma radiation field from the riverbed. The concentration of coolant in the water samples never exceeded 0.02 mg/L, the lower limit of detection. The mortality of crayfish, held in cages where the riverbed was covered with the largest deposits of coolant, was not significantly different from that in the control cages upstream of the outfall. No evidence of fish kill was found. (author)

  5. Primary coolant circuits in FBR type reactors

    International Nuclear Information System (INIS)

    Kutani, Masushiro.

    1985-01-01

    Purpose: To eliminate the requirement of a pump for the forcive circulation of primary coolants and avoid the manufacturing difficulty of equipments. Constitution: In primary coolant circuits of an LMFBR type reactor having a recycling path forming a closed loop between a reactor core and a heat exchanger, coolants recycled through the recycling path are made of a magnetic fluid comprising liquid sodium incorporated with fine magnetic powder, and an electromagnet is disposed to the downstream of the heat exchanger. In the above-mentioned structure, since the magnetic fluid as the primary coolants losses its magnetic property when heated in the reactor core but recovers the property at a lower temperature after the completion of the heat exchange, the magnetic fluid can forcively be flown through the recycling path under the effect of the electromagnet disposed to the down stream of the heat exchanger to thereby forcively recycle the primary coolants. (Kawakami, Y.)

  6. BN600 reactivity definition

    International Nuclear Information System (INIS)

    Zheltyshev, V.; Ivanov, A.

    2000-01-01

    Since 1980, the fast BN600 reactor with sodium coolant has been operated at Beloyarsk Nuclear Power Plant. The periodic monitoring of the reactivity modifications should be implemented in compliance with the standards and regulations applied in nuclear power engineering. The reactivity measurements are carried out in order to confirm the basic neutronic features of a BN600 reactor. The reactivity measurements are aimed to justify that nuclear safety is provided in course of the in-reactor installation of the experimental core components. Two reactivity meters are to be used on BN600 operation: 1. Digital on-line reactivity calculated under stationary reactor operation on power (approximation of the point-wise kinetics is applied). 2. Second reactivity meter used to define the reactor control rod operating components efficiency under reactor startup and take account of the changing efficiency of the sensor, however, this is more time-consumptive than the on-line reactivity meter. The application of two reactivity meters allows for the monitoring of the reactor reactivity under every operating mode. (authors)

  7. VIDE: The Void IDentification and Examination toolkit

    Science.gov (United States)

    Sutter, P. M.; Lavaux, G.; Hamaus, N.; Pisani, A.; Wandelt, B. D.; Warren, M.; Villaescusa-Navarro, F.; Zivick, P.; Mao, Q.; Thompson, B. B.

    2015-03-01

    We present VIDE, the Void IDentification and Examination toolkit, an open-source Python/C++ code for finding cosmic voids in galaxy redshift surveys and N-body simulations, characterizing their properties, and providing a platform for more detailed analysis. At its core, VIDE uses a substantially enhanced version of ZOBOV (Neyinck 2008) to calculate a Voronoi tessellation for estimating the density field and performing a watershed transform to construct voids. Additionally, VIDE provides significant functionality for both pre- and post-processing: for example, VIDE can work with volume- or magnitude-limited galaxy samples with arbitrary survey geometries, or dark matter particles or halo catalogs in a variety of common formats. It can also randomly subsample inputs and includes a Halo Occupation Distribution model for constructing mock galaxy populations. VIDE uses the watershed levels to place voids in a hierarchical tree, outputs a summary of void properties in plain ASCII, and provides a Python API to perform many analysis tasks, such as loading and manipulating void catalogs and particle members, filtering, plotting, computing clustering statistics, stacking, comparing catalogs, and fitting density profiles. While centered around ZOBOV, the toolkit is designed to be as modular as possible and accommodate other void finders. VIDE has been in development for several years and has already been used to produce a wealth of results, which we summarize in this work to highlight the capabilities of the toolkit. VIDE is publicly available at http://bitbucket.org/cosmicvoids/vide_public and http://www.cosmicvoids.net.

  8. Neutron Physics aspects of using lead as a coolant in Fast Reactors

    International Nuclear Information System (INIS)

    Kiefhaber, E.

    1991-02-01

    The use of lead as a coolant for fast reactors is being considered as an attractive alternative in the USSR, especially with respect to its inherent safety features. In order to come to an own assessment at KfK, some investigations have been performed concerning a comparison of the nuclear characteristics of fast reactors with lead and sodium cooling. The studies have shown, that the nuclear and thermal hydraulic design calculations do not face special problems and that the nuclear characteristics of both types of cores do not differ essentially, except for the coolant density or void effect, which is more favourable for smaller sized lead cooled cores. A proper safety assessment of lead cooled cores will however require more detailed safety studies. Crucial points of lead cooling are the strong corrosion of austenitic steels in lead and the unknown behavior of ferritic steels in lead and under irradiation

  9. Void fraction prediction of NUPEC PSBT tests by CATHARE code

    International Nuclear Information System (INIS)

    Del Nevo, A.; Michelotti, L.; Moretti, F.; Rozzia, D.; D'Auria, F.

    2011-01-01

    The current generation of thermal-hydraulic system codes benefits of about sixty years of experiments and forty years of development and are considered mature tools to provide best estimate description of phenomena and detailed reactor system representations. However, there are continuous needs for checking the code capabilities in representing nuclear system, for drawing attention to their weak points, for identifying models which need to be refined for best-estimate calculations. Prediction of void fraction and Departure from Nucleate Boiling (DNB) in system thermal-hydraulics is currently based on empirical approaches. The database carried out by Nuclear Power Engineering Corporation (NUPEC), Japan addresses these issues. It is suitable for supporting the development of new computational tools based on more mechanistic approaches (i.e. three-field codes, two-phase CFD, etc.) as well as for validating current generation of thermal-hydraulic system codes. Selected experiments belonging to this database are used for the OECD/NRC PSBT benchmark. The paper reviews the activity carried out by CATHARE2 code on the basis of the subchannel (four test sections) and presents rod bundle (different axial power profile and test sections) experiments available in the database in steady state and transient conditions. The results demonstrate the accuracy of the code in predicting the void fraction in different thermal-hydraulic conditions. The tests are performed varying the pressure, coolant temperature, mass flow and power. Sensitivity analyses are carried out addressing nodalization effect and the influence of the initial and boundary conditions of the tests. (author)

  10. The sink strengths of voids and the expected swelling for both random and ordered void distributions

    International Nuclear Information System (INIS)

    Quigley, T.M.; Murphy, S.M.; Bullough, R.; Wood, M.H.

    1981-10-01

    The sink strength of a void has been obtained when the void is a member of a random or ordered distribution of voids. The former sink strength derivation has employed the embedding model and the latter the cellular model. In each case the spatially varying size-effect interaction between the intrinsic point defects and the voids has been included together with the presence of other sink types in addition to the voids. The results are compared with previously published sink strengths that have made use of an approximate representation for the size-effect interactions, and indicate the importance of using the exact form of the interaction. In particular the bias for interstitials compared with vacancies of small voids is now much reduced and contamination of the surfaces of such voids no longer appears essential to facilitate the nucleation and growth of the voids. These new sink strengths have been used, in conjunction with recently published dislocation sink strengths, to calculate the expected swelling of materials containing network dislocations and voids. Results are presented for both the random and the void lattice situations. (author)

  11. Flow boiling test of GDP replacement coolants

    International Nuclear Information System (INIS)

    Park, S.H.

    1995-01-01

    The tests were part of the CFC replacement program to identify and test alternate coolants to replace CFC-114 being used in the uranium enrichment plants at Paducah and Portsmouth. The coolants tested, C 4 F 10 and C 4 F 8 , were selected based on their compatibility with the uranium hexafluoride process gas and how well the boiling temperature and vapor pressure matched that of CFC-114. However, the heat of vaporization of both coolants is lower than that of CFC-114 requiring larger coolant mass flow than CFC-114 to remove the same amount of heat. The vapor pressure of these coolants is higher than CFC-114 within the cascade operational range, and each coolant can be used as a replacement coolant with some limitation at 3,300 hp operation. The results of the CFC-114/C 4 F 10 mixture tests show boiling heat transfer coefficient degraded to a minimum value with about 25% C 4 F 10 weight mixture in CFC-114 and the degree of degradation is about 20% from that of CFC-114 boiling heat transfer coefficient. This report consists of the final reports from Cudo Technologies, Ltd

  12. Coolant cleanup method in a nuclear reactor

    International Nuclear Information System (INIS)

    Kubota, Masayoshi; Nishimura, Shigeoki; Takahashi, Sankichi; Izumi, Kenkichi; Motojima, Kenji.

    1983-01-01

    Purpose : To effectively adsorb to remove low molecular weight organic substances from iron exchange resins for use in the removal of various radioactive nucleides contained in reactor coolants. Method : Reactor coolants are recycled by a main recyling pump in a nuclear reactor and a portion of the coolants is cooled and, thereafter, purified in a coolant desalter. While on the other hand, high pressure steams generated from the reactor are passed through a turbine, cooled in a condensator, eliminated with claddings or the likes by the passage through a filtration desalter using powderous ion exchange resins and then further passed through a desalter (filled with granular ion exchange resins). For instance, an adsorption and removing device for organic substances (resulted through the decomposition of ion exchange resins) precoated with activated carbon powder or filled with granular activated carbon is disposed at the downstream for each of the desalters. In this way, the organic substances in the coolants are eliminated to prevent the reduction in the desalting performance of the ion exchange resins caused by the formation of complexes between organic substances and cobalt in the coolants, etc. In this way, the coolant cleanup performance is increased and the amount of wasted ion exchange resins can be decreased. (Horiuchi, T.)

  13. Continuous surveillance of reactor coolant circuit integrity

    International Nuclear Information System (INIS)

    1986-01-01

    Continuous surveillance is important to assuring the integrity of a reactor coolant circuit. It can give pre-warning of structural degradation and indicate where off-line inspection should be focussed. These proceedings describe the state of development of several techniques which may be used. These involve measuring structural vibration, core neutron noise, acoustic emission from cracks, coolant leakage, or operating parameters such as coolant temperature and pressure. Twenty three papers have been abstracted and indexed separately for inclusion in the data base

  14. Evaluation of the Air Void Analyzer

    Science.gov (United States)

    2013-07-01

    concrete using image analysis: Petrography of cementitious materials. ASTM STP 1215. S.M. DeHayes and D. Stark, eds. Philadelphia, PA: American...Administration (FHWA). 2006. Priority, market -ready technologies and innovations: Air Void Analyzer. Washington D.C. PDF file. Germann Instruments (GI). 2011...tests and properties of concrete and concrete-making materials. STP 169D. West Conshohocken, PA: ASTM International. Magura, D.D. 1996. Air void

  15. Using voids to unscreen modified gravity

    Science.gov (United States)

    Falck, Bridget; Koyama, Kazuya; Zhao, Gong-Bo; Cautun, Marius

    2018-04-01

    The Vainshtein mechanism, present in many models of gravity, is very effective at screening dark matter haloes such that the fifth force is negligible and general relativity is recovered within their Vainshtein radii. Vainshtein screening is independent of halo mass and environment, in contrast to e.g. chameleon screening, making it difficult to test. However, our previous studies have found that the dark matter particles in filaments, walls, and voids are not screened by the Vainshtein mechanism. We therefore investigate whether cosmic voids, identified as local density minima using a watershed technique, can be used to test models of gravity that exhibit Vainshtein screening. We measure density, velocity, and screening profiles of stacked voids in cosmological N-body simulations using both dark matter particles and dark matter haloes as tracers of the density field. We find that the voids are completely unscreened, and the tangential velocity and velocity dispersion profiles of stacked voids show a clear deviation from Λ cold dark matter at all radii. Voids have the potential to provide a powerful test of gravity on cosmological scales.

  16. Simulation of dust voids in complex plasmas

    Science.gov (United States)

    Goedheer, W. J.; Land, V.

    2008-12-01

    In dusty radio-frequency (RF) discharges under micro-gravity conditions often a void is observed, a dust free region in the discharge center. This void is generated by the drag of the positive ions pulled out of the discharge by the electric field. We have developed a hydrodynamic model for dusty RF discharges in argon to study the behaviour of the void and the interaction between the dust and the plasma background. The model is based on a recently developed theory for the ion drag force and the charging of the dust. With this model, we studied the plasma inside the void and obtained an understanding of the way it is sustained by heat generated in the surrounding dust cloud. When this heating mechanism is suppressed by lowering the RF power, the plasma density inside the void decreases, even below the level where the void collapses, as was recently shown in experiments on board the International Space Station. In this paper we present results of simulations of this collapse. At reduced power levels the collapsed central cloud behaves as an electronegative plasma with corresponding low time-averaged electric fields. This enables the creation of relatively homogeneous Yukawa balls, containing more than 100 000 particles. On earth, thermophoresis can be used to balance gravity and obtain similar dust distributions.

  17. Condition monitoring of main coolant pumps, Dhruva

    International Nuclear Information System (INIS)

    Prasad, V.; Satheesh, C.; Acharya, V.N.; Tikku, A.C.; Mishra, S.K.

    2002-01-01

    Full text: Dhruva is a 100 MW research reactor with natural uranium fuel, heavy water as moderator and primary coolant. Three Centrifugal pumps circulate the primary coolant across the core and the heat exchangers. Each pump is coupled to a flywheel (FW) assembly in order to meet operational safety requirements. All the 3 main coolant pump (MCP) sets are required to operate during operation of the reactor. The pump-sets are in operation since the year 1984 and have logged more than 1,00,000 hrs. Frequent breakdowns of its FW bearings were experienced during initial years of operation. Condition monitoring of these pumps, largely on vibration based parameters, was initiated on regular basis. Break-downs of main coolant pumps reduced considerably due to the fair accurate predictions of incipient break-downs and timely maintenance efforts. An effort is made in this paper to share the experience

  18. Coolant processing device for nuclear reactor

    International Nuclear Information System (INIS)

    Kizawa, Hideo; Funakoshi, Toshio; Izumoji, Yoshiaki

    1981-01-01

    Purpose: To reduce an entire facility cost by concentrating and isolating tritium accumulated in coolants, removing the tritium out of the system, and returning hydrogen gas generated at a reactor accident to a recombiner in a closed loop by the switching of a valve. Constitution: Coolant from a reactor cooling system processed by a chemical volume control system facility (CVCS) and coolant drain from various devices processed by a liquid waste disposing system facility (LWDS) are fed to a tritium isolating facility, in which they are isolated into concentrated tritium water and dilute tritium water. The concentrated tritium water is removed out of the system and stored. The dilute tritium water is reused as supply water for coolant. If an accident occurs to cause hydrogen to be generated, a closed loop is formed between the containment vessel and the recombiner, the hydrogen is recombined with oxygen in the air of the closed loop to be thus returned to water. (Kamimura, M.)

  19. Fatigue management considering LWR coolant environments

    International Nuclear Information System (INIS)

    Park, Heung Bae; Jin, Tae eun

    2000-01-01

    Design fatigue curve for structural material in the ASME Boiler and Pressure Vessel Code do not explicitly address the effects of reactor coolant environments on fatigue life. Environmentally assisted cracking (EAC) of low-alloy steels in light water reactor (LWR) coolant environments has been a concern ever since the early 1970's. And, recent fatigue test data indicate a significant decrease in fatigue lives of carbon steels, low-alloy steels and austenitic stainless steels in LWR coolant environments. For these reasons, fatigue of major components has been identified as a technical issue remaining to be resolved for life management and license renewal of nuclear power plants. In the present paper, results of recent investigations by many organizations are reviewed to provide technical justification to support the development of utility approach regarding the management of fatigue considering LWR coolant environments for the purpose of life management and license renewal of nuclear power plants. (author)

  20. Selection of nuclear reactor coolant materials

    International Nuclear Information System (INIS)

    Shi Lisheng; Wang Bairong

    2012-01-01

    Nuclear material is nuclear material or materials used in nuclear industry, the general term, it is the material basis for the construction of nuclear power, but also a leader in nuclear energy development, the two interdependent and mutually reinforcing. At the same time, nuclear materials research, development and application of the depth and breadth of science and technology reflects a nation and the level of the nuclear power industry. Coolant also known as heat-carrier agent, is an important part of the heart nuclear reactor, its role is to secure as much as possible to the economic output in the form fission energy to heat the reactor to be used: the same time cooling the core, is controlled by the various structural components allowable temperature. This paper described the definition of nuclear reactor coolant and characteristics, and then addressed the requirements of the coolant material, and finally were introduced several useful properties of the coolant and chemical control. (authors)

  1. Standardized sampling system for reactor coolants

    International Nuclear Information System (INIS)

    Divine, J.R.; Munson, L.F.; Nelson, J.L.; McDowell, R.L.; Jankowski, M.W.

    1982-09-01

    A three-pronged approach was developed to reach the objectives of acceptable coolant sampling, assessment of occupational exposure from corrosion products, and model development for the transport and buildup of corrosion products. Emphasis is on sampler design

  2. Reactor coolant pump seals: improving their performance

    International Nuclear Information System (INIS)

    Pothier, N.E.; Metcalfe, R.

    1986-06-01

    Large CANDU plants are benefitting from transient-resistant four-year reliable reactor coolant pump seal lifetimes, a direct result of AECL's 20-year comprehensive seal improvement program involving R and D staff, manufacturers, and plant designers and operators. An overview of this program is presented, which covers seal modification design, testing, post-service examination, specialized maintenance and quality control. The relevancy of this technology to Light Water Reactor Coolant Pump Seals is also discussed

  3. Coolant clean up system in nuclear reactor

    International Nuclear Information System (INIS)

    Tajima, Fumio; Iwami, Hiroshi.

    1981-01-01

    Purpose: To decrease the amount of main steams and improve the plant heat efficiency by the use of condensated water as coolants for not-regenerative heat exchangers in a coolant clean up system of a nuclear reactor. Constitution: In a coolant clean up system of a nuclear reactor, a portion of condensates is transferred to the shell of a non-regenerative heat exchanger by way of a condensate pump for non-regenerative heat exchanger through a branched pipeway provided to the outlet of a condensate desalter for using the condensates as the coolants for the shell of the heat exchanger and the condensates are then returned to the inlet of a feedwater heater after the heat exchange. The branched flow rate of the condensates is controlled by the flow rate control valve mounted in the pipeway. Condensates passed through the heat exchanger and the condensates not passed through the heat exchanger are mixed and heated in a heater and then fed to the nuclear reactor. In a case where no feedwater is necessary to the nuclear reactor such as upon shutdown of the reactor, the condensates are returned by way of feedwater bypass pipeway to the condensator. By the use of the condensates as the coolants for the heat exchanger, the main steam loss can be decreased and the thermal load for the auxiliary coolant facility can be reduced. (Kawakami, Y.)

  4. 38 CFR 3.207 - Void or annulled marriage.

    Science.gov (United States)

    2010-07-01

    ... 38 Pensions, Bonuses, and Veterans' Relief 1 2010-07-01 2010-07-01 false Void or annulled marriage... Void or annulled marriage. Proof that a marriage was void or has been annulled should consist of: (a... marriage void, together with such other evidence as may be required for a determination. (b) Annulled. A...

  5. Software quality assurance plan for void fraction instrument

    International Nuclear Information System (INIS)

    Gimera, M.

    1994-01-01

    Waste Tank SY-101 has been the focus of extensive characterization work over the past few years. The waste continually generates gases, most notably hydrogen, which are periodically released from the waste. Gas can be trapped in tank waste in three forms: as void gas (bubbles), dissolved gas, or absorbed gas. Void fraction is the volume percentage of a given sample that is comprised of void gas. The void fraction instrument (VFI) acquires the data necessary to calculate void fraction. This document covers the product, Void Fraction Data Acquisition Software. The void fraction software being developed will have the ability to control the void fraction instrument hardware and acquire data necessary to calculate the void fraction in samples. This document provides the software quality assurance plan, verification and validation plan, and configuration management plan for developing the software for the instrumentation that will be used to obtain void fraction data from Tank SY-101

  6. Post-void residual urine under 150 ml does not exclude voiding dysfunction in women

    DEFF Research Database (Denmark)

    Khayyami, Yasmine; Klarskov, Niels; Lose, Gunnar

    2016-01-01

    INTRODUCTION AND HYPOTHESIS: It has been claimed that post-void residual urine (PVR) below 150 ml rules out voiding dysfunction in women with stress urinary incontinence (SUI) and provides license to perform sling surgery. The cut-off of 150 ml seems arbitrary, not evidence-based, and so we sough...

  7. Nucleation of voids - the impurity effect

    International Nuclear Information System (INIS)

    Chen, I-W; Taiwo, A.

    1984-01-01

    Nucleation of voids under irradiation in multicomponent alloys remains an unsolved theoretical problem. Of particular interest are the effects of nonequilibrium solute segregation phenomena on the critical nucleus and the nucleation rate. The resolution of the multicomponent nucleation in a dissipative system also has broader implication to the field of irreversible thermodynamics. The present paper describes a recent study of solute segregation effects in void nucleation. We begin with a thermodynamic model for a nonequilibrium void with interfacial segregation. The thermodynamic model is coupled with kinetic considerations of solute/solvent diffusion under a bias, which is itself related to segregation by the coating effect, to assess the stability of void embryos. To determine nucleation rate, we develop a novel technique by extending the most probable path method in statistical mechanics for nonequilibrium steady state to simulate large fluctuation with nonlinear dissipation. The path of nucleation is determined by solving an analogous problem on particle trajectory in classical dynamics. The results of both the stability analysis and the fluctuation analysis establish the paramount significance of the impurity effect via the mechanism of nonequilibrium segregation. We conclude that over-segregation is probably the most general cause for the apparently low nucleation barriers that are responsible for nearly ubiquitous occurrence of void swelling in common metals

  8. The Metallicity of Void Dwarf Galaxies

    Science.gov (United States)

    Kreckel, K.; Croxall, K.; Groves, B.; van de Weygaert, R.; Pogge, R. W.

    2015-01-01

    The current ΛCDM cosmological model predicts that galaxy evolution proceeds more slowly in lower density environments, suggesting that voids are a prime location to search for relatively pristine galaxies that are representative of the building blocks of early massive galaxies. To test the assumption that void galaxies are more pristine, we compare the evolutionary properties of a sample of dwarf galaxies selected specifically to lie in voids with a sample of similar isolated dwarf galaxies in average density environments. We measure gas-phase oxygen abundances and gas fractions for eight dwarf galaxies (Mr > -16.2), carefully selected to reside within the lowest density environments of seven voids, and apply the same calibrations to existing samples of isolated dwarf galaxies. We find no significant difference between these void dwarf galaxies and the isolated dwarf galaxies, suggesting that dwarf galaxy chemical evolution proceeds independent of the large-scale environment. While this sample is too small to draw strong conclusions, it suggests that external gas accretion is playing a limited role in the chemical evolution of these systems, and that this evolution is instead dominated mainly by the internal secular processes that are linking the simultaneous growth and enrichment of these galaxies.

  9. Evaluation Methodology for Void Swelling Susceptibility of APR1400 Reactor Vessel Internals for U.S. NRC Design Certification

    Energy Technology Data Exchange (ETDEWEB)

    Kweon, Hyeong Do; Lee, Do Hwan [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The APR1400 RVI (Reactor Vessel Internals) operates in harsh conditions, such as long term exposure to neutron irradiation, high temperatures, reactor coolant environment, and other operating loads. Therefore, even though the RVI components are mainly made of austenitic stainless steel which is well known to have good mechanical and corrosion-resistive properties, these operating conditions. The aging is characterized by a chromium depletion along grain boundaries of austenitic stainless steel, a decrease in ductility and fracture toughness of the steel, an increase in yield and ultimate strength of the steel, and a potential volume change due to void formation in the steel. For these reasons, under certain conditions of stress, temperature, and level of irradiation, the void swelling which is one of the challenging degradation mechanisms affecting the integrity of the RVI may appear at specific locations of the RVI, especially due to high neutron fluence and high temperature under localized gamma heating and low velocity of coolant flow. To assess the effects of operating neutron fluences, temperatures and stresses on the material properties changes and the susceptibility to the void swelling, the evaluation methodology of the APR1400 RVI components for U.S. NRC Design Certification was suggested in this paper. The approach to the evaluation is summarized as follows: 1. RVI component list of the APR1400 is collected. 2. Initial screening to determine the evaluation scope is completed using the design values of fluences. 3. Functionality assessments (radiation transport analysis, CFD analysis, structural analysis) are sequentially performed. 4. Susceptibility to the void swelling is identified through ANSYS/USERMAT module. KHNP believes that the proposed methodology which is based on the EPRI works for operating reactors is the best way to evaluate the void swelling for new reactors such as the APR1400.

  10. Failure by void coalescence in metallic materials containing primary and secondary voids subject to intense shearing

    DEFF Research Database (Denmark)

    Nielsen, Kim Lau; Tvergaard, Viggo

    2011-01-01

    Failure under intense shearing at close to zero stress triaxiality is widely observed for ductile metallic materials, and is identified in experiments as smeared-out dimples on the fracture surface. Numerical cell-model studies of equal sized voids have revealed that the mechanism governing...... this shear failure mode boils down to the interaction between primary voids which rotate and elongate until coalescence occurs under severe plastic deformation of the internal ligaments. The objective of this paper is to analyze this failure mechanism of primary voids and to study the effect of smaller...... secondary damage that co-exists with or nucleation in the ligaments between larger voids that coalesce during intense shearing. A numerical cell-model study is carried out to gain a parametric understanding of the overall material response for different initial conditions of the two void populations...

  11. Consideration of hot channel factors in design for providing operating margins on coolant channel outlet temperature

    International Nuclear Information System (INIS)

    Sharma, V.K.; Surendar, C.; Bapat, C.N.

    1994-01-01

    The Indian Pressurized Heavy Water Reactors (IPHWR) are horizontal pressure tube reactors using natural uranium oxide fuel in the form of short (495 mm) clusters. The fuel clusters in the Zr-Nb pressure tubes are cooled by high pressure, high temperature and subcooled circulating heavy water. Coolant flow distribution to individual channels is designed to match the power distribution so as to obtain uniform coolant outlet temperature. However, during operation, the coolant outlet temperature in individual channels deviate from their nominal value due to: tolerances in process design; effects of grid frequency on the pump speed; deviation in channel powers from the nominal values due to on-power fuelling and movement of reactivity devices, and so on. Thus an operating margin, between the highest permissible and nominal coolant outlet temperatures, is required taking into account various hot channel factors that contribute to higher coolant outlet temperatures. The paper discusses the methodology adopted to assess various hot channel factors which would provide optimum operating margins while ensuring sub-cooling. (author)

  12. Experimental and numerical investigation of the coolant mixing during fast deboration transients

    International Nuclear Information System (INIS)

    Hoehne, T.; Rohde, U.; Weiss, F.P.

    1999-01-01

    For the analysis of boron dilution transients and main steam line break scenarios the modeling of the coolant mixing inside the reactor vessel is important, because the reactivity insertion strongly depends on boron acid concentration or the coolant temperature distribution. Calculations for steady state flow conditions for the VVER-440 were performed with a CFD code (CFX-4). The comparison with experimental data and an analytical mixing model which is implemented in the neutron-kinetic code DYN3D showed a good agreement for near-nominal conditions. First experiments at the Rossendorf Mixing Test Facility ROCOM were performed simulating the start-up of the first main coolant pump. The reference reactor for the geometrically 1:5 scaled Plexiglas model is the German Konvoi type PWR. After demonstrating the capability of the CFD code to simulate these complicated flow transients, calculations were performed for the start-up of the first pump in a VVER-440 type reactor. These calculations are a first step of understanding the coolant mixing in the RPV of a VVER-440 type reactor under transient conditions. The results of the calculation show a very complex flow in the downcomer. A high downcomer of VVER-440 and the existence of the lower control rod chamber support coolant mixing is concluded. (author)

  13. Void growth suppression by dislocation impurity atmospheres

    International Nuclear Information System (INIS)

    Weertman, J.; Green, W.V.

    1976-01-01

    A detailed calculation is given of the effect of an impurity atmosphere on void growth under irradiation damage conditions. Norris has proposed that such an atmosphere can suppress void growth. The hydrostatic stress field of a dislocation that is surrounded by an impurity atmosphere was found and used to calculate the change in the effective radius of a dislocation line as a sink for interstitials and vacancies. The calculation of the impurity concentration in a Cottrell cloud takes into account the change in hydrostatic pressure produced by the presence of the cloud itself. It is found that void growth is eliminated whenever dislocations are surrounded by a condensed atmosphere of either oversized substitutional impurity atoms or interstitial impurity atoms. A condensed atmosphere will form whenever the average impurity concentration is larger than a critical concentration

  14. Comparative study of void fraction models

    International Nuclear Information System (INIS)

    Borges, R.C.; Freitas, R.L.

    1985-01-01

    Some models for the calculation of void fraction in water in sub-cooled boiling and saturated vertical upward flow with forced convection have been selected and compared with experimental results in the pressure range of 1 to 150 bar. In order to know the void fraction axial distribution it is necessary to determine the net generation of vapour and the fluid temperature distribution in the slightly sub-cooled boiling region. It was verified that the net generation of vapour was well represented by the Saha-Zuber model. The selected models for the void fraction calculation present adequate results but with a tendency to super-estimate the experimental results, in particular the homogeneous models. The drift flux model is recommended, followed by the Armand and Smith models. (F.E.) [pt

  15. From Voids to Yukawaballs And Back

    International Nuclear Information System (INIS)

    Land, V.; Goedheer, W. J.

    2008-01-01

    When dust particles are introduced in a radio-frequency discharge under micro-gravity conditions, usually a dust free void is formed due to the ion drag force pushing the particles away from the center. Experiments have shown that it is possible to close the void by reducing the power supplied to the discharge. This reduces the ion density and with that the ratio between the ion drag force and the opposing electric force. We have studied the behavior of a discharge with a large amount of dust particles (radius 3.4 micron) with our hydrodynamic model, and simulated the closure of the void for conditions similar to the experiment. We also approached the formation of a Yukawa ball from the other side, starting with a discharge at low power and injecting batches of dust, while increasing the power to prevent extinction of the discharge. Eventually the same situation could be reached.

  16. Investigation of coolant mixture in pressurized water reactors at the Rossendorf mixing test facility ROCOM

    International Nuclear Information System (INIS)

    Grunwald, G.; Hoehne, T.; Prasser, H.M.; Richter, K.; Weiss, F.P.

    1999-01-01

    During the so-called boron dilution or cold water transients at pressurized water reactors too weakly borated water or too cold water, respectively, might enter the reactor core. This results in the insertion of positive reactivity and possibly leads to a power excursion. If the source of unborated or subcooled water is not located in all coolant loops but in selected ones only, the amount of reactivity insertion depends on the coolant mixing in the downcomer and lower plenum of the reactor pressure vessel (RPV). Such asymmetric disturbances of the coolant temperature or boron concentration might e.g. be the result of a failure of the chemical and volume control system (CVCS) or of a main steam line break (MSLB) that does only affect selected steam generators (SG). For the analysis of boron dilution or MSLB accidents coupled neutron kinetics/thermo-hydraulic system codes have been used. To take into account coolant mixing phenomena in these codes in a realistic manner, analytical mixing models might be included. These models must be simple and fast running on the one hand, but must well describe the real mixing conditions on the other hand. (orig.)

  17. LWR primary coolant pipe rupture test rig

    International Nuclear Information System (INIS)

    Yoshitoshi, Shyoji

    1978-01-01

    The rupture test rig for primary coolant pipes is constructed in the Japan Atomic Energy Research Institute to verify the reliability of the primary coolant pipes for both PWRs and BWRs. The planned test items consisted of reaction force test, restraint test, whip test, jet test and continuous release test. A pressure vessel of about 4 m 3 volume, a circulating pump, a pressurizer, a heater, an air cooler and the related instrumentation and control system are included in this test rig. The coolant test condition is 160 kg/cm 2 g, 325 deg C for PWR test, and 70 kg/cm 2 g, saturated water and steam for BWR test, 100 ton of test load for the ruptured pipe bore of 8B Schedule 160, and 20 lit/min. discharge during 20 h for continuous release of coolant. The maximum pit internal pressure was estimated for various pipe diameters and time under the PWR and BWR conditions. The spark rupturing device was adopted for the rupture mechanics in this test rig. The computer PANAFACOM U-300 is used for the data processing. This test rig is expected to operate in 1978 effectively for the improvement of reliability of LWR primary coolant pipes. (Nakai, Y.)

  18. Critical velocities for deflagration and detonation triggered by voids in a REBO high explosive

    Energy Technology Data Exchange (ETDEWEB)

    Herring, Stuart Davis [Los Alamos National Laboratory; Germann, Timothy C [Los Alamos National Laboratory; Jensen, Niels G [Los Alamos National Laboratory

    2010-01-01

    The effects of circular voids on the shock sensitivity of a two-dimensional model high explosive crystal are considered. We simulate a piston impact using molecular dynamics simulations with a Reactive Empirical Bond Order (REBO) model potential for a sub-micron, sub-ns exothermic reaction in a diatomic molecular solid. The probability of initiating chemical reactions is found to rise more suddenly with increasing piston velocity for larger voids that collapse more deterministically. A void with radius as small as 10 nm reduces the minimum initiating velocity by a factor of 4. The transition at larger velocities to detonation is studied in a micron-long sample with a single void (and its periodic images). The reaction yield during the shock traversal increases rapidly with velocity, then becomes a prompt, reliable detonation. A void of radius 2.5 nm reduces the critical velocity by 10% from the perfect crystal. A Pop plot of the time-to-detonation at higher velocities shows a characteristic pressure dependence.

  19. Partial discharges in ellipsoidal and spheroidal voids

    DEFF Research Database (Denmark)

    Crichton, George C; Karlsson, P. W.; Pedersen, Aage

    1989-01-01

    Transients associated with partial discharges in voids can be described in terms of the charges induced on the terminal electrodes of the system. The relationship between the induced charge and the properties which are usually measured is discussed. The method is illustrated by applying it to a s......Transients associated with partial discharges in voids can be described in terms of the charges induced on the terminal electrodes of the system. The relationship between the induced charge and the properties which are usually measured is discussed. The method is illustrated by applying...

  20. Voids and overdensities of coupled Dark Energy

    International Nuclear Information System (INIS)

    Mainini, Roberto

    2009-01-01

    We investigate the clustering properties of dynamical Dark Energy even in association of a possible coupling between Dark Energy and Dark Matter. We find that within matter inhomogeneities, Dark Energy migth form voids as well as overdensity depending on how its background energy density evolves. Consequently and contrarily to what expected, Dark Energy fluctuations are found to be slightly suppressed if a coupling with Dark Matter is permitted. When considering density contrasts and scales typical of superclusters, voids and supervoids, perturbations amplitudes range from |δ φ | ∼ O(10 −6 ) to |δ φ | ∼ O(10 −4 ) indicating an almost homogeneous Dark Energy component

  1. Radar application in void and bar detection

    International Nuclear Information System (INIS)

    Amry Amin Abas; Mohamad Pauzi Ismail; Suhairy Sani

    2003-01-01

    Radar is one of the new non-destructive testing techniques for concrete and structures inspection. Radar is a non-ionizing electromagnetic wave that can penetrate deep into concrete or soil in about several tenths of meters. Method of inspection using radar enables us to perform high resolution detection, imaging and mapping of subsurface concrete and soil condition. This paper will discuss the use of radar for void and bar detection and sizing. The samples used in this paper are custom made samples and comparison will be made to validate the use of radar in detecting, locating and also size determination of voids and bars. (Author)

  2. Measurement of void fractions by nuclear techniques

    International Nuclear Information System (INIS)

    Hernandez G, A.; Vazquez G, J.; Diaz H, C.; Salinas R, G.A.

