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)

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

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.

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

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

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)

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)

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

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

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)

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.

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

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.)

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

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

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)

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

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)

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)

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)

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)

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

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

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)

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

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)

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)

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.)

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

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

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.

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.

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)

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

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%

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)

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)

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.

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

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.

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

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.

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)

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)

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

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)

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

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.

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

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)

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

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

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.

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

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

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)

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

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

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)

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)

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

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

Comparison of MCNP and WIMS-AECL/RFSP calculations against critical heavy water experiments in ZED-2 with CANFLEX-LVRF and CANFLEX-LEU fuels

International Nuclear Information System (INIS)

Bromley, B. P.; Watts, D. G.; Pencer, J.; Zeller, M.; Dweiri, Y.

2009-01-01

This paper summarizes calculations of MCNP5 and WIMS-AECL/RFSP compared against measurements in coolant void substitution experiments in the ZED-2 critical facility with CANFLEX R-LEU/RU (Low Enriched Uranium, Recovered Uranium) reference fuels and CANFLEX-LVRF (Low Void Reactivity Fuel) test fuel, and H 2 O/air coolants. Both codes are tested for the prediction of the change in reactivity with complete voiding of all fuel channels, and that for a checkerboard voiding pattern. Understanding these phenomena is important for the ACR-1000 R reactor. Comparisons are also made for the prediction of the axial and radial neutron flux distributions, as measured by copper foil activation. The experimental data for these comparisons were obtained from critical mixed lattice / substitution experiments in AECL's ZED-2 critical facility using CANFLEX-LEU/RU and CANFLEX-LVRF fuel in a 24-cm square lattice pitch at 25 degrees C. Substitution analyses were performed to isolate the properties (buckling, bare critical lattice dimensions) of the CANFLEX-LVRF fuel. This data was then used to further test the lattice physics codes. These comparisons establish biases/uncertainties and errors in the calculation of k eff , coolant void reactivity, checkerboard coolant void reactivity, and flux distributions. Results show small to modest biases in void reactivity and very good agreement for flux distributions. The importance of boundary conditions and the modeling of un-moderated fuel in the critical experiments are demonstrated. This comparison study provides data that supports code validation and gives good confidence in the reactor physics tools used in the design and safety analysis of the ACR-1000 reactor. (authors)

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.)

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)

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.

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.

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

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

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)

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

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

Analysis of DCA experimental data

International Nuclear Information System (INIS)

Min, B. J.; Kim, S. Y.; Ryu, S. J.; Seok, H. C.

2000-01-01

The lattice characteristics of DCA are calculated with WIMS-ATR code to validate WIMS-AECL code for the lattice analysis of CANDU core by using experimental data of DCA at JNC. Analytical studies of some critical experiments had been performed to analyze the effects of fuel composition. Different items of reactor physics such as local power peaking factor (LPF), effective multiplication factor (Keff) and coolant void reactivity were calculated for two coolant void fractions (0% and 100%). LPFs calculated by WIMS-ATR code are in close agreement with the experimental results. LPFs calculated by WIMS-AECL code with WINFRITH and ENDF/B-V libraries have similar values for both libraries but the differences between experimental data and calculated results by WIMS-AECL code are larger than those of WIMS-ATR code. The maximum difference between the values calculated by WIMS-ATR and experimental values of LPFs are within 1.3%. The coupled code systems WIMS-ATR and CITATION used in this analysis predict Keff within 1% ΔK and coolant void reactivity within 4 % ΔK/K in all cases. The coolant void reactivity of uranium fuel is found to be positive. To validate WIMS-AECL code, the core characteristics of DCA shall be calculated by WIMS-AECL and CITATION codes in the future

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

Minor actinide burning in dedicated lead-bismuth cooled fast reactors

International Nuclear Information System (INIS)

Hejzlar, P.; Driscoll, M.J.; Kazimi, M.S.; Todreas, N.E.

2001-01-01

The destruction of minor actinides (MA) in dedicated burners is of contemporary interest in Europe and Japan because it requires the deployment of smaller number of special transmutation facilities. A major fraction of Pu from spent LWR fuel can be then burned in PWRs (or fast reactors) using dedicated fertile-free fuel assemblies. However, the design of MA burning fast spectrum cores poses significant challenges because of deterioration of key safety parameters, in particular of the coolant void coefficient. This study proposes the concept of an lead-bismuth eutectic (LBE)-cooled dedicated MA burner having metallic fuel (MA-Pu-Zr) and streaming assemblies to attain acceptable coolant void worth performance. It is shown that a large 1800 MWth fertile-free core containing 37 wt% TRU with very high fraction of MA(59 wt%) from LWR spent fuel can be burned in a first cycle for 700 EFPDs with a very small reactivity swing: less than β eff . Moreover, the reactivity void worth is negative for a fully voided core when all surrounding coolant is kept at reference density. However, the core reactivity increases as coolant density falls from the reference value of 10.25 to 6 g/cm 3 . Because its coolant density coefficient value is less than that of a sodium cooled IFR, the concept provides good potential for the achievement of self-regulation characteristics in unprotected events, provided that small negative fuel temperature feedback can be maintained. (authors)