    1997-01-01

    In this work it is done a general analysis of those techniques used to determine void fractions and it is chosen a nuclear technique to be used in the heat transfer circuit of the Physics Department of the Basic Sciences Management. The used methods for the determination of void fractions are: radioactive absorption, acoustic techniques, average velocity measurement, electromagnetic flow measurement, optical methods, oscillating absorption, nuclear magnetic resonance, relation between pressure and flow oscillation, infrared absorption methods, sound neutron analysis. For the case of this work it will be treated about the radioactive absorption method which is based in the gamma rays absorption. (Author)

  3. Fuel -coolant interactions in LWRs and LMFBRs: relationships and distinctions

    Energy Technology Data Exchange (ETDEWEB)

    Duffey, R B; Lellouche, G S [Nuclear Safety and Analysis Department, Electric Power Research Institute, Palo Alto, CA (United States)

    1979-10-15

    The question of fuel-coolant interaction and of potential vapor explosion is raised here. lt is the contention of the authors that there is in fact no need to study this question vis a vis Light Water Reactors (LWR) except from an academic point of view since it does not impact on safety considerations. As for LMFBRs, the design basis whole core accidents for LWRs are derived from the fundamental concern of maintaining core geometry to provide for convective cooling. However, the important distinction is that the core is in its most reactive configuration, and core and fuel rearrangement is therefore not of such concern. The author's thesis is that even if the probability of steam explosion following core melt were two orders of magnitude greater than currently assumed (10{sup -2}) the total LWR risk would increase only by a factor of 2-6 for BWRs and less a factor of 10 for PWRs

  4. Fusion Blanket Coolant Section Criteria, Methodology, and Results

    Energy Technology Data Exchange (ETDEWEB)

    DeMuth, J. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Meier, W. R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Jolodosky, A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Frantoni, M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Reyes, S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-10-02

    The focus of this LDRD was to explore potential Li alloys that would meet the tritium breeding and blanket cooling requirements but with reduced chemical reactivity, while maintaining the other attractive features of pure Li breeder/coolant. In other fusion approaches (magnetic fusion energy or MFE), 17Li- 83Pb alloy is used leveraging Pb’s ability to maintain high TBR while lowering the levels of lithium in the system. Unfortunately this alloy has a number of potential draw-backs. Due to the high Pb content, this alloy suffers from very high average density, low tritium solubility, low system energy, and produces undesirable activation products in particular polonium. The criteria considered in the selection of a tritium breeding alloy are described in the following section.

  5. Nuclear reactor of pressurized liquid coolant type

    International Nuclear Information System (INIS)

    Costes, D.

    1976-01-01

    The reactor comprises a vertical concrete pressure vessel, a bell-housing having an open lower end and disposed coaxially with the interior of the pressure vessel so as to delimit therewith a space filled with gas under pressure for the thermal insulation of the internal vessel wall, a pressurizing device for putting the coolant under pressure within the bell-housing and comprising a volume of control gas in contact with a large free surface of coolant in order that an appreciable variation in volume of liquid displaced within the coolant circuit inside the bell-housing should correspond to a small variation in pressure of the control gas. 9 claims, 3 drawing figures

  6. ENVIRONMENTALLY REDUCING OF COOLANTS IN METAL CUTTING

    Directory of Open Access Journals (Sweden)

    Veijo KAUPPINEN

    2012-11-01

    Full Text Available Strained environment is a global problem. In metal industries the use of coolant has become more problematic in terms of both employee health and environmental pollution. It is said that the use of coolant forms approximately 8 - 16 % of the total production costs.The traditional methods that use coolants are now obviously becoming obsolete. Hence, it is clear that using a dry cutting system has great implications for resource preservation and waste reduction. For this purpose, a new cooling system is designed for dry cutting. This paper presents the new eco-friendly cooling innovation and the benefits gained by using this method. The new cooling system relies on a unit for ionising ejected air. In order to compare the performance of using this system, cutting experiments were carried out. A series of tests were performed on a horizontal turning machine and on a horizontal machining centre.

  7. Iron crud supply device to reactor coolant

    International Nuclear Information System (INIS)

    Baba, Takao.

    1993-01-01

    In a device for supplying iron cruds into reactor coolants in a BWR type power plant, a system in which feed water containing iron cruds is supplied to the reactor coolants after once passing through an ion exchange resin is disposed. As a result, iron cruds having characteristics similar with those of naturally occurring iron cruds in the plant are obtained and they react with ionic radioactivity, to form composite oxides. Then, iron cruds having high performance of being secured to the surface of a fuel cladding tube can be supplied to the reactor coolants, thereby enabling to greatly reduce the density of reactor water ionic radioactivity. In its turn, dose rate on the surface of pipelines can be reduced, thereby enabling to reduce operators' radiation exposure dose in the plant. Further, contamination of a condensate desalting device due to iron cruds can be prevented, and further, the density of the iron cruds supplied can easily be controlled. (N.H.)

  8. Limits to fuel/coolant mixing

    International Nuclear Information System (INIS)

    Corradini, M.L.; Moses, G.A.

    1985-01-01

    The vapor explosion process involves the mixing of fuel with coolant prior to the explosion. A number of analysts have identified limits to the amount of fuel/coolant mixing that could occur within the reactor vessel following a core melt accident. Past models are reviewed and a sim plified approach is suggested to estimate the upper limit on the amount of fuel/coolant mixing pos sible. The approach uses concepts first advanced by Fauske in a different way. The results indicat that water depth is an important parameter as well as the mixing length scale D /SUB mix/ , and for large values of D /SUB mix/ the fuel mass mixed is limited to <7% of the core mass

  9. The evolution of voids in the adhesion approximation

    Science.gov (United States)

    Sahni, Varun; Sathyaprakah, B. S.; Shandarin, Sergei F.

    1994-08-01

    We apply the adhesion approximation to study the formation and evolution of voids in the universe. Our simulations-carried out using 1283 particles in a cubical box with side 128 Mpc-indicate that the void spectrum evolves with time and that the mean void size in the standard Cosmic Background Explorer Satellite (COBE)-normalized cold dark matter (CDM) model with H50 = 1 scals approximately as bar D(z) = bar Dzero/(1+2)1/2, where bar Dzero approximately = 10.5 Mpc. Interestingly, we find a strong correlation between the sizes of voids and the value of the primordial gravitational potential at void centers. This observation could in principle, pave the way toward reconstructing the form of the primordial potential from a knowledge of the observed void spectrum. Studying the void spectrum at different cosmological epochs, for spectra with a built in k-space cutoff we find that the number of voids in a representative volume evolves with time. The mean number of voids first increases until a maximum value is reached (indicating that the formation of cellular structure is complete), and then begins to decrease as clumps and filaments erge leading to hierarchical clustering and the subsequent elimination of small voids. The cosmological epoch characterizing the completion of cellular structure occurs when the length scale going nonlinear approaches the mean distance between peaks of the gravitaional potential. A central result of this paper is that voids can be populated by substructure such as mini-sheets and filaments, which run through voids. The number of such mini-pancakes that pass through a given void can be measured by the genus characteristic of an individual void which is an indicator of the topology of a given void in intial (Lagrangian) space. Large voids have on an average a larger measure than smaller voids indicating more substructure within larger voids relative to smaller ones. We find that the topology of individual voids is strongly epoch dependent

  10. Dynamic void behavior in polymerizing polymethyl methacrylate cement.

    Science.gov (United States)

    Muller, Scott D; McCaskie, Andrew W

    2006-02-01

    Cement mantle voids remain controversial with respect to survival of total hip arthroplasty. Void evolution is poorly understood, and attempts at void manipulation can only be empirical. We induced voids in a cement model simulating the constraints of the proximal femur. Intravoid pressure and temperature were recorded throughout polymerization, and the initial and final void volumes were measured. Temperature-dependent peak intravoid pressures and void volume increases were observed. After solidification, subatmospheric intravoid pressures were observed. The magnitude of these observations could not be explained by the ideal gas law. Partial pressures of the void gas at peak pressures demonstrated a dominant effect of gaseous monomer, thereby suggesting that void growth is a pressure-driven phenomenon resulting from temperature-dependent evaporation of monomer into existing trapped air voids.

  11. Friction stir welding process to repair voids in aluminum alloys

    Science.gov (United States)

    Rosen, Charles D. (Inventor); Litwinski, Edward (Inventor); Valdez, Juan M. (Inventor)

    1999-01-01

    The present invention provides an in-process method to repair voids in an aluminum alloy, particularly a friction stir weld in an aluminum alloy. For repairing a circular void or an in-process exit hole in a weld, the method includes the steps of fabricating filler material of the same composition or compatible with the parent material into a plug form to be fitted into the void, positioning the plug in the void, and friction stir welding over and through the plug. For repairing a longitudinal void (30), the method includes machining the void area to provide a trough (34) that subsumes the void, fabricating filler metal into a strip form (36) to be fitted into the trough, positioning the strip in the trough, and rewelding the void area by traversing a friction stir welding tool longitudinally through the strip. The method is also applicable for repairing welds made by a fusing welding process or voids in aluminum alloy workpieces themselves.

  12. Main coolant pump testing at Ontario Hydro

    International Nuclear Information System (INIS)

    Hartlen, R.

    1991-01-01

    This article describes Ontario Hydro Research Division's experience with a computerized data acquisition and analysis system for monitoring mechanical vibration in reactor coolant pumps. The topics covered include bench-marking of the computer system and the coolant pumps, signatures of normal and malfunctioning pumps, analysis of data collected by the monitoring system, simulation of faults, and concerns that have been expressed about data interpretation, sensor types and locations, alarm/shutdown limits and confirmation of nondestructive examination testing. This presentation consists of overheads only

  13. Comparative design study of FR plants with various coolants. 1. Studies on Na coolant FR, Pb-Bi coolant FR, gas coolant FR

    International Nuclear Information System (INIS)

    Konomura, Mamoru; Shimakawa, Yoshio; Hori, Toru; Kawasaki, Nobuchika; Enuma, Yasuhiro; Kida, Masanori; Kasai, Shigeo; Ichimiya, Masakazu

    2001-01-01

    In Phase I of the Feasibility Studies on the Commercialized Fast Reactor (FR) Cycle System, plant designs on FR were performed with various coolants. This report describes the plant designs on FR with sodium, lead-bismuth, CO 2 gas and He gas coolants. A construction cost of 0.2 million yen/kWe was set up as a design goal. The result is as follows: The sodium reactor has a capability to obtain the goal, and lead-bismuth and gas reactors may satisfy the goal with further improvements. (author)

  14. On-Line Coolant Chemistry Analysis

    International Nuclear Information System (INIS)

    LM Bachman

    2006-01-01

    Impurities in the gas coolant of the space nuclear power plant (SNPP) can provide valuable indications of problems in the reactor and an overall view of system health. By monitoring the types and amounts of these impurities, much can be implied regarding the status of the reactor plant. However, a preliminary understanding of the expected impurities is important before evaluating prospective detection and monitoring systems. Currently, a spectroscopy system is judged to hold the greatest promise for monitoring the impurities of interest in the coolant because it minimizes the number of entry and exit points to the plant and provides the ability to detect impurities down to the 1 ppm level

  15. Leak detection device for reactor coolant

    International Nuclear Information System (INIS)

    Oshima, Koichiro.

    1990-01-01

    In a light water cooled reactor, if reactor coolants are leaked from pipelines in a pipeline chamber, activated products (N-16) are diffused together to an atmosphere in the pipeline chamber. N-16 is sucked from an extracting tube which is always sucking the atmosphere in the pipeline chamber to a sucking blower. Then, β-rays released from N-16 are monitored by a radiation monitor in a measuring chamber which is radiation-shielded from the pipeline chamber. Accordingly, since the radiation monitor can detect even slight leakage, the slight leakage of reactor coolants in the pipelines can be detected at an early stage. (I.N.)

  16. Reactor coolant pump for a nuclear reactor

    International Nuclear Information System (INIS)

    Burkhardt, W.; Richter, G.

    1976-01-01

    An improvement is proposed concerning the easier disengagement of the coupling at the reactor coolant pump for a nuclear reactor transporting a pressurized coolant. According to the invention the disengaging coupling consists of two parts separated by screws. At least one of the screws contains a propellent charge ananged within a bore and provided with a speed-dependent ignition device in such a way that by separation of the screws at overspeeds the coupling is disengaged. The sub-claims are concerned with the kind of ignition ot the propellent charge. (UWI) [de

  17. Theoretical Calculations of the Effect on Lattice Parameters of Emptying the Coolant Channels in a D{sub 2}O- Moderated and Cooled Natural Uranium Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Weissglas, P [The Swedish State Power Board, Stockholm (Sweden)

    1960-11-15

    The purpose of the present study was to evaluate theoretically the effect of coolant boiling and subsequent void formation in a pressurized D{sub 2}O moderated and cooled reactor. The fuel rods were arranged in a cluster geometry and clad in Zr-2. The coolant was separated from the moderator by a Zr-2 shroud. In this geometry the following problems have been given special attention: l) calculation of the effective resonance integral, 2) thermal disadvantage factors, 3) fast fission effects, 4) leakage effects, 5) changes in epithermal absorption. No account has up to now been taken of the variation of these effects with position in the reactor and burnup. Some comparisons of the theoretical methods and measurements have been attempted. It is concluded that at the present time it is not possible to calculate the void coefficient with any accuracy but it may be possible to give an upper limit from theoretical consideration.

  18. The Metallicity of Void Dwarf Galaxies

    NARCIS (Netherlands)

    Kreckel, K.; Croxall, K.; Groves, B.; van de Weygaert, R.; Pogge, R. W.

    The current ΛCDM cosmological model predicts that galaxy evolution proceeds more slowly in lower density environments, suggesting that voids are a prime location to search for relatively pristine galaxies that are representative of the building blocks of early massive galaxies. To test the

  19. "Dark energy" in the Local Void

    Science.gov (United States)

    Villata, M.

    2012-05-01

    The unexpected discovery of the accelerated cosmic expansion in 1998 has filled the Universe with the embarrassing presence of an unidentified "dark energy", or cosmological constant, devoid of any physical meaning. While this standard cosmology seems to work well at the global level, improved knowledge of the kinematics and other properties of our extragalactic neighborhood indicates the need for a better theory. We investigate whether the recently suggested repulsive-gravity scenario can account for some of the features that are unexplained by the standard model. Through simple dynamical considerations, we find that the Local Void could host an amount of antimatter (˜5×1015 M ⊙) roughly equivalent to the mass of a typical supercluster, thus restoring the matter-antimatter symmetry. The antigravity field produced by this "dark repulsor" can explain the anomalous motion of the Local Sheet away from the Local Void, as well as several other properties of nearby galaxies that seem to require void evacuation and structure formation much faster than expected from the standard model. At the global cosmological level, gravitational repulsion from antimatter hidden in voids can provide more than enough potential energy to drive both the cosmic expansion and its acceleration, with no need for an initial "explosion" and dark energy. Moreover, the discrete distribution of these dark repulsors, in contrast to the uniformly permeating dark energy, can also explain dark flows and other recently observed excessive inhomogeneities and anisotropies of the Universe.

  20. Void porosity measurements in coastal structures

    NARCIS (Netherlands)

    Bosma, C.; Verhagen, H.J.; D'Angremond, K.; Sint Nicolaas, W.

    2002-01-01

    The paper describes the use of two fundamental design parameters, the void porosity and layer thickness in rock armour constructions. These design parameters are very sensible for factors such as the boundary definition of a rock layer, rock production properties, intrinsic properties and

  1. Analytical and sampling problems in primary coolant circuits of PWR-type reactors

    International Nuclear Information System (INIS)

    Illy, H.

    1980-10-01

    Details of recent analytical methods on the analysis and sampling of a PWR primary coolant are given in the order as follows: sampling and preparation; analysis of the gases dissolved in the water; monitoring of radiating substances; checking of boric acid concentration which controls the reactivity. The bibliography of this work and directions for its use are published in a separate report: KFKI-80-48 (1980). (author)

  2. Nonlocal plasticity effects on interaction of different size voids

    DEFF Research Database (Denmark)

    Tvergaard, Viggo; Niordson, Christian Frithiof

    2004-01-01

    A nonlocal elastic-plastic material model is used to show that the rate of void growth is significantly reduced when the voids are small enough to be comparable with a characteristic material length. For a very small void in the material between much larger voids the competition between...... dimensional array of spherical voids. It is shown that the high growth rate of very small voids predicted by conventional plasticity theory is not realistic when the effect of a characteristic length, dependent on the dislocation structure, is accounted for. (C) 2003 Elsevier Ltd. All rights reserved....

  3. Corrective actions to gas accumulation in safety injection system pipings of PWRs and gas void detection method

    International Nuclear Information System (INIS)

    Maki, Nobuo

    2000-01-01

    In the US, gas accumulation events of safety injection systems of PWRs during plant operation are continuously reported. As the events may result in loss of safety function, the USNRC is alerting licensees by Information Notices. The cause of the events is coolant leakage to interfacing systems with lower pressure, or gas dissolution of primary coolant by partial pressure drop. In this study, it was clarified by the evaluation of the cause of the events of US plants, gas accumulation in piping between an accumulator and Residual Heat Removal System should be quantitatively investigated regarding Japanese plants. Also, effectiveness of ultrasonic testing which is used for monthly gas accumulation surveillance in US plants was demonstrated using a model loop. In addition, the method was confirmed applicable by an experiment carried out at INSS to detect cavitation voids in piping systems. (author)

  4. Dependence of hotspot initiation on void distribution in high explosive crystals simulated with molecular dynamics

    Science.gov (United States)

    Herring, Stuart Davis

    Microscopic defects may dramatically affect the susceptibility of high explosives to shock initiation. Such defects redirect the shock's energy and become hotspots (concentrations of stress and heat) that can initiate chemical reactions. Sufficiently large or numerous defects may produce a self-sustaining deflagration or even detonation from a shock notably too weak to detonate defect-free samples. The effects of circular or spherical voids on the shock sensitivity of a model (two- or three-dimensional) high explosive crystal are considered. We simulate a piston impact using molecular dynamics with a Reactive Empirical Bond Order (REBO) model potential for a sub-micron, sub-ns exothermic reaction in a diatomic molecular solid. In both dimensionalities, the probability of initiating chemical reactions rises more suddenly with increasing piston velocity for larger voids that collapse more deterministically. A void of even 10 nm radius (˜39 interatomic spacings) reduces the minimum initiating velocity by a factor of 4 (8 in 3D). The transition at larger velocities to detonation is studied in micron-long samples with a single void (and its periodic images). Reactions during the shock traversal increase rapidly with velocity, then become a reliable detonation. In 2D, a void of radius 2.5 nm reduces the critical velocity by 10% from the perfect crystal; a Pop plot of the detonation delays at higher velocities shows a characteristic pressure dependence. 3D samples are more likely to react but less to detonate. In square lattices of voids, reducing the (common) void radius or increasing the porosity without changing the other parameter causes the hotspots to consume the material faster and detonation to occur sooner and at lower velocities. Early behavior is seen to follow a very simple ignition and growth model; the pressure exponents are more realistic than with single voids. The hotspots collectively develop a broad pressure wave (a sonic, diffuse deflagration front

  5. Coolant cleanup system for BWR type reactor

    International Nuclear Information System (INIS)

    Kinoshita, Shoichiro; Araki, Hidefumi.

    1993-01-01

    The cleanup system of the present invention removes impurity ions and floating materials accumulated in a reactor during evaporation of coolants in the nuclear reactor. That is, coolants pass pipelines from a pressure vessel using pressure difference between a high pressure in the pressure vessel and a low pressure at the upstream of a condensate filtration/desalting device of a condensate/feed water system as a driving source, during which cations and floating materials are removed in a high temperature filtration/desalting device and coolants flow into the condensate/feedwater system. Impurities containing anions are removed here by the condensates filtration/desalting device. Then, they return to the pressure vessel while pressurized and heated by a condensate pump, a feed water pump and a feed water heater. At least pumps, a heat exchanger for heating, a filtration/desalting device for removing anions and pipelines connecting them used exclusively for the coolant cleanup system are no more necessary. (I.S.)

  6. Fission product release into the primary coolant

    International Nuclear Information System (INIS)

    Apperson, C.E.

    1977-01-01

    The analytic evaluation of steady state primary coolant activity is discussed. The reported calculations account for temperature dependent fuel failure in two particle types and arbitrary radioactive decay chains. A matrix operator technique implemented in the SUVIUS code is used to solve the simultaneous equations. Results are compared with General Atomic Company's published results

  7. The 10B(n,α)7Li reaction in PWR coolants: calculations of the effect on coolant pH and on decreases in 10B isotopic fractions

    International Nuclear Information System (INIS)

    Polley, M.V.

    1988-07-01

    Boron is used as a chemical shim in PWRs for reactivity control and is added in the form of boric acid to the primary coolant. The 10 B(n,α) 7 Li reaction leads to a continuous increase in 7 Li in the primary coolant and to a continuous decrease in 10 B the isotope of boron responsible for control of reactivity. The rate of increase in coolant pH due to 7 Li production is calculated for the Sizewell 'B' PWR to enable judgements to be made on the frequency of sampling and removal of lithium required to maintain the pH of the primary coolant within the desired limits. Calculations are contrasted for the cases of natural boron and 100% 10 B chemical shims, for both a normal cycle and an extended 18 month cycle. Calculations of 10 B depletion over 30 years of operation as a function of the quantity of boron discharged to waste are also presented. 10 B isotopic fractions are calculated for the reactor coolant (RC), boric acid tanks (BATs) and refuelling water storage tank (RWST) assuming rapid mixing of BAT and RC boron for tritium control and other reasons. Such predictions enable assessments of the reactor physics implications of 10 B consumption to be made. (author)

  8. Application of radcal gamma thermometer assemblies for coolant monitoring in Ringhals W-PWRs

    International Nuclear Information System (INIS)

    Smith, R.D.; Romslo, K.; Moen, Oe.

    1982-07-01

    A study has been carried out investigating how Radcal Gamma Thermometers (RGTs) can be used for coolant inventory and core cooling monitoring in the Ringhals Westinghouse PWRs. The study concludes that two types of RGT rods would be required to come up with a complete solution covering both coolant inventory and core cooling monitoring. Above-core RGT rods will be installed in the guide tubes housing the outlet thermocouples. The Above-Core RGT rod is designed with 8 sensors where 4 are located in the upper head and 4 in the plenum. This rod will give an early warning about loss of coolant or void formation in the space from top of fuel to the reactor lid. A ninth thermocouple in this rod will measure the core outlet temperature as did the thermocouple the RGT rod replaced. The Above-Core RGT rods will give an early warning about approach to Inadequate Core Cooling (ICC) by measuring the collapsed water level inside the thermocouple guide tube. Four such rods are recommended per reactor. In-Core RGT rods are inserted from the seal table. These rods will give the information required for intelligent accident management in case ICC has developed. The signals obtainable from the rods will give direct information about fuel decay heat, core heat transfer conditions, core temperature and core coolant water level. The In-Core RGT rods can be used for local power monitoring during normal operation. Such a system can be shown to be economically motivated from a reactor operation point of view due to increased sensor lifetime, more accurate local power measurements, simpler physics corrections to signals, lower exposure to maintenance personnel. The signal transmission to the control room has been discussed, and ways have been indicated for presenting the information available to the operators. (Authors)

  9. Void fraction prediction in saturated flow boiling

    International Nuclear Information System (INIS)

    Francisco J Collado

    2005-01-01

    Full text of publication follows: An essential element in thermal-hydraulics is the accurate prediction of the vapor void fraction, or fraction of the flow cross-sectional area occupied by steam. Recently, the author has suggested to calculate void fraction working exclusively with thermodynamic properties. It is well known that the usual 'flow' quality, merely a mass flow rate ratio, is not at all a thermodynamic property because its expression in function of thermodynamic properties includes the slip ratio, which is a parameter of the process not a function of state. By the other hand, in the classic and well known expression of the void fraction - in function of the true mass fraction of vapor (also called 'static' quality), and the vapor and liquid densities - does not appear the slip ratio. Of course, this would suggest a direct procedure for calculating the void fraction, provided we had an accurate value of the true mass fraction of vapor, clearly from the heat balance. However the classic heat balance is usually stated in function of the 'flow' quality, what sounds really contradictory because this parameter, as we have noted above, is not at all a thermodynamic property. Then we should check against real data the actual relationship between the thermodynamic properties and the applied heat. For saturated flow boiling just from the inlet of the heated tube, and not having into account the kinetic and potential terms, the uniform applied heat per unit mass of inlet water and per unit length (in short, specific linear heat) should be closely related to a (constant) slope of the mixture enthalpy. In this work, we have checked the relation between the specific linear heat and the thermodynamic enthalpy of the liquid-vapor mixture using the actual mass fraction. This true mass fraction is calculated using the accurate measurements of the outlet void fraction taken during the Cambridge project by Knights and Thom in the sixties for vertical and horizontal

  10. Tank SY-101 void fraction instrument functional design criteria

    International Nuclear Information System (INIS)

    McWethy, L.M.

    1994-01-01

    This document presents the functional design criteria for design, analysis, fabrication, testing, and installation of a void fraction instrument for Tank SY-101. This instrument will measure the void fraction in the waste in Tank SY-101 at various elevations

  11. Risk management of low air void asphalt concrete mixtures.

    Science.gov (United States)

    2013-07-01

    Various forms of asphalt pavement distress, such as rutting, shoving and bleeding, can be attributed, in many cases, to low air voids in : the mixtures during production and placement. The occurrence of low air void contents during plant production m...

  12. Analysis of Coolant Options for Advanced Metal Cooled Nuclear Reactors

    National Research Council Canada - National Science Library

    Can, Levent

    2006-01-01

    .... The overall focus of this study is the build up of induced radioactivity in the coolant of metal cooled reactors as well as the evaluation of other physical and chemical properties of such coolants...

  13. Plasticity size effects in voided crystals

    DEFF Research Database (Denmark)

    Hussein, M. I.; Borg, Ulrik; Niordson, Christian Frithiof

    singularities in an elastic material. The lattice resistance to dislocation motion, dislocation nucleation, dislocation interaction with obstacles and annihilation are incorporated through a set of constitutive rules. Over the range of length scales investigated, both the discrete dislocation and strain......The shear and equi-biaxial straining responses of periodic voided single crystals are analysed using discrete dislocation plasticity and a continuum strain gradient crystal plasticity theory. In the discrete dislocation formulation the dislocations are all of edge character and are modelled as line...... predictions of the two formulations for all crystal types and void volume fractions considered when the material length scale in the non-local plasticity model chosen to be $0.325\\mu m$ (around ten times the slip plane spacing in the discrete dislocation models)....

  14. Plasticity size effects in voided crystals

    DEFF Research Database (Denmark)

    Hussein, M.I.; Borg, Ulrik; Niordson, Christian Frithiof

    2008-01-01

    as line singularities in an elastic material. The lattice resistance to dislocation motion, dislocation nucleation, dislocation interaction with obstacles and annihilation are incorporated through a set of constitutive rules. Over the range of length scales investigated, both the discrete dislocation......The shear and equi-biaxial straining responses of periodic voided single crystals are analysed using discrete dislocation plasticity and a continuum strain gradient crystal plasticity theory. In the discrete dislocation formulation, the dislocations are all of edge character and are modelled...... between predictions of the two formulations for all crystal types and void volume fractions considered when the material length scale in the non-local plasticity model is chosen to be 0.325 mu m (about 10 times the slip plane spacing in the discrete dislocation models)....

  15. Effect of void cluster on ductile failure evolution

    DEFF Research Database (Denmark)

    Tvergaard, Viggo

    2016-01-01

    The behavior of a non-uniform void distribution in a ductile material is investigated by using a cell model analysis to study a material with a periodic pattern of void clusters. The special clusters considered consist of a number of uniformly spaced voids located along a plane perpendicular...

  16. Void probability scaling in hadron nucleus interactions

    International Nuclear Information System (INIS)

    Ghosh, Dipak; Deb, Argha; Bhattacharyya, Swarnapratim; Ghosh, Jayita; Bandyopadhyay, Prabhat; Das, Rupa; Mukherjee, Sima

    2002-01-01

    Heygi while investigating with the rapidity gap probability (that measures the chance of finding no particle in the pseudo-rapidity interval Δη) found that a scaling behavior in the rapidity gap probability has a close correspondence with the scaling of a void probability in galaxy correlation study. The main aim in this paper is to study the scaling behavior of the rapidity gap probability

  17. Neutron gauging to detect voids in polyurethane

    International Nuclear Information System (INIS)

    Tsang, F.Y.; Alger, D.M.; Brugger, R.M.

    1978-01-01

    Thermal-neutron radiography and fast-neutron gauging measurements were made to evaluate the feasibility of detecting voids in a polyurethane block placed between steel plates. This sandwich of polyurethane and steel simulates the walls of a canister being designed to hold explosive devices. The polyurethane would act as a shock absorber in the canister. A large fabrication cost saving would result by casting the polyurethane, but a nondestructive testing (NDT) method is needed to determine the uniformity of the polyurethane fill. The radiography measurements used a beam of thermal neutrons, while the gauging used filtered beams of 24 keV and fission spectrum neutrons. For the 83-mm-thick polyurethane and 130-mm-thick steel matrix, the thermal-neutron radiography was able to detect only those voids equal to about one-half the polyurethane thickness. The gauging detected voids in the path of the neutron beam of a few millimetres thickness in seconds to minutes. The gauging is feasible as an NDT method for the canister application

  18. Transcutaneous sacral neurostimulation for irritative voiding dysfunction.

    Science.gov (United States)

    Walsh, I K; Johnston, R S; Keane, P F

    1999-01-01

    Patients with irritative voiding dysfunction are often unresponsive to standard clinical treatment. We evaluated the response of such individuals to transcutaneous electrical stimulation of the third sacral nerve. 32 patients with refractory irritative voiding dysfunction (31 female and 1 male; mean age 47 years) were recruited to the study. Ambulatory transcutaneous electrical neurostimulation was applied bilaterally to the third sacral dermatomes for 1 week. Symptoms of frequency, nocturia, urgency, and bladder pain were scored by each patient throughout and up to 6 months following treatment. The mean daytime frequency was reduced from 11.3 to 7.96 (p = 0.01). Nocturia episodes were reduced from a mean of 2.6 to 1.8 (p = 0.01). Urgency and bladder pain mean symptom scores were reduced from 5.97 to 4.89 and from 1.48 to 0.64, respectively. After stopping therapy, symptoms returned to pretreatment levels within 2 weeks in 40% of the patients and within 6 months in 100%. Three patients who continued with neurostimulation remained satisfied with this treatment modality at 6 months. Transcutaneous third sacral nerve stimulation may be an effective and noninvasive ambulatory technique for the treatment of patients with refractory irritative voiding dysfunction. Following an initial response, patients may successfully apply this treatment themselves to ensure long-term relief.

  19. Void fraction measurements using neutron radiography

    International Nuclear Information System (INIS)

    Glickstein, S.S.; Vance, W.H.; Joo, H.

    1992-01-01

    Real-time neutron radiography is being evaluated for studying the dynamic behavior of two phase flow and for measuring void fraction in vertical and inclined water ducts. This technique provides a unique means of visualizing the behavior of fluid flow inside thick metal enclosures. To simulate vapor conditions encountered in a fluid flow duct, an air-water flow system was constructed. Air was injected into the bottom of the duct at flow rates up to 0.47 I/s (1 cfm). The water flow rate was varied between 0--3.78 I/m (0--1 gpm). The experiments were performed at the Pennsylvania State University nuclear reactor facility using a real-time neutron radiography camera. With a thermal neutron flux on the order of 10 6 n/cm 2 /s directed through the thin duct dimension, the dynamic behavior of the air bubbles was clearly visible through 5 cm (2 in.) thick aluminum support plates placed on both sides of the duct wall. Image analysis techniques were employed to extract void fractions from the data which was recorded on videotape. This consisted of time averaging 256 video frames and measuring the gray level distribution throughout the region. The distribution of the measured void fraction across the duct was determined for various air/water mixtures. Details of the results of experiments for a variety of air and water flow conditions are presented

  20. Vesicoureteral reflux in children: comparison of contrast - enhanced voiding ultrasonography with radiographic voiding cystourethrography - preliminary report

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Chong Hyun; Kim, Hyun Joo; Goo, Hyun Woo; Kim, Hungy; Lee, Jung Joo; Kim, Ellen Ai-Rhan; Kim, Ki Soo; Park, Young Seo; Pi, Soo Young [Ulsan Univ. College of Medicine, Seoul (Korea, Republic of)

    2001-01-01

    To compared the usefullness of contrst-enhanced voiding ultrasonogrphy (US) with that of radiogrphic voiding cystourethrography (VCUG) for the diagnosis of vesicoureteral reflux (VUR) in children. Ninety-five kidney-ureter units of 47 patients referred for investigation of VUR underwent contrast -enhanced voiding US followed by radiographic VCUG. After baseline US examination of the urinaru tract, residual urine in the bladder was drained through an inserted Foley catheter and the bladder was gravityfilled at a height of 1 m with normal saline. A galactose-based, microbubble-containning echo-enhancing agent (Lvovist; Dchering, Berlin, Germany) was then administered. The amount of this was approximately 10% of bldder capacity, and VUR was diagnosed when microbubbles appeared in the ureter or pelvocalyceal system. Using radiographic VCUG as a reference point, the accuracy with which contrst-enhanced voiding US detected VUR was calcilated. In 87 of 95 kidney-ureter units (91.6%), the two methods showed similiar results regarding the diagnosis or exclusion of VUR, which was detected by both in 12 units, but by neither in 75. VUR was shown to occcur in a total of 20 units, but in eight of these by one method only. In two units, VUR detected by contrast-enhanced voiding US was was not demostarted by radiographic VCUG; in six units, the resverse was true. In the detection of VUR, contrast-enhanced voiding us showed a sensitivity of 66.7%, a sprcificity of 97.4%, a positive predictive value of 85.7%, and a negative predictive value of 92.6%. Contrst-enhanced voiding US is highly specific and has high positive and nagative predictive values; its sensitivity, however, is not sufficiently high. The modality appears to be a useful diagnostic tool for the detection of VUR without exposure to ionizing radiation, though to be certain of its value, more experience of its use its first required.

  1. Vesicoureteral reflux in children: comparison of contrast - enhanced voiding ultrasonography with radiographic voiding cystourethrography - preliminary report

    International Nuclear Information System (INIS)

    Yoon, Chong Hyun; Kim, Hyun Joo; Goo, Hyun Woo; Kim, Hungy; Lee, Jung Joo; Kim, Ellen Ai-Rhan; Kim, Ki Soo; Park, Young Seo; Pi, Soo Young

    2001-01-01

    To compared the usefullness of contrst-enhanced voiding ultrasonogrphy (US) with that of radiogrphic voiding cystourethrography (VCUG) for the diagnosis of vesicoureteral reflux (VUR) in children. Ninety-five kidney-ureter units of 47 patients referred for investigation of VUR underwent contrast -enhanced voiding US followed by radiographic VCUG. After baseline US examination of the urinaru tract, residual urine in the bladder was drained through an inserted Foley catheter and the bladder was gravityfilled at a height of 1 m with normal saline. A galactose-based, microbubble-containning echo-enhancing agent (Lvovist; Dchering, Berlin, Germany) was then administered. The amount of this was approximately 10% of bldder capacity, and VUR was diagnosed when microbubbles appeared in the ureter or pelvocalyceal system. Using radiographic VCUG as a reference point, the accuracy with which contrst-enhanced voiding US detected VUR was calcilated. In 87 of 95 kidney-ureter units (91.6%), the two methods showed similiar results regarding the diagnosis or exclusion of VUR, which was detected by both in 12 units, but by neither in 75. VUR was shown to occcur in a total of 20 units, but in eight of these by one method only. In two units, VUR detected by contrast-enhanced voiding US was was not demostarted by radiographic VCUG; in six units, the resverse was true. In the detection of VUR, contrast-enhanced voiding us showed a sensitivity of 66.7%, a sprcificity of 97.4%, a positive predictive value of 85.7%, and a negative predictive value of 92.6%. Contrst-enhanced voiding US is highly specific and has high positive and nagative predictive values; its sensitivity, however, is not sufficiently high. The modality appears to be a useful diagnostic tool for the detection of VUR without exposure to ionizing radiation, though to be certain of its value, more experience of its use its first required

  2. High-resolution simulations of cylindrical void collapse in energetic materials: Effect of primary and secondary collapse on initiation thresholds

    Science.gov (United States)

    Rai, Nirmal Kumar; Schmidt, Martin J.; Udaykumar, H. S.

    2017-04-01

    Void collapse in energetic materials leads to hot spot formation and enhanced sensitivity. Much recent work has been directed towards simulation of collapse-generated reactive hot spots. The resolution of voids in calculations to date has varied as have the resulting predictions of hot spot intensity. Here we determine the required resolution for reliable cylindrical void collapse calculations leading to initiation of chemical reactions. High-resolution simulations of collapse provide new insights into the mechanism of hot spot generation. It is found that initiation can occur in two different modes depending on the loading intensity: Either the initiation occurs due to jet impact at the first collapse instant or it can occur at secondary lobes at the periphery of the collapsed void. A key observation is that secondary lobe collapse leads to large local temperatures that initiate reactions. This is due to a combination of a strong blast wave from the site of primary void collapse and strong colliding jets and vortical flows generated during the collapse of the secondary lobes. The secondary lobe collapse results in a significant lowering of the predicted threshold for ignition of the energetic material. The results suggest that mesoscale simulations of void fields may suffer from significant uncertainty in threshold predictions because unresolved calculations cannot capture the secondary lobe collapse phenomenon. The implications of this uncertainty for mesoscale simulations are discussed in this paper.

  3. A core performance study on an actinide recycling 'zero-sodium-void worth' core

    International Nuclear Information System (INIS)

    Kawashima, M.; Nakagawa, M.; Yamaoka, M.; Kasahara, F.

    1994-01-01

    A core performance study was made for an absorber-type parfait core (A-APC) as one of 'Zero-sodium-void-worth' core concepts. This evaluation study pursued different two aspects; one for transuranic (TRU) management strategy, and another for a loss-of-coolant anticipated transient behavior considering the unique core configuration. The results indicated that this core has a large flexibility for actinide recycling in terms of self-sufficiency and minor actinide burning. The result also showed that this core has kept a large mitigation potential for ULOF events as well as a simple flat core concept, reflecting detailed three dimensional core bowing behavior for the A-APC configuration. (author)

  4. Freeform Deposition Method for Coolant Channel Closeout

    Science.gov (United States)

    Gradl, Paul R. (Inventor); Reynolds, David Christopher (Inventor); Walker, Bryant H. (Inventor)

    2017-01-01

    A method is provided for fabricating a coolant channel closeout jacket on a structure having coolant channels formed in an outer surface thereof. A line of tangency relative to the outer surface is defined for each point on the outer surface. Linear rows of a metal feedstock are directed towards and deposited on the outer surface of the structure as a beam of weld energy is directed to the metal feedstock so-deposited. A first angle between the metal feedstock so-directed and the line of tangency is maintained in a range of 20-90.degree.. The beam is directed towards a portion of the linear rows such that less than 30% of the cross-sectional area of the beam impinges on a currently-deposited one of the linear rows. A second angle between the beam and the line of tangency is maintained in a range of 5-65 degrees.

  5. CAREM-25: considerations about primary coolant chemistry

    International Nuclear Information System (INIS)

    Chocron, Mauricio; Iglesias, Alberto M.; Raffo Calderon, Maria C.; Villegas, Marina

    2000-01-01

    World operating experience, in conjunction with basic studies has been modifying chemistry specifications for the primary coolant of water cooled nuclear reactors along with the reactor type and structural materials involved in the design. For the reactor CAREM-25, the following sources of information have been used: 1) Experience gained by the Chemistry Department of the National Atomic Energy Commission (CNEA, Argentina); 2) Participation of the Chemistry Department (CNEA) in international cooperation projects; 3) Guidelines given by EPRI, Siemens-KWU, AECL, etc. Given the main objectives: materials integrity, low radiation levels and personnel safety, which are in turn a balance between the lowest corrosion and activity transport achievable and considering that the CAREM-25 is a pressurized vessel integrated reactor, a group of guidelines for the chemistry and additives for the primary coolant have been given in the present work. (author)

  6. Recovery studies for plutonium machining oil coolant

    International Nuclear Information System (INIS)

    Navratil, J.D.; Baldwin, C.E.

    1977-01-01

    Lathe coolant oil, contaminated with plutonium and having a carbon tetrachloride diluent, is generated in plutonium machining areas at Rocky Flats. A research program was initiated to determine the nature of plutonium in this mixture of oil and carbon tetrachloride. Appropriate methods then could be developed to remove the plutonium and to recycle the oil and carbon tetrachloride. Studies showed that the mixtures of spent oil and carbon tetrachloride contained particulate plutonium and plutonium species that are soluble in water or in oil and carbon tetrachloride. The particulate plutonium was removed by filtration; the nonfilterable plutonium was removed by adsorption on various materials. Laboratory-scale tests indicated the lathe-coolant oil mixture could be separated by distilling the carbon tetrachloride to yield recyclable products

  7. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1980-01-01

    This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the 13 N content in the containment atmosphere. 13 N is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/ 13 N+ 4 He. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium 13 N concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

  8. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1980-01-01

    This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/Nl3+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

  9. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1979-08-01

    The present paper deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process H1+016 → N13+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m -3 and 7 kBq m -3 for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge (Li) flow detector assembly operated at elevated pressure. (Auth.)

  10. Measurements of void fraction by an improved multi-channel conductance void meter

    International Nuclear Information System (INIS)

    Song, Chul-Hwa; Chung, Moon Ki; No, Hee Cheon

    1998-01-01

    An improved multi-channel Conductance Void Meter (CVM) was developed to measure a void fraction. Its measuring principle is basically based upon the differences of electrical conductance of a two-phase mixture due to the variation of void fraction around a sensor. The sensor is designed to be flush-mounted to the inner wall of the test section to avoid the flow disturbances. The signal processor with three channels is specially designed so as to minimize the inherent error due to the phase difference between channels. It is emphasized that the guard electrodes are electrically shielded in order not to affect the measurements of two-phase mixture conductance, but to make the electric fields evenly distributed in a measuring volume. Void fraction is measured for bubbly and slug flow regimes in a vertical air-water loop, and statistical signal processing techniques are applied to show that CVM has a good dynamic resolution which is required to investigate the structural developments of bubbly flow and the propagation of void waves in a flow channel. (author)

  11. Void growth to coalescence in a non-local material

    DEFF Research Database (Denmark)

    Niordson, Christian Frithiof

    2008-01-01

    of different material length parameters in a multi-parameter theory is studied, and it is shown that the important length parameter is the same as under purely hydrostatic loading. It is quantified how micron scale voids grow less rapidly than larger voids, and the implications of this in the overall strength...... of the material is emphasized. The size effect on the onset of coalescence is studied, and results for the void volume fraction and the strain at the onset of coalescence are presented. It is concluded that for cracked specimens not only the void volume fraction, but also the typical void size is of importance...... to the fracture strength of ductile materials....

  12. Reactor coolant pump seal leakage monitoring

    International Nuclear Information System (INIS)

    Stevens, D.M.; Spencer, J.W.; Morris, D.J.; James, W.; Shugars, H.G.

    1986-01-01

    Problems with reactor coolant pump seals have historically accounted for a large percentage of unscheduled outages. Studies performed for the Electric Power Research Institute (EPRI) have shown that the replacement of coolant pump seals has been one of the leading causes of nuclear plant unavailability over the last ten years. Failures of coolant pump seals can lead to primary coolant leakage rates of 200-500 gallons per minute into the reactor building. Airborne activity and high surface contamination levels following these failures require a major cleanup effort and increases the time and personnel exposure required to refurbish the pump seals. One of the problems in assessing seal integrity is the inability to accurately measure seal leakage. Because seal leakage flow is normally very small, it cannot be sensed directly with normal flow instrumentation, but must be inferred from several other temperature and flow measurements. In operating plants the leakage rate has been quantified with a tipping-bucket gauge, a device which indicates when one quart of water has been accumulated. The tipping-bucket gauge has been used for most rainfall-intensity monitoring. The need for a more accurate and less expensive gauge has been addressed. They have developed a drop-counter precipitation sensor has been developed and optimized. The applicability of the drop-counter device to the problem of measuring seal leakage is being investigated. If a review of system specification and known drop-counter performance indicates that this method is feasible for measuring seal leak rates, a drop-counter gauge will be fabricated and tested in the laboratory. If laboratory tests are successful the gauge will be demonstrated in a pump test loop at Ontario Hydro and evaluated under simulated plant conditions. 3 references, 2 figures

  13. Enhancing resistance to burnout via coolant chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Tu, J. P.; Dinh, T. N.; Theofanous, T. G. [Univ. of California, Santa Barbara (United States)

    2003-07-01

    Boiling Crisis (BC) on horizontal, upwards-facing copper and steel surfaces under the influence of various coolant chemistries relevant to reactor containment waters is considered. In addition to Boric Acid (BA) and TriSodium Phosphate (TSP), pure De-Ionized Water (DIW) and Tap Water (TW) are included in experiments carried out in the BETA facility. The results are related to a companion paper on the large scale ULPU facility.

  14. Minimizing secondary coolant blowdown in HANARO

    International Nuclear Information System (INIS)

    Park, Y. C.; Woo, J. S.; Ryu, J. S.; Cho, Y. G.; Lim, N. Y.

    2000-01-01

    There is about 80m 3 /h loss of the secondary cooling water by evaporation, windage and blowdown during the operation of HANARO, 30MW research reactor. The evaporation and the windage is necessary loss to maintain the performance of cooling tower, but the blowdown is artificial lose to get rid of the foreign material and to maintain the quality of the secondary cooling water. Therefore, minimizing the blowdown loss was studied. It was confirmed, through the relation of the number of cycle and the loss rate of secondary coolant, that the number of cycle is saturated to 12 without blowdown because of the windage loss. When the secondary coolant is treated by high Ca-hardness treatment program (the number of cycle > 10) to maintain the number of cycle around 12 without blowdown, only the turbidity exceeds the limit. By adding filtering system it was confirmed, through the relation of turbidity and filtering rate of secondary cooling water, that the turbidity is reduced below the limit (5 deg.) by 2% of filtering rate without blowdown. And it was verified, through the performance test of back-flow filtering unit, that this unit gets rid of foreign material up to 95% of the back-flow and that the water can be reused as coolant. Therefore, the secondary cooling water can be treated by the high Ca-hardness program and filter system without blowdown

  15. Reactor coolant pumps for nuclear reactors

    International Nuclear Information System (INIS)

    Harand, E.; Richter, G.; Tschoepel, G.

    1975-01-01

    A brake for the pump rotor of a main coolant pump or a shutoff member on the pump are provided in order to prevent excess speeds of the pump rotor. Such excess speeds may occur in PWR type reactors with water at a pressure below, e.g., 150 bars if there is leakage from a coolant line associated with the main coolant pump. As a brake, a centrifugal brake depending upon the pump speed or a brake ring arranged on the pump housing and acting on the pump rotor, which ring would be activated by pressure differentials in the pump, may be used. If the pressure differences between suction and pressure sockets are very small, a controlled hydraulic increase of the pressure force on the brake may also be provided. Furthermore, a turbine brake may be provided. A slide which is automatically movable in closing position along the pump rotor axis is used as a shutoff element. It is of cylindrical configuration and is arranged concentrically with the rotor axis. (DG) [de

  16. Design of automotive engine coolant hoses

    Directory of Open Access Journals (Sweden)

    Hrishikesh D BACHCHHAV

    2018-03-01

    Full Text Available In this paper, we are present the performance of engine coolant hoses (radiator hoses used in passenger cars by checking various physical behaviours such as hose leakage, hose burst, hose collapse or any mechanical damage as studied-thru design guidelines, CFD analysis and product validation testing and also check pressure drop of the hoses when engine will be running. The design term is more likely used for technical part modelling using CAD tool. Later on, we will focus on the transformation of the part design to process design. The process design term is more likely used for "tooling design" for manufacturing of the product using CAD Tool. Then inlet hose carries coolant from engine to radiator inlet tank, then coolant circulated in radiator and passed through radiator outlet tank to water pump of engine with the help of outlet hose. After that …nding any leakage, Burst, damage or collapse of hose and pressure drop of the hose with the help of design checklist, CFD Analysis and product validation testing.

  17. A method of simulating voids in experimental studies of boiling water reactors

    International Nuclear Information System (INIS)

    Down, H.J.; Dickie, J.; Fox, W.N.

    1963-11-01

    The coolant density in boiling water reactors may vary from 3 at pressures up to 1000 p.s.i. In order to study the effect of reduced water density on reactivity in unpressurized experimental systems, the effective water density is reduced by packing small beads of highly expanded polystyrene into the fuel clusters and flooding the interstices with water. Coolant densities of from 0.4 to 0.6 gm/cm 3 may be produced with the introduction of only about 0.4 gm/cm 3 of non-hydrogeneous material. This memorandum describes the production, properties and handling of polystyrene beads and the tests carried out to establish the validity of the technique. (author)

  18. Voiding dysfunction in children aged five to 15 years

    Directory of Open Access Journals (Sweden)

    Karaklajić Dragana

    2004-01-01

    Full Text Available Voiding dysfunction in children was analyzed in 91 patients in a period from January 1st to October 1st 1998. Most of the patients had functional voiding disorder (92.31%, and only 7.69% manifested monosymptomatic night enuresis. The number of girls was bigger in the group of patients with voiding dysfunction while the boys were predominant in the group with mono-symptomatic nocturnal enuresis. More than a half of children with functional voiding disorder had repeated urinal infections (58.23%, incontinence (93.49%, need for urgent voiding (68.13%, and vesicoureteral reflux (47.61%. The most common type of voiding dysfunction was urge syndrome/urge incontinence. The incidence of dysfunctional voiding disorder was more often in children with scaring changes of kidney which were diagnosed by static scintigraphy.

  19. On the observability of coupled dark energy with cosmic voids

    Science.gov (United States)

    Sutter, P. M.; Carlesi, Edoardo; Wandelt, Benjamin D.; Knebe, Alexander

    2015-01-01

    Taking N-body simulations with volumes and particle densities tuned to match the sloan digital sky survey DR7 spectroscopic main sample, we assess the ability of current void catalogues to distinguish a model of coupled dark matter-dark energy from Λ cold dark matter cosmology using properties of cosmic voids. Identifying voids with the VIDE toolkit, we find no statistically significant differences in the ellipticities, but find that coupling produces a population of significantly larger voids, possibly explaining the recent result of Tavasoli et al. In addition, we use the universal density profile of Hamaus et al. to quantify the relationship between coupling and density profile shape, finding that the coupling produces broader, shallower, undercompensated profiles for large voids by thinning the walls between adjacent medium-scale voids. We find that these differences are potentially measurable with existing void catalogues once effects from survey geometries and peculiar velocities are taken into account.

  20. Void growth and coalescence in metals deformed at elevated temperature

    DEFF Research Database (Denmark)

    Klöcker, H.; Tvergaard, Viggo

    2000-01-01

    For metals deformed at elevated temperatures the growth of voids to coalescence is studied numerically. The voids are assumed to be present from the beginning of deformation, and the rate of deformation considered is so high that void growth is dominated by power law creep of the material, without...... any noticeable effect of surface diffusion. Axisymmetric unit cell model computations are used to study void growth in a material containing a periodic array of voids, and the onset of the coalescence process is defined as the stage where plastic flow localizes in the ligaments between neighbouring...... voids. The focus of the study is on various relatively high stress triaxialties. In order to represent the results in terms of a porous ductile material model a set of constitutive relations are used, which have been proposed for void growth in a material undergoing power law creep....

  1. NABUB a non-saturated model of coolant boiling in a fast reactor sub-assembly

    International Nuclear Information System (INIS)

    Brook, A.J.; Mills, D.S.

    1975-08-01

    A theoretical model is described of sodium boiling in a fast reactor sub-assembly in which the usual assumptions of a saturated vapour are not made. Instead, vapour pressure is calculated in a perfect gas basis, which enables some allowance to be made for the possible presence of non-condensables, which may inhibit the condensation f the vapour. Indications are given of the circumstances under which such inhibition might be expected to show the most marked effects, and some sample results ontained by the code are presented. These show that the coolant voiding pattern is most sensitive to restrictions on the condensing flux in the 100 to 200w/cm 2 range. If unrestricted condensation is assumed, the results of the code are in excellent agreement with more conventional saturation models. (author)

  2. Development of Coolant Radioactivity Interpretation Code

    International Nuclear Information System (INIS)

    Kim, Kiyoung; Jung, Youngsuk; Kim, Kyounghyun; Kim, Jangwook

    2013-01-01

    In Korea, the coolant radioactivity analysis has been performed by using the computer codes of foreign companies such as CADE (Westinghouse), IODYNE and CESIUM (ABB-CE). However, these computer codes are too conservative and have involved considerable errors. Furthermore, since these codes are DOS-based program, their easy operability is not satisfactory. Therefore it is required development of an enhanced analysis algorithm applying an analytical method reflecting the change of operational environments of domestic nuclear power plants and a fuel failure evaluation software considering user' conveniences. We have developed a nuclear fuel failure evaluation code able to estimate the number of failed fuel rods and the burn-up of failed fuels during nuclear power plant operation cycle. A Coolant Radio-activity Interpretation Code (CRIC) for LWR has been developed as the output of the project 'Development of Fuel Reliability Enhanced Technique' organized by Korea Institute of Energy Technology Evaluation and Planning (KETEP). The CRIC is Windows based-software able to evaluate the number of failed fuel rods and the burn-up of failed fuel region by analyzing coolant radioactivity of LWR in operation. The CRIC is based on the model of fission products release commonly known as 'three region model' (pellet region, gap region, and coolant region), and we are verifying the CRIC results based on the cases of domestic fuel failures. CRIC users are able to estimate the number of failed fuel rods, burn-up and regions of failed fuel considered enrichment and power distribution of fuel region by using operational cycle data, coolant activity data, fuel loading pattern, Cs-134/Cs-137 ratio according to burn-up and U-235 enrichment provided in the code. Due to development of the CRIC, it is secured own unique fuel failure evaluation code. And, it is expected to have the following significant meaning. This is that the code reflecting a proprietary technique for quantitatively

  3. Patient dose reduction during voiding cystourethrography

    International Nuclear Information System (INIS)

    Ward, Valerie L.

    2006-01-01

    Voiding cystourethrography (VCUG) is a commonly performed examination in a pediatric uroradiology practice. This article contains suggestions on how the radiation dose to a child from VCUG can be made ''as low as reasonably achievable'' (ALARA). The pediatric radiologist should consider the appropriateness of the clinical indication before performing VCUG and utilize radiation exposure techniques and parameters during VCUG to reduce radiation exposure to a child. The medical physicist and fluoroscope manufacturer can also work together to optimize a pulsed-fluoroscopy unit and further reduce the radiation exposure. Laboratory and clinical research is necessary to investigate methods that reduce radiation exposures during VCUG, and current research is presented here. (orig.)

  4. Small-break loss-of-coolant accidents in the updated PIUS 600 advanced reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Boyack, B.E.; Steiner, J.L.; Harmony, S.C. [Los Alamos National Lab., Albuquerque, NM (United States)] [and others

    1995-09-01

    The PIUS advanced reactor is a 640-MWe pressurized water reactor developed by Asea Brown Boveri (ABB). A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity is normally controlled by coolant boron concentration and the temperature of the moderator coolant. ABB submitted the PIUS design to the US Nuclear Regulatory Commission (NRC) for preapplication review, and Los Alamos supported the NRC`s review effort. Baseline analyses of small-break initiators at two locations were performed with the system neutronic and thermal-hydraulic analysis code TRAC-PF1/MOD2. In addition, sensitivity studies were performed to explore the robustness of the PIUS concept to severe off-normal conditions having a very low probability of occurrence.

  5. Experimental and numerical investigation of the coolant mixing during fast deboration transients

    International Nuclear Information System (INIS)

    Hoehne, T.; Rohde, U.; Weiss, F.P.

    1999-01-01

    For the analysis of boron dilution transients and main steam line break scenarios the modeling of the coolant mixing inside the reactor vessel is important, because the reactivity insertion strongly depends on boron acid concentration or the coolant temperature distribution. Calculations for steady state flow conditions for the WWER-440 were performed with a CFD code (CFX-4). For this calculation the RPV from the cold legs inlet through the downcomer, the lower plenum and the lower core support plate was nodulized in detail. The comparison with experimental data and analytical mixing model which is implemented in the neutron kinetic code DYN3D showed a good agreement for near-nominal conditions (all MCPs are running). The comparison between the CFD-results and the analytical model revealed differences for MSLB conditions[1]. (Authors)

  6. Sacral Herpes Zoster Associated with Voiding Dysfunction in a Young Patient with Scrub Typhus.

    Science.gov (United States)

    Hur, Jian

    2015-06-01

    When a patient presents with acute voiding dysfunction without a typical skin rash, it may be difficult to make a diagnosis of herpes zoster. Here, we present a case of scrub typhus in a 25-year-old man with the complication of urinary dysfunction. The patient complained of loss of urinary voiding sensation and constipation. After eight days, he had typical herpes zoster eruptions on the sacral dermatomes and hypalgesia of the S1-S5 dermatomes. No cases of dual infection with varicella zoster virus and Orientia tsutsugamushi were found in the literature. In the described case, scrub typhus probably induced sufficient stress to reactivate the varicella zoster virus. Early recognition of this problem is imperative for prompt and appropriate management, as misdiagnosis can lead to long-term urinary dysfunction. It is important that a diagnosis of herpes zoster be considered, especially in patients with sudden onset urinary retention.

  7. The module CCM for the simulation of the thermal-hydraulic situation within a coolant channel

    International Nuclear Information System (INIS)

    Hoeld, A.

    2000-01-01

    A coolant channel module (Cc) will be presented which aim is to simulate, in a very general way, the thermal-hydraulic behaviour of single- and two-phase fluids flowing along a heated (or cooled) vertical, inclined or horizontal coolant channel. It is based on a theoretical drift-flux supported 3-equation mixture-fluid model describing the steady state and transient behaviour of characteristic thermal-hydraulic parameters of a single- and two-phase flow within such a channel. The module can be applied as an element within an overall theoretical model for large and complex plant assemblies (PWR and BWR core channels, parallel channels in 3D cores, primary and secondary sides of different steam generators types etc.). The model refers to a general (basic) coolant channel (BC) which can consists of different flow regimes. The BC has thus to be subdivided accordingly into a number of subchannels (SC-s). All of them can belong, however, to only two types of SC-s (single-phase fluid with subcooled water or superheated steam or a two-phase flow regime). For both of them the possibility of variable entrance or outlet positions has to be considered. For discretization purposes the BC (and thus also the SC-s) have to be subdivided into a number of (BC and SC) nodes, discretizing thus the conservation equations for mass, energy and momentum along these nodes by applying a very general spatial procedure, namely a 'modified finite volume method'. A special quadratic polygon approximation method (PAX procedure) helps then to establish a connection between nodal boundary and mean nodal parameters. Considering their constitutive equations (among them an adequate drift-flux correlation package) yields finally a set of non-linear algebraic and non-linear ordinary differential equations for the characteristic parameters of each of these SC nodes (mass flow, pressure drop, coolant temperature and/or void fraction). Based on this theory a code package (CCM) could be established

  8. Morphological Segregation in the Surroundings of Cosmic Voids

    Energy Technology Data Exchange (ETDEWEB)

    Ricciardelli, Elena; Tamone, Amelie [Laboratoire d’Astrophysique, École Polytechnique Fédérale de Lausanne (EPFL), 1290 Sauverny (Switzerland); Cava, Antonio [Observatoire de Genève, Université de Genève, 51 Ch. des Maillettes, 1290 Versoix (Switzerland); Varela, Jesus, E-mail: elena.ricciardelli@epfl.ch [Centro de Estudios de Física del Cosmos de Aragón (CEFCA), Plaza San Juan 1, E-44001 Teruel (Spain)

    2017-09-01

    We explore the morphology of galaxies living in the proximity of cosmic voids, using a sample of voids identified in the Sloan Digital Sky Survey Data Release 7. At all stellar masses, void galaxies exhibit morphologies of a later type than galaxies in a control sample, which represent galaxies in an average density environment. We interpret this trend as a pure environmental effect, independent of the mass bias, due to a slower galaxy build-up in the rarefied regions of voids. We confirm previous findings about a clear segregation in galaxy morphology, with galaxies of a later type being found at smaller void-centric distances with respect to the early-type galaxies. We also show, for the first time, that the radius of the void has an impact on the evolutionary history of the galaxies that live within it or in its surroundings. In fact, an enhanced fraction of late-type galaxies is found in the proximity of voids larger than the median void radius. Likewise, an excess of early-type galaxies is observed within or around voids of a smaller size. A significant difference in galaxy properties in voids of different sizes is observed up to 2 R {sub void}, which we define as the region of influence of voids. The significance of this difference is greater than 3 σ for all the volume-complete samples considered here. The fraction of star-forming galaxies shows the same behavior as the late-type galaxies, but no significant difference in stellar mass is observed in the proximity of voids of different sizes.

  9. A NEW STATISTICAL PERSPECTIVE TO THE COSMIC VOID DISTRIBUTION

    International Nuclear Information System (INIS)

    Pycke, J-R; Russell, E.

    2016-01-01

    In this study, we obtain the size distribution of voids as a three-parameter redshift-independent log-normal void probability function (VPF) directly from the Cosmic Void Catalog (CVC). Although many statistical models of void distributions are based on the counts in randomly placed cells, the log-normal VPF that we obtain here is independent of the shape of the voids due to the parameter-free void finder of the CVC. We use three void populations drawn from the CVC generated by the Halo Occupation Distribution (HOD) Mocks, which are tuned to three mock SDSS samples to investigate the void distribution statistically and to investigate the effects of the environments on the size distribution. As a result, it is shown that void size distributions obtained from the HOD Mock samples are satisfied by the three-parameter log-normal distribution. In addition, we find that there may be a relation between the hierarchical formation, skewness, and kurtosis of the log-normal distribution for each catalog. We also show that the shape of the three-parameter distribution from the samples is strikingly similar to the galaxy log-normal mass distribution obtained from numerical studies. This similarity between void size and galaxy mass distributions may possibly indicate evidence of nonlinear mechanisms affecting both voids and galaxies, such as large-scale accretion and tidal effects. Considering the fact that in this study, all voids are generated by galaxy mocks and show hierarchical structures in different levels, it may be possible that the same nonlinear mechanisms of mass distribution affect the void size distribution.

  10. Fuel-coolant interactions: preliminary experiments on the effect of gases dissolved in the 'coolant'

    International Nuclear Information System (INIS)

    Asher, R.C.; Davies, D.; Jones, P.G.

    1976-12-01

    A simple apparatus has been used to study fuel-coolant interactions under reasonably well controlled conditions. Preliminary experiments have used water as the 'coolant' and molten tin at 800 0 C as the 'fuel' and have investigated how the violence of the interaction is affected by dissolving gases (oxygen, nitrogen, carbon dioxide and nitrous oxide) in the water. It was found that saturating the water with carbon dioxide or nitrous oxide completely suppresses the violent interaction. Experiments in which the concentrations of these gases were varied showed that a certain critical concentration was needed; below this concentration the dissolved gas has no significant effect but above it the suppression is

  11. User's guide to EPIC, a computer program to calculate the motion of fuel and coolant subsequent to pin failure in an LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Pizzica, P.A.; Garner, P.L.; Abramson, P.B.

    1979-10-01

    The computer code EPIC models fuel and coolant motion which results from internal fuel pin pressure (from fission gas or fuel vapor) and possibly from the generation of sodium vapor pressure in the coolant channel subsequent to pin failure in a liquid-metal fast breeder reactor. The EPIC model is restricted to conditions where fuel pin geometry is generally preserved and is not intended to treat the total disruption of the pin structure. The modeling includes the ejection of molten fuel from the pin into a coolant channel with any amount of voiding through a clad breach which may be of any length or which may extend with time. One-dimensional Eulerian hydrodynamics is used to treat the motion of fuel and fission gas inside a molten fuel cavity in the fuel pin as well as the mixture of two-phase sodium and fission gas in the coolant channel. Motion of fuel in the coolant channel is tracked with a type of particle-in-cell technique. EPIC is a Fortran-IV program requiring 400K bytes of storage on the IBM 370/195 computer. 21 refs., 2 figs.

  12. Development of small, fast reactor core designs using lead-based coolant

    International Nuclear Information System (INIS)

    Cahalan, J. E.; Hill, R. N.; Khalil, H. S.; Wade, D. C.

    1999-01-01

    A variety of small (100 MWe) fast reactor core designs are developed, these include compact configurations, long-lived (15-year fuel lifetime) cores, and derated, natural circulation designs. Trade studies are described which identify key core design issues for lead-based coolant systems. Performance parameters and reactivity feedback coefficients are compared for lead-bismuth eutectic (LBE) and sodium-cooled cores of consistent design. The results of these studies indicate that the superior neutron reflection capability of lead alloys reduces the enrichment and burnup swing compared to conventional sodium-cooled systems; however, the discharge fluence is significantly increased. The size requirement for long-lived systems is constrained by reactivity loss considerations, not fuel burnup or fluence limits. The derated lead-alloy cooled natural circulation cores require a core volume roughly eight times greater than conventional compact systems. In general, reactivity coefficients important for passive safety performance are less favorable for the larger, derated configurations

  13. EDF PWRs primary coolant purification strategies

    International Nuclear Information System (INIS)

    Gressier, Frederic; Mascarenhas, Darren; Taunier, Stephane; Le-Calvar, Marc; Bretelle, Jean-Luc; Ranchoux, Gilles

    2012-09-01

    In order to achieve a good physico-chemical quality of the primary coolant fluid, the primary water is continuously treated by the Chemical and Volume Control System (CVCS). This system is composed of a treatment chain containing filters and ion-exchange resins. In the EDF design, an upstream filter is placed before the resin so as to prevent it from being saturated with insoluble particles. Then, the fluid passes through several resin beds (up to 3 depending on the configuration) and again through a downstream filter that prevents resin fines dissemination into the reactor coolant. Much work has been conducted in the last 5 years on the homogenisation of products and usage on French EDF NPP primary coolant treatment, while taking into account the compromise between source term reduction, liquid and solid waste, and buying and disposal costs. Two national markets have been created, and two operational documents for chemists on site have been published: a filtration guideline and an ion-exchange resin guideline. Both documents give general information about the products used, how are they characterized and selected for national market (technical requirements, standards and tests), how they should be used and what are the change-out criteria. They are also periodically updated based on feedback from sites. The positive impact on resin and filter lifetime (extension of some, limitation of others), homogenisation of products and usage will be presented. Moreover, EDF is constantly in the process of improving the current purification methods, as well as researching the use of existing and novel technologies. In this field, recent experiments on short loading of resin during reactor shutdown has been tested on site with success. In addition, work is done on silica free filters, filter consumption and filter chemical release. An overview of these optimization methods will be given. (authors)

  14. Coolant degassing device for PWR type reactors

    International Nuclear Information System (INIS)

    Kita, Kaoru; Takezawa, Kazuaki; Minemoto, Masaki.

    1982-01-01

    Purpose: To efficiently decrease the rare gas concentration in primary coolants, as well as shorten the degassing time required for the periodical inspection in the waste gas processing system of a PWR type reactor. Constitution: Usual degassing method by supplying hydrogen or nitrogen to a volume control tank is replaced with a method of utilizing a degassing tower (method of flowing down processing liquid into the filled tower from above while uprising streams from the bottom of the tower thereby degassing the gases dissolved in the liquid into the steams). The degassing tower is combined with a hydrogen separator or hydrogen recombiner to constitute a waste gas processing system. (Ikeda, J.)

  15. Microstructural characterization of primary coolant pipe steel

    International Nuclear Information System (INIS)

    Miller, M.K.; Bentley, J.

    1986-01-01

    Atom probe field-ion microscopy, analytical electron microscopy, and optical microscopy have been used to investigate the changes that occur in the microstructure of cast CF 8 primary coolant pipe stainless steel after long term thermal aging. The cast duplex microstructure consisted of austenite with 15% delta-ferrite. Investigation of the aged material revealed that the ferrite spinodally decomposed into a fine scaled network of α and α'. A fine G-phase precipitate was also observed in the ferrite. The observed degradation in mechanical properties is probably a consequence of the spinodal decomposition in the ferrite

  16. Calorimetric and reactor coolant system flow uncertainty

    International Nuclear Information System (INIS)

    Bates, L.; McLean, T.

    1991-01-01

    This paper describes a methodology for the quantification of errors associated with the determination of a feedwater flow, secondary power, and Reactor Coolant System (RCS) flow used at the Trojan Nuclear Plant to ensure compliance with regulatory requirements. The sources of error in Plant indications and process measurement are identified and tracked, using examples, through the mathematical processes necessary to calculate the uncertainty in the RCS flow measurement. An error of approximately 1.4 percent is calculated for secondary power. This error results, along with the consideration of other errors, in an uncertainty of approximately 3 percent in the RCS flow determination

  17. Relationship between voided volume and the urge to void among patients with lower urinary tract symptoms.

    Science.gov (United States)

    Blaivas, Jerry G; Tsui, Johnson F; Amirian, Michael; Ranasinghe, Buddima; Weiss, Jeffrey P; Haukka, Jari; Tikkinen, Kari A O

    2014-12-01

    The aim of this study was to explore the relationship between voided volume (VV) and urge to void among patients with lower urinary tract symptoms. Consecutive adult patients (aged 23-90 years) were enrolled, and completed a 24 h bladder diary and the Urgency Perception Scale (UPS). Patients were categorized as urgency or non-urgency based on the Overactive Bladder Symptom Score. The relationship between UPS and VV (based on the bladder diary) was analyzed by Spearman's rho and proportional odds model. In total, 1265 micturitions were evaluated in 117 individuals (41 men, 76 women; 56 individuals in the urgency and 61 in the non-urgency group). The mean (± SD) VV and UPS were 192 ± 127 ml and 2.4 ± 1.2 ml in the urgency group and 173 ± 124 ml and 1.7 ± 1.1 ml in the non-urgency group, respectively. Spearman's rho (between UPS and VV) was 0.21 [95% confidence interval (CI) 0.13-029, p < 0.001] for the urgency group, 0.32 (95% CI 0.25-0.39, p < 0.001) for the non-urgency group, and 0.28 (95% CI 0.23-0.33, p < 0.001) for the total cohort. Urgency patients had higher UPS [odds ratio (OR) 3.1, 95% CI 2.5-3.8]. Overall, each additional 50 ml VV increased the odds of having a higher UPS with OR 1.2 (95% CI 1.2-1.3). The relationship between VV and UPS score was similar in both groups (p = 0.548 for interaction). Although urgency patients void with a higher UPS score, among both urgency and non-urgency patients there is only a weak correlation between VV and the urge to void. This suggests that there are factors other than VV that cause the urge to void.