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

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)

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

FAST and SAFE Passive Safety Devices for Sodium-cooled Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Hartanto, Donny; Kim, Chihyung; Kim, In-Hyung; Kim, Yonghee [KAIST, Daejeon (Korea, Republic of)

2015-05-15

The major factor is the impact of the neutron spectral hardening. The second factor that affects the CVR is reduced capture by the coolant when the coolant voiding occurs. To improve the CVR, many ideas and concepts have been proposed, which include introduction of an internal blanket, spectrum softening, or increasing the neutron leakage. These ideas may reduce the CVR, but they deteriorate the neutron economy. Another potential solution is to adopt a passive safety injection device such as the ARC (autonomous reactivity control) system, which is still under development. In this paper, two new concepts of passive safety devices are proposed. The devices are called FAST (Floating Absorber for Safety at Transient) and SAFE (Static Absorber Feedback Equipment). Their purpose is to enhance the negative reactivity feedback originating from the coolant in fast reactors. SAFE is derived to balance the positive reactivity feedback due to sodium coolant temperature increases. It has been demonstrated that SAFE allows a low-leakage SFR to achieve a self-shutdown and self-controllability even though the generic coolant temperature coefficient is quite positive and the coolant void reactivity can be largely managed by the new FAST device. It is concluded that both FAST and SAFE devices will improve substantially the fast reactor safety and they deserve more detailed investigations.

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.

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.

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)

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.

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

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.

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)

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.

Safety aspect of long-life small safe power reactors

International Nuclear Information System (INIS)

Zaki, S.; Sekimoto, H.

1995-01-01

Safety aspects of several design options of long-life small safe fast power reactors using nitride fuel and lead-bismuth as coolant are discussed. In the present study hypothetical accidents are simulated for these reactors, i.e., unprotected simultaneous ULOF (total loss of primary pumping system) and UTOP (rod run out transient over power) accidents, caused by the simultaneous withdrawal of all control rods. The proposed designs have some important safety characteristics as low reactivity swing (only 0.2-0.25$), and negative coolant void coefficient over whole burnup period. Effectively negative value of all components of reactivity during an accident is observed. The safety performances of the balance, pancake, and tall slender type of core, each of them satisfy reactivity and negative coolant void coefficient constraint, against the above accident are compared. The simulation results show that all of the design options can survive the above accidents without the help of reactor scram and without the need of operator actions. (author)

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

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.)

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.

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)

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

Analysis of a main steam isolation value closure anticipated transient without scram in a boiling water reactor

International Nuclear Information System (INIS)

Liaw, T.J.; Pan, C.; Chen, G.S.

1989-01-01

Anticipated transient without scram (ATWS) could be a major accident sequence with possible core melt and containment damage in a boiling water reactor (BWR). The behavior of a BWR/6 during a main steam isolation valve closure ATWS is investigated using the best-estimate computer program, RETRAN-02. The effects of both makeup coolant and boron injection on the reactor behavior are studied. It is found that the BWR/6 behaves similarly to the BWR/2 and BWR/4. Without boron injection and makeup coolant, the reactor loses its coolant inventory very quickly and the reactor power drops rapidly to ∼ 16% of rated power due to negative void reactivity. With coolant makeup from the high-pressure core spray and the reactor core isolation cooling systems, the rector reaches a quasi-steady-state condition after an initially rapidly changing transient. The dome pressure, downcomer water level, and core power oscillate around a mean value; the average core power is ∼ 15%, which is approximately equal to the power needed to heat and evaporate the subcooled makeup coolant. Lower boron concentrations in the core tend to complicate reactor behavior due to the combination of two competing phenomena: the negative boron reactivity and the positive reactivity caused by a void collapse

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

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

Burnable poison option for DUPIC fuel

International Nuclear Information System (INIS)

Choi, Hang Bok; Cupta, H. P.