  18. Effect of helium on void formation in nickel

    International Nuclear Information System (INIS)

    Brimhall, J.L.; Simonen, E.P.

    1977-01-01

    This study examines the influence of helium on void formation in self-ion irradiated nickel. Helium was injected either simultaneously with, or prior to, the self-ion bombardment. The void microstructure was characterized as a function of helium deposition rate and the total heavy-ion dose. In particular, at 575 0 C and 5 X 10 -3 displacements per atom per second the void density is found to be proportional to the helium deposition rate. The dose dependence of swelling is initially dominated by helium driven nucleation. The void density rapidly saturates after which swelling continues with increasing dose only from void growth. It is concluded that helium promotes void nucleation in nickel with either helium implantation technique, pre-injection or simultaneous injection. Qualitative differences, however, are recognized. (Auth.)

  19. Void migration, coalescence and swelling in fusion materials

    International Nuclear Information System (INIS)

    Cottrell, G.A.

    2003-01-01

    A recent analysis of the migration of voids and bubbles, produced in neutron irradiated fusion materials, is outlined. The migration, brought about by thermal hopping of atoms on the surface of a void, is normally a random Brownian motion but, in a temperature gradient, can be slightly biassed up the gradient. Two effects of such migrations are the transport of voids and trapped transmutation helium atoms to grain boundaries, where embrittlement may result; and the coalescence of migrating voids, which reduces the number of non-dislocation sites available for the capture of knock-on point defects and thereby enables the dislocation bias process to maintain void swelling. A selection of candidate fusion power plant armour and structural metals have been analysed. The metals most resistant to void migration and its effects are tungsten and molybdenum. Steel and beryllium are least so and vanadium is intermediate

  20. Nucleation and growth of voids by radiation. Pt. 2

    International Nuclear Information System (INIS)

    Mayer, R.M.; Brown, L.M.

    1980-01-01

    The original model of Brown, Kelly and Mayer [1] for the nucleation of interstitial loops has been extended to take into account the following: (i) mobility of the vacancies, (ii) generation and migration of gas atoms during irradiation, (iii) nucleation and growth of voids, and (iv) vacancy emission from voids and clusters at high temperatures. Using chemicalrate equations, additional expressions are formulated for the nucleation and growth of vacancy loops and voids. (orig.)

  1. Dislocation and void segregation in copper during neutron irradiation

    DEFF Research Database (Denmark)

    Singh, Bachu Narain; Leffers, Torben; Horsewell, Andy

    1986-01-01

    ); the irradiation experiments were carried out at 250 degree C. The irradiated specimens were examined by transmission electron microscopy. At both doses, the irradiation-induced structure was found to be highly segregated; the dislocation loops and segments were present in the form of irregular walls and the voids...... density, the void swelling rate was very high (approximately 2. 5% per dpa). The implications of the segregated distribution of sinks for void formation and growth are briefly discussed....

  2. Multirods burst tests under loss-of-coolant conditions

    International Nuclear Information System (INIS)

    Kawasaki, S.; Uetsuka, H.; Furuta, T.

    1983-01-01

    In order to know the upper limit of coolant flow area restriction in a fuel assembly under loss-of-coolant accidents in LWRs, burst tests of fuel bundles were performed. Each bundle consisted of 49 rods(7x7 rods), and bursts were conducted in flowing steam. In some cases, 4 rods were replaced by control rods with guide tubes in a bundle. After the burst, the ballooning behavior of each rod and the degree of coolant flow area restriction in the bundle were measured. Ballooning behavior of rods and degree of coolant flow channel restriction in bundles with control rods were not different from those without control rods. The upper limit of coolant flow channel restriction under loss-of-coolant conditions was estimated to be about 80%. (author)

  3. Reactor auxiliary cooling facility and coolant supplying method therefor

    Energy Technology Data Exchange (ETDEWEB)

    Ando, Koji; Kinoshita, Shoichiro

    1996-06-07

    A reactor auxiliary cooling facility of the present invention comprises a coolant recycling line for recycling coolants by way of a reactor auxiliary coolant pump and a cooling load, a gravitational surge tank for supplying coolants to the coolant recycling line and a supplemental water supplying line for supplying a supply the supplemental water to the tank. Then, a pressurization-type supply water surge tank is disposed for operating the coolant recycling line upon performing an initial system performance test in parallel with the gravitational surge tank. With such a constitution, the period of time required from the start of the installation of reactor auxiliary cooling facilities to the completion of the system performance test can be shortened at a reduced cost without enlarging the scale of the facility. (T.M.)

  4. Reactor auxiliary cooling facility and coolant supplying method therefor

    International Nuclear Information System (INIS)

    Ando, Koji; Kinoshita, Shoichiro.

    1996-01-01

    A reactor auxiliary cooling facility of the present invention comprises a coolant recycling line for recycling coolants by way of a reactor auxiliary coolant pump and a cooling load, a gravitational surge tank for supplying coolants to the coolant recycling line and a supplemental water supplying line for supplying a supply the supplemental water to the tank. Then, a pressurization-type supply water surge tank is disposed for operating the coolant recycling line upon performing an initial system performance test in parallel with the gravitational surge tank. With such a constitution, the period of time required from the start of the installation of reactor auxiliary cooling facilities to the completion of the system performance test can be shortened at a reduced cost without enlarging the scale of the facility. (T.M.)

  5. Effect of parameter variation of reactor coolant pump on loss of coolant accident consequence

    International Nuclear Information System (INIS)

    Dang Gaojian; Huang Daishun; Gao Yingxian; He Xiaoqiang

    2015-01-01

    In this paper, the analyses were carried out on Ling'ao nuclear power station phase II to study the consequence of the loss of coolant accident when the homologous characteristic curves and free volumes of the reactor coolant pump changed. Two different pumps used in the analysis were 100D (employed on Ling'ao nuclear power station phase II) and ANDRITZ. The thermal characteristics in the large break LOCA accident were analyzed using CATHRE GB and CONPATE4, and the reactor coolant system hydraulics load during blow-clown phase of LOCA accident was analyzed using ATHIS and FORCET. The calculated results show that the homologous characteristic curves have great effect on the thermal characteristics of reactor core during the reflood phase of the large break LOCA accident. The maximum cladding surface temperatures are quite different when the pump's homologous characteristic curves change. On the other hand, the pump's free volume changing results in the variation of the LOCA rarefaction wave propagation, and therefore, the reactor coolant system hydraulic load in LOCA accident would be different. (authors)

  6. Void growth to coalescence in a non-local material

    DEFF Research Database (Denmark)

    Niordson, Christian Frithiof

    of different material length parameters in a multi-parameter theory is studied, and it is shown that the important length parameter is the same as under purely hydrostatic loading. It is quantified how micron scale voids grow less rapidly than larger voids, and the implications of this in the overall strength...... of the material is emphasized. It is concluded that for cracked specimens not only the void volume fraction, but also the typical void size is of importance to the fracture strength of ductile materials....

  7. Void shrinkage in stainless steel during high energy electron irradiation

    International Nuclear Information System (INIS)

    Singh, B.N.; Foreman, A.J.E.

    1976-03-01

    During irradiation of thin foils of an austenitic stainless steel in a high voltage electron microscope, steadily growing voids have been observed to suddenly shrink and disappear at the irradiation temperature of 650 0 Cthe phenomenon has been observed in specimens both with and withoutimplanted helium. Possible mechanisms for void shrinkage during irradiation are considered. It is suggested that the dislocation-pipe-diffusion of vacancies from or of self-interstitial atoms to the voids can explain the shrinkage behaviour of voids observed during our experiments. (author)

  8. Natural circulation in reactor coolant system

    International Nuclear Information System (INIS)

    Han, J.T.

    1987-01-01

    Reactor coolant system (RCS) natural circulation in a PWR is the buoyancy-driven coolant circulation between the core and the upper-plenum region (in-vessel circulation) with or without a countercurrent flow in the hot leg piping between the vessel and steam generators (ex-vessel circulation). This kind of multidimensional bouyancy-driven flow circulation serves as a means of transferring the heat from the core to the structures in the upper plenum, hot legs, and possibly steam generators. As a result, the RCS piping and other pressure boundaries may be heated to high temperatures at which the structural integrity is challenged. RCS natural circulation is likely to occur during the core uncovery period of the TMLB' accident in a PWR when the vessel upper plenum and hot leg are already drained and filled with steam and possibly other gaseous species. RCS natural circulation is being studied for the Surry plant during the TMLB' accident in which station blackout coincides with the loss of auxiliary feedwater and no operator actions. The effects of the multidimensional RCS natural circulation during the TMLB' accident are discussed

  9. CFD analyses of coolant channel flowfields

    Science.gov (United States)

    Yagley, Jennifer A.; Feng, Jinzhang; Merkle, Charles L.

    1993-01-01

    The flowfield characteristics in rocket engine coolant channels are analyzed by means of a numerical model. The channels are characterized by large length to diameter ratios, high Reynolds numbers, and asymmetrical heating. At representative flow conditions, the channel length is approximately twice the hydraulic entrance length so that fully developed conditions would be reached for a constant property fluid. For the supercritical hydrogen that is used as the coolant, the strong property variations create significant secondary flows in the cross-plane which have a major influence on the flow and the resulting heat transfer. Comparison of constant and variable property solutions show substantial differences. In addition, the property variations prevent fully developed flow. The density variation accelerates the fluid in the channels increasing the pressure drop without an accompanying increase in heat flux. Analyses of the inlet configuration suggest that side entry from a manifold can affect the development of the velocity profile because of vortices generated as the flow enters the channel. Current work is focused on studying the effects of channel bifurcation on the flow field and the heat transfer characteristics.

  10. Efficiency of water coolant for DEMO divertor

    International Nuclear Information System (INIS)

    Fetzer, Renate; Igitkhanov, Yuri; Bazylev, Boris

    2015-01-01

    Up to now, water-cooled divertor concepts have been developed for limited incident fluxes without taking into account transient power loadings. In this paper we analyzed the efficiency of water as a coolant for the particular PFC tungsten monoblock shield with a cooling tube made from Cu alloy (Cu OFHC) as a laminate adjacent to W and a low activation martensitic steel (Eurofer) as inner tube contacting the coolant. Thermal analysis is carried out by using the code MEMOS, which simulates W armour damage under the repetitive ELM heat loads. We consider cooling conditions which allow one to keep relatively high material temperatures (in the range 300–600 °C) thus minimizing Eurofer embrittlement under neutron irradiation. Expected DEMO I and DEMO II heat loads including type I ELMs are found to cause melting of the W surface during unmitigated ELMs. By mitigation of the ELMs melting of W is avoided. DEMO I operation under these conditions is save for cooling at water pressure 15.5 MPa and temperature 325 °C, while for DEMO II with mitigated ELMs the critical heat flux is exceeded and safe operation is not provided.

  11. Chemistry of liquid metal coolants and sensors

    International Nuclear Information System (INIS)

    Gnanasekaran, T.

    2015-01-01

    Liquid sodium is the coolant of choice for the current generation fast breeder reactors. When sodium contains low levels of dissolved non-metallic impurities, it is highly compatible with structural steels. When the dissolved oxygen level is high, corrosion and mass transfer in sodium-steel circuits are enhanced and this involves formation of NaxMyOz type of species (M = alloying components in steels). Experience has shown that this enhancement of corrosion in a sodium circuit with all austenitic steel structural materials would not be encountered if oxygen level in sodium is below ~ 5ppm. For understanding this observation, a complete knowledge on the phase diagrams of Na-M-O systems and the thermochemical data of all relevant NaxMyOz compounds is essential. This presentation would highlight the work carried out at IGCAR on the chemistry of liquid sodium and heavy liquid metal coolants. Work carried out on various sensors for their use in these liquid metal circuits would be described and their current status would be discussed

  12. Efficiency of water coolant for DEMO divertor

    Energy Technology Data Exchange (ETDEWEB)

    Fetzer, Renate, E-mail: renate.fetzer@kit.edu; Igitkhanov, Yuri; Bazylev, Boris

    2015-10-15

    Up to now, water-cooled divertor concepts have been developed for limited incident fluxes without taking into account transient power loadings. In this paper we analyzed the efficiency of water as a coolant for the particular PFC tungsten monoblock shield with a cooling tube made from Cu alloy (Cu OFHC) as a laminate adjacent to W and a low activation martensitic steel (Eurofer) as inner tube contacting the coolant. Thermal analysis is carried out by using the code MEMOS, which simulates W armour damage under the repetitive ELM heat loads. We consider cooling conditions which allow one to keep relatively high material temperatures (in the range 300–600 °C) thus minimizing Eurofer embrittlement under neutron irradiation. Expected DEMO I and DEMO II heat loads including type I ELMs are found to cause melting of the W surface during unmitigated ELMs. By mitigation of the ELMs melting of W is avoided. DEMO I operation under these conditions is save for cooling at water pressure 15.5 MPa and temperature 325 °C, while for DEMO II with mitigated ELMs the critical heat flux is exceeded and safe operation is not provided.

  13. Requirements of coolants in nuclear reactors

    International Nuclear Information System (INIS)

    Abass, O. A. M.

    2014-11-01

    This study discussed the purposes and types of coolants in nuclear reactors to generate electricity. The major systems and components associated with nuclear reactors are cooling system. There are two major cooling systems utilized to convert the heat generated in the fuel into electrical power. The primary system transfers the heat from the fuel to the steam generator, where the secondary system begins. The steam formed in the steam generator is transferred by the secondary system to the main turbine generator, where it s converted into electricity after passing through the low pressure turbine. There are various coolants used in nuclear reactors-light water, heavy water and liquid metal. The two major types of water-cooled reactors are pressurized water reactors (PWR) and boiling water reactors (BWR) but pressurized water reactors are more in the world. Also discusses this study the reactors and impact of the major nuclear accidents, in the April 1986 disaster at the Chernobyl nuclear power plant in Ukraine was the product operators, and in the March 2011 at the Fukushima nuclear power plant in Japan was the product of earthquake of magnitude 9.0, the accidents caused the largest uncontrolled radioactive release into the environment.(Author)

  14. Upper internals of PWR with coolant flow separator

    International Nuclear Information System (INIS)

    Chevereau, G.; Heuze, A.

    1989-01-01

    The upper internals for a PWR has a collecting volume for the coolant merging from the core and an apparatus for separating the flow of coolant. This apparatus has a guide for the control rods, a lower plate perforated to allow the coolant through from the core, an upper plate also perforated to allow the coolant through to the collecting volume and a peripheral binding ring joining the two plates. Each guide comprises an envelope without holes and joined perceptibly tight to the plates [fr

  15. Coolant make-up device for BWR type reactor

    International Nuclear Information System (INIS)

    Sasagawa, Hiroshi.

    1994-01-01

    In a coolant make-up device, an opening of a pressure equalizing pipeline in a pressure vessel is disposed in coolants above a reactor core and below a usual fluctuation range of a reactor vessel water level. Further, a float check valve is disposed to the pressure equalizing pipeline for preventing coolants in the pressure vessel flowing into the pipeline. If the water level in the pressure vessel is lowered than the setting position for the float check valve, the float drops by its own weight to open the opening of the pressure equalizing pipeline. Then, steams in the pressure vessel are flown into the pipeline, to equalize the pressure between a coolant storage tank and the pressure vessel of the reactor. Coolants in the coolant storage tank is injected to the pressure vessel by way of the water injection pipeline due to the difference of the pressure head between the water level in the coolants storage tank and the water level in the pressure vessel. If the coolants are lowered than the setting position for the float check value, the float check valve does not close unless the water level is recovered to the setting position for the float valve and, accordingly, the coolant make-up is continued. (N.H.)

  16. Void fraction instrument acceptance test procedure

    International Nuclear Information System (INIS)

    Stokes, T.I.; Pearce, K.L.

    1994-01-01

    This document presents the results of the acceptance test for the mechanical and electrical features (not specifically addressed by the software ATP) of the void fraction instrument (VFI). Acceptance testing of the VFI, control console, and decontamination spray assembly was conducted in the 306E building high bay and area adjacent to the facility. The VFI was tested in the horizontal position supported in multiple locations on rolling tables. The control console was located next to the VFI pneumatic control assembly. The VFI system was operated exactly as is expected in the tank farm, with the following exceptions: power was provided from a building outlet and the VFI was horizontal. The testing described in this document verifies that the mechanical and electrical features are operating as designed and that the unit is ready for field service

  17. How to perform the perfect voiding cystourethrogram

    International Nuclear Information System (INIS)

    Agrawalla, Seema; Pearce, Rowena; Goodman, T.Robin

    2004-01-01

    The voiding cystourethrogram (VCUG) examination is a difficult investigation to perform and is a stressful experience for patients and their parents, as well as for the radiologists, technicians and paediatric radiology nurses involved in the examination. Despite the VCUG being one of the most commonly performed fluoroscopic procedures in paediatric radiology practice, there is no general consensus as to the best way to perform this investigation. This is particularly concerning when one considers the potentially high gonadal radiation dose children may receive. Because of this, we have undertaken a comprehensive literature review of various aspects of the test in order to determine the best way to perform the VCUG in modern paediatric radiology practice. (orig.)

  18. Effect of metallurgical variables on void swelling

    International Nuclear Information System (INIS)

    Johnston, W.G.; Lauritzen, T.; Rosolowski, J.H.; Turkalo, A.M.

    1976-01-01

    The mechanism of void swelling is reviewed briefly and the anticipated effects of metallurgical variables are described. Experimental results showing the effects of metallurgical variables are reviewed, most of the work being done by simulation methods employing charged particle bombardments to simulate reactor damage. Although the early emphasis was on structural variables such as grain size, cold work and precipitates to control swelling, it now seems that the practical reduction of swelling will be achieved by modifying alloy composition. Void swelling is strongly influenced by the relative amounts of Fe, Cr, and Ni in an alloy; the amount of swelling can be varied by three orders of magnitude by changing the relative amounts of the three elements in an austenitic ternary alloy. The effect of composition on swelling of a simple ferritic alloy will also be described. The swelling of a simple austenitic alloy of Fe, Cr, and Ni can be reduced by certain minor element additions. The most effective swelling inhibitors are Si, Ti, Zr, and Nb, and combinations of Si and Ti are synergetic. Swelling reductions of two orders of magnitude have been achieved with combined additions. Predictions of swelling in commercial solid solution alloys are made on the basis of the present knowledge of the effects of major composition and minor element additions. The predictions agree with experimental results. For more complex commercial alloys, predictions are made for the effects on swelling of heat treatments that cause changes in matrix composition. In some cases, heat treatment is expected to change the peak swelling by more than a factor of ten, and to shift the peak swelling temperature by almost 100 0 C. Sensitivity of swelling to detailed matrix composition places emphasis on the need for developing understanding of the stability of structure and local composition in an irradiation environment

  19. Post Operative Voiding Efficacy after Anterior Colporrhaphy

    Directory of Open Access Journals (Sweden)

    Behnoosh Miladpoor

    2010-02-01

    Full Text Available The aim of this study was to determine the most effective and suitable time to remove the urinary catheter (Foley after anterior and posterior colporrhaphy surgery. Patients who experience anterior Colporrhaphy operation for genuine stress incontinency or pelvic organ prolapsed will have post operative voiding dysfunction. These patients need postoperative drainage. One of the methods preferred for this purpose is to apply Foley Catheter, but there is no particular regimen available for the exact time of catheter removal in these patients. We have tried to find out the best time to remove Foley catheter after which the repeated Foley catheter is not required or minimized. One hundred and eighty nine patients who have been undergone Colporrhaphy have been selected randomly and divided into three groups' as 1, 2 and 4 days of catheter removal. The number of patients in each group was 62, 63 and 64 respectively. In all three groups, before removing urinary catheter, it was clamped every 4 hrs, for 3 times. After removing of Foley, the patients were guided for urination; the voiding and residual volume was measured. In the patients with an increase of residual volume, the  repeated Foley requirement was increased. However,  5.6 % of the patients with residual volume of ≤ 33 percent and 23.9% of the patients with residual volume between 33 to 68 percent, and finally  64.8% of the patients with residual volume of ≥ 68% had repeated Foley insertion. When considering the number of days, 85, 65 and 35.7 percent of the patients needed repeated Foley after 1, 2, and 4 days of catheter removal respectively. Interestingly, in the third group ( 4 days of the catheter removal with residual volume of ≤ 33% the repeated Foley requirement was nil, with no increase risk of urinary infection. We suggest that the best time to remove the urinary Foley catheter after anterior and posterior Colporrhaphy is the day four.

  20. 14C Behaviour in PWR coolant

    International Nuclear Information System (INIS)

    Sims, Howard; Dickinson Shirley; Garbett, Keith

    2012-09-01

    Although 14 C is produced in relatively small amounts in PWR coolant, it is important to know its fate, for example whether it is released by gaseous discharge, removed by absorption on ion exchange (IX) resins or deposited on the fuel pin surfaces. 14 C can exist in a range of possible chemical forms: inorganic carbon compounds (probably mainly CO 2 ), elemental carbon, and organic compounds such as hydrocarbons. This paper presents results from a preliminary survey of the possible reactions of 14 C in PWR coolant. The main conclusions of the study are: - A combination of thermal and radiolytic reactions controls the chemistry of 14 C in reactor coolant. A simple chemical kinetic model predicts that CH 3 OH would be the initial product from radiolytic reactions of 14 C following its formation from 17 O. CH 3 OH is predicted to arise as a result of reactions of OH . with CH 4 and CH 3 , and it persists because there is no known radiation chemical reduction mechanism. - Thermodynamic considerations show that CH 3 OH can be thermally reduced to CH 4 in PWR conditions, although formation of CO 2 from small organics is the most thermodynamically favourable outcome. Such reactions could be catalysed on active nickel surfaces in the primary circuit. - Limited plant data would suggest that CH 4 is the dominant form in PWR and CO 2 in BWR. This implies that radiation chemistry may be important in determining the speciation. - Addition of acetate does not affect the amount of 14 C formed, but the addition of large amounts of stable carbon would lead to a large range of additional products, some of which would be expected to deposit on fuel pin surfaces as high molecular weight hydrocarbons. However, the subsequent thermal decomposition reactions of these products are not known. - Acetate addition may represent a small input of 12 C compared with organic material released from CVCS resins, although the importance of this may depend on whether that is predominantly soluble

  1. Voids and the Cosmic Web: cosmic depression & spatial complexity

    NARCIS (Netherlands)

    van de Weygaert, Rien; Shandarin, S.; Saar, E.; Einasto, J.

    2016-01-01

    Voids form a prominent aspect of the Megaparsec distribution of galaxies and matter. Not only do theyrepresent a key constituent of the Cosmic Web, they also are one of the cleanest probesand measures of global cosmological parameters. The shape and evolution of voids are highly sensitive tothe

  2. A variational void coalescence model for ductile metals

    KAUST Repository

    Siddiq, Amir; Arciniega, Roman; El Sayed, Tamer

    2011-01-01

    We present a variational void coalescence model that includes all the essential ingredients of failure in ductile porous metals. The model is an extension of the variational void growth model by Weinberg et al. (Comput Mech 37:142-152, 2006

  3. Theory of void swelling, irradiation creep and growth

    International Nuclear Information System (INIS)

    Wood, M.H.; Bullough, R.; Hayns, M.R.

    Recent progress in our understanding of the fundamental mechanisms involved in swelling, creep and growth of materials subjected to irradiation is reviewed. The topics discussed are: the sink types and their strengths in the lossy continuum; swelling and void distribution analysis, including recent work on void nucleation; and, irradiation creep and growth of zirconium and zircaloy are taken as an example

  4. Void nucleation in spheroidized steels during tensile deformation

    International Nuclear Information System (INIS)

    Fisher, J.R. Jr.

    1980-04-01

    An investigation was conducted to determine the effects of various mechanical and material parameters on void formation at cementite particles in axisymmetric tensile specimens of spheroidized plain carbon steels. Desired microstructures for each of three steel types were obtained. Observations of void morphology with respect to various microstructural features were made using optical and scanning electron microscopy

  5. Coolant mixing in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hoehne, T; Grunwald, G

    1998-10-01

    The behavior of PWRs during cold water or boron dilution transients is strongly influenced by the distribution of coolant temperature and boron concentration at the core inlet. This distribution is the needed input to 3-dimensional neutron kinetics to calculate the power distribution in the core. It mainly depends on how the plugs of cold or unborated water formed in a single loop are mixed in the downcomer and in the lower plenum. To simulate such mixture phenomena requires the application of 3-dimensional CFD (computational fluid dynamics) codes. The results of the simulation have to be validated against mixture experiments at scaled facilities. Therefore, in the framework of a research project funded by BMBF, the institute creates a 1:5 mixture facility representing first the geometry of a German pressurized water reactor and later the European Pressurized Water Reactor (EPR) geometry. The calculations are based on the CFD Code CFX-4. (orig.)

  6. Slow coolant phaseout could worsen warming

    Science.gov (United States)

    Reese, April

    2018-03-01

    In the summer of 2016, temperatures in Phalodi, an old caravan town on a dry plain in northwestern India, reached a blistering 51°C—a record high during a heat wave that claimed more than 1600 lives across the country. Wider access to air conditioning (AC) could have prevented many deaths—but only 8% of India's 249 million households have AC. As the nation's economy booms, that figure could rise to 50% by 2050. And that presents a dilemma: As India expands access to a life-saving technology, it must comply with international mandates—the most recent imposed just last fall—to eliminate coolants that harm stratospheric ozone or warm the atmosphere.

  7. Automated surveillance of reactor coolant pump performance

    International Nuclear Information System (INIS)

    Gross, K.C.; Singer, R.M.; Humenik, K.E.

    1992-01-01

    An artificial intelligence based expert system has been developed for continuous surveillance and diagnosis of centrifugal-type reactor coolant pump (RCP) performance and operability. The expert system continuously monitors digitized signals from a variety of physical variables (speed, vibration level, motor power, discharge pressure) associated with RCP performance for annunciation of the incipience or onset of off-normal operation. The system employs an extremely sensitive pattern-recognition technique, the sequential probability ratio test (SPRT) for rapid identification of pump operability degradation. The sequential statistical analysis of the signal noise has been shown to provide the theoretically shortest sampling time to detect disturbances and thus has the potential of providing incipient fault detection information to operators sufficiently early to avoid forced plant shutdowns. The sensitivity and response time of the expert system are analyzed in this paper using monte carlo simulation techniques

  8. Power module assemblies with staggered coolant channels

    Science.gov (United States)

    Herron, Nicholas Hayden; Mann, Brooks S; Korich, Mark D

    2013-07-16

    A manifold is provided for supporting a power module assembly with a plurality of power modules. The manifold includes a first manifold section. The first face of the first manifold section is configured to receive the first power module, and the second face of the first manifold section defines a first cavity with a first baseplate thermally coupled to the first power module. The first face of the second manifold section is configured to receive the second power module, and the second face of the second manifold section defines a second cavity with a second baseplate thermally coupled to the second power module. The second face of the first manifold section and the second face of the second manifold section are coupled together such that the first cavity and the second cavity form a coolant channel. The first cavity is at least partially staggered with respect to second cavity.

  9. Reactor coolant flow measurements at Point Lepreau

    International Nuclear Information System (INIS)

    Brenciaglia, G.; Gurevich, Y.; Liu, G.

    1996-01-01

    The CROSSFLOW ultrasonic flow measurement system manufactured by AMAG is fully proven as reliable and accurate when applied to large piping in defined geometries for such applications as feedwater flows measurement. Its application to direct reactor coolant flow (RCF) measurements - both individual channel flows and bulk flows such as pump suction flow - has been well established through recent work by AMAG at Point Lepreau, with application to other reactor types (eg. PWR) imminent. At Point Lepreau, Measurements have been demonstrated at full power; improvements to consistently meet ±1% accuracy are in progress. The development and recent customization of CROSSFLOW to RCF measurement at Point Lepreau are described in this paper; typical measurement results are included. (author)

  10. Reactor coolant system and containment aqueous chemistry

    International Nuclear Information System (INIS)

    Torgerson, D.F.

    1986-01-01

    Fission products released from fuel during reactor accidents can be subject to a variety of environments that will affect their ultimate behavior. In the reactor coolant system (RCS), for example, neutral or reducing steam conditions, radiation, and surfaces could all have an effect on fission product retention and chemistry. Furthermore, if water is encountered in the RCS, the high temperature aqueous chemistry of fission products must be assessed to determine the quantity and chemical form of fission products released to the containment building. In the containment building, aqueous chemistry will determine the longer-term release of volatile fission products to the containment atmosphere. Over the past few years, the principles of physical chemistry have been rigorously applied to the various chemical conditions described above. This paper reviews the current state of knowledge and discusses the future directions of chemistry research relating to the behavior of fission products in the RCS and containment

  11. Alignment of galaxy spins in the vicinity of voids

    International Nuclear Information System (INIS)

    Slosar, Anže; White, Martin

    2009-01-01

    We provide limits on the alignment of galaxy orientations with the direction to the void center for galaxies lying near the edges of voids. We locate spherical voids in volume limited samples of galaxies from the Sloan Digital Sky Survey using the HB inspired void finder and investigate the orientation of (color selected) spiral galaxies that are nearly edge-on or face-on. In contrast with previous literature, we find no statistical evidence for departure from random orientations. Expressed in terms of the parameter c, introduced by Lee and Pen to describe the strength of such an alignment, we find that c0.11(0.13) at 95% (99.7%) confidence limit within a context of a toy model that assumes a perfectly spherical voids with sharp boundaries

  12. Stability of void lattices under irradiation: a kinetic model

    International Nuclear Information System (INIS)

    Benoist, P.; Martin, G.

    1975-01-01

    Voids are imbedded in a homogeneous medium where point defects are uniformly created and annihilated. As shown by a perturbation calculation, the proportion of the defects which are lost on the cavities goes through a maximum, when the voids are arranged on a translation lattice. If a void is displaced from its lattice site, its growth rate becomes anisotropic and is larger in the direction of the vacant site. The relative efficiency of BCC versus FCC void lattices for the capture of point defects is shown to depend on the relaxation length of the point defects in the surrounding medium. It is shown that the rate of energy dissipation in the crystal under irradiation is maximum when the voids are ordered on the appropriate lattice

  13. Stability of void lattices under irradiation: a kinetic model

    International Nuclear Information System (INIS)

    Benoist, P.; Martin, G.

    1975-01-01

    Voids are imbedded in a homogeneous medium where point defects are uniformly created and annihilated. As shown by a perturbation calculation, the proportion of the defects which are lost on the cavities goes through a maximum, when the voids are arranged on a translation lattice. If a void is displaced from its lattice site, its growth the rate becomes anisotropic and is larger in the direction of the vacant site. The relative efficiency of BCC versus FCC void lattices for the capture of point defects is shown to depend on the relaxation length of the point defects in the surrounding medium. It is shown that the rate of energy dissipation in the crystal under irradiation is maximum when the voids are ordered on the appropriate lattice [fr

  14. Structural control of void formation in dual phase steels

    DEFF Research Database (Denmark)

    Azuma, Masafumi

    The objective of this study is to explore the void formation mechanisms and to clarify the influence of the hardness and structural parameters (volume fraction, size and morphology) of martensite particles on the void formation and mechanical properties in dual phase steels composed of ferrite...... and (iii) strain localization. The critical strain for void formation depends on hardness of the martensite, but is independent of the volume fraction, shape, size and distribution of the martensite. The strain partitioning between the martensite and ferrite depends on the volume fraction and hardness...... of the martensite accelerates the void formation in the martensite by enlarging the size of voids both in the martensite and ferrite. It is suggested that controlling the hardness and structural parameters associated with the martensite particles such as morphology, size and volume fraction are the essential...

  15. Coolant rate distribution in horizontal steam generator under natural circulation

    International Nuclear Information System (INIS)

    Blagovechtchenski, A.; Leontieva, V.; Mitrioukhin, A.

    1997-01-01

    In the presentation the major factors determining the conditions of NCC (Natural Coolant Circulation) in the primary circuit and in particular conditions of coolant rate distribution on the horizontal tubes of PGV-1000 in NPP with VVER-1000 under NCC are considered

  16. Channel type reactors with supercritical water coolant. Russian experience

    International Nuclear Information System (INIS)

    Kuznetsov, Y.N.; Gabaraev, B.A.

    2003-01-01

    Transition to coolant of supercritical parameters allows for principle engineering-andeconomic characteristics of light-water nuclear power reactors to be substantially enhanced. Russian experience in development of channel-type reactors with supercritical water coolant has demonstrated advantages and practical feasibility of such reactors. (author)

  17. Automatic coolant flow control device for a nuclear reactor assembly

    Science.gov (United States)

    Hutter, Ernest

    1986-01-01

    A device which controls coolant flow through a nuclear reactor assembly comprises a baffle means at the exit end of said assembly having a plurality of orifices, and a bimetallic member in operative relation to the baffle means such that at increased temperatures said bimetallic member deforms to unblock some of said orifices and allow increased coolant flow therethrough.

  18. Coolant rate distribution in horizontal steam generator under natural circulation

    Energy Technology Data Exchange (ETDEWEB)

    Blagovechtchenski, A.; Leontieva, V.; Mitrioukhin, A. [St. Petersburg State Technical Univ. (Russian Federation)

    1997-12-31

    In the presentation the major factors determining the conditions of NCC (Natural Coolant Circulation) in the primary circuit and in particular conditions of coolant rate distribution on the horizontal tubes of PGV-1000 in NPP with VVER-1000 under NCC are considered. 5 refs.

  19. Fuel coolant interaction experiment by direct electrical heating method

    International Nuclear Information System (INIS)

    Takeda, Tsuneo; Hirano, Kenmei

    1979-01-01

    In the PCM (Power Cooling Mismatch) experiments, the FCI (Fuel Coolant Interaction) test is one of necessary tests in order to predict various phenomena that occur during PCM in the core. A direct electrical heating method is used for the FCI tests for fuel pellet temperature of over 1000 0 C. Therefore, preheating is required before initiating the direct electrical heating. The fuel pin used in the FCI tests is typical LWR fuel element, which is surrounded by coolant water. It is undersirable to heat up the coolant water during preheating of the fuel pin. Therefore, a zirconia (ZrO 2 ) pellet which is similar to a UO 2 pellet in physical and chemical properties is used. Electric property (electric conductivity) of ZrO 2 is particularly suitable for direct electrical heating as in the case of UO 2 . In this experiment, ZrO 2 pellet (melting point 2500 0 C) melting was achieved by use of both preheating and direct electrical heating. Temperature changes of coolant and fuel surface, as well as the pressure change of coolant water, were measured. The molten fuel interacted with the coolant and generated shock waves. A portion of this molten fuel fragmented into small particles during this interaction. The peak pressure of the observed shock wave was about 35 bars. The damaged fuel pin was photographed after disassembly. This report shows the measured coolant pressure changes and the coolant temperature changes, as well as photographs of damaged fuel pin and fuel fragments. (author)

  20. Coolant rate distribution in horizontal steam generator under natural circulation

    Energy Technology Data Exchange (ETDEWEB)

    Blagovechtchenski, A; Leontieva, V; Mitrioukhin, A [St. Petersburg State Technical Univ. (Russian Federation)

    1998-12-31

    In the presentation the major factors determining the conditions of NCC (Natural Coolant Circulation) in the primary circuit and in particular conditions of coolant rate distribution on the horizontal tubes of PGV-1000 in NPP with VVER-1000 under NCC are considered. 5 refs.

  1. Stress Analysis of Fuel Rod under Axial Coolant Flow

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Hai Lan; Lee, Young Shin; Lee, Hyun Seung [Chungnam National University, Daejeon (Korea, Republic of); Park, Num Kyu; Jeon, Kyung Rok [Kerea Nuclear Fuel., Daejeon (Korea, Republic of)

    2010-05-15

    A pressurized water reactor(PWR) fuel assembly, is a typical bundle structure, which uses light water as a coolant in most commercial nuclear power plants. Fuel rods that have a very slender and long clad are supported by fuel assembly which consists of several spacer grids. A coolant is a fluid which flows through device to prevent its overheating, transferring the heat produced by the device to other devices that use or dissipate it. But at the same time, the coolant flow will bring out the fluid induced vibration(FIV) of fuel rods and even damaged the fuel rod. This study has been conducted to investigate the flow characteristics and nuclear reactor fuel rod stress under effect of coolant. Fluid structure interaction(FSI) analysis on nuclear reactor fuel rod was performed. Fluid analysis of the coolant which flow along the axial direction and structural analysis under effect of flow velocity were carried out under different output flow velocity conditions

  2. Method of charging instruments into liquid metal coolant

    International Nuclear Information System (INIS)

    Yamazaki, Hiroshi

    1980-01-01

    Purpose: To alleviate the thermal shock of a reactor charging machine when charging the machine into liquid metal coolant after the machine is preheated in cover gas. Method: When a reactor fueling machine reaches at the lowermost portion the position immediately above liquid metal coolant surface level, the machine is stopped moving down. The reactor fueling machine is heated at the lowermost portion by thermal radiation from the surface of the liquid metal coolant. After the machine is thus preheated in cover gas, it is again steadily moved down by a winch and charged into the liquid metal coolant. Therefore, the thermal shock of the machine becomes low when charging the machine into the liquid metal coolant to eliminate the damage and deformation at the machine. (Yoshihara, H.)

  3. Stress Analysis of Fuel Rod under Axial Coolant Flow

    International Nuclear Information System (INIS)

    Jin, Hai Lan; Lee, Young Shin; Lee, Hyun Seung; Park, Num Kyu; Jeon, Kyung Rok

    2010-01-01

    A pressurized water reactor(PWR) fuel assembly, is a typical bundle structure, which uses light water as a coolant in most commercial nuclear power plants. Fuel rods that have a very slender and long clad are supported by fuel assembly which consists of several spacer grids. A coolant is a fluid which flows through device to prevent its overheating, transferring the heat produced by the device to other devices that use or dissipate it. But at the same time, the coolant flow will bring out the fluid induced vibration(FIV) of fuel rods and even damaged the fuel rod. This study has been conducted to investigate the flow characteristics and nuclear reactor fuel rod stress under effect of coolant. Fluid structure interaction(FSI) analysis on nuclear reactor fuel rod was performed. Fluid analysis of the coolant which flow along the axial direction and structural analysis under effect of flow velocity were carried out under different output flow velocity conditions

  4. The effect of coolant quantity on local fuel–coolant interactions in a molten pool

    International Nuclear Information System (INIS)

    Cheng, Songbai; Matsuba, Ken-ichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Tohru; Tobita, Yoshiharu

    2015-01-01

    Highlights: • We investigate local fuel–coolant interactions in a molten pool. • As water volume increases, limited pressurization and mechanical energy observed. • Only a part of water is evaporated and responsible for the pressurization. - Abstract: Studies on local fuel–coolant interactions (FCI) in a molten pool are important for severe accident analyses of sodium-cooled fast reactors (SFRs). Motivated by providing some evidence for understanding this interaction, in this study several experimental tests, with comparatively larger difference in coolant volumes, were conducted by delivering a given quantity of water into a simulated molten fuel pool (formed with a low-melting-point alloy). Interaction characteristics including the pressure-buildup as well as mechanical energy release and its conversion efficiency are evaluated and compared. It is found that as water quantity increases, a limited pressure-buildup and the resultant mechanical energy release are observable. The performed analyses also suggest that only a part of water is probably vaporized during local FCIs and responsible for the pressurization and mechanical energy release, especially for those cases with much larger water volumes

  5. Comparative assessment of thermophysical and thermohydraulic characteristics of lead, lead-bismuth and sodium coolants for fast reactors

    International Nuclear Information System (INIS)

    2002-06-01

    All prototype, demonstration and commercial liquid metal cooled fast reactors (LMFRs) have used liquid sodium as a coolant. Sodium cooled systems, operating at low pressure, are characterised by very large thermal margins relative to the coolant boiling temperature and a very low structural material corrosion rate. In spite of the negligible thermal energy stored in the liquid sodium available for release in case of leakage, there is some safety concern because of its chemical reactivity with respect to air and water. Lead, lead-bismuth or other alloys of lead, appear to eliminate these concerns because the chemical reactivity of these coolants with respect to air and water is very low. Some experts believe that conceptually, these systems could be attractive if high corrosion activity inherent in lead, long term materials compatibility and other problems will be resolved. Extensive research and development work is required to meet this goal. Preliminary studies on lead-bismuth and lead cooled reactors and ADS (accelerator driven systems) have been initiated in France, Japan, the United States of America, Italy, and other countries. Considerable experience has been gained in the Russian Federation in the course of development and operation of reactors cooled with lead-bismuth eutectic, in particular, propulsion reactors. Studies on lead cooled fast reactors are also under way in this country. The need to exchange information on alternative fast reactor coolants was a major consideration in the recommendation by the Technical Working Group on Fast Reactors (TWGFRs) to collect, review and document the information on lead and lead-bismuth alloy coolants: technology, thermohydraulics, physical and chemical properties, as well as to make an assessment and comparison with respective sodium characteristics

  6. Experimental investigations of two-phase mixture level swell and axial void fraction distribution under high pressure, low heat flux conditions in rod bundle geometry

    International Nuclear Information System (INIS)

    Anklam, T.M.; White, M.D.

    1981-01-01

    Experimental data is reported from a series of quasi-steady-state two-phase mixture level swell and void fraction distribution tests. Testing was performed at ORNL in the Thermal Hydraulic Test Facility - a large electrically heated test loop configured to produce conditions similar to those expected in a small break loss of coolant accident. Pressure was varied from 2.7 to 8.2 MPa and linear power ranged from 0.33 to 1.95 kW/m. Mixture swell was observed to vary linearly with the total volumetric vapor generation rate over the power range of primary interest in small break analysis. Void fraction data was fit by a drift-flux model and both the drift-velocity and concentration parameter were observed to decrease with increasing pressure

  7. Radionuclide voiding cystography in intrarenal reflux detection

    International Nuclear Information System (INIS)

    Rizzoni, G.; Perale, R.; Bui, F.; Pitter, M.; Pavanello, L.; Boscolo, R.; Passerini Glazel, G.; Macri, C.

    1986-01-01

    In order to evaluate the possibility of detecting intra-renal reflux (IRR) with a more sensitive procedure, 48 children with recurrent urinary tract infections underwent intravenous urography (IVU) and voiding cystourethrogram (VCU) using a solution containing contrast medium and sup(99m)Tc-sulfur colloid particles which are known to persist in the renal parenchyma for a long time. Scintigraphic images were taken at 5 and 20 hours after VCU. 18 children had no vesico-ureteral reflux, 11 showed unilateral and 19 bilateral VUR, which was therefore present in 49 renal units. Among the 49 renal refluxing units (RRUs) IRR was detected radiologically in 8; of these isotopic activity in the renal area was present in all 6 RRUs who were examined at 20 hours. Of the remaining 41 RRUs with no radiologically detectable IRR 24 were evaluated at 20 hours and 5 (21%) showed renal radioactivity. Renal scars were significantly more frequent in kidneys with radioisotopic activity at 20 hours. The results of this study indicate that radionuclide cystography using sup(99m)Tc-sulfur colloid is a reliable procedure for demonstrating IRR, and to this end is more sensitive than X-ray VCU. Radionuclide cystography with sulfur colloid particles should therefore be considered a simple and useful complementary procedure, which is more sensitive than X-ray VCU in the diagnosis and follow-up of IRR

  8. RELAXATION OF BLAZAR-INDUCED PAIR BEAMS IN COSMIC VOIDS

    Energy Technology Data Exchange (ETDEWEB)

    Miniati, Francesco [Physics Department, Wolfgang-Pauli-Strasse 27, ETH-Zuerich, CH-8093 Zuerich (Switzerland); Elyiv, Andrii, E-mail: fm@phys.ethz.ch [Institut d' Astrophysique et de Geophysique, Universite de Liege, B-4000 Liege (Belgium)

    2013-06-10

    The stability properties of a low-density ultrarelativistic pair beam produced in the intergalactic medium (IGM) by multi-TeV gamma-ray photons from blazars are analyzed. The problem is relevant for probes of magnetic field in cosmic voids through gamma-ray observations. In addition, dissipation of such beams could considerably affect the thermal history of the IGM and structure formation. We use a Monte Carlo method to quantify the properties of the blazar-induced electromagnetic shower, in particular the bulk Lorentz factor and the angular spread of the pair beam generated by the shower, as a function of distance from the blazar itself. We then use linear and nonlinear kinetic theory to study the stability of the pair beam against the growth of electrostatic plasma waves, employing the Monte Carlo results for our quantitative estimates. We find that the fastest growing mode, like any perturbation mode with even a very modest component perpendicular to the beam direction, cannot be described in the reactive regime. Due to the effect of nonlinear Landau damping, which suppresses the growth of plasma oscillations, the beam relaxation timescale is found to be significantly longer than the inverse Compton loss time. Finally, density inhomogeneities associated with cosmic structure induce loss of resonance between the beam particles and plasma oscillations, strongly inhibiting their growth. We conclude that relativistic pair beams produced by blazars in the IGM are stable on timescales that are long compared with the electromagnetic cascades. There appears to be little or no effect of pair beams on the IGM.

  9. Evaluation of alternate secondary (and tertiary) coolants for the molten-salt breeder reactor

    International Nuclear Information System (INIS)

    Kelmers, A.D.; Baes, C.F.; Bettis, E.S.; Brynestad, J.; Cantor, S.; Engel, J.R.; Grimes, W.R.; McCoy, H.E.; Meyer, A.S.

    1976-04-01

    The three most promising coolant selections for an MSBR have been identified and evaluated in detail from the many coolants considered for application either as a secondary coolant in 1000-MW(e) MSBR configurations using only one coolant, or as secondary and tertiary coolants in an MSBR dual coolant configuration employing two different coolants. These are, as single secondary coolants: (1) a ternary sodium--lithium--beryllium fluoride melt; (2) the sodium fluoroborate--sodium fluoride eutectic melt, the present reference design secondary coolant. In the case of the dual coolant configuration, the preferred system is molten lithium--beryllium fluoride (Li 2 BeF 4 ) as the secondary coolant and helium gas as the tertiary coolant

  10. Partial discharges within two spherical voids in an epoxy resin

    International Nuclear Information System (INIS)

    Illias, H A; Mokhlis, H; Tunio, M A; Chen, G; Bakar, A H A

    2013-01-01

    A void in a dielectric insulation material may exist due to imperfection in the insulation manufacturing or long term stressing. Voids have been identified as one of the common sources of partial discharge (PD) activity within an insulation system, such as in cable insulation and power transformers. Therefore, it is important to study PD phenomenon within void cavities in insulation. In this work, a model of PD activity within two spherical voids in a homogeneous dielectric material has been developed using finite element analysis software to study the parameters affecting PD behaviour. The parameters that have been taken into account are the void surface conductivity, electron generation rate and the inception and extinction fields. Measurements of PD activity within two spherical voids in an epoxy resin under ac sinusoidal applied voltage have also been performed. The simulation results have been compared with the measurement data to validate the model and to identify the parameters affecting PD behaviour. Comparison between measurements of PD activity within single and two voids in a dielectric material have also been made to observe the difference of the results under both conditions. (paper)

  11. The association of age of toilet training and dysfunctional voiding

    Directory of Open Access Journals (Sweden)

    Hodges SJ

    2014-10-01

    Full Text Available Steve J Hodges, Kyle A Richards, Ilya Gorbachinsky, L Spencer KraneDepartment of Urology, Wake Forest University, Winston-Salem, NC, USAObjective: To determine whether age of toilet training is associated with dysfunctional voiding in children.Materials and methods: We compared patients referred to the urologic clinics for voiding dysfunction with age-matched controls without urinary complaints. Characteristics including age and reason for toilet training, method of training, and encopresis or constipation were compared between both groups.Results: Initiation of toilet training prior to 24 months and later than 36 months of age were associated with dysfunctional voiding. However, dysfunctional voiding due to late toilet training was also associated with constipation.Conclusion: Dysfunctional voiding may be due to delayed emptying of the bowel and bladder by children. The symptoms of dysfunctional voiding are more common when toilet training early, as immature children may be less likely to empty in a timely manner, or when training late due to (or in association with constipation.Keywords: voiding dysfunction, constipation

  12. A sharp interface model for void growth in irradiated materials

    Science.gov (United States)

    Hochrainer, Thomas; El-Azab, Anter

    2015-03-01

    A thermodynamic formalism for the interaction of point defects with free surfaces in single-component solids has been developed and applied to the problem of void growth by absorption of point defects in irradiated metals. This formalism consists of two parts, a detailed description of the dynamics of defects within the non-equilibrium thermodynamic frame, and the application of the second law of thermodynamics to provide closure relations for all kinetic equations. Enforcing the principle of non-negative entropy production showed that the description of the problem of void evolution under irradiation must include a relationship between the normal fluxes of defects into the void surface and the driving thermodynamic forces for the void surface motion; these thermodynamic forces are identified for both vacancies and interstitials and the relationships between these forces and the normal point defect fluxes are established using the concepts of transition state theory. The latter theory implies that the defect accommodation into the surface is a thermally activated process. Numerical examples are given to illustrate void growth dynamics in this new formalism and to investigate the effect of the surface energy barriers on void growth. Consequences for phase field models of void growth are discussed.

  13. On nonlinear excitation of voids in dusty plasmas

    International Nuclear Information System (INIS)

    Nebbat, E.; Annou, R.; Bharuthram, R.

    2007-01-01

    The void, which is a dust-free region inside the dust cloud in the plasma, results from a balance of the electrostatic force and the ion-drag force on a dust particulate that has numerous forms, some of which are based on models whereas others are driven from first principles. To explain the generation of voids, K. Avinash, A. Bhattacharjee, and S. Hu [Phys. Rev. Lett. 90, 075001 (2003)] proposed a time-dependent nonlinear model that describes the void as a result of an instability. We augment this model by incorporating the grain drift and reintroducing the velocity convective term as well as by replacing the modeled ion-drag force by a more accurate one. The analysis is conducted in a spherical configuration. It is revealed that the void formation is a threshold phenomenon, i.e., it depends on the grain size. Furthermore, the void possesses a sharp boundary beyond which the dust density decreases and may present a corrugated aspect. For big size grains, the use of both ion-drag forces leads to voids of the same dimension, though for grains of small sizes, the Avinash force drives voids of a higher dimension. The model shows good agreement with the experiment

  14. Void nucleation at elevated temperatures under cascade-damage irradiation

    International Nuclear Information System (INIS)

    Semenov, A.A.; Woo, C.H.

    2002-01-01

    The effects on void nucleation of fluctuations respectively due to the randomness of point-defect migratory jumps, the random generation of free point defects in discrete packages, and the fluctuating rate of vacancy emission from voids are considered. It was found that effects of the cascade-induced fluctuations are significant only at sufficiently high total sink strength. At lower sink strengths and elevated temperatures, the fluctuation in the rate of vacancy emission is the dominant factor. Application of the present theory to the void nucleation in annealed pure copper neutron-irradiated at elevated temperatures with doses of 10 -4 -10 -2 NRT dpa showed reasonable agreement between theory and experiment. This application also predicts correctly the temporal development of large-scale spatial heterogeneous microstructure during the void nucleation stage. Comparison between calculated and experimental void nucleation rates in neutron-irradiated molybdenum at temperatures where vacancy emission from voids is negligible showed reasonable agreement as well. It was clearly demonstrated that the athermal shrinkage of relatively large voids experimentally observable in molybdenum at such temperatures may be easily explained in the framework of the present theory

  15. Magnetic resonance voiding cystography in the diagnosis of vesicoureteral reflux: comparative study with voiding cystourethrography.

    Science.gov (United States)

    Lee, Sang Kwon; Chang, Yongmin; Park, Noh Hyuck; Kim, Young Hwan; Woo, Seongku

    2005-04-01

    To evaluate the feasibility of magnetic resonance voiding cystography (MRVC) compared with voiding cystourethrography (VCUG) for detecting and grading vesicoureteral reflux (VUR). MRVC was performed upon 20 children referred for investigation of reflux. Either coronal T1-weighted spin-echo (SE) or gradient-echo (GE) (fast multiplanar spoiled gradient-echo (FMPSPGR) or turbo fast low-angle-shot (FLASH)) images were obtained before and after transurethral administration of gadolinium solution, and immediately after voiding. The findings of MRVC were compared with those of VCUG and technetium-99m ((99m)Tc) dimercaptosuccinic acid (DMSA) single-photon emission computed tomography (SPECT) performed within 6 months of MRVC. VUR was detected in 23 ureterorenal units (16 VURs by both methods, 5 VURs by VCUG, and 2 VURs by MRVC). With VCUG as the standard of reference, the sensitivity of MRVC was 76.2%; the specificity, 90.0%; the positive predictive value, 88.9%; and the negative predictive value, 78.3%. There was concordance between two methods regarding the grade of reflux in all 16 ureterorenal units with VUR detected by both methods. Of 40 kidneys, MRVC detected findings of renal damage or reflux nephropathy in 13 kidneys, and (99m)Tc DMSA renal SPECT detected findings of reflux nephropathy in 17 kidneys. Although MRVC is shown to have less sensitivity for VUR than VCUG, MRVC may represent a method of choice offering a safer nonradiation test that can additionally evaluate the kidneys for changes related to reflux nephropathy. Copyright 2005 Wiley-Liss, Inc.

  16. Neutronic performance of high molecular weight coolants for a prismatic VHTR

    International Nuclear Information System (INIS)

    Schriener, T. M.; El-Genk, M. S.

    2008-01-01

    A neutronic model is developed of a prismatic Very High Temperature Reactor (VHTR) to investigate the effects on the excess reactivity and operation cycle length of replacing helium with binary gas mixtures of He-Ne, He-N 2 , or He-Xe as reactor coolants and working fluids in the direct Closed Brayton Cycle (CBC) for energy conversion. Also investigated is the neutron activation of these binary gas mixtures in the VHTR. The motivation for using the heavy binary mixtures is the smaller size and the fewer number of stages of the CBC turbo-machinery. The present analysis uses the Monte Carlo code MCNPX 2.6D at typical operating conditions (500-1000 degrees and 7.12 MPa) in the VHTR. He-Ne (15 g/mol) is the best neutronically, but not thermal-hydraulically, followed by He-N 2 . Although He-Ne has ∼13.6% lower heat transfer coefficient than helium, it insignificantly affects the initial excess reactivity and the operation life cycle and experiences no neutrons activation. On the other hand, He-N 2 has 4.4% higher heat transfer coefficient than helium and experiences insignificant neutron activation in the reactor, but decreases the initial excess reactivity by ∼5.2% and the operation cycle length by 6.7%. He-Xe (15 g/mol) has 8% higher heat transfer coefficient than helium, but decreases the initial excess reactivity by 18.2% and the operational cycle length by 17%. In addition, neutron activation of xenon produces a significant source term, requiring shielding of the CBC loop and could contaminate the turbo-machinery with long-lived radioactive cesium. Thus, He-Xe is not recommended as a reactor coolant, but could be used as working fluid in a CBC loop that is indirectly coupled to helium cooled VHTR. (authors)

  17. Breakup of jet and drops during premixing phase of fuel coolant interactions

    Energy Technology Data Exchange (ETDEWEB)

    Haraldsson, Haraldur Oskar

    2000-05-01

    performed. The coolant temperature was found to significantly affect the shape and size of the debris. The maximum fragment size was found to increase with reduction in coolant temperature. No effect of coolant voiding on the fragment size distribution was observed. A series of high temperature melt jet experiments were performed, in the MIRA-20L experimental facility. Three types of oxidic melts, namely; CaO-B{sub 2}O{sub 3}, MnO-TiO{sub 2} and WO{sub 3}-CaO were employed as melt jet liquid. The melt jet fragmentation was classified into two regimes, the hydrodynamic-controlled regime and the solidification-controlled regime. The delineation between those regimes was observed from the size characteristic and morphology of the solidified debris which was formed. The temperature of the coolant was the primary parameter in determining which regime the jet fragmentation would fall into. It was found, at low subcooling, the fragments are relatively large and irregular compared to smaller particles produced at higher subcooling. The melt density was found to have significant effect on the particle size. The mass mean size of the debris changes proportional to the square root of the coolant to melt density ratio. A systematic investigation of the performance of statistical distributions which may be used to describe the size distributions of fragments obtained from molten fuel coolant interaction (MFCI) experiments was performed. The statistical analysis of the debris produced in both experiments showed that the sequential fragmentation theory fits best the particle distribution produced during the jet fragmentation process. In the second part of the second chapter, analysis of the jet breakup experiments are performed. The low temperature jet fragmentation experiments are simulated with a recently developed Multiphase Eulerian Lagrangian Method. The effect of particle diameter and particle drag on the jet dynamics and penetration behavior is investigated. The third part of the

  18. Breakup of jet and drops during premixing phase of fuel coolant interactions

    International Nuclear Information System (INIS)

    Haraldsson, Haraldur Oskar

    2000-05-01

    performed. The coolant temperature was found to significantly affect the shape and size of the debris. The maximum fragment size was found to increase with reduction in coolant temperature. No effect of coolant voiding on the fragment size distribution was observed. A series of high temperature melt jet experiments were performed, in the MIRA-20L experimental facility. Three types of oxidic melts, namely; CaO-B 2 O 3 , MnO-TiO 2 and WO 3 -CaO were employed as melt jet liquid. The melt jet fragmentation was classified into two regimes, the hydrodynamic-controlled regime and the solidification-controlled regime. The delineation between those regimes was observed from the size characteristic and morphology of the solidified debris which was formed. The temperature of the coolant was the primary parameter in determining which regime the jet fragmentation would fall into. It was found, at low subcooling, the fragments are relatively large and irregular compared to smaller particles produced at higher subcooling. The melt density was found to have significant effect on the particle size. The mass mean size of the debris changes proportional to the square root of the coolant to melt density ratio. A systematic investigation of the performance of statistical distributions which may be used to describe the size distributions of fragments obtained from molten fuel coolant interaction (MFCI) experiments was performed. The statistical analysis of the debris produced in both experiments showed that the sequential fragmentation theory fits best the particle distribution produced during the jet fragmentation process. In the second part of the second chapter, analysis of the jet breakup experiments are performed. The low temperature jet fragmentation experiments are simulated with a recently developed Multiphase Eulerian Lagrangian Method. The effect of particle diameter and particle drag on the jet dynamics and penetration behavior is investigated. The third part of the second chapter

  19. Track 5: safety in engineering, construction, operations, and maintenance. Reactor physics design, validation, and operating experience. 5. A Negative Reactivity Feedback Device for Actinide Burner Cores

    International Nuclear Information System (INIS)

    Driscoll, M.J.; Hejzlar, P.

    2001-01-01

    Lead-bismuth eutectic (LBE) cooled reactors are of considerable interest because they may be useful for destruction of actinides in a cost-effective manner, particularly cores fueled predominantly with minor actinides, which gain reactivity with burnup. However, they also pose several design challenges: 1. a small (and perhaps even slightly positive) Doppler feedback; 2. small effective delayed neutron yield; 3. a small negative feedback from axial fuel expansion; 4. positive coolant void and temperature coefficients for conventional designs. This has motivated a search for palliative measures, leading to conceptualization of the reactivity feedback device (RFD). The RFD consists of an in-core flask containing helium gas, tungsten wool, and a small reservoir of LBE that communicates with vertical tubes housing neutron absorber floats. The upper part of these guide tubes contains helium gas that is vented into a separate, cooler ex-core helium gas plenum. The principle of operation is as follows: 1. The tungsten wool, hence the helium gas in the in-core plenum, is heated by gammas and loses heat to the walls by convection and conduction (radiation is feeble for monatomic gases and, in any event, intercepted by the tungsten wool). An energy balance determines the gas temperature, hence, pressure, which is 10 atm here. The energy loss rate can be adjusted by using xenon or a gas mixture in place of helium. The tungsten wool mass, which is 1 vol% wool here, can also be increased to increase gamma heating and further retard convection; alternatively, a Dewar flask could be used in place of the additional wool. 2. An increase in core power causes a virtually instantaneous increase in gamma flux, hence, gas heatup: The thermal time constant of the tungsten filaments and their surrounding gas film is ∼40 μs. 3. The increased gas temperature is associated with an increased gas pressure, which forces more liquid metal into the float guide tubes: LBE will rise ∼100 cm

  20. A void fraction model for annular two-phase flow

    Energy Technology Data Exchange (ETDEWEB)

    Tandon, T.N.; Gupta, C.P.; Varma, H.K.

    1985-01-01

    An analytical model has been developed for predicting void fraction in two-phase annular flow. In the analysis, the Lockhart-Martinelli method has been used to calculate two-phase frictional pressure drop and von Karman's universal velocity profile is used to represent the velocity distribution in the annular liquid film. Void fractions predicted by the proposed model are generally in good agreement with a available experimental data. This model appears to be as good as Smith's correlation and better than the Wallis and Zivi correlations for computing void fraction.

  1. Radiation-induced void swelling in metals and alloys

    International Nuclear Information System (INIS)

    Zelinskij, V.F.; Neklyudov, I.M.; Ozhigov, L.S.; Reznichenko, Eh.A.; Rozhkov, V.V.; Chernyaeva, T.T.

    1979-01-01

    Main regularities in the development of radiation-induced void swelling are considered. Special attention is paid to consideration of a possibility to obtain information on material behaviour under conditions of reactor irradiation proceeding from the data of simulation experiments and to methods of rate control, for the processes which occur in material during irradiation and further annealing by the way of rationalized alloying, of thermomechanical treatment and programmed change of irradiation conditions under operation. Problems of initiation and growth of voids in irradiated materials are discussed as well as the ways to decrease the rate of radiation-induced void swelling

  2. Numerical simulation of void growth under dynamic loading

    International Nuclear Information System (INIS)

    Iqbal, A.

    1996-01-01

    Following a brief general review of developments in material behavior under high strain rates, a cylindrical cell surrounding a spherical void in OFHC copper is numerically simulated by Zerri-Armstrong model. This simulation results show that the plastic deformation tends to be concentrated in the vicinity of voids either in the axial or transverse direction depending upon the stress state. This event is associated with the accelerated void through accompanying coalescence causing ductile fracture. A3-node triangular mesh generation code used as input for finite element code is developed by a 'Central Generation' technique. (author)

  3. Local void and slip model used in BODYFIT-2PE

    International Nuclear Information System (INIS)

    Chen, B.C.J.; Chien, T.H.; Kim, J.H.; Lellouche, G.S.

    1983-01-01

    A local void and slip model has been proposed for a two-phase flow without the need of fitting any empirical parameters. This model is based on the assumption that all bubbles have reached their terminal rise velocities in the two-phase region. This simple model seems to provide reasonable calculational results when compared with the experimental data and other void and slip models. It provides a means to account for the void and slip of a two-phase flow on a local basis. This is particularly suitable for a fine mesh thermal-hydraulic computer program such as BODYFIT-2PE

  4. Triboengineering problems of lead coolant in innovative fast reactors

    International Nuclear Information System (INIS)

    Beznosov, A.V.; Novozhilova, O.O.; Shumilkov, A.I.; Lvov, A.V.; Bokova, T.A.; Makhov, K.A.

    2013-01-01

    Graphical abstract: Models of experimental sites for research of processes tribology in heavy liquid metal coolant. -- Highlights: • The contact a pair of heavy liquid metal coolant for reactors on fast neutrons. • The hydrostatic bearings main circulation pumps. • Oxide coating and degree of wear of friction surfaces in heavy liquid metal coolant. -- Abstract: So far, there are plenty of works dedicated to studying the phenomenon of friction. However, there are none dedicated to functioning of contact pairs in heavy liquid-metal coolants for fast neutron, reactor installations (Kogaev and Drozdov, 1991; Modern Tribology, 2008; Drozdov et al., 1986). At the Nizhny Novgorod State Technical University, such research is conducted in respect to friction, bearings of main circulating pumps, interaction of sheaths of neutron absorber rods with their covers, of the reactor control and safety system, refueling systems, and interaction of coolant flows with, channel borders. As a result of experimental studies, the characteristic of friction pairs in the heavy, liquid metal coolant shows the presence dependences of oxide film on structural materials of the wear. The inapplicability of existing calculation methods for assessing the performance of the bearing nodes, in the heavy liquid metal coolant is shown

  5. ISS Internal Active Thermal Control System (IATCS) Coolant Remediation Project

    Science.gov (United States)

    Morrison, Russell H.; Holt, Mike

    2005-01-01

    The IATCS coolant has experienced a number of anomalies in the time since the US Lab was first activated on Flight 5A in February 2001. These have included: 1) a decrease in coolant pH, 2) increases in inorganic carbon, 3) a reduction in phosphate buffer concentration, 4) an increase in dissolved nickel and precipitation of nickel salts, and 5) increases in microbial concentration. These anomalies represent some risk to the system, have been implicated in some hardware failures and are suspect in others. The ISS program has conducted extensive investigations of the causes and effects of these anomalies and has developed a comprehensive program to remediate the coolant chemistry of the on-orbit system as well as provide a robust and compatible coolant solution for the hardware yet to be delivered. The remediation steps include changes in the coolant chemistry specification, development of a suite of new antimicrobial additives, and development of devices for the removal of nickel and phosphate ions from the coolant. This paper presents an overview of the anomalies, their known and suspected system effects, their causes, and the actions being taken to remediate the coolant.

  6. Device for preventing coolant outflow in a reactor

    International Nuclear Information System (INIS)

    Nemoto, Kiyomitsu; Mochizuki, Keiichi.

    1975-01-01

    Object: To prevent outflow of coolant from a reactor vessel even in an occurrence of leaking trouble at a low position in a primary cooling system or the like in the reactor vessel. Structure: An inlet at the foremost end of a coolant inlet pipe inserted into a reactor vessel is arranged at a level lower than a core, and a check valve is positioned at a level higher than the core in a rising portion of the inlet. In normal condition, the check valve is pushed up by discharge pressure of a main circulating pump and remains closed, and hence, producing no flow loss of coolant, sodium. However, when a trouble such as rupture occurs at the lower position in the primary cooling system, the attractive force for allowing the coolant to back-flow outside the reactor vessel and the load force of the coolant within the reactor vessel cause the check valve to actuate, as a consequence of which a liquid level of the coolant downwardly moves to the position of the check valve to intake the cover gases into a gas intake, thereby cutting off a flow passage of the coolant to stop outflow thereof. (Kamimura, M.)

  7. Development of lead-bismuth coolant technology for nuclear device

    International Nuclear Information System (INIS)

    Kamata, Kin-ya; Kitano, Teruaki; Ono, Mikinori

    2004-01-01

    Liquid lead-bismuth is a promising material as a future fast reactor coolant or an intensive neutron source material for accelerator driven transmutation system (ADS). To develop nuclear plants and their installations using lead-bismuth coolant for practical use, both coolant technologies, inhabitation process of steels and quality control of coolant, and total operation system for liquid lead-bismuth plants are required. Based on the experience of liquid metal coolant, Mitsui Engineering and Shipbuilding Co., Ltd. (MES) has completed the liquid lead-bismuth forced circulation loop and has acquired various engineering data on main components including economizer. As a result of tis operation, MES has developed key technologies of lead-bismuth coolant such as controlling of oxygen content in lead-bismuth and a purification of lead-bismuth coolant. MES participated in the national project, ''The Development of Accelerator Driven Transmutation System'', together with JAERI (Japan Atomic Energy Research Institute) and started corrosion test for beam window of ADS. (author)

  8. Liquid metal coolant disposal from UKAEA reactors at Dounreay

    International Nuclear Information System (INIS)

    Adam, E.R.

    1997-01-01

    As part of the United Kingdom's Fast Reactor Development programme two reactors were built and operated at Dounreay in the North of Scotland. DFR (Dounreay Fast Reactor) was operated from 1959-1977 and PFR (Prototype Fast Reactor) was operated from 1974-1994. Both reactors are currently undergoing Stage 1 Decommissioning and are installing plant to dispose of the bulk coolant (DFR ∼ 60 tonne; PFR ∼ 1500 tonne). The coolant (NaK) remaining at DFR is mainly in the primary circuit which contains in excess of 500 TBq of Cs137. Disposal of 40 tonnes of secondary coolant has already been carried out. The paper will describe the processes used to dispose of this secondary circuit coolant and how it is intended the remaining primary circuit coolant will be handled. The programme to process the primary coolant will also be described which involves the conversion of the liquid metal to caustic and its decontamination. No PFR coolant Na has been disposed off to date. The paper will describe the current decommissioning programme activities relating to liquid metal disposal and treatment describing the materials to be disposed of and the issue of decontamination of the effluents. (author)

  9. Power supplyer for reactor coolant recycling pump

    International Nuclear Information System (INIS)

    Nara, Hiroshi; Okinaka, Yo.

    1991-01-01

    The present invention concerns a variable voltage/variable frequency static power source (static power source) used as a power source for a coolants recycling pump motor of a nuclear power plant. That is, during lower power operation such as start up or shutdown in which stoppage of the power source gives less effect to a reactor core, power is supplied from a power system, a main power generator connected thereto or a high voltage bus in the plant or a common high voltage bus to the static power source. However, during rated power operation, power is supplied from the output of an axially power generator connected with a main power generator having an extremely great inertia moment to the static power device. With such a constitution, the static power device is not stopped by the lowering of the voltage due to a thunderbolt falling accident or the like to a power-distribution line suddenly occurred in the power system. Accordingly, reactor core flowrate is free from rapid decrease caused by the reduction of rotation speed of the recycling pump. Accordingly, disadvantgages upon operation control in the reactor core is not caused. (I.S.)