1996-08-01

The mechanisms of positive coolant void reactivity of CANDU natural uranium and DUPIC fuel have been studied. The design study of DUPIC fuel was performed using the burnable poison material in the center pin to reduce the coolant void reactivity. The amount of burnable poison was determined such that the prompt inverse period of DUPIC fuel upon full coolant voiding is the same as that of natural uranium fuel at equilibrium burnup. A parametric study on various burnable poisons has shown that natural dysprosium has more merit over other materials because it uniformly controls the void reactivity throughout the burnup with reasonable amount of poison. Additional studies on the option of using scattering or absorber material in the center pin position and the option using variable fuel density were performed. In any case of option using variable fuel density were performed. In any case of options to reduce the void reactivity, it was found that either the discharge burnup and/or the relative linear pin power are sacrificed. A preliminary study was performed for the evaluation of reference DUPIC fuel performance especially represented by Stress Corrosion Cracking(SCC) parameters which is mainly influenced by the refueling operations. For the reference 2-bundle shift refueling scheme, the predicted ramped power and power increment of the reference DUPIC fuel are below the SCC thresholds of CANDU natural uranium fuel. For a 4-bundle shift refueling scheme, the envelopes of element ramped power and power increment upon refueling are 8% and 44% higher than those of a 2-bundle shift refueling scheme on the average, respectively, but still have margins to the failure thresholds of natural uranium fuel. 23 tabs., 25 figs., 20 refs. (Author)

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.

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

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)

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

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.)

A Simplified Supercritical Fast Reactor with Thorium Fuel

Directory of Open Access Journals (Sweden)

Peng Zhang

2014-01-01

Full Text Available Super-Critical water-cooled Fast Reactor (SCFR is a feasible option for the Gen-IV SCWR designs, in which much less moderator and thus coolant are needed for transferring the fission heat from the core compared with the traditional LWRs. The fast spectrum of SCFR is useful for fuel breeding and thorium utilization, which is then beneficial for enhancing the sustainability of the nuclear fuel cycle. A SCFR core is constructed in this work, with the aim of simplifying the mechanical structure and keeping negative coolant void reactivity during the whole core life. A core burnup simulation scheme based on Monte Carlo lattice homogenization is adopted in this study, and the reactor physics analysis has been performed with DU-MOX and Th-MOX fuel. The main issues discussed include the fuel conversion ratio and the coolant void reactivity. The analysis shows that thorium-based fuel can provide inherent safety for SCFR without use of blanket, which is favorable for the mechanical design of SCFR.

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)

Design and safety aspect of lead and lead-bismuth cooled long-life small safe fast reactors for various core configurations

International Nuclear Information System (INIS)

Zaki, S.; Sekimoto, Hiroshi

1995-01-01

Design and safety aspects of long-life small safe fast reactors using liquid lead or lead-bismuth coolant with metallic or nitride fuel are discussed. Neutronic analyses are performed to investigate the effect of core height to diameter ratio (H/D) on design performance of the proposed reactors. All reactors are subjected to the constraint of 12 years operation without refueling and shuffling with constant 150 MWt reactor power and also to the requirement of maximum excess reactivity during burnup to be less than 0.1%Δk. The results show that the pancake design with H/D of ∼2/3 gives the most negative coolant void coefficient under the requirements for excess reactivity. Modified designs with the central region axially fulfilled with fertile material are proposed to improve the coolant void coefficient. Thermal-hydraulic analysis results show the possibility to operate the reactors up to the end of life without changing their orifice pattern, necessary pumping power for the proposed design smaller than the conventional large sodium cooled FBR, and the natural circulation contribution of 25-40% at the normal operating condition. The reactivity feedback coefficients are also estimated and appeared to be negative for all the components including the coolant density coefficient. (author)

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...

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.

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

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

Pre-conceptual core design of a small modular fast reactor cooled by supercritical CO{sub 2}

Energy Technology Data Exchange (ETDEWEB)

Liu, Baolin; Cao, Liangzhi; Wu, Hongchun [School of Nuclear Science and Technology, Xi’an Jiaotong University, No 28, Xianning West Road, Xi’an 710049, Shaanxi (China); Yuan, Xianbao, E-mail: ztsbaby@163.com [School of Nuclear Science and Technology, Xi’an Jiaotong University, No 28, Xianning West Road, Xi’an 710049, Shaanxi (China); College of Mechanical & Power Engineering, China Three Gorges University, No 8, Daxue Road, Yichang 443002, Hubei (China); Wang, Kunpeng [Nuclear and Radiation Safety Center, PO Box 8088, Beijing 100082 (China)