  10. Characterization of reactor coolant by XRF

    Energy Technology Data Exchange (ETDEWEB)

    Legreid, G.; Beverskog, B. [OECD Halden Reactor Project (Norway)

    2002-07-01

    The analyzes of membrane filters is of utmost importance in characterizing the coolant chemistry in nuclear power plants. Traditional analyzes of filters includes oxidative digestion followed by instrumental analyzes. XRF (X-ray Fluorescence spectrometry) can analyze without digestion of the filters. The method is much faster and demands only a cutting step as sample preparation. By use of XRF the analytical laboratory at the Halden Reactor Project will get increased capacity, which makes it possible to analyze more samples and improve the characterization of the water. The method has shown to give more stable results than other methods in use, and has proved to have good precision. New calibration methods have been developed and tested successfully against other methods. A round robin test, attending seven laboratories from nuclear power plants, was initiated by the Halden Project to verify the instrument. The test of standard cation exchange filters showed that conventional filter digestion results in too low values. The XRF methodology shows very good agreement with the standard values. The round robin test for particle filters could not confirm that filter digestion results in too low values. This was mainly due to lack of standard particle filters and large scatter in the reported data. (author)

  11. Characterization of reactor coolant by XRF

    International Nuclear Information System (INIS)

    Legreid, G.; Beverskog, B.

    2002-01-01

    The analyzes of membrane filters is of utmost importance in characterizing the coolant chemistry in nuclear power plants. Traditional analyzes of filters includes oxidative digestion followed by instrumental analyzes. XRF (X-ray Fluorescence spectrometry) can analyze without digestion of the filters. The method is much faster and demands only a cutting step as sample preparation. By use of XRF the analytical laboratory at the Halden Reactor Project will get increased capacity, which makes it possible to analyze more samples and improve the characterization of the water. The method has shown to give more stable results than other methods in use, and has proved to have good precision. New calibration methods have been developed and tested successfully against other methods. A round robin test, attending seven laboratories from nuclear power plants, was initiated by the Halden Project to verify the instrument. The test of standard cation exchange filters showed that conventional filter digestion results in too low values. The XRF methodology shows very good agreement with the standard values. The round robin test for particle filters could not confirm that filter digestion results in too low values. This was mainly due to lack of standard particle filters and large scatter in the reported data. (author)

  12. Numerical model simulation of atmospheric coolant plumes

    International Nuclear Information System (INIS)

    Gaillard, P.

    1980-01-01

    The effect of humid atmospheric coolants on the atmosphere is simulated by means of a three-dimensional numerical model. The atmosphere is defined by its natural vertical profiles of horizontal velocity, temperature, pressure and relative humidity. Effluent discharge is characterised by its vertical velocity and the temperature of air satured with water vapour. The subject of investigation is the area in the vicinity of the point of discharge, with due allowance for the wake effect of the tower and buildings and, where application, wind veer with altitude. The model equations express the conservation relationships for mometum, energy, total mass and water mass, for an incompressible fluid behaving in accordance with the Boussinesq assumptions. Condensation is represented by a simple thermodynamic model, and turbulent fluxes are simulated by introduction of turbulent viscosity and diffusivity data based on in-situ and experimental water model measurements. The three-dimensional problem expressed in terms of the primitive variables (u, v, w, p) is governed by an elliptic equation system which is solved numerically by application of an explicit time-marching algorithm in order to predict the steady-flow velocity distribution, temperature, water vapour concentration and the liquid-water concentration defining the visible plume. Windstill conditions are simulated by a program processing the elliptic equations in an axisymmetrical revolution coordinate system. The calculated visible plumes are compared with plumes observed on site with a view to validate the models [fr

  13. Reactor coolant pump monitoring and diagnostic system

    International Nuclear Information System (INIS)

    Singer, R.M.; Gross, K.C.; Walsh, M.; Humenik, K.E.

    1990-01-01

    In order to reliably and safely operate a nuclear power plant, it is necessary to continuously monitor the performance of numerous subsystems to confirm that the plant state is within its prescribed limits. An important function of a properly designed monitoring system is the detection of incipient faults in all subsystems (with the avoidance of false alarms) coupled with an information system that provides the operators with fault diagnosis, prognosis of fault progression and recommended (either automatic or prescriptive) corrective action. In this paper, such a system is described that has been applied to reactor coolant pumps. This system includes a sensitive pattern-recognition technique based upon the sequential probability ratio test (SPRT) that detects incipient faults from validated signals, an expert system embodying knowledge bases on pump and sensor performance, extensive hypertext files containing operating and emergency procedures as well as pump and sensor information and a graphical interface providing the operator with easily perceived information on the location and character of the fault as well as recommended corrective action. This system is in the prototype stage and is currently being validated utilizing data from a liquid-metal cooled fast reactor (EBR-II). 3 refs., 4 figs

  14. Reactor Coolant Temperature Measurement using Ultrasonic Technology

    Energy Technology Data Exchange (ETDEWEB)

    Jung, JaeCheon [KEPCO International Nuclear graduate School, Ulsan (Korea, Republic of); Seo, YongSun; Bechue, Nicholas [Krohne Messtechnik GmbH, Duisburg (Germany)

    2016-10-15

    In NPP, the primary piping temperature is detected by four redundant RTDs (Resistance Temperature Detectors) installed 90 degrees apart on the RCS (Reactor Coolant System) piping circumferentially. Such outputs however, if applied to I and C systems would not give balanced results. The discrepancy can be explained by either thermal stratification or improper arrangement of thermo-wells and RTDs. This phenomenon has become more pronounced in the hot-leg piping than in the cold-leg. Normally, the temperature difference among channels is in the range of 1°F in Korean nuclear power Plants. Consequently, a more accurate pipe average temperate measurement technique is required. Ultrasonic methods can be used to measure average temperatures with relatively higher accuracy than RTDs because the sound wave propagation in the RCS pipe is proportional to the average temperature around pipe area. The inaccuracy of RCS temperature measurement worsens the safety margin for both DNBR and LPD. The possibility of this discrepancy has been reported with thermal stratification effect. Proposed RCS temperature measurement system based on ultrasonic technology offers a countermeasure to cope with thermal stratification effect on hot-leg piping that has been an unresolved issue in NPPs. By introducing ultrasonic technology, the average internal piping temperature can be measured with high accuracy. The inaccuracy can be decreased less than ±1℉ by this method.

  15. Alternative protections for loss of coolant accidents

    International Nuclear Information System (INIS)

    Estevez, E.A.

    1997-01-01

    One way to mitigate a small loss of coolant accident (LOCA) is by depressurizing the primary system, in order to turn the accident into a sequence where water is fed to a low pressure system. It can be achieved by two different ways: by incorporating a valve system (ADS - Automatic Depressurization System) to the design, which helps to diminish the pressure, obtaining a bigger LOCA, or by extracting heat from the system. Our analysis is centered in integrated reactors. The first characterization performed was on CAREM reactor. The idea was then to observe its behavior with LOCAs for different thermal power relations, water volume and rupture area. A simple depressurization model is presented, which enables us to find the parameter relationships which characterize this process, from which some particular cases will arise. ADS implementation is then analyzed, giving the criteria for the triggering time. A study on its reliability and the probability of a spurious opening is made, taking into account independent and dependent failures. An analysis on heat extraction as alternative for depressurizing is also made. Finally, the different reasons to choose between ADS or heat extraction as alternative are given, and the meaning of the parameters found are discussed. An alternative to classify LOCAs, instead of the traditional classification, by fracture size, is suggested. (author)

  16. Method of decontaminating primary coolant circuits

    International Nuclear Information System (INIS)

    Ishibashi, Masaru; Sumi, Masao.

    1981-01-01

    Purpose: To eliminate hard contaminated layers as well as soft contaminated layers without injuring substrate materials, upon decontamination of radiation contaminated portions in equipments and pipeways constituting primary coolant circuits. Constitution: High pressure water from a high pressure pump is jetted out from the nozzle of a spray gun to the radiation contaminated portions in equipments, for example, to the surface of water chamber in a vapor evaporator. High pressure pure water or aqueous boric acid is jetted out from the periphery and boric oxide particles (of about 1 - 100 μ particle size) are jetted out from the center of the nozzle of the spray gun. The particles (blasting material) jetted out together with the high pressure water impinge on the contaminated surfaces to remove the contaminated layers. Upon impingement, the high pressure water acts as the shock absorber for the blasting material and, after the impingement, it flows down to the bottom of the water chamber, and the blasting material is dissolved in the high pressure water. (Horiuchi, T.)

  17. Void consolidation during open-die forging for ultralarge rotor shafts. (1. Formulation of void-closing behavior)

    International Nuclear Information System (INIS)

    Ono, Shin-ichi; Minami, Katsuyuki; Ochiai, Tomoyuki; Iwadate, Tadao; Nakata, Shin-ichi.

    1995-01-01

    Open-die forging experiments using different die geometries under hot isothermal conditions and three-dimensional simulations using rigid-plastic finite-element method were performed to formulate a void-closing behavior using only two factors; the integral of hydrostatic stress and the equivalent strain. First, upsetting, side-upsetting and V-shape die cogging of several cylinders with a spherical void at the center are carried out and the information on the void volume reduction is obtained. Seconds, the same forgings, but without voids is treated numerically and the development of stress and strain at the location of voids is investigated. Then, by combining these results, and using regression analysis, it is found that the void volume reduction is expressed as a polynomial function of the two factors. When the polynomial function is used, various forging methods can be evaluated quantitatively in terms of void-closing behavior. Therefore it is beneficial to optimize the forging process for a large rotor shaft. (author)

  18. Steam as turbine blade coolant: Experimental data generation

    Energy Technology Data Exchange (ETDEWEB)

    Wilmsen, B.; Engeda, A.; Lloyd, J.R. [Michigan State Univ., East Lansing, MI (United States)

    1995-10-01

    Steam as a coolant is a possible option to cool blades in high temperature gas turbines. However, to quantify steam as a coolant, there exists practically no experimental data. This work deals with an attempt to generate such data and with the design of an experimental setup used for the purpose. Initially, in order to guide the direction of experiments, a preliminary theoretical and empirical prediction of the expected experimental data is performed and is presented here. This initial analysis also compares the coolant properties of steam and air.

  19. Evaluation of primary coolant leaks and assessment of detection methods

    International Nuclear Information System (INIS)

    Cassette, P.; Giroux, C.; Roche, H.; Seveon, J.J.

    1984-11-01

    A review of French PWR situation concerning primary coolant leaks is presented, including a description of operating technical specifications, of the collecting system of primary coolant leakage into the containment and of the detection methods. It is mainly based on a compilation over three years, 1981 to 1983, of almost all occurred leaks, their natures, causes, consequences and methods used for their detection. By analysing these data it is possible to evaluate the efficiency of the primary coolant leak detection system and the problems raised by the compliance with the criteria defined in the operating technical specifications

  20. Evaluation of primary coolant leaks and assessment of detection methods

    International Nuclear Information System (INIS)

    Cassette, P.; Giroux, C.; Roche, H.; Seveon, J.J.

    1986-01-01

    A review of the French PWR situation concerning primary coolant leaks is presented, including a description of operating technical specifications, of the collecting system of primary coolant leakage into the containment and of the detection methods. It is mainly based on a compilation over three years, 1981 to 1983, of almost all actual leaks, their natures, causes, consequences and methods used for their detection. By analysing these data it is possible to evaluate the efficiency of the primary coolant leak detection system and the problems raised by compliance with the criteria defined in the operating technical specifications

  1. Evaluation of filtration and distillation methods for recycling automotive coolant

    International Nuclear Information System (INIS)

    Randall, P.M.; Gavaskar, A.R.

    1992-01-01

    Government regulations and high waste disposal cost of spent automotive coolant have driven the vehicle maintenance industry to explore on-site recycling. The USEPA in cooperation with the New Jersey Department of Environmental Protection (NJDEP) and the New Jersey Department of Transportation (NJDOT) evaluated two commercially available technologies that have potential for reducing the volume of spent automotive coolant. The objective of this study was to evaluate the quality of the recycled coolant, the pollution prevention potential, and the economic feasibility of the technologies

  2. Low-activation lead coolant for advanced small modular NPP

    International Nuclear Information System (INIS)

    Khorasanov, G.L.; Ivanov, A.P.; Blokhin, A.I.

    2001-01-01

    The purpose of the paper is in studying perspectives of a new heavy liquid metal coolant for a small fast reactor (FR) concept. To reduce the post irradiation activity of the coolant the using of lead isotope, Pb-206, instead of natural lead, Pb-nat, is offered. In this case the accumulation of such hazardous radionuclides, as Po-210, Bi-208, Bi-207, essentially decreases. The interval of the lead-206 coolant cost which does not exceed 20% of the overall FR cost is estimated. The possibility of lead-206 obtaining for FR needs with the centrifugal separation technique is pointed out. (author)

  3. KANDY - a numerical model to describe phenomena, which - in a heated and voided fuel element of an LMFBR - may occur

    International Nuclear Information System (INIS)

    Thurnay, K.

    1984-02-01

    Kandy is a model developed to describe the essential destructionphenomena of the fuel elements of an LMFBR. The fuel element is assumed to be a voided one, in which the heat generation is still going on. The main process to be modeled is the melting/bursting/evaporating of parts of the fuel pins and the subsequent dislocation of these materials in the coolant channel. The work presented summarizes the assumptions constituting the model, develops the corresponding equations of motion and describes the procedure, turning these into a system of difference-equations ready for coding. As a final part results of a testcase calculation with the Kandy-code are presentend and interpreted. (orig.) [de

  4. Analysis of Differences in Void Coefficient Predictions for Mixed-Oxide-Fueled Tight-Pitch Light Water Reactor Cells

    International Nuclear Information System (INIS)

    Unesaki, Hironobu; Shiroya, Seiji; Kanda, Keiji; Cathalau, Stephane; Carre, Franck-Olivier; Aizawa, Otohiko; Takeda, Toshikazu

    2000-01-01

    Analysis of the benchmark problems on the void coefficient of mixed-oxide (MOX)-fueled tight-pitch cells has been performed using the Japanese SRAC code system with the JENDL-3.2 library and the French APOLLO-2 code with the CEA93 library based on JEF-2.2. The benchmark problems have been specified to investigate the physical phenomena occurring during the progressive voidage of MOX-fueled tight-pitch lattices, such as high conversion light water reactor lattices, and to evaluate the impact of nuclear data and calculational methods. Despite the most recently compiled nuclear data libraries and the sophisticated calculation schemes employed in both code systems, the k ∞ and void reactivity values obtained by the two code systems show considerable discrepancy especially in the highly voided state. The discrepancy of k ∞ values shows an obvious dependence on void fraction and also has been shown to be sensitive to the isotopic composition of plutonium. The observed discrepancies are analyzed by being decomposed into contributing isotopes and reactions and have been shown to be caused by a complicated balance of both negative and positive components, which are mainly attributable to differences in a limited number of isotopes including 239 Pu, 241 Pu, 16 O, and stainless steel

  5. Void formation in ODS EUROFER produced by hot isostatic pressing

    International Nuclear Information System (INIS)

    Ortega, Y.; Monge, M.A.; Castro, V. de; Munoz, A.; Leguey, T.; Pareja, R.

    2009-01-01

    Positron annihilation experiments were performed on oxide dispersion strengthened (ODS) and non-ODS EUROFER prepared by mechanical alloying and hot isostatic pressing. The results revealed the presence of small voids in these materials in the as-HIPed conditions. Their evolution under isochronal annealing experiments was investigated. The coincidence Doppler broadening spectra of ODS EUROFER exhibited a characteristic signature attributed to positron annihilation in Ar-decorated voids at the oxide particle/matrix interfaces. The variation of the positron annihilation parameters with the annealing temperature showed three stages: up to 623 K, between 823 and 1323 K, and above 1323 K. In the temperature range 823-1323 K void coarsening had effect. Above 1323 K some voids annealed out, but others, associated to oxide particles and small precipitates, survived to annealing at 1523 K. Transmission electron microscopy observations were also performed to verify the characteristics of the surviving defects after annealing at 1523 K.

  6. Void formation in ODS EUROFER produced by hot isostatic pressing

    Energy Technology Data Exchange (ETDEWEB)

    Ortega, Y. [Departamento de Fisica, Universidad Carlos III de Madrid, 28911 Leganes (Spain)], E-mail: yanicet@fis.ucm.es; Monge, M.A. [Departamento de Fisica, Universidad Carlos III de Madrid, 28911 Leganes (Spain); Castro, V. de [Department of Materials, University of Oxford, Oxford OX1 3PH (United Kingdom); Munoz, A.; Leguey, T.; Pareja, R. [Departamento de Fisica, Universidad Carlos III de Madrid, 28911 Leganes (Spain)

    2009-04-30

    Positron annihilation experiments were performed on oxide dispersion strengthened (ODS) and non-ODS EUROFER prepared by mechanical alloying and hot isostatic pressing. The results revealed the presence of small voids in these materials in the as-HIPed conditions. Their evolution under isochronal annealing experiments was investigated. The coincidence Doppler broadening spectra of ODS EUROFER exhibited a characteristic signature attributed to positron annihilation in Ar-decorated voids at the oxide particle/matrix interfaces. The variation of the positron annihilation parameters with the annealing temperature showed three stages: up to 623 K, between 823 and 1323 K, and above 1323 K. In the temperature range 823-1323 K void coarsening had effect. Above 1323 K some voids annealed out, but others, associated to oxide particles and small precipitates, survived to annealing at 1523 K. Transmission electron microscopy observations were also performed to verify the characteristics of the surviving defects after annealing at 1523 K.

  7. Comment on theories for helium-assisted void nucleation

    International Nuclear Information System (INIS)

    Russell, K.C.

    1976-01-01

    Voids form by agglomeration of irradiation-induced vacancies which remain after preferential absorption of self interstitials at dislocation lines. Helium which is formed by (n,α) transmutations and, in simulation studies, may be ion-implanted, often plays an important, but puzzling role. In some materials, very few voids form in the absence of helium, even after intense irradiation. In many other materials , voids form readily under a variety of irradiation conditions, even in the absence of helium. Why some materials require helium - typically in the 10 -6 apa (atom per atom) range - and others do not, and the reason for that particular level are by no means clear. The physics of void nucleation, particularly the role of helium, have been the subject of several theoretical papers. This note presents a critique of these theories, and then briefly outlines a new analysis which is not subject to their limitations. (Auth.)

  8. A variational void coalescence model for ductile metals

    KAUST Repository

    Siddiq, Amir

    2011-08-17

    We present a variational void coalescence model that includes all the essential ingredients of failure in ductile porous metals. The model is an extension of the variational void growth model by Weinberg et al. (Comput Mech 37:142-152, 2006). The extended model contains all the deformation phases in ductile porous materials, i.e. elastic deformation, plastic deformation including deviatoric and volumetric (void growth) plasticity followed by damage initiation and evolution due to void coalescence. Parametric studies have been performed to assess the model\\'s dependence on the different input parameters. The model is then validated against uniaxial loading experiments for different materials. We finally show the model\\'s ability to predict the damage mechanisms and fracture surface profile of a notched round bar under tension as observed in experiments. © Springer-Verlag 2011.

  9. (100) faceted anion voids in electron irradiated fluorite

    International Nuclear Information System (INIS)

    Johnson, E.

    1979-01-01

    High fluence electron irradiation of fluorite crystals in the temperature range 150 to 320 K results in formation of a simple cubic anion void superlattice. Above 320 K the damage structure changes to a random distribution of large [001] faceted anion voids. This voidage behaviour, similar to that observed in a range of irradiated metals, is discussed in terms points defect rather than conventional colour centre terminology. (Auth.)

  10. Uroflowmetry in neurologically normal children with voiding disorders

    DEFF Research Database (Denmark)

    Jensen, K M; Nielsen, K.K.; Kristensen, E S

    1985-01-01

    of neurological deficits underwent a complete diagnostic program including intravenous urography, voiding cystography and cystoscopy as well as spontaneous uroflowmetry, cystometry-emg and pressure-flow-emg study. The incidence of dyssynergia was 22%. However, neither the flow curve pattern nor single flow...... variables were able to identify children with dyssynergia. Consequently uroflowmetry seems inefficient in the screening for dyssynergia in neurological normal children with voiding disorders in the absence of anatomical bladder outlet obstruction....

  11. Effect of Dark Energy Perturbation on Cosmic Voids Formation

    Science.gov (United States)

    Endo, Takao; Nishizawa, Atsushi J.; Ichiki, Kiyotomo

    2018-05-01

    In this paper, we present the effects of dark energy perturbation on the formation and abundance of cosmic voids. We consider dark energy to be a fluid with a negative pressure characterised by a constant equation of state w and speed of sound c_s^2. By solving fluid equations for two components, namely, dark matter and dark energy fluids, we quantify the effects of dark energy perturbation on the sizes of top-hat voids. We also explore the effects on the size distribution of voids based on the excursion set theory. We confirm that dark energy perturbation negligibly affects the size evolution of voids; c_s^2=0 varies the size only by 0.1% as compared to the homogeneous dark energy model. We also confirm that dark energy perturbation suppresses the void size when w -1 (Basse et al. 2011). In contrast to the negligible impact on the size, we find that the size distribution function on scales larger than 10 Mpc/h highly depends on dark energy perturbation; compared to the homogeneous dark energy model, the number of large voids of radius 30Mpc is 25% larger for the model with w = -0.9 and c_s^2=0 while they are 20% less abundant for the model with w = -1.3 and c_s^2=0.

  12. Improving Coolant Effectiveness through Drill Design Optimization in Gundrilling

    Science.gov (United States)

    Woon, K. S.; Tnay, G. L.; Rahman, M.

    2018-05-01

    Effective coolant application is essential to prevent thermo-mechanical failures of gun drills. This paper presents a novel study that enhances coolant effectiveness in evacuating chips from the cutting zone using a computational fluid dynamic (CFD) method. Drag coefficients and transport behaviour over a wide range of Reynold numbers were first established through a series of vertical drop tests. With these, a CFD model was then developed and calibrated with a set of horizontal drilling tests. Using this CFD model, critical drill geometries that lead to poor chip evacuation including the nose grind contour, coolant hole configuration and shoulder dub-off angle in commercial gun drills are identified. From this study, a new design that consists a 20° inner edge, 15° outer edge, 0° shoulder dub-off and kidney-shaped coolant channel is proposed and experimentally proven to be more superior than all other commercial designs.

  13. Transient behaviour of main coolant pump in nuclear power plants

    International Nuclear Information System (INIS)

    Delja, A.

    1986-01-01

    A basic concept of PWR reactor coolant pump thermo-hydraulic modelling in transient and accident operational condition is presented. The reactor coolant pump is a component of the nuclear steam supply system which forces the coolant through the reactor and steam generator, maintaining design heat transfer condition. The pump operating conditions have strong influence on the flow and thermal behaviour of NSSS, both in the stationary and nonstationary conditions. A mathematical model of the reactor coolant pump is formed by using dimensionless homologous relations in the four-quadrant regimes: normal pump, turbine, dissipation and reversed flow. Since in some operational regimes flow of mixture, liquid and steam may occur, the model has additional correction members for two-phase homologous relations. Modular concept has been used in developing computer program. The verification is performed on the simulation loss of offsite power transient and obtained results are presented. (author)

  14. Flow rate control systems for coolants for BWR type reactors

    International Nuclear Information System (INIS)

    Igarashi, Yoko; Kato, Naoyoshi.

    1981-01-01

    Purpose: To increase spontaneous recycling flow rate of coolants in BWR type reactors when the water level in the reactor decreases, by communicating a downcomer with a lower plenum. Constitution: An opening is provided to the back plate disposed at the lower end of a reactor core shroud for communicating a downcomer with a lower plenum, and an ON-OFF valve actuated by an operation rod is provided to the opening. When abnormal water level or pressure in the reactor is detected by a level metal or pressure meter, the operation rod is driven to open the ON-OFF valve, whereby coolants fed from a jet pump partially flows through the opening to increase the spontaneous recycling flow rate of the coolants. This can increase the spontaneous recycling flow rate of the coolants upon spontaneous recycling operation, thereby maintaining the reactor safety and the fuel soundness. (Moriyama, K.)

  15. Multidimensional analysis of fluid flow in the loft cold leg blowdown pipe during a loss-of-coolant experiment

    International Nuclear Information System (INIS)

    Demmie, P.N.; Hofmann, K.R.

    1979-03-01

    A computer analysis of fluid flow in the Loss-of-Fluid Test (LOFT) cold leg blowdown pipe during a loss-of-coolant experiment (LOCE) was performed using the computer program K-FIX/MOD1. The purpose of this analysis was to evaluate the capability of K-FIX/MOD1 to calculate theoretical fluid quantity distributions in the blowdown pipe during a LOCE for possible application to the analysis of LOFT experimental data, the determination of mass flow, or the development of data reduction models. A rectangular section of a portion of the LOFT blowdown pipe containing measurement Station BL-1 was modeled using time-dependent boundary conditions. Fluid quantities were calculated during a simulation of the first 26 s of LOFT LOCE L1-4. Sensitivity studies were made to determine changes in void fractions and velocities resulting from specific changes in the inflow boundary conditions used for this simulation

  16. In reactor performance of defected zircaloy-clad U3Si fuel elements in pressurized and boiling water coolants

    International Nuclear Information System (INIS)

    Feraday, M.A.; Allison, G.M.; Ambler, J.F.R.; Chalder, G.H.; Lipsett, J.J.

    1968-05-01

    The results of two in-reactor defect tests of Zircaloy-clad U 3 Si are reported. In the first test, a previously irradiated element (∼5300 MWd/ tonne U) was defected then exposed to first pressurized water then boiling water at ∼270 o C. In the second test, an unirradiated element containing a central void was defected, waterlogged, then exposed to pressurized water for 50 minutes. Both tests were terminated because of high activity in the loop coolant detected by both gamma and delayed neutron monitors. Post-irradiation examination showed that both elements had suffered major sheath failures which were attributed to the volume increase accompanying the formation of large quantities of corrosion product formed by the reaction of water with the hot central part of the fuel. It was concluded that the corrosion resistance of U 3 Si at 300 o C is not seriously affected by irradiation, but the corrosion rate increases rapidly with temperature. (author)

  17. Simulation of a loss of coolant accident

    International Nuclear Information System (INIS)

    1987-06-01

    An essential component of nuclear safety activities is the analysis of postulated accidents which are taken as a design basis for a facility. This analysis is usually carried out by using complex computer codes to simulate the behaviour of the plant and to calculate vital plant parameters, which are then compared with the design limits. Since these simulations cannot be verified at the plant itself, computer codes must be validated by comparing the results of calculations with experimental data obtained in test facilities. With this objective in mind, the Central Research Institute for Physics (CRIP) of the Hungarian Academy of Sciences designed and constructed the PMK-NVH (Paks Model Circuit) test facility, a scaled down model of the WWER-440 Paks nuclear power plant. Hungary with the aim of strengthening the international co-operation on nuclear safety, made the PMK-NVH facility available to the IAEA to conduct a standard problem exercise. In this exercise, experimental data from the simulation of a 7.4% break loss of coolant accident were compared with analytical predictions of the behaviour of the facility calculated with computer codes. This document presents a complete overview of the Standard Problem Exercise, including description of the facility, the experiment, the codes and models used by the participants and a detailed intercomparison of calculated and experimental results. It is recognized that code assessment is a long process which involves many inter-related steps, therefore, no general conclusion on optimum code or best model was reached. However, the exercise was recognized as an important contributor to code validation

  18. Fusion-reactor blanket and coolant material compatibility

    International Nuclear Information System (INIS)

    Jeppson, D.W.; Keough, R.F.

    1981-01-01

    Fusion reactor blanket and coolant compatibility tests are being conducted to aid in the selection and design of safe blanket and coolant systems for future fusion reactors. Results of scoping compatibility tests to date are reported for blanket material and water interactions at near operating temperatures. These tests indicate the quantitative hydrogen release, the maximum temperature and pressures produced and the rates of interactions for selected blanket materials

  19. Review on research of small break loss of coolant accident

    International Nuclear Information System (INIS)

    Bo Jinhai; Wang Fei

    1998-01-01

    The Small Break Loss of Coolant Accident (SBLOCA) and its research art-of -work are reviewed. A typical SBLOCA process in Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) and the influence of break size, break location and reactor coolant pump on the process are described. The existing papers are classified in two categories: experimental and numerical modeling, with the primary experimental apparatuses in the world listed and the research works on SBLOCA summarized

  20. Comparative analysis of coolants for FBR of future nuclear power

    International Nuclear Information System (INIS)

    Toshinsky, G.I.; Grigoryev, O.G.; Pylchenkov, E.H.; Skorikov, D.E.; Komkova, O.I.

    2001-01-01

    Selection of a fast reactor (FR) coolant for future nuclear reactors is a complex task that has not a single solution. Safety requirements are expected to grow in the future. The requirements to FR are reconsidered. Gradual transition from the FR as a builder up of plutonium to the FR as an economically effective energy source, is taking place. Among all types of coolants viable for FR, LMC (light molten salt coolants) cover the most complete range of requirements to advanced reactors and have a complete database. Sodium and lead-bismuth coolant (LBC) are selected because there is a complete package of technologies for their handling. Heavy liquid metal coolant (HLMC), being at a disadvantage of heat transfer rate in relation to sodium, make it possible to give the inherent safety properties to the reactor and, as a result, to simplify essentially the reactor design and its safety systems. This results in capital and costs reduction. Neutronic characteristics of HLMC cooled reactors make possible to transmute their own minor actinides (MA) safely, and LBC cooled reactors are able to transmute LWR'MA with high safety characteristics. Basing on the comparison carried out, it can be concluded, that both LBC and sodium are perspective coolants for future FR

  1. Comparative analysis of coolants for FBR of future nuclear power

    Energy Technology Data Exchange (ETDEWEB)

    Toshinsky, G.I.; Grigoryev, O.G.; Pylchenkov, E.H.; Skorikov, D.E.; Komkova, O.I. [State Scientific Center of Russian Federation, Institute for Physics and Power Engineering named after Academician A.I. Leipusky, Kaluga Region (Russian Federation)

    2001-07-01

    Selection of a fast reactor (FR) coolant for future nuclear reactors is a complex task that has not a single solution. Safety requirements are expected to grow in the future. The requirements to FR are reconsidered. Gradual transition from the FR as a builder up of plutonium to the FR as an economically effective energy source, is taking place. Among all types of coolants viable for FR, LMC (light molten salt coolants) cover the most complete range of requirements to advanced reactors and have a complete database. Sodium and lead-bismuth coolant (LBC) are selected because there is a complete package of technologies for their handling. Heavy liquid metal coolant (HLMC), being at a disadvantage of heat transfer rate in relation to sodium, make it possible to give the inherent safety properties to the reactor and, as a result, to simplify essentially the reactor design and its safety systems. This results in capital and costs reduction. Neutronic characteristics of HLMC cooled reactors make possible to transmute their own minor actinides (MA) safely, and LBC cooled reactors are able to transmute LWR'MA with high safety characteristics. Basing on the comparison carried out, it can be concluded, that both LBC and sodium are perspective coolants for future FR.

  2. Urinary tract infection after voiding cystourethrogram.

    Science.gov (United States)

    Johnson, E K; Malhotra, N R; Shannon, R; Jacobson, D L; Green, J; Rigsby, C K; Holl, J L; Cheng, E Y

    2017-08-01

    Reported rates of post-procedural urinary tract infection (ppUTI) after voiding cystourethrogram (VCUG) are highly variable (0-42%). This study aimed to determine the risk of ppUTI after cystogram, and evaluate predictors of ppUTI. A retrospective cohort study of children undergoing VCUG or radionuclide cystogram (henceforth 'cystogram') was conducted. Children with neurogenic bladder who underwent cystogram in the operating room and without follow-up at the study institution were excluded. Incidence of symptomatic ppUTI within 7 days after cystogram was recorded. Predictors of ppUTI were evaluated using univariate statistics. A total of 1108 children (54% female, median age 1.1 years) underwent 1203 cystograms: 51% were on periprocedural antibiotics, 75% had a pre-existing urologic diagnosis (i.e., vesicoureteral reflux (VUR) or hydronephrosis; not UTI alone), and 18% had a clinical UTI within 30 days before cystogram. Of the cystograms, 41% had an abnormal cystogram and findings included VUR (82%), ureterocele (6%), and diverticula (6%). Twelve children had a ppUTI (1.0%; four girls, five uncircumcised boys, three circumcised boys; median age 0.9 years). Factors significantly associated with diagnosis of a ppUTI (Summary fig.) included: pre-existing urologic diagnosis prior to cystogram (12/12, 100% of patients with ppUTI), abnormal cystogram results (11/12, 92%), and use of periprocedural antibiotics (11/12, 92%). All 11 children with an abnormal cystogram had VUR ≥ Grade III. However, among all children with VUR ≥ Grade III, 4% (11/254) had a ppUTI. This is the largest study to date that has examined incidence and risk factors for ppUTI after cystogram. The retrospective nature of the study limited capture of some clinical details. This study demonstrated that the risk of ppUTI after a cystogram is very low (1.0% in this cohort). Having a pre-existing urologic diagnosis such as VUR or hydronephrosis was associated with ppUTI; therefore, children with

  3. Reactive Arthritis

    Directory of Open Access Journals (Sweden)

    Eren Erken

    2013-06-01

    Full Text Available Reactive arthritis is an acute, sterile, non-suppurative and inflammatory arthropaty which has occured as a result of an infectious processes, mostly after gastrointestinal and genitourinary tract infections. Reiter syndrome is a frequent type of reactive arthritis. Both reactive arthritis and Reiter syndrome belong to the group of seronegative spondyloarthropathies, associated with HLA-B27 positivity and characterized by ongoing inflammation after an infectious episode. The classical triad of Reiter syndrome is defined as arthritis, conjuctivitis and urethritis and is seen only in one third of patients with Reiter syndrome. Recently, seronegative asymmetric arthritis and typical extraarticular involvement are thought to be adequate for the diagnosis. However, there is no established criteria for the diagnosis of reactive arthritis and the number of randomized and controlled studies about the therapy is not enough. [Archives Medical Review Journal 2013; 22(3.000: 283-299

  4. Reliability Impact of Stockpile Aging: Stress Voiding; TOPICAL

    International Nuclear Information System (INIS)

    ROBINSON, DAVID G.

    1999-01-01

    The objective of this research is to statistically characterize the aging of integrated circuit interconnects. This report supersedes the stress void aging characterization presented in SAND99-0975, ''Reliability Degradation Due to Stockpile Aging,'' by the same author. The physics of the stress voiding, before and after wafer processing have been recently characterized by F. G. Yost in SAND99-0601, ''Stress Voiding during Wafer Processing''. The current effort extends this research to account for uncertainties in grain size, storage temperature, void spacing and initial residual stress and their impact on interconnect failure after wafer processing. The sensitivity of the life estimates to these uncertainties is also investigated. Various methods for characterizing the probability of failure of a conductor line were investigated including: Latin hypercube sampling (LHS), quasi-Monte Carlo sampling (qMC), as well as various analytical methods such as the advanced mean value (Ah/IV) method. The comparison was aided by the use of the Cassandra uncertainty analysis library. It was found that the only viable uncertainty analysis methods were those based on either LHS or quasi-Monte Carlo sampling. Analytical methods such as AMV could not be applied due to the nature of the stress voiding problem. The qMC method was chosen since it provided smaller estimation error for a given number of samples. The preliminary results indicate that the reliability of integrated circuits due to stress voiding is very sensitive to the underlying uncertainties associated with grain size and void spacing. In particular, accurate characterization of IC reliability depends heavily on not only the frost and second moments of the uncertainty distribution, but more specifically the unique form of the underlying distribution

  5. OECD/DOE/CEA VVER-1000 Coolant Transient Benchmark. Summary Record of the Third Workshop (V1000-CT3)

    International Nuclear Information System (INIS)

    2005-01-01

    The overall objective of the VVER-1000 coolant transient (V1000CT) benchmark is to assess computer codes used in the safety analysis of VVER power plants, specifically for their use in analysis of reactivity transients in a VVER-1000. The V1000CT benchmark consists of two phases: V1000CT-1 is a simulation of the switching on of one main coolant pump (MCP) when the other three MCPs are in operation, and V1000CT-2 concerns calculation of coolant mixing tests and main steam line break (MSLB) scenarios. Each of the two phases contains three exercises. The reference problem chosen for simulation in Phase 1 is a MCP switching on when the other three main coolant pumps are in operation in a VVER-1000. This event is characterized by rapid increase in the flow through the core resulting in a coolant temperature decrease, which is spatially dependent. This leads to insertion of spatially distributed positive reactivity due to the modelled feedback mechanisms and non-symmetric power distribution. Simulation of the transient requires evaluation of core response from a multi-dimensional perspective (coupled three-dimensional neutronics/core thermal-hydraulics) supplemented by a one-dimensional simulation of the remainder of the reactor coolant system. Three exercises are defined in the framework of Phase 1: a) Exercise 1 - Point kinetics plant simulation; b) Exercise 2 - Coupled 3-D neutronics/core thermal-hydraulics response evaluation; c) Exercise 3 - Best-estimate coupled 3-D core/plant system transient modelling. In addition to the measured (experiment) scenario, extreme calculation scenarios were defined in the frame of Exercise 3 for better testing 3-D neutronics/thermal-hydraulics techniques. The proposals concerned: rod ejection simulations with scram set points at two different power levels. The technical topics presented at this workshop were: Review of the benchmark activities after the 2. Workshop; - Discussion of participant's feedback and introduced modifications

  6. Breaking the vicious circle: Onabotulinum toxin A in children with therapy-refractory dysfunctional voiding

    NARCIS (Netherlands)

    L.A. 't Hoen (Lisette); J. van den Hoek (Joop); K.P. Wolffenbuttel (Katja); F. van der Toorn; J.R. Scheepe (Jeroen)

    2015-01-01

    textabstractIntroduction An increased activity of the external urethral sphincter or pelvic floor muscles during voluntary voiding leads to dysfunctional voiding. Frequently reported symptoms are urinary incontinence, urinary tract infections and high post-void residuals. Dysfunctional voiding is a

  7. Automated air-void system characterization of hardened concrete: Helping computers to count air-voids like people count air-voids---Methods for flatbed scanner calibration

    Science.gov (United States)

    Peterson, Karl

    Since the discovery in the late 1930s that air entrainment can improve the durability of concrete, it has been important for people to know the quantity, spacial distribution, and size distribution of the air-voids in their concrete mixes in order to ensure a durable final product. The task of air-void system characterization has fallen on the microscopist, who, according to a standard test method laid forth by the American Society of Testing and Materials, must meticulously count or measure about a thousand air-voids per sample as exposed on a cut and polished cross-section of concrete. The equipment used to perform this task has traditionally included a stereomicroscope, a mechanical stage, and a tally counter. Over the past 30 years, with the availability of computers and digital imaging, automated methods have been introduced to perform the same task, but using the same basic equipment. The method described here replaces the microscope and mechanical stage with an ordinary flatbed desktop scanner, and replaces the microscopist and tally counter with a personal computer; two pieces of equipment much more readily available than a microscope with a mechanical stage, and certainly easier to find than a person willing to sit for extended periods of time counting air-voids. Most laboratories that perform air-void system characterization typically have cabinets full of prepared samples with corresponding results from manual operators. Proponents of automated methods often take advantage of this fact by analyzing the same samples and comparing the results. A similar iterative approach is described here where scanned images collected from a significant number of samples are analyzed, the results compared to those of the manual operator, and the settings optimized to best approximate the results of the manual operator. The results of this calibration procedure are compared to an alternative calibration procedure based on the more rigorous digital image accuracy

  8. Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

    Science.gov (United States)

    Jones, Robert D.

    1978-01-01

    A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level.

  9. Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Jones, R.D.

    1978-01-01

    A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level

  10. An investigation on the material effect on the result of fuel coolant interactions in the TROI experiments

    International Nuclear Information System (INIS)

    Park, I. K.; Kim, J. H.; Min, B. T.; Hong, S. W.

    2008-01-01

    One of the findings from the TROI experiments is that the results of the fuel coolant interaction (FCI) are strongly dependent on the composition of the corium, which is composed of UO 2 , ZrO 2 , Zr, steel. TEXAS- V simulation for the TROI experiments indicated that a relatively low void fraction seems to have resulted in a strong steam explosion and the low voided mixture must be induced by big size particles. The particle sizes of the non-explosive TROI tests were analyzed because the explosive tests do not represent the particles during mixing. It indicates that the debris size seems to reflect the material difference, and the trend is the same as the debris size in the TEXAS-V simulation. TEXAS-V calculation for the alumina/water system indicates that the conductivity is also related to the material effect on the FCI result. The heat loss evaluation using a single sphere film boiling model shows that a reasonable conductivity and particle size give a reliable estimation for the FCI result. Thus reliable values for the physical properties such as the surface tension and a better understanding for the breakup process would be necessary for a more convincible nuclear safety analysis. (authors)

  11. ON THE STAR FORMATION PROPERTIES OF VOID GALAXIES

    Energy Technology Data Exchange (ETDEWEB)

    Moorman, Crystal M.; Moreno, Jackeline; White, Amanda; Vogeley, Michael S. [Department of Physics, Drexel University, 3141 Chestnut Street, Philadelphia, PA 19104 (United States); Hoyle, Fiona [Pontifica Universidad Catolica de Ecuador, 12 de Octubre 1076 y Roca, Quito (Ecuador); Giovanelli, Riccardo; Haynes, Martha P., E-mail: crystal.m.moorman@drexel.edu [Center for Radiophysics and Space Research, Space Sciences Building, Cornell University Ithaca, NY 14853 (United States)

    2016-11-10

    We measure the star formation properties of two large samples of galaxies from the SDSS in large-scale cosmic voids on timescales of 10 and 100 Myr, using H α emission line strengths and GALEX FUV fluxes, respectively. The first sample consists of 109,818 optically selected galaxies. We find that void galaxies in this sample have higher specific star formation rates (SSFRs; star formation rates per unit stellar mass) than similar stellar mass galaxies in denser regions. The second sample is a subset of the optically selected sample containing 8070 galaxies with reliable H i detections from ALFALFA. For the full H i detected sample, SSFRs do not vary systematically with large-scale environment. However, investigating only the H i detected dwarf galaxies reveals a trend toward higher SSFRs in voids. Furthermore, we estimate the star formation rate per unit H i mass (known as the star formation efficiency; SFE) of a galaxy, as a function of environment. For the overall H i detected population, we notice no environmental dependence. Limiting the sample to dwarf galaxies still does not reveal a statistically significant difference between SFEs in voids versus walls. These results suggest that void environments, on average, provide a nurturing environment for dwarf galaxy evolution allowing for higher specific star formation rates while forming stars with similar efficiencies to those in walls.

  12. Electromigration of intergranular voids in metal films for microelectronic interconnects

    CERN Document Server

    Averbuch, A; Ravve, I

    2003-01-01

    Voids and cracks often occur in the interconnect lines of microelectronic devices. They increase the resistance of the circuits and may even lead to a fatal failure. Voids may occur inside a single grain, but often they appear on the boundary between two grains. In this work, we model and analyze numerically the migration and evolution of an intergranular void subjected to surface diffusion forces and external voltage applied to the interconnect. The grain-void interface is considered one-dimensional, and the physical formulation of the electromigration and diffusion model results in two coupled fourth-order one-dimensional time-dependent PDEs. The boundary conditions are specified at the triple points, which are common to both neighboring grains and the void. The solution of these equations uses a finite difference scheme in space and a Runge-Kutta integration scheme in time, and is also coupled to the solution of a static Laplace equation describing the voltage distribution throughout the grain. Since the v...

  13. How institutional voids influence Brazilian foreign direct investment in Angola

    Directory of Open Access Journals (Sweden)

    Renato Virches

    2017-04-01

    Full Text Available How do institutional voids influence emerging market multinationals (EMNEs foreign direct investment (FDI in developing countries? In this article we respond to this question by examining Brazilian FDI in Angola as our analytical setting. We focus on the host country’s institutions and its institutional voids as essential factors that attract the FDI of EMNES to developing countries. The research indicates that Brazilian companies fill in much of these voids within the market intermediaries, often creating a point of competitive advantage, and also creating advantages in relation to FDI from other economies that invest in Angola. The scarce literature on FDI in Africa has been largely dedicated to the analysis of Chinese investment in the region. We aim to complement recent research on the influence of the host country’s institutions on the behavior of FDI in developing countries, explaining how some EMNEs are able to use the institutional voids of developing countries as market opportunities. Our findings should provide also implications for EMNEs managers from other emerging markets by providing a better understanding of how Brazilian multinationals expand their business in less developed countries, handle institutional voids and manage relationships with local and foreign institutions in the host country.

  14. Understanding void fraction in steady state and dynamic environments

    International Nuclear Information System (INIS)

    Chexal, B.; Maulbetsch, J.; Harrison, J.; Petersen, C.; Jensen, P.; Horowitz, J.

    1997-01-01

    Understanding void fraction behavior in steady-state and dynamic environments is important to accurately predict the thermal-hydraulic behavior of two-phase or two-component systems. The Chexal-Lellouche (C-L) void fraction mode described herein covers the full range of pressures, flows, void fractions, and fluid types (steam-water, air-water, and refrigerants). A drift flux model formulation is used which covers the complete range of concurrent and countercurrent flows. The (1996) model revises the earlier C-L void fraction correlation, improves the capability of the model in countercurrent flow based on the incorporation of additional data, and improves the characteristics of the correlation that are important in transient programs. The model has been qualified with data from a number of steady state two-phase and two-component tests, and has been incorporated into the transient analysis code RELAP5 and RETRAN-3D and evaluated with a variety of transient and steady state tests. A 'plug-in' module for the void fraction correlation has been developed and implemented in RELAP5 and RETRAN-3D. The module is available as source code for inclusion into other thermal-hydraulic programs and can be used in any program that utilizes the same interface variables

  15. Effect of voids-controlled vacancy supersaturations on B diffusion

    International Nuclear Information System (INIS)

    Marcelot, O.; Claverie, A.; Cristiano, F.; Cayrel, F.; Alquier, D.; Lerch, W.; Paul, S.; Rubin, L.; Jaouen, H.; Armand, C.

    2007-01-01

    We present here preliminary results on boron diffusion in presence of pre-formed voids of different characteristics. The voids were fabricated by helium implantation followed by annealing allowing the desorption of He prior to boron implantation. We show that under such conditions boron diffusion is always largely reduced and can even be suppressed in some cases. Boron diffusion suppression can be observed in samples not containing nanovoids in the boron-rich region. It is suggested that direct trapping of Si(int)s by the voids is not the mechanism responsible for the reduction of boron diffusion in such layers. Alternatively, our experimental results suggest that this reduction of diffusivity is more probably due to the competition between two Ostwald ripening phenomena taking place at the same time: in the boron-rich region, the competitive growth of extrinsic defects at the origin of TED and, in the void region, the Ostwald ripening of the voids which involves large supersaturations of Vs

  16. Effect of voids-controlled vacancy supersaturations on B diffusion

    Energy Technology Data Exchange (ETDEWEB)

    Marcelot, O. [CEMES/CNRS, 29 rue Jeanne Marvig, 31055 Toulouse (France)]. E-mail: marcelot@cemes.fr; Claverie, A. [CEMES/CNRS, 29 rue Jeanne Marvig, 31055 Toulouse (France); Cristiano, F. [LAAS/CNRS, 7 av. du Col. Roche, 31077 Toulouse (France); Cayrel, F. [LMP, Universite de Tours, 16 rue Pierre et Marie Curie, BP 7155, 37071 Tours (France); Alquier, D. [LMP, Universite de Tours, 16 rue Pierre et Marie Curie, BP 7155, 37071 Tours (France); Lerch, W. [Mattson Thermal Products GmbH, Daimlerstr. 10, D-89160 Dornstadt (Germany); Paul, S. [Mattson Thermal Products GmbH, Daimlerstr. 10, D-89160 Dornstadt (Germany); Rubin, L. [Axcelis Technologies, 108 Cherry Hill Drive, Beverly MA 01915 (United States); Jaouen, H. [STMicroelectronics, 850 rue Jean Monnet, 38926 Crolles (France); Armand, C. [LNMO/INSA, Service analyseur ionique, 135 av. de Rangueil, 31077 Toulouse (France)

    2007-04-15

    We present here preliminary results on boron diffusion in presence of pre-formed voids of different characteristics. The voids were fabricated by helium implantation followed by annealing allowing the desorption of He prior to boron implantation. We show that under such conditions boron diffusion is always largely reduced and can even be suppressed in some cases. Boron diffusion suppression can be observed in samples not containing nanovoids in the boron-rich region. It is suggested that direct trapping of Si(int)s by the voids is not the mechanism responsible for the reduction of boron diffusion in such layers. Alternatively, our experimental results suggest that this reduction of diffusivity is more probably due to the competition between two Ostwald ripening phenomena taking place at the same time: in the boron-rich region, the competitive growth of extrinsic defects at the origin of TED and, in the void region, the Ostwald ripening of the voids which involves large supersaturations of Vs.

  17. Controlling Interfacial Separation in Porous Structures by Void Patterning

    Science.gov (United States)

    Ghareeb, Ahmed; Elbanna, Ahmed

    Manipulating interfacial response for enhanced adhesion or fracture resistance is a problem of great interest to scientists and engineers. In many natural materials and engineering applications, an interface exists between a porous structure and a substrate. A question that arises is how the void distribution in the bulk may affect the interfacial response and whether it is possible to alter the interfacial toughness without changing the surface physical chemistry. In this paper, we address this question by studying the effect of patterning voids on the interfacial-to-the overall response of an elastic plate glued to a rigid substrate by bilinear cohesive material. Different patterning categories are investigated; uniform, graded, and binary voids. Each case is subjected to upward displacement at the upper edge of the plate. We show that the peak force and maximum elongation at failure depend on the voids design and by changing the void size, alignment or gradation we may control these performance measures. We relate these changes in the measured force displacement response to energy release rate as a measure of interfacial toughness. We discuss the implications of our results on design of bulk heterogeneities for enhanced interfacial behavior.

  18. Critical Void Volume Fraction fc at Void Coalescence for S235JR Steel at Low Initial Stress Triaxiality

    Science.gov (United States)

    Grzegorz Kossakowski, Paweł; Wciślik, Wiktor

    2017-10-01

    The paper is concerned with the nucleation, growth and coalescence of microdefects in the form of voids in S235JR steel. The material is known to be one of the basic steel grades commonly used in the construction industry. The theory and methods of damage mechanics were applied to determine and describe the failure mechanisms that occur when the material undergoes deformation. Until now, engineers have generally employed the Gurson-Tvergaard- Needleman model. This material model based on damage mechanics is well suited to define and analyze failure processes taking place in the microstructure of S235JR steel. It is particularly important to determine the critical void volume fraction fc , which is one of the basic parameters of the Gurson-Tvergaard-Needleman material model. As the critical void volume fraction fc refers to the failure stage, it is determined from the data collected for the void coalescence phase. A case of multi-axial stresses is considered taking into account the effects of spatial stress state. In this study, the parameter of stress triaxiality η was used to describe the failure phenomena. Cylindrical tensile specimens with a circumferential notch were analysed to obtain low values of initial stress triaxiality (η = 0.556 of the range) in order to determine the critical void volume fraction fc . It is essential to emphasize how unique the method applied is and how different it is from the other more common methods involving parameter calibration, i.e. curve-fitting methods. The critical void volume fraction fc at void coalescence was established through digital image analysis of surfaces of S235JR steel, which involved studying real, physical results obtained directly from the material tested.

  19. Primary coolant recycling device for FBR type reactor

    International Nuclear Information System (INIS)

    Kanbe, Mitsuru; Tokiwai, Moriyasu

    1998-01-01

    A primary coolants (liquid sodium) recycling device comprises a plurality of recycling pumps. The recycling pumps are operated while using, as a power source, electric power generated by a thermoelectric power generation system by utilizing heat stored in the coolants. The thermoelectric power generation system comprises a thermo-electric conversion module, heat collecting heat pipes as a high temperature side heat conduction means and heat dissipating pipes as a low temperature side heat conduction means. The heat of coolants is transferred to the surface of the high temperature side of each thermo-electric conversion elements of the thermal power generation system by the heat collecting heat pipes. The heat on the low temperature side of each of the thermo-electric conversion elements is removed by the heat dissipating pipes. Accordingly, temperature difference is caused between both surfaces of the thermo-electric conversion elements. Even upon loss of a main power source due to stoppage of electricity, electric power is generated by utilizing heat of coolants, so that the recycling pumps circulate coolants to cool a reactor core continuously. (I.N.)

  20. Analyses of Decrease in Reactor Coolant Flow Rate in SMART

    International Nuclear Information System (INIS)

    Kim, Hyung Rae; Bae, Kyoo Hwan; Choi, Suhn

    2011-01-01

    SMART is a small integral reactor, which is under development at KAERI to get the standard design approval by the end of 2011. SMART works like a pressurized light-water reactor in principle though it is more compact than large commercial reactors. SMART houses major components such as steam generators, a pressurizer, and reactor coolant pumps inside the reactor pressure vessel. Due to its compact design, SMART adopts a canned-motor type reactor coolant pump which has much smaller rotational inertia than the ones used in commercial reactors. As a consequence, the reactor coolant pump has very short coastdown time and reactor coolant flow rate decreases more severely compared to commercial reactors. The transients initiated by reduction of reactor coolant flow rate have been analyzed to ensure that SMART can be safely shutdown on such transients. The design basis events in this category are complete loss of flow, single pump locked rotor with loss of offsite power, and single pump shaft break with loss of offsite power

  1. Full reactor coolant system chemical decontamination qualification programs

    Energy Technology Data Exchange (ETDEWEB)

    Miller, P.E. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

    1995-03-01

    Corrosion and wear products are found throughout the reactor coolant system (RCS), or primary loop, of a PWR power plant. These products circulate with the primary coolant through the reactor where they may become activated. An oxide layer including these activated products forms on the surfaces of the RCS (including the fuel elements). The amount of radioactivity deposited on the different surface varies and depends primarily on the corrosion rate of the materials concerned, the amount of cobalt in the coolant and the chemistry of the coolant. The oxide layer, commonly called crud, on the surfaces of nuclear plant systems leads to personnel radiation exposure. The level of the radiation fields from the crud increases with time from initial plant startup and typically levels off after 4 to 6 cycles of plant operation. Thereafter, significant personnel radiation exposure may be incurred whenever major maintenance is performed. Personnel exposure is highest during refueling outages when routine maintenance on major plant components, such as steam generators and reactor coolant pumps, is performed. Administrative controls are established at nuclear plants to minimize the exposure incurred by an individual and the plant workers as a whole.

  2. The installation welding of pressure water reactor coolant piping

    International Nuclear Information System (INIS)

    Deng Feng

    2010-01-01

    Large pressure water reactor nuclear power plants are constructing in our country. There are three symmetry standard loops in reactor coolant system. Each loop possesses a steam generator and a primary poop, in which one of the loops is equipped with a pressurizer. These components are connected with reactor pressure vessel by installation welding of the coolant piping. The integrity of reactor coolant pressure boundary is the second barrier to protect the radioactive substance from release to outside, so the safe operation of nuclear power plant is closely related to the quality of coolant piping installation welding. The heavy tube with super low carbon content austenitic stainless steel is selected for coolant piping. This kind of material has good welding behavior, but the poor thermal conductivity, the big liner expansion coefficient and the big welding deformation will cause bigger welding stress. To reduce the welding deformation, to control the dimension precision, to reduce the residual stress and to ensure the welding quality the installation sequence should be properly designed and the welding technology should be properly controlled. (authors)

  3. The relationship between temperament, gender, and childhood dysfunctional voiding.

    Science.gov (United States)

    Colaco, Marc; Dobkin, Roseanne D; Sterling, Matthew; Schneider, Dona; Barone, Joseph

    2013-08-01

    Dysfunctional voiding (DV) is an extremely common pediatric complaint. The goal of this study was to examine the relationship between DV and childhood temperament. Information about the voiding behaviors and temperaments of 50 children was examined using a case-control model. Caregivers were asked to fill out the Children's Behavior Questionnaire in order to rate their child on the dimensions of surgency, negative affect, and effortful control. The relationship between DV and these dimensions was then evaluated. Males with DV were found to have lower effortful control than males with normal voiding habits. Females with DV did not demonstrate a difference in effortful control, but did demonstrate a higher rate of surgency. The results suggest that temperament does have an association with DV. These findings are in line with temperamental associations with other externalizing trouble behaviors and may inform potential treatment strategies for DV.

  4. Prediction of pool void fraction by new drift flux correlation

    International Nuclear Information System (INIS)

    Kataoka, I.; Ishii, M.

    1986-06-01

    A void fraction for a bubbling or boiling pool system is one of the important parameters in analyzing heat and mass transfer processes. Using the drift flux formulation, correlations for the pool void fraction have been developed in collaboration with a large number of experimental data. It has been found that the drift velocity in a pool system depends upon vessel diameter, system pressure, gas flux and fluid physical properties. The results show that the relative velocity and void fraction can be quite different from those predicted by conventional correlations. In terms of the rise velocity, four different regimes are identified. These are bubbly, churn-turbulent, slug and cap bubble regimes. The present correlations are shown to agree with the experimental data over wide ranges of parameters such as vessel diameter, system pressure, gas flux and physical properties. 39 refs., 41 figs

  5. Nucleation from a cluster of inclusions, leading to void coalescense

    DEFF Research Database (Denmark)

    Tvergaard, Viggo

    2017-01-01

    A cell model analysis is used to study the nucleation and subsequent growth of voids from a non-uniform distribution of inclusions in a ductile material. Nucleation is modeled as either stress controlled or strain controlled. The special clusters considered consist of a number of uniformly spaced...... inclusions located along a plane perpendicular to the maximum principal tensile stress. A plane strain approximation is used, where the inclusions are parallel cylinders perpendicular to the plane. Clusters with different numbers of inclusions are compared with the nucleation and growth from a single...... inclusion, such that the total initial volume of the inclusions is the same for the clusters and the single inclusion. After nucleation, local void coalescence inside the clusters is accounted for, since this makes it possible to compare the rate of growth of the single larger void that results from...

  6. Internal Nano Voids in Yttria-Stabilised Zirconia (YSZ Powder

    Directory of Open Access Journals (Sweden)

    Chen Barad

    2017-12-01

    Full Text Available Porous yttria-stabilised zirconia ceramics have been gaining popularity throughout the years in various fields, such as energy, environment, medicine, etc. Although yttria-stabilised zirconia is a well-studied material, voided yttria-stabilised zirconia powder particles have not been demonstrated yet, and might play an important role in future technology developments. A sol-gel synthesis accompanied by a freeze-drying process is currently being proposed as a method of obtaining sponge-like nano morphology of embedded faceted voids inside yttria-stabilised zirconia particles. The results rely on a freeze-drying stage as an effective and simple method for generating nano-voided yttria-stabilised zirconia particles without the use of template-assisted additives.

  7. Multiple void formation in plasmas containing multispecies charged grains

    International Nuclear Information System (INIS)

    Liu, Y. H.; Chen, Z. Y.; Bogaerts, A.; Yu, M. Y.

    2006-01-01

    Self-organized separation of charged-dust species in two-dimensional dusty plasmas is studied by means of molecular-dynamics simulation. The multispecies dust grains, interacting through a screened Coulomb potential with a long-range attractive component, are confined by an external quadratic potential and subjected to a radially outward ion drag force. It is found that, in general, the species are spatially separated by bandlike dust-free (or void) regions, and grains of the same species tend to populate a common shell. At large ion drag and/or large plasma screening, a central disklike void as well as concentric bandlike voids separating the different species appear. Because of the outward drag and the attractive component of the dust-dust interaction forces, highly asymmetrical states consisting of species-separated dust clumps can also exist despite the fact that all the forces are either radial or central

  8. Void distributions in liquid BiBr{sub 3}

    Energy Technology Data Exchange (ETDEWEB)

    Maruyama, K [Faculty of Science, Niigata University, Niigata 950-2181 (Japan); Endo, H [Faculty of Science, Kyoto University, Kyoto 606-8224 (Japan); Hoshino, H [Faculty of Education, Hirosaki University, Hirosaki 036-8560 (Japan); Kawakita, Y [Faculty of Sciences, Kyushu University, Fukuoka 810-8560 (Japan); Kohara, S; Itou, M [Japan Synchrotron Radiation Research Institute(JASRI), Sayo-cho 679-5198 (Japan)

    2008-02-15

    The X-ray diffraction experiments and the reverse Monte Carlo analysis for liquid BiBr{sub 3} have been performed to clarify the distribution of Bi and Br ions around voids, comparing with previous results derived in the neutron diffraction experiments. The hexagonal cages involving voids are formed by the corner-sharing of the trigonal pyramidal BiBr{sub 3} blocks. The neighboring cages are linked together in highly correlated fashion. The observed pre-peak in S(Q) at 1.3A{sup -1} is related to the pre-peak of the void-based S'{sub CC} (Q) due to an intermediate chemical order in the structure. The pre-peak intensity increases with increasing temperature. This characteristic change for the pre-peak intensity is discussed by considering the modifications of the topology and stacking in the hexagonal cages.

  9. Two-dimensional void reconstruction by neutron transmission

    International Nuclear Information System (INIS)

    Zakaib, G.D.; Harms, A.A.; Vlachopoulos, J.

    1978-01-01

    Contemporary algebraic reconstruction methods are utilized in investigating the two-dimensional void distribution in a water analog from neutron transmission measurements. It is sought to ultimately apply these techniques to the determination of time-averaged void distribution in two-phase flow systems as well as for potential usage in neutron radiography. Initially, projection data were obtained from a digitized model of a hypothetical two-phase representation and later from neutron beam traverses across a voided methacrylate plastic model. From 10 to 15 views were incorporated, and decoupling of overlapped measurements was utilized to afford greater resolution. In general, the additive Algebraic Reconstruction Technique yielded the best reconstructions, with others showing promise for noisy data. Results indicate the need for some further development of the method in interpreting real data

  10. Voiding patterns in men evaluated by a questionnaire survey

    DEFF Research Database (Denmark)

    Sommer, P; Nielsen, K K; Bauer, T

    1990-01-01

    A questionnaire on obstructive and irritative voiding symptoms was sent to 572 men aged between 20 and 79 years, selected at random from the National Register; 337 questionnaires were completed. None of the responders had consulted a doctor because of voiding symptoms. There was a significant...... voiding symptoms in men aged 60 to 79 years without subjective prostatism was the same as in patients admitted with prostatism, although most of the men had milder symptoms. Only nocturia and urge incontinence were more prevalent in patients admitted with prostatism. About 20% of men in the oldest decades...... had symptoms equal in severity to those found in men undergoing prostatectomy; 29% and 11% of men in the eighth decade [corrected] had nocturia twice and 3 times or more respectively; 19% complained of urge incontinence. More information on possible treatment is needed....

  11. Noninvasive Medical Tools for Evaluating Voiding Pattern in Real Life

    Directory of Open Access Journals (Sweden)

    Kwonsoo Chun

    2017-04-01

    Full Text Available Voiding dysfunction is a common disease that contributes to a lower quality of life and has an increased prevalence in the elderly population. Noninvasive and objective methods such as uroflowmetry (UFM and voiding diaries (VDs are essential for exact diagnosis and effective treatment of this condition because patients with different causes of voiding dysfunction can complain of the same lower urinary tract symptoms. Further, different treatment options can be determined based on the diagnosis made from these symptoms. In order to improve the quality of UFM and VDs and to provide a convenient testing environment, several advances have been made by previous investigators. In this study, we investigate the history and technological mechanisms of UFM and VDs. We also aim to review UFM from the viewpoint of clinical and at-home uses, including the recently proposed toilet-shaped UFM and electronic VDs.

  12. Loss of urinary voiding sensation due to herpes zoster.

    Science.gov (United States)

    Hiraga, Akiyuki; Nagumo, Kiyomi; Sakakibara, Ryuji; Kojima, Shigeyuki; Fujinawa, Naoto; Hashimoto, Tasuku

    2003-01-01

    A case of sacral herpes zoster infection in a 56-year-old man with the complication of loss of urinary voiding sensation is presented. He had typical herpes zoster eruption on the left S2 dermatome, hypalgesia of the S1-S4 dermatomes, and absence of urinary voiding sensation. There was no other urinary symptom at the first medical examination. Urinary complications associated with herpes zoster are uncommon, but two types, acute cystitis and acute retention, have been recognized. No cases of loss of urinary voiding sensation due to herpes zoster have been reported. In this case, hypalgesia of the sacral dermatomes was mild compared to the marked loss of urethral sensation. This inconsistency is explained by the hypothesis that the number of urethral fibers is very small as compared to that of cutaneous fibers, therefore, urethral sensation would be more severely disturbed than cutaneous sensation. Copyright 2003 Wiley-Liss, Inc.

  13. Effects of two-phase mixing and void drift models on subchannel void fraction predictions in vertical bundles

    Energy Technology Data Exchange (ETDEWEB)

    Leung, K.H. [McMaster Univ., Hamilton, Ontario (Canada)], E-mail: leungk4@mcmaster.ca

    2009-07-01

    The evaluation of the subchannel code ASSERT against the OECD/NEA BFBT benchmark data demonstrated that at low pressures, the void fraction in the corner and side subchannels of a vertical bundle was over-predicted. Preliminary results suggest that this was due to the use of Carlucci's empirical correlation for void drift beyond its applicable range of pressure. Further examination indicates that the choice of the mixing and void drift models has a negligible effect on the error of the subchannel void fraction predictions. A single, isolated subchannel was simulated and results suggest that the root cause behind the over-prediction is inadequate mixing at the sides and corners of the bundle. Increasing the magnitude of the void drift coefficients in Carlucci's model at low pressure was found to improve the overall accuracy of the predictions. A simple correlation relating {omega} to the outlet pressure was found to increase the number of points falling within experimental error by 1.0%. (author)

  14. Effects of two-phase mixing and void drift models on subchannel void fraction predictions in vertical bundles

    International Nuclear Information System (INIS)

    Leung, K.H.

    2009-01-01

    The evaluation of the subchannel code ASSERT against the OECD/NEA BFBT benchmark data demonstrated that at low pressures, the void fraction in the corner and side subchannels of a vertical bundle was over-predicted. Preliminary results suggest that this was due to the use of Carlucci's empirical correlation for void drift beyond its applicable range of pressure. Further examination indicates that the choice of the mixing and void drift models has a negligible effect on the error of the subchannel void fraction predictions. A single, isolated subchannel was simulated and results suggest that the root cause behind the over-prediction is inadequate mixing at the sides and corners of the bundle. Increasing the magnitude of the void drift coefficients in Carlucci's model at low pressure was found to improve the overall accuracy of the predictions. A simple correlation relating Ω to the outlet pressure was found to increase the number of points falling within experimental error by 1.0%. (author)

  15. Supernovae observations in a 'meatball' universe with a local void

    International Nuclear Information System (INIS)

    Kainulainen, Kimmo; Marra, Valerio

    2009-01-01

    We study the impact of cosmic inhomogeneities on the interpretation of observations. We build an inhomogeneous universe model without dark energy that can confront supernova data and yet is reasonably well compatible with the Copernican principle. Our model combines a relatively small local void, that gives apparent acceleration at low redshifts, with a meatball model that gives sizable lensing (dimming) at high redshifts. Together these two elements, which focus on different effects of voids on the data, allow the model to mimic the concordance model.

  16. Supernovae observations in a 'meatball' universe with a local void

    Energy Technology Data Exchange (ETDEWEB)

    Kainulainen, Kimmo; Marra, Valerio [Department of Physics, University of Jyvaeskylae, PL 35 (YFL), FIN-40014 Jyvaeskylae, Finland and Helsinki Institute of Physics, University of Helsinki, PL 64, FIN-00014 Helsinki (Finland)

    2009-12-15

    We study the impact of cosmic inhomogeneities on the interpretation of observations. We build an inhomogeneous universe model without dark energy that can confront supernova data and yet is reasonably well compatible with the Copernican principle. Our model combines a relatively small local void, that gives apparent acceleration at low redshifts, with a meatball model that gives sizable lensing (dimming) at high redshifts. Together these two elements, which focus on different effects of voids on the data, allow the model to mimic the concordance model.

  17. Supernovae observations in a ``meatball'' universe with a local void

    Science.gov (United States)

    Kainulainen, Kimmo; Marra, Valerio

    2009-12-01

    We study the impact of cosmic inhomogeneities on the interpretation of observations. We build an inhomogeneous universe model without dark energy that can confront supernova data and yet is reasonably well compatible with the Copernican principle. Our model combines a relatively small local void, that gives apparent acceleration at low redshifts, with a meatball model that gives sizable lensing (dimming) at high redshifts. Together these two elements, which focus on different effects of voids on the data, allow the model to mimic the concordance model.

  18. Cluster-void degeneracy breaking: Modified gravity in the balance

    Science.gov (United States)

    Sahlén, Martin; Silk, Joseph

    2018-05-01

    Combining galaxy cluster and void abundances is a novel, powerful way to constrain deviations from general relativity and the Λ CDM model. For a flat w CDM model with growth of large-scale structure parametrized by the redshift-dependent growth index γ (z )=γ0+γ1z /(1 +z ) of linear matter perturbations, combining void and cluster abundances in future surveys with Euclid and the four-meter multiobject spectroscopic telescope could improve the figure of merit for (w ,γ0,γ1) by a factor of 20 compared to individual abundances. In an ideal case, improvement on current cosmological data is a figure of merit factor 600 or more.

  19. Void redistribution in sand under post-earthquake loading

    International Nuclear Information System (INIS)

    Boulanger, R.W.; Truman, S.P.