2016-04-15

Abstracts: A Small Modular fast reactor cooled by Supercritical CO{sub 2} (SMoSC) is pre-conceptually designed through three-dimensional coupled neutronics/thermal-hydraulics analysis. The power rating of the SMoSC is designed to be 300 MW{sub th} to meet the energy demand of small electrical grids. The excellent thermal properties of supercritical CO{sub 2} (S-CO{sub 2}) are employed to obtain a high thermal efficiency of about 40% with an electric output of 120 MWe. MOX fuel is utilized in the core design to improve fuel efficiency. The tube-in-duct (TID) assembly is applied to get lower coolant volume fraction and reduce the positive coolant void reactivity. According to the coupled neutronics/thermal-hydraulics calculations, the coolant void reactivity is kept negative throughout the whole core life. With a specific power density of 9.6 kW/kg and an average discharge burnup of 70.1 GWd/tHM, the SmoSC can be operated for 20 Effective Full Power Years (EFPYs) without refueling.

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.)

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)

Comparison of power pulses from homogeneous and time-average-equivalent models

International Nuclear Information System (INIS)

De, T.K.; Rouben, B.

1995-01-01

The time-average-equivalent model is an 'instantaneous' core model designed to reproduce the same three dimensional power distribution as that generated by a time-average model. However it has been found that the time-average-equivalent model gives a full-core static void reactivity about 8% smaller than the time-average or homogeneous models. To investigate the consequences of this difference in static void reactivity in time dependent calculations, simulations of the power pulse following a hypothetical large-loss-of-coolant accident were performed with a homogeneous model and compared with the power pulse from the time-average-equivalent model. The results show that there is a much smaller difference in peak dynamic reactivity than in static void reactivity between the two models. This is attributed to the fact that voiding is not complete, but also to the retardation effect of the delayed-neutron precursors on the dynamic flux shape. The difference in peak reactivity between the models is 0.06 milli-k. The power pulses are essentially the same in the two models, because the delayed-neutron fraction in the time-average-equivalent model is lower than in the homogeneous model, which compensates for the lower void reactivity in the time-average-equivalent model. (author). 1 ref., 5 tabs., 9 figs

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

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...

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.)

Modeling of LVRF critical experiments in ZED-2 using WIMS9A/PANTHER and MCNP5

International Nuclear Information System (INIS)

Sissaoui, M.T.; Carlson, P.A.; Lebenhaft, J.R.

2009-01-01

The accuracy of WIMS9A/PANTHER and MCNP5 in modeling D 2 O-moderated, and H 2 O-, D 2 O- or air-cooled, doubly heterogeneous lattices of fuel clusters was demonstrated using Low Void Reactivity Fuel (LVRF) substitution experiments in the ZED-2 critical facility. MCNP5 with ENDF/B-VI (Release 5) underpredicted k eff but gave excellent coolant void reactivity (CVR) bias values. WIMS9A/PANTHER with JEF-2.2 overpredicted k eff and underpredicted the CVR bias relative to MCNP5 by 100-200 pcm. Both codes reproduced the measured axial and radial flux shapes accurately

The Steam Generating Heavy Water Reactor

International Nuclear Information System (INIS)

Middleton, J.E.

1975-01-01

An account is given of the SGHWR, the prototype of which was built by the United Kingdom Atomic Energy Authority at Winfrith, under the following headings: Introduction; origin of the SGHWR concept; conceptual design (choice of reactor type, steam cycle, reactor coolant system, nuclear behaviour, fuel design, core design, and protective, auxiliary and containment systems); operation and control (integrity of core cooling, reactivity control, power trimming, long term reactivity control, xenon override, load following, power shaping, spatial stability control, void coefficient); protective systems (breached coolant circuit trip, intact coolant circuits trip, power set-back trip); dynamic characteristics; reactor control; station control (decoupled control system, coupled control system, rate of response); Winfrith prototype (design and safety philosophy, conceptual features and parameters, reactor coolant system, protective systems, emergency core cooling, core structure, fuel design, vented containment). (U.K.)

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.

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...

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

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.)

Safety design of Pb-Bi-cooled direct contact boiling water fast reactor (PBWFR)

International Nuclear Information System (INIS)

Takahashi, Minoru; Uchida, Shoji; Yamada, Yumi; Koyama, Kazuya

2008-01-01

In Pb-Bi-cooled direct contact boiling water small fast reactor (PBWFR), steam is generated by direct contact of feedwater with primary Pb-Bi coolant above the core, and Pb-Bi coolant is circulated by steam lift pump in chimneys. Safety design has been developed to show safety features of PBWFR. Negative void reactivity is inserted even if whole of the core and upper plenum are voided hypothetically by steam intrusion from above. The control rod ejection due to coolant pressure is prevented using in-vessel type control rod driving mechanism. At coolant leak from reactor vessel and feedwater pipes, Pb-Bi coolant level in the reactor vessel required for decay heat removal is kept using closed guard vessel. Dual pipes for feedwater are employed to avoid leak of water. Although there is no concern of loss of flow accident due to primary pump trip, feedwater pump trip initiates loss of coolant flow (LOF). Injection of high pressure water slows down the flow coast down of feedwater at the LOF event. The unprotected loss of flow and heat sink (ATWS) has been evaluated, which shows that the fuel temperatures are kept lower than the safety limits. (author)

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.

Comparison of MCNP calculations against measurements in moderator temperature experiments with CANFLEX-LEU in ZED-2

International Nuclear Information System (INIS)

Watts, D.G.; Adams, F.P.; Zeller, M.B.; Bromley, B.P.

2008-01-01

This paper summarizes sample calculations of MCNP5 compared against measurements of moderator temperature coefficient experiments in the ZED-2 critical facility with CANFLEX-LEU fuel. MCNP5 is tested for key parameters associated with various reactor physics phenomena of interest for CANDU/ACR-1000) reactors, including reactivity changes with coolant density, moderator density, and moderator temperature, and also normalized flux distributions. The experimental data for these comparisons were obtained from critical experiments in AECL's ZED-2 critical facility using CANFLEX-LEU fuel in a 24-cm square lattice pitch. These comparisons establish biases/uncertainties in the calculation of k-eff, coolant void reactivity, and moderator temperature coefficient of reactivity. Results show very little bias in the moderator temperature coefficient of reactivity, and very good agreement in the calculation of normalized flux distributions. (author)

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.)

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)

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.)

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.

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)

Modeling of LVRF Critical Experiments in ZED-2 Using WIMS9A/PANTHER and MCNP5

International Nuclear Information System (INIS)

Sissaoui, M.T.; Lebenhaft, J.R; Carlson, P.A.

2008-01-01

The accuracy of WIMS9A/PANTHER and MCNP5 in modeling D 2 O-moderated, and H 2 O-, D 2 O- or air-cooled, doubly heterogeneous lattices of fuel clusters was demonstrated using Low Void Reactivity Fuel (LVRF) substitution experiments in the ZED-2 critical facility. MCNP5 with ENDF/B-VI (Release 5) under-predicted k eff but gave excellent coolant void reactivity (CVR) bias values. WIMS9A/PANTHER with JEF-2.2 over-predicted k eff and under-predicted the CVR bias relative to MCNP5 by 100 pcm to 200 pcm. Both codes reproduced the measured axial and radial flux shapes accurately. (authors)

Comparison of MCNP and WIMS-AECL/RFSP calculations with high temperature substitution experiments in ZED-2 using CANFLEX-L VRF

International Nuclear Information System (INIS)

Pencer, J.; Bromley, B.P.; Watts, D.G.; Carlson, P.; Rauket, A.; Zeller, M.

2009-01-01

This paper summarizes comparisons of calculation results from MCNP5 and WIMS-AECL / RFSP with experimental results obtained from the Zero Energy Deuterium (ZED-2) critical facility, examining CANFLEX Low Void Reactivity Fuel (CANFLEX-LVRF) in heated channels, substituted into a reference lattice and cooled under ACR-like coolant conditions, with H 2 O, air, or CO 2 as an air substitute. CANFLEX-LVRF shares features in common with the ACR-1000 fuel, notably an increase in enrichment (over natural uranium) in the outer elements of the fuel bundle, and presence of a neutron absorber in the central element. The reference and substituted fuel channels were arranged in a 24.5-cm hexagonal lattice in order to provide neutron similarity to the 24-cm square lattice pitch of the ACR-1000. These results therefore provide useful data for validation of the reactor physics toolset for use in ACR-1000 applications. For the mixed lattices, results for both MCNP5 and WIMS-AECL / RFSP show small biases in k eff , ranging from -7 mk to -5 mk, small biases in coolant void reactivity, ranging from -1 mk to +0.5 mk, and good agreement for copper activation rate distributions (based on calculated neutron flux). Bare core MCNP and WIMS-AECL stand-alone results, based on substitution analysis, also show small biases in k eff , ranging from -6 mk to -0.4 mk, and small biases in coolant void reactivity, ranging from -0.3 mk to +3.7 mk. This validation exercise thus gives good agreement between measurement and calculation and provides confidence in the accuracy of the physics toolset. (author)

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

Visualization Study of Melt Dispersion Behavior for SFR with a Metallic Fuel under Severe Accidents

Energy Technology Data Exchange (ETDEWEB)

Heo, Hyo Heo; Park, Seong Dae; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Jerng, Dong Wook [Jungang Univ., Seoul (Korea, Republic of)

2015-05-15

The safety strategy provides negative reactivity driven by the melt dispersal, so it could reduce the possibility of the recriticality event under a severe triple or more fault scenario for SFR. Since the behavior of the melt dispersion is unpredictable, it depends on the accident condition, particularly core region. While the voided coolant channel region is usually developed in the inner core, the unvoided coolant channel region is formed in the outer core. It is important to confirm the fuel dispersion with the core region, but there are not sufficient existing studies for them. From the existing studies, the coolant vapor pressure is considered as one of driving force to move the melt towards outside of the core. There is a complexity of the phenomena during intermixing of the melt with the coolant after the horizontal melt injections. It is too difficult to understand the several combined mechanisms related to the melt dispersion and the fragmentation. The specific conditions to be well dispersed for the molten metallic fuel were discussed in the experiments with the simulant materials. The each melt behavior was compared to evaluate the melt dispersion under the coolant void condition and the boiling condition.

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)

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.

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

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.

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)

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)

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.

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.

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

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

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)

Fuel assembly

International Nuclear Information System (INIS)

Kurihara, Kunitoshi; Azekura, Kazuo.

1992-01-01

In a reactor core of a heavy water moderated light water cooled pressure tube type reactor, no sufficient effects have been obtained for the transfer width to a negative side of void reactivity change in a region of a great void coefficient. Then, a moderation region divided into upper and lower two regions is disposed at the central portion of a fuel assembly. Coolants flown into the lower region can be discharged to the cooling region from an opening disposed at the upper end portion of the lower region. Light water flows from the lower region of the moderator region to the cooling region of the reactor core upper portion, to lower the void coefficient. As a result, the reactivity performance at low void coefficient, i.e., a void reaction rate is transferred to the negative side. Thus, this flattens the power distribution in the fuel assembly, increases the thermal margin and enables rapid operaiton and control of the reactor core, as well as contributes to the increase of fuel burnup ratio and reduction of the fuel cycle cost. (N.H.)

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.

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 ...

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

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

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)

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)

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)

Study on chemical reactivity control of liquid sodium. Research program

International Nuclear Information System (INIS)

Saito, Jun-ichi; Ara, Kuniaki; Sugiyama, Ken-ichiro; Kitagawa, Hiroshi; Oka, Nobuki; Yoshioka, Naoki

2007-01-01

Liquid sodium has the excellent properties as coolant of the fast breeder reactor (FBR). On the other hand, it reacts high with water and oxygen. So an innovative technology to suppress the reactivity is desired. The purpose of this study is to control the chemical reactivity of liquid sodium by dispersing the nanometer-size metallic particles (we call them Nano-particles) into liquid sodium. We focus on the atomic interaction between Nano-particles and sodium atoms. And we try to apply it to suppress the chemical reactivity of liquid sodium. Liquid sodium dispersing Nano-particles is named 'Nano-fluid'. Research programs of this study are the Nano-particles production, the evaluation of reactivity suppression of liquid sodium and the feasibility study to FBR plant. In this paper, the research programs and status are described. The important factors for particle production were understood. In order to evaluate the chemical reactivity of Nano-fluid the research programs were planned. The feasibility of the application of Nano-fluid to the coolant of FBR plant was evaluated preliminarily from the viewpoint of design and operation. (author)

Current Status of the Transmutation Reactor Technology and Preliminary Evaluation of Transmutation Performance of the KALIMER Core

Energy Technology Data Exchange (ETDEWEB)

Hong, Ser Gi; Sim, Yoon Sub; Kim, Yeong Il; Kim, Young Gyum; Lee, Byung Woon; Song, Hoon; Lee, Ki Bog; Jang, Jin Wook; Lee, Dong Uk

2005-08-15

Recently the most countries using the nuclear power plants for electricity generation have been faced with the problem of the preparation of the repository for the disposition of the nuclear waste generated from LWR. It was well-known that the issues related with long term risk of the radioactive wastes for the future generations are due only to 1% of the total waste. This small fraction of 1% consists of transuranic (TRU) nuclides such as Pu, Np, Am, Cm and the long lived fission products such as Tc and I. For the transuranic (TRU) nuclides, their half lives range from several years to several hundred thousands years and hence their radioactive toxicity can be lasted over very long time period. This has made the change of the rule of the fast spectrum reactor from the economical use of uranium resource through breeding to the reduction of the nuclear waste through the transmutation. The purpose of this study is to obtain the basic knowledge on the nuclear transmutation technology and to suggest the technical solution ways for the future technology development and enhancement through a survey of the state-of-art of the international research on the nuclear transmutation. The increase of the transmutation rate requires the reduction of the breeding ratio. In fact, the transmutation rate is determined by the breeding ratio. The reduction of the breeding ratio can be achieved by reducing the U-238 content in fuel or increasing the neutron leakage through core boundary or absorbing the neutrons by using some absorbers. However, the reduction of the U-238 content results in the degradation of the fuel Doppler coefficient that is one of the most important safety-related parameters and the reduction of the effective delayed neutron fraction that is related with the controllability of the reactor core. Also, the increase of the transmutation rate can lead to the increase of the coolant void reactivity worth unless some ways to reduce the coolant void reactivity are not

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)

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

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.

A 3D stylized half-core CANDU benchmark problem

International Nuclear Information System (INIS)

Pounders, Justin M.; Rahnema, Farzad; Serghiuta, Dumitru; Tholammakkil, John

2011-01-01

A 3D stylized half-core Canadian deuterium uranium (CANDU) reactor benchmark problem is presented. The benchmark problem is comprised of a heterogeneous lattice of 37-element natural uranium fuel bundles, heavy water moderated, heavy water cooled, with adjuster rods included as reactivity control devices. Furthermore, a 2-group macroscopic cross section library has been developed for the problem to increase the utility of this benchmark for full-core deterministic transport methods development. Monte Carlo results are presented for the benchmark problem in cooled, checkerboard void, and full coolant void configurations.

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)

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...

All heavy metals closed-cycle analysis on water-cooled reactors of uranium and thorium fuel cycle systems

International Nuclear Information System (INIS)

Permana, Sidik; Sekimoto, Hiroshi; Waris, Abdul; Takaki, Naoyuki

2009-01-01

Uranium and Thorium fuels as the basis fuel of nuclear energy utilization has been used for several reactor types which produce trans-uranium or trans-thorium as 'by product' nuclear reaction with higher mass number and the remaining uranium and thorium fuels. The utilization of recycled spent fuel as world wide concerns are spent fuel of uranium and plutonium and in some cases using recycled minor actinide (MA). Those fuel schemes are used for improving an optimum nuclear fuel utilization as well to reduce the radioactive waste from spent fuels. A closed-cycle analysis of all heavy metals on water-cooled cases for both uranium and thorium fuel cycles has been investigated to evaluate the criticality condition, breeding performances, uranium or thorium utilization capability and void reactivity condition. Water-cooled reactor is used for the basic design study including light water and heavy water-cooled as an established technology as well as commercialized nuclear technologies. A developed coupling code of equilibrium fuel cycle burnup code and cell calculation of SRAC code are used for optimization analysis with JENDL 3.3 as nuclear data library. An equilibrium burnup calculation is adopted for estimating an equilibrium state condition of nuclide composition and cell calculation is performed for calculating microscopic neutron cross-sections and fluxes in relation to the effect of different fuel compositions, different fuel pin types and moderation ratios. The sensitivity analysis such as criticality, breeding performance, and void reactivity are strongly depends on moderation ratio and each fuel case has its trend as a function of moderation ratio. Heavy water coolant shows better breeding performance compared with light water coolant, however, it obtains less negative or more positive void reactivity. Equilibrium nuclide compositions are also evaluated to show the production of main nuclides and also to analyze the isotopic composition pattern especially

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...

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)

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)

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)

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

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.

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)

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)

Modelling of 28-element UO2 flux-map critical experiments in ZED-2 using WIMS9A/PANTHER

International Nuclear Information System (INIS)

Sissaoui, M.T.; Kozier, K.S.; Labrie, J.P.

2011-01-01

The accuracy of WIMS9A/PANTHER in modelling D 2 O-moderated, and H 2 O- or air-cooled, doubly heterogeneous lattices of fuel clusters has been demonstrated using 28-element UO 2 flux-map critical experiments in the ZED-2 facility. Presented here are the predicted k eff values, coolant void reactivity biases, and the radial and axial flux shapes.

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

Neutron dynamics of fast-spectrum dedicated cores for waste transmutation

International Nuclear Information System (INIS)

Massara, S.

2002-04-01

Among different scenarios achieving minor actinide transmutation, the possibility of double strata scenarios with critical, fast spectrum, dedicated cores must be checked and quantified. In these cores, the waste fraction has to be at the highest level compatible with safety requirements during normal operation and transient conditions. As reactivity coefficients are poor in such critical cores (low delayed neutron fraction and Doppler feed-back, high coolant void coefficient), their dynamic behaviour during transient conditions must be carefully analysed. Three nitride-fuel configurations have been analysed: two liquid metal-cooled (sodium and lead) and a particle-fuel helium-cooled one. A dynamic code, MAT4 DYN, has been developed during the PhD thesis, allowing the study of loss of flow, reactivity insertion and loss of coolant accidents, and taking into account two fuel geometries (cylindrical and spherical) and two thermal-hydraulics models for the coolant (incompressible for liquid metals and compressible for helium). Dynamics calculations have shown that if the fuel nature is appropriately chosen (letting a sufficient margin during transients), this can counterbalance the bad state of reactivity coefficients for liquid metal-cooled cores, thus proving the interest of this kind of concept. On the other side, the gas-cooled core dynamics is very badly affected by the high value of the helium void coefficient (which is a consequence of the choice of a hard spectrum), this effect being amplified by the very low thermal inertia of particle-fuel design. So, a new kind of concept should be considered for a helium-cooled fast-spectrum dedicated core. (authors)

Neutron dynamics of fast-spectrum dedicated cores for waste transmutation; Etude et amelioration du comportement cinetique de coeurs rapides a la transmutation de dechets a vie longue

Energy Technology Data Exchange (ETDEWEB)

Massara, S

2002-04-01

Among different scenarios achieving minor actinide transmutation, the possibility of double strata scenarios with critical, fast spectrum, dedicated cores must be checked and quantified. In these cores, the waste fraction has to be at the highest level compatible with safety requirements during normal operation and transient conditions. As reactivity coefficients are poor in such critical cores (low delayed neutron fraction and Doppler feed-back, high coolant void coefficient), their dynamic behaviour during transient conditions must be carefully analysed. Three nitride-fuel configurations have been analysed: two liquid metal-cooled (sodium and lead) and a particle-fuel helium-cooled one. A dynamic code, MAT4 DYN, has been developed during the PhD thesis, allowing the study of loss of flow, reactivity insertion and loss of coolant accidents, and taking into account two fuel geometries (cylindrical and spherical) and two thermal-hydraulics models for the coolant (incompressible for liquid metals and compressible for helium). Dynamics calculations have shown that if the fuel nature is appropriately chosen (letting a sufficient margin during transients), this can counterbalance the bad state of reactivity coefficients for liquid metal-cooled cores, thus proving the interest of this kind of concept. On the other side, the gas-cooled core dynamics is very badly affected by the high value of the helium void coefficient (which is a consequence of the choice of a hard spectrum), this effect being amplified by the very low thermal inertia of particle-fuel design. So, a new kind of concept should be considered for a helium-cooled fast-spectrum dedicated core. (authors)

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.

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.

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.)

Change in CANDU-6 reactivity following a power reduction at low PHT purity

International Nuclear Information System (INIS)

Whitlock, J.J.; Soulard, M.R.; Baudouin, A.

1995-01-01

The reactivity effect of a power reduction in CANDU-6 is examined using a three-dimensional, steady-state, coupled neutronics/thermalhydraulics methodology, starting from a global irradiation distribution matched to site data. The power reduction is sufficient to suppress coolant boiling in the fuel channels, and thus the significant parameters affecting reactivity are an increase in coolant density and a decrease in fuel temperature. These individual components are estimated using infinite-lattice-cell methodology. The effect of using newer methodology, particularly for the thermalhydraulic analysis, is examined by comparison with previous simulations. (author). 10 refs., 7 tabs., 1 fig

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)

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)

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.)

Evaluation of reactivity and Xe behavior during daily load following operation

International Nuclear Information System (INIS)

Sakamoto, Yasunori; Araki, Tsuneyasu; Yamamoto, Fumiaki

1992-01-01

A boiling water reactor (BWR) has an excellent load following capability provided by a core flow control, which is used for changing a reactor power level and for compensating the subsequent Xe concentration change. The core characteristics during load following operations are investigated in detail, using our reactor core simulator. Comparisons of changes of the Doppler reactivity, the void reactivity and the Xe reactivity during transients are performed. Also the features of Xe transient during load following operations are shown. It has been shown that the core flow change required to compensate the Xe reactivity change produces much greater change of the void reactivity than that required for power level changes, and that the resulting local power change in the lower part of the core is greater than that in the upper part, because the Xe concentration change in the lower part is hardly compensated by the core flow control. Also the effects of power level changes, cycle patterns, and initial concentration of Xe and I on the Xe transient behavior have been investigated. (author)

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.)

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...

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

Conceptual designing of a reduced moderation pressurized water reactor by use of MVP and MVP-BURN

International Nuclear Information System (INIS)

Kugo, T.

2001-01-01

A conceptual design of a seed-blanket assembly PWR core with a complicated geometry and a strong heterogeneity has been carried forward by use of the continuous