    1996-01-01

    A mechanism for void redistribution in an infinite slope under post-earthquake loading conditions is described by consideration of the in situ loading paths that can occur under post-earthquake conditions and the results of triaxial tests designed to represent specific in situ post-earthquake loading paths. The mechanism is illustrated by application to an example problem. Void redistribution is shown to be a phenomena that may be more pronounced at the field scale than at the laboratory scale. (author). 12 refs., 4 figs

  20. Influence of void on image quality of industrial SPECT

    International Nuclear Information System (INIS)

    Park, J G; Jung, S H; Kim, J B; Moon, J; Kim, C H

    2013-01-01

    Industrial single photon emission computed tomography (SPECT) is a promising technique to determine the dynamic behavior of industrial process media and has been developed in the Korea Atomic Energy Research Institute. The present study evaluated the influence of a void, which is presence in multiphase reactors of industrial process, on the image quality of an industrial SPECT. The results are very encouraging; that is, the performance of the industrial SPECT system is little influenced by the presence of a void, which means that industrial SPECT is an appropriate tool to estimate the dynamic characteristics of the process media in a water-air phase bubble column with a static gas sparger

  1. Transient void fraction measurements in rod bundle geometries

    International Nuclear Information System (INIS)

    Chan, A.M.C.

    1998-01-01

    A new gamma densitometer with a Ba-133 source and a Nal(TI) scintillator operated in the count mode has been designed for transient void fraction measurements in the RD-14M heated channels containing a seven-element heater bundle. The device was calibrated dynamically in the laboratory using an air-water flow loop. The void fraction measured was found to compare well with values obtained using the trapped-water method. The device was also found to follow very well the passage of air slugs in pulsating flow with slug passing frequencies of up to about 1.5 hz. (author)

  2. Void swelling behaviour of austenitic stainless steel during electron irradiation

    International Nuclear Information System (INIS)

    Sheng Zhongqi; Xiao Hong; Peng Feng; Ti Zhongxin

    1994-04-01

    The irradiation swelling behaviour of 00Cr17Ni14Mo2 austenitic stainless steel (AISI 316L) was investigated by means of high voltage electron microscope. Results showed that in solution annealed condition almost no swelling incubation period existed, and the swelling shifted from the transition period to the steady-state one when the displacement damage was around 40 dpa. In cold rolled condition there was evidently incubation period, and when the displacement damage was up to 84 dpa the swelling still remained in the transition period. The average size and density of voids in both conditions were measured, and the factors, which influenced the void swelling, were discussed. (3 figs.)

  3. Recommended reactor coolant water chemistry requirements for WWER-1000 units with 235U higher enriched fuel

    International Nuclear Information System (INIS)

    Dobrevski, I.; Zaharieva, N.

    2011-01-01

    The last decade worldwide experience of PWRs and WWERs confirms the trends for the improvement of the nuclear power industry electricity production through the implementation of high burn-up or high fuel duty, which are usually accompanied with the usage of UO 2 fuel with higher content of 235 U - 4.0% - 4.5% (5.0%). It was concluded that the onset of sub-cooled nucleate boiling (SNB) on the fuel cladding surfaces and the initial excess reactivity of the core are the primary and basic factors accompanying the implementation of uranium fuel with higher 235 U content, aiming extended fuel cycles and higher burn-up of the fuel in Pressurized Water Reactors. As main consequences of the presence of these factors the modifications of chemical / electrochemical environments of nuclear fuel cladding- and reactor coolant system- surfaces are evaluated. These conclusions are the reason for: 1) The determination of the choices of the type of fuel cladding materials in respect with their enough corrosion resistance to the specific fuel cladding environment, created by the presence of SNB; 2) The development and implementation of primary circuit water chemistry guidelines ensuring the necessary low corrosion rates of primary circuit materials and limitation of cladding deposition and out-of-core radioactivity buildup; 3) Implementation of additional neutron absorbers which allow enough decrease of the initial concentration of H 3 BO 3 in coolant, so that its neutralization will be possible with the permitted alkalising agent concentrations. In this paper the specific features of WWER-1000 units in Bulgarian Nuclear Power Plant; use of 235 U higher enriched fuel in the WWER-1000 reactors in the Kozloduy NPP; coolant water chemistry and radiochemistry plant data during the power operation period of the Kozloduy NPP Unit 5, 15 th fuel cycle; evaluation of the approaches and results by the conversion of the WWER-1000 Units at the Kozloduy NPP to the uranium fuel with 4.3% 235 U as

  4. Influence of coolant motion on structure of hardened steel element

    Directory of Open Access Journals (Sweden)

    A. Kulawik

    2008-08-01

    Full Text Available Presented paper is focused on volumetric hardening process using liquid low melting point metal as a coolant. Effect of convective motion of the coolant on material structure after hardening is investigated. Comparison with results obtained for model neglecting motion of liquid is executed. Mathematical and numerical model based on Finite Element Metod is described. Characteristic Based Split (CBS method is used to uncouple velocities and pressure and finally to solve Navier-Stokes equation. Petrov-Galerkin formulation is employed to stabilize convective term in heat transport equation. Phase transformations model is created on the basis of Johnson-Mehl and Avrami laws. Continuous cooling diagram (CTPc for C45 steel is exploited in presented model of phase transformations. Temporary temperatures, phases participation, thermal and structural strains in hardening element and coolant velocities are shown and discussed.

  5. Actively controlling coolant-cooled cold plate configuration

    Science.gov (United States)

    Chainer, Timothy J.; Parida, Pritish R.

    2015-07-28

    A method is provided to facilitate active control of thermal and fluid dynamic performance of a coolant-cooled cold plate. The method includes: monitoring a variable associated with at least one of the coolant-cooled cold plate or one or more electronic components being cooled by the cold plate; and dynamically varying, based on the monitored variable, a physical configuration of the cold plate. By dynamically varying the physical configuration, the thermal and fluid dynamic performance of the cold plate are adjusted to, for example, optimally cool the one or more electronic components, and at the same time, reduce cooling power consumption used in cooling the electronic component(s). The physical configuration can be adjusted by providing one or more adjustable plates within the coolant-cooled cold plate, the positioning of which may be adjusted based on the monitored variable.

  6. Sloshing of coolant in a seismically isolated reactor

    International Nuclear Information System (INIS)

    Wu, T.S.; Guildys, J.; Seidensticker, R.W.

    1988-01-01

    During a seismic event, the liquid coolant inside the reactor vessel has sloshing motion which is a low-frequency phenomenon. In a reactor system incorporated with seismic isolation, the isolation frequency usually is also very low. There is concern on the potential amplification of sloshing motion of the liquid coolant. This study investigates the effects of seismic isolation on the sloshing of liquid coolant inside the reactor vessel of a liquid metal cooled reactor. Based on a synthetic ground motion whose response spectra envelop those specified by the NRC Regulator Guide 1.60, it is found that the maximum sloshing wave height increases from 18 in. to almost 30 in. when the system is seismically isolated. Since higher sloshing wave may introduce severe impact forces and thermal shocks to the reactor closure and other components within the reactor vessel, adequate design considerations should be made either to suppress the wave height or to reduce the effects caused by high waves

  7. Fracture mechanics evaluation for at typical PWR primary coolant pipe

    International Nuclear Information System (INIS)

    Tanaka, T.; Shimizu, S.; Ogata, Y.

    1997-01-01

    For the primary coolant piping of PWRs in Japan, cast duplex stainless steel which is excellent in terms of strength, corrosion resistance, and weldability has conventionally been used. The cast duplex stainless steel contains the ferrite phase in the austenite matrix and thermal aging after long term service is known to change its material characteristics. It is considered appropriate to apply the methodology of elastic plastic fracture mechanics for an evaluation of the integrity of the primary coolant piping after thermal aging. Therefore we evaluated the integrity of the primary coolant piping for an initial PWR plant in Japan by means of elastic plastic fracture mechanics. The evaluation results show that the crack will not grow into an unstable fracture and the integrity of the piping will be secured, even when such through wall crack length is assumed to equal the fatigue crack growth length for a service period of up to 60 years

  8. Design and development of remotely operated coolant channel cutting machine

    International Nuclear Information System (INIS)

    Suthar, R.L.; Sinha, A.K.; Srikrishnamurty, G.

    1994-01-01

    One of the coolant tubes of Narora Atomic Power Station (NAPS) reactor needs to be removed. To remove a coolant tube, four cutting operations, (liner tube cutting, end-fitting cutting, machining of seal weld of bellow ring and finally coolant tube cutting) are required to be carried out. A remotely operated cutting machine to carry out all these operations has been designed and developed by Central Workshops. This machine is able to cut at the exact location because of numerically controlled axial and radial travel of tool. Only by changing the tool head and tool holder, same machine can be used for various types of cutting/machining operations. This report details the design, manufacture, assembly and testing work done on the machine. (author). 4 figs

  9. Radioactivity analysis of KAMINI reactor coolant from regulatory perspectives

    International Nuclear Information System (INIS)

    Srinivasan, T.K.; Sulthan, Bajeer; Sarangapani, R.; Jose, M.T.; Venkatraman, B.; Thilagam, L.

    2016-01-01

    KAMINI (a 30kWt) research reactor is operated for neutron radiography of fuel subassemblies and pyro devices and activation analysis of various samples. The reactor is fueled by 233 U and DM water is used as the coolant. During reactor operation, fission product noble gasses (FPNGs) such as 85m Kr, 87 Kr, 88 Kr, 135 Xe, 135m Xe and 138 Xe are detected in the coolant water. In order to detect clad failure, the water is sampled during reactor operation at regular intervals as per the technical specifications. In the present work, analysis of measured activities in coolant samples collected during reactor operation at 25 kWt are presented and compared with computed values obtained using ORIGEN (Isotope Generation) code

  10. Method of eliminating cruds in the primary coolants of reactors

    International Nuclear Information System (INIS)

    Tamura, Takaaki.

    1984-01-01

    Purpose: To eliminate cruds in the primary coolants by using rind of onions or peanuts. Method: Since cruds contained in the reactor primary coolants increase the radioactive exposure to reactor operators, they have been intended to remove by ion exchange resins. In this invention, rind of onions or peanuts are crushed into an adequate particle size and packed into an absorption column instead of ion exchange resins into which primary coolants are circulated. The powderous onions or peanuts rind contain glucoside such as cosmosiin and has an effect of cationic exchanger, they satisfactorily catch heavy metals such as Fe and Cu. They have an excellent filtering effect even under a high pH condition and are excellent in economical point of view. They can be decrease the volume of the absorption column, reduce their devolume after use through corrosion and easily subjected to waste procession through oxidizing combustion in liquid. (Nakamoto, H.)

  11. Physical model and calculation code for fuel coolant interactions

    International Nuclear Information System (INIS)

    Goldammer, H.; Kottowski, H.

    1976-01-01

    A physical model is proposed to describe fuel coolant interactions in shock-tube geometry. According to the experimental results, an interaction model which divides each cycle into three phases is proposed. The first phase is the fuel-coolant-contact, the second one is the ejection and recently of the coolant, and the third phase is the impact and fragmentation. Physical background of these phases are illustrated in the first part of this paper. Mathematical expressions of the model are exposed in the second part. A principal feature of the computational method is the consistent application of the fourier-equation throughout the whole interaction process. The results of some calculations, performed for different conditions are compiled in attached figures. (Aoki, K.)

  12. Fracture mechanics evaluation for at typical PWR primary coolant pipe

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, T. [Kansai Electric Power Company, Osaka (Japan); Shimizu, S.; Ogata, Y. [Mitsubishi Heavy Industries, Ltd., Kobe (Japan)

    1997-04-01

    For the primary coolant piping of PWRs in Japan, cast duplex stainless steel which is excellent in terms of strength, corrosion resistance, and weldability has conventionally been used. The cast duplex stainless steel contains the ferrite phase in the austenite matrix and thermal aging after long term service is known to change its material characteristics. It is considered appropriate to apply the methodology of elastic plastic fracture mechanics for an evaluation of the integrity of the primary coolant piping after thermal aging. Therefore we evaluated the integrity of the primary coolant piping for an initial PWR plant in Japan by means of elastic plastic fracture mechanics. The evaluation results show that the crack will not grow into an unstable fracture and the integrity of the piping will be secured, even when such through wall crack length is assumed to equal the fatigue crack growth length for a service period of up to 60 years.

  13. LOFT advanced densitometer for nuclear loss-of-coolant experiments

    International Nuclear Information System (INIS)

    Johnson, L.O.; Lassahn, G.D.; Wood, D.B.

    1979-01-01

    A ''nuclear hardened'' gamma densitometer, a device which uses radiation attenuation to measure fluid density in the presence of a background radiation field, is described. Data from the nuclear hardened gamma densitometer are acquired by time sampling the coolant fluid piping and fluid attenuated source energy spectrum. The data are used to calculate transient coolant fluid cross sectional average density to analyze transient mass flow and other thermal-hydraulic characteristics during the Loss-of-Fluid Test (LOFT) loss-of-coolant experiments. The nuclear hardened gamma densitometer uses a pulse height analysis or energy discrimination, pulse counting technique which makes separation of the gamma radiation source signal from the reactor generated gamma radiation background noise signal possible by processing discrete pulses which retain their pulse amplitude information

  14. Reactivity feedback coefficients Pakistan research reactor-1 using PRIDE code

    Energy Technology Data Exchange (ETDEWEB)

    Mansoor, Ali; Ahmed, Siraj-ul-Islam; Khan, Rustam [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Inam-ul-Haq [Comsats Institute of Information Technology, Islamabad (Pakistan). Dept. of Physics

    2017-05-15

    Results of the analyses performed for fuel, moderator and void's temperature feedback reactivity coefficients for the first high power core configuration of Pakistan Research Reactor - 1 (PARR-1) are summarized. For this purpose, a validated three dimensional model of PARR-1 core was developed and confirmed against the reference results for reactivity calculations. The ''Program for Reactor In-Core Analysis using Diffusion Equation'' (PRIDE) code was used for development of global (3-dimensional) model in conjunction with WIMSD4 for lattice cell modeling. Values for isothermal fuel, moderator and void's temperature feedback reactivity coefficients have been calculated. Additionally, flux profiles for the five energy groups were also generated.

  15. Experimental interaction of magma and “dirty” coolants

    Science.gov (United States)

    Schipper, C. Ian; White, James D. L.; Zimanowski, Bernd; Büttner, Ralf; Sonder, Ingo; Schmid, Andrea

    2011-03-01

    The presence of water at volcanic vents can have dramatic effects on fragmentation and eruption dynamics, but little is known about how the presence of particulate matter in external water will further alter eruptions. Volcanic edifices are inherently “dirty” places, where particulate matter of multiple origins and grainsizes typically abounds. We present the results of experiments designed to simulate non-explosive interactions between molten basalt and various “coolants,” ranging from homogeneous suspensions of 0 to 30 mass% bentonite clay in pure water, to heterogeneous and/or stratified suspensions including bentonite, sand, synthetic glass beads and/or naturally-sorted pumice. Four types of data are used to characterise the interactions: (1) visual/video observations; (2) grainsize and morphology of resulting particles; (3) heat-transfer data from a network of eight thermocouples; and (4) acoustic data from three force sensors. In homogeneous coolants with ~20% sediment, heat transfer is by forced convection and conduction, and thermal granulation is less efficient, resulting in fewer blocky particles, larger grainsizes, and weaker acoustic signals. Many particles are droplet-shaped or/and “vesicular,” containing bubbles filled with coolant. Both of these particle types indicate significant hydrodynamic magma-coolant mingling, and many of them are rewelded into compound particles. The addition of coarse material to heterogeneous suspensions further slows heat transfer thus reducing thermal granulation, and variable interlocking of large particles prevents efficient hydrodynamic mingling. This results primarily in rewelded melt piles and inefficient distribution of melt and heat throughout the coolant volume. Our results indicate that even modest concentrations of sediment in water will significantly limit heat transfer during non-explosive magma-water interactions. At high concentrations, the dramatic reduction in cooling efficiency and increase in

  16. On line monitoring of temperatures of coolant channels by thermal imaging in a laboratory set-up fabricated for the detection of leakage of coolants

    Energy Technology Data Exchange (ETDEWEB)

    Mukherjee, S; Ghosh, J K [Bhabha Atomic Research Centre, Bombay (India). Radiometallurgy Div.; Patel, R J [Bhabha Atomic Research Centre, Mumbai (India). Refuelling Technology Division

    1994-12-31

    Leakage from coolant channels in Pressurised Heavy Water Reactors (PHWR) increases the temperatures of the faulty channels. Measurement of temperatures of the coolant channels is, therefore, one way to detect the leaking channel. Thermal imaging technique offers a unique means for this detection providing a fast, non-contact, on-line measurement. An experiment was carried out for the detection of leakage of coolants through the seal plugs of the coolant channels in PHWR using an experimental setup under the simulated conditions of temperature and pressure of the coolant channels inside the reactor and using an infrared imaging system. The experimental details and the observations have been presented. 7 figs.

  17. On line monitoring of temperatures of coolant channels by thermal imaging in a laboratory set-up fabricated for the detection of leakage of coolants

    International Nuclear Information System (INIS)

    Mukherjee, S.; Ghosh, J.K.; Patel, R.J.

    1994-01-01

    Leakage from coolant channels in Pressurised Heavy Water Reactors (PHWR) increases the temperatures of the faulty channels. Measurement of temperatures of the coolant channels is, therefore, one way to detect the leaking channel. Thermal imaging technique offers a unique means for this detection providing a fast, non-contact, on-line measurement. An experiment was carried out for the detection of leakage of coolants through the seal plugs of the coolant channels in PHWR using an experimental setup under the simulated conditions of temperature and pressure of the coolant channels inside the reactor and using an infrared imaging system. The experimental details and the observations have been presented. 7 figs

  18. Integrated Fuel-Coolant Interaction (IFCI 6.0) code

    International Nuclear Information System (INIS)

    Davis, F.J.; Young, M.F.

    1994-04-01

    The integrated Fuel-Coolant interaction (IFCI) computer code is being developed at Sandia National Laboratories to investigate the fuel-coolant interaction (FCI) problem at large scale using a two-dimensional, four-field hydrodynamic framework and physically based models. IFCI will be capable of treating all major FCI processes in an integrated manner. This document is a product of the effort to generate a stand-alone version of IFCI, IFCI 6.0. The User's Manual describes in detail the hydrodynamic method and physical models used in IFCI 6.0. Appendix A is an input manual, provided for the creation of working decks

  19. Reactor coolant pump shaft seal behavior during blackout conditions

    International Nuclear Information System (INIS)

    Mings, W.J.

    1985-01-01

    The United States Nuclear Regulatory Commission has classified the problem of reactor coolant pump seal failures as an unresolved safety issue. This decision was made in large part due to experimental results obtained from a research program developed to study shaft seal performance during station blackout and reported in this paper. Testing and analysis indicated a potential for pump seal failure under postulated blackout conditions leading to a loss of primary coolant with a concomitant danger of core uncovery. The work to date has not answered all the concerns regarding shaft seal failure but it has helped scope the problem and focus future research needed to completely resolve this issue

  20. Effects of different rod spacers (helical types) on coolant crossmixing

    International Nuclear Information System (INIS)

    Zhukov, A.V.; Sviridenko, E.Ya.; Matyukhin, N.M.; Rymkevich, K.S.; Ushakov, P.A.

    1981-11-01

    The results of investigations (electromagnetic measuring method) on coolant cross mixing in rod clusters with spiral wire spacers with different winding directions, with alternating unfinned and finned rods (case 'fin to rod'), as well as in rod clusters with much space between the rods, (case 'fin to fin') are reported. The local fluid dynamics parameters (distribution of the transversal and longitudinal velocity component) that define the physical processes of the coolant exchange in the rod clusters with helical spacers are explained. The investigation results for different helical spacer types are compared with each other. (orig.) [de

  1. The 1994 loss of coolant incident at Pickering NGS

    Energy Technology Data Exchange (ETDEWEB)

    Charlebois, P R; Clarke, T R; Goodman, R M; McEwan, W F [Ontario Hydro, Pickering, ON (Canada). Pickering Generating Station; Cuttler, J M [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

    1996-12-31

    Fracture of the rubber diaphragm in a liquid relief valve initiated events leading to a loss of coolant in Unit 2, on December 10. The valve failed open, filling the bleed condenser. The reactor shut itself down. When pressure recovered, two spring-loaded safety relief valves opened and one of them chattered. The shock and pulsations cracked the inlet pipe to the chattering valve, and the subsequent loss of coolant triggered the emergency core cooling system. The incident was terminated by operator action. No abnormal radioactivity was released. The four reactor units of Pickering A remained shut down until the corrective actions were completed in April/May 1995. (author). 4 figs.

  2. LWR and HTGR coolant dynamics: the containment of severe accidents

    International Nuclear Information System (INIS)

    Theofanous, T.G.; Gherson, P.; Nourbakhsh, H.P.; Hu, K.; Iyer, K.; Viskanta, R.; Lommers, L.

    1983-07-01

    This is the final report of a project containing three major tasks. Task I deals with the fundamental aspects of energetic fuel/coolant interactions (steam explosions) as they pertain to LWR core melt accidents. Task II deals with the applied aspects of LWR core melt accident sequences and mechanisms important to containment response, and includes consideration of energetic fuel/coolant interaction events, as well as non-explosive ones, corium material disposition and eventual coolability, and containment pressurization phenomena. Finally, Task III is concerned with HTGR loss of forced circulation accidents. This report is organized into three major parts corresponding to these three tasks respectively

  3. Knock-limited performance of several internal coolants

    Science.gov (United States)

    Bellman, Donald R; Evvard, John C

    1945-01-01

    The effect of internal cooling on the knock-limited performance of an-f-28 fuel was investigated in a CFR engine, and the following internal coolants were used: (1) water, (2), methyl alcohol-water mixture, (3) ammonia-methyl alcohol-water mixture, (4) monomethylamine-water mixture, (5) dimethylamine-water mixture, and (6) trimethylamine-water mixture. Tests were run at inlet-air temperatures of 150 degrees and 250 degrees F. to indicate the temperature sensitivity of the internal-coolant solutions.

  4. Impedance calculations for power cables to primary coolant pump motors

    International Nuclear Information System (INIS)

    Hegerhorst, K.B.

    1977-01-01

    The LOFT primary system motor generator sets are located in Room B-239 and are connected to the primary coolant pumps by means of a power cable. The calculated average impedance of this cable is 0.005323 ohms per unit resistance and 0.006025 ohms per unit reactance based on 369.6 kVA and 480 volts. The report was written to show the development of power cable parameters that are to be used in the SICLOPS (Simulation of LOFT Reactor Coolant Loop Pumping System) digital computer program as written in LTR 1142-16 and also used in the pump coastdowns for the FSAR Analysis

  5. Steam generator for a pressurized-water coolant nuclear reactor

    International Nuclear Information System (INIS)

    Schroeder, H.J.; Berger, W.

    1975-01-01

    A description is given of a steam generator which has a vertical cylindrical housing having a steam output outlet, a horizontal tube sheet closing the lower end of this housing, and an inverted U-shaped tube bundle inside of the housing and having vertical inlet and outlet legs with their ends mounted in the tube sheet. Beneath the tube sheet there are inlet and outlet manifolds for the respective ends of the tube bundle so that pressurized-water coolant from a pressurized-water coolant nuclear reactor can be circulated through the tube bundle

  6. Probabilistic analyses of failure in reactor coolant piping

    International Nuclear Information System (INIS)

    Holman, G.S.

    1984-01-01

    LLNL is performing probabilistic reliability analyses of PWR and BWR reactor coolant piping for the NRC Office of Nuclear Regulatory Research. Specifically, LLNL is estimating the probability of a double-ended guillotine break (DEGB) in the reactor coolant loop piping in PWR plants, and in the main stream, feedwater, and recirculation piping of BWR plants. In estimating the probability of DEGB, LLNL considers two causes of pipe break: pipe fracture due to the growth of cracks at welded joints (direct DEGB), and pipe rupture indirectly caused by the seismically-induced failure of critical supports or equipment (indirect DEGB)

  7. Sensitivity calculation of the coolant temperature regarding the thermohydraulic parameters

    International Nuclear Information System (INIS)

    Andrade Lima, F.R. de; Silva, F.C. da; Thome Filho, Z.D.; Alvim, A.C.M.; Oliveira Barroso, A.C. de.

    1985-01-01

    It's studied the application of the Generalized Perturbation Theory (GPT) in the sensitivity calculation of thermalhydraulic problems, aiming at verifying the viability of the extension of the method. For this, the axial distribution, transient, of the coolant temperature in a PWR channel are considered. Perturbation expressions are developed using the GPT formalism, and a computer code (Tempera) is written, to calculate the channel temperature distribution and the associated importance function, as well as the effect of the thermalhydraulic parameters variations in the coolant temperature (sensitivity calculation). The results are compared with those from the direct calculation. (E.G.) [pt

  8. Control rod drive mechanism stator loss of coolant test

    International Nuclear Information System (INIS)

    Besel, L.; Ibatuan, R.

    1977-04-01

    This report documents the stator loss of coolant test conducted at HEDL on the lead unit Control Rod Drive Mechanism (CRDM) in February, 1977. The purpose of the test was to demonstrate scram capability of the CRDM with an uncooled stator and to obtain a time versus temperature curve of an uncooled stator under power. Brief descriptions of the test, hardware used, and results obtained are presented in the report. The test demonstrated that the CRDM could be successfully scrammed with no anomalies in both the two-phase and three-phase stator winding hold conditions after the respective equilibrium stator temperatures had been obtained with no stator coolant

  9. Effect of the critical size of initial voids on stress-induced migration

    International Nuclear Information System (INIS)

    Aoyagi, Minoru

    2004-01-01

    The stress-induced migration phenomenon is one of the problems related to the reliability of metal interconnections in semiconductor devices. This phenomenon causes voids and fractures in interconnections. The basic feature of this phenomenon is vacancy migration to minute initial voids. Expanding initial voids grow into larger voids and fractures. The purpose of this work is to theoretically clarify the effects of residual thermal stress and void surface stress on the behavior of the initial voids which exist immediately after a passivation process. Using a spherical metal sample with a spherical void under external stress, vacancy absorption or emission was investigated between the void surface and the sample surface. The behavior of vacancies and atoms was also investigated in interconnections under residual thermal stress. We show that the void or sample surface becomes a vacancy sink or source, depending on the mutual relationship between the surface stress due to the surface-free energy and the residual thermal stress. We also reveal that the initial voids, which exist immediately after a passivation process, grow into larger voids and fractures when the size of the initial voids exceeds the critical size. If the size of the initial void can be controlled to below the critical size, voids and fractures do not occur

  10. OECD/DOE/CEA VVER-1000 Coolant Transient Benchmark. Summary Record of the First Workshop (V1000-CT1)

    International Nuclear Information System (INIS)

    2003-01-01

    The first workshop for the VVER-1000 Coolant Transient Benchmark TT Benchmark was hosted by the Commissariat a l'Energie Atomique, Centre d'Etudes de Saclay, France. The V1000CT benchmark defines standard problems for validation of coupled three-dimensional (3-D) neutron-kinetics/system thermal-hydraulics codes for application to Soviet-designed VVER-1000 reactors using actual plant data without any scaling. The overall objective is to access computer codes used in the safety analysis of VVER power plants, specifically for their use in reactivity transient simulations in a VVER-1000. The V1000CT benchmark consists of two phases: V1000CT-1 - simulation of the switching on of one main coolant pump (MCP) while the other three MCP are in operation, and V1000CT- 2 - calculation of coolant mixing tests and Main Steam Line Break (MSLB) scenario. Further background information on this benchmark can be found at the OECD/NEA benchmark web site . The purpose of the first workshop was to review the benchmark activities after the Starter Meeting held last year in Dresden, Germany: to discuss the participants' feedback and modifications introduced in the Benchmark Specifications on Phase 1; to present and to discuss modelling issues and preliminary results from the three exercises of Phase 1; to discuss the modelling issues of Exercise 1 of Phase 2; and to define work plan and schedule in order to complete the two phases

  11. Enhancing Reactivity in Structural Energetic Materials

    Science.gov (United States)

    Glumac, Nick

    2017-06-01

    In many structural energetic materials, only a small fraction of the metal oxidizes, and yet this provides a significant boost in the overall energy release of the system. Different methodologies to enhance this reactivity include alloying and geometric modifications of microstructure of the reactive material (RM). In this presentation, we present the results of several years of systematic study of both chemical (alloy) and mechanical (geometry) effects on reactivity for systems with typical charge to case mass ratios. Alloys of aluminum with magnesium and lithium are considered, as these are common alloys in aerospace applications. In terms of geometric modifications, we consider surface texturing, inclusion of dense additives, and inclusion of voids. In all modifications, a measurable influence on output is observed, and this influence is related to the fragment size distribution measured from the observed residue. Support from DTRA is gratefully acknowledged.

  12. THE PREDICTION OF VOID VOLUME IN SUBCOOLED NUCLEATE POOL BOILING

    Energy Technology Data Exchange (ETDEWEB)

    Duke, E. E. [General Dynamics, San Diego, CA (United States)

    1963-11-15

    A three- step equation was developed that adequately describes the average volume of vapor occurring on a horizontal surface due to nucleate pool boiling of subcooled water. Since extensive bubble frequency data are lacking, the data of others were combined with experimental observations to make predictions of void volume at ambient pressure with various degrees of subcooling. (auth)

  13. Nucleation of voids and other irradiation-produced defect aggregates

    International Nuclear Information System (INIS)

    Wiedersich, H.; Katz, J.L.

    1976-01-01

    The nucleation of defect clusters in crystalline solids from radiation-produced defects is different from the usual nucleation processes in one important aspect: the condensing defects, interstitial atoms and vacancies, can mutually annihilate and are thus similar to matter and antimatter. The nucleation process is described as the simultaneous reaction of vacancies and interstitials (and gas atoms if present) with embryos of all sizes. The reaction rates for acquisition of point defects (and gas atoms) are calculated from their respective jump frequencies and concentrations in the supersaturated system. The reaction rates for emission of point defects are derived from the free energies of the defect clusters in the thermodynamic equilibrium system, i.e., the system without excess point defects. This procedure differs from that used in conventional nucleation theory and permits the inclusion of the ''antimatter'' defect into the set of reaction-rate equations in a straightforward manner. The method is applied to steady-state nucleation, during irradiation, of both dislocation loops and voids in the absence and in the presence of immobile and mobile gas. The predictions of the nucleation theory are shown to be in qualitative agreement with experimental observations, e.g., void densities increase with increasing displacement rates; gases such as helium enhance void nucleation; at low displacement rates and at high temperatures the presence of gas is essential to void formation. For quantitative predictions, the theory must be extended to include the termination of nucleation

  14. Predictive efficacy of radioisotope voiding cystography for renal outcome

    International Nuclear Information System (INIS)

    Kim, Seok Ki; Lee, Dong Soo; Kim, Kwang Myeung; Choi, Whang; Chung, June Key; Lee, Myung Chul

    2000-01-01

    As vesicoureteral reflux (VUR) could lead to renal functional deterioration when combined with urinary tract infection, we need to decide whether operative anti-reflux treatment should be performed at the time of diagnosis of VUR. Predictive value of radioisotope voiding cystography (RIVCG) for renal outcome was tested. In 35 children (18 males, 17 females), radiologic voiding cystoure-thrography (VCU), RIVCG and DMSA scan were performed. Change in renal function was evaluated using the follow-up DMSA scan, ultrasonography, and clinical information. Discriminant analysis was performed using individual or integrated variables such as reflux amount and extent at each phase of voiding on RIVCG, in addition to age, gender and cortical defect on DMSA scan at the time of diagnosis. Discriminant function was composed and its performance was examined. Reflux extent at the filling phase and reflux amount and extent at postvoiding phase had a significant prognostic value. Total reflux amount was a composite variable to predict prognosis. Discriminant function composed of reflux extent at the filling phase and reflux amount and extent at postvoiding phase showed better positive predictive value and specificity than conventional reflux grading. RIVCG could predict renal outcome by disclosing characteristic reflux pattern during various voiding phases.=20

  15. Decay of charge deposited on the wall of gaseous void

    DEFF Research Database (Denmark)

    McAllister, Iain Wilson

    1992-01-01

    sustained in the void. However, the absolute value of this field is reduced with an increase in bulk permittivity. It is concluded that the present choice of a point charge to simulate the wall charge has the disadvantage that such a source is associated with a field singularity, and thus it is not possible...

  16. Electric field theory and the fallacy of void capacitance

    DEFF Research Database (Denmark)

    McAllister, Iain Wilson

    1991-01-01

    The concept of the capacitance of a gaseous void is discussed as applied to electrical insulation science. The most pertinent aspect of the capacitance definition is that of reference to a single-valued potential difference between surfaces. This implies that these surfaces must be surfaces...

  17. The cosmic web in CosmoGrid void regions

    NARCIS (Netherlands)

    Rieder, Steven; van de Weygaert, Rien; Cautun, Marius; Beygu, Burcu; Portegies Zwart, Simon

    2016-01-01

    We study the formation and evolution of the cosmic web, using the high-resolution CosmoGrid ΛCDM simulation. In particular, we investigate the evolution of the large-scale structure around void halo groups, and compare this to observations of the VGS-31 galaxy group, which consists of three

  18. Monte Carlo validation of self shielding and void effect calculations

    International Nuclear Information System (INIS)

    Tellier, H.; Coste, M.; Raepsaet, C.; Soldevila, M.; Van der Gucht, C.

    1995-01-01

    The self shielding validation and the void effect are studied with Monte Carlo method. The satisfactory comparison obtained between the APOLLO 2 results of the self shielding effect and the TRIPOLI and MCNP results allows us to be confident in the multigroup transport code. (K.A.)

  19. Extrapolation of ZPR sodium void measurements to the power reactor

    International Nuclear Information System (INIS)

    Beck, C.L.; Collins, P.J.; Lineberry, M.J.; Grasseschi, G.L.

    1976-01-01

    Sodium-voiding measurements of ZPPR assemblies 2 and 5 are analyzed with ENDF/B Version IV data. Computations include directional diffusion coefficients to account for streaming effects resulting from the plate structure of the critical assembly. Bias factors for extrapolating critical assembly data to the CRBR design are derived from the results of this analysis

  20. Detection of Vesico-Ureteric Reflux Using Voiding Hippuran Ureterograms

    Energy Technology Data Exchange (ETDEWEB)

    Sutherland, J. B.; Palser, R. [Section of Nuclear Medicine, Manitoba Cancer Treatment and Research Foundation, Winnipeg General Hospital, Winnipeg (Canada); Macpherson, R. I. [Children' s Hospital of Winnipeg, Winnipeg (Canada)

    1971-02-15

    Initial results of a technique for the demonstration of vesico-ureteric reflux in children are described. Hippuran-{sup 131}I (15 {mu}Ci) (ortho-iodohippurate) is injected intravenously. A standard renogram is obtained. Additional collimation is added to the recording probes and they are positioned to record the radioactivity from the mid-ureteric region. Recordings of normal and abnormal peristaltic activity during per-ora hydration of the patient are thus obtained. When the child is willing to void, he is placed upright on a bed pan, the probes positioned to record over the lower ureteric region and recordings are made while the child voids. All data are recorded on a 512-channel analyser operated in the multi-scaler node. Data are punched out on paper tape and, after an 11 point computer smoothing program, are displayed graphically. These recordings show different patterns in normal children and those with vesico-ureteric reflux. There are several advantages to this technique over the standard radiological and other radionuclide voiding cystoureterograms. The results are compared with contrast voiding cystourethrograms in both normal children and those with vesico-ureteric reflux. (author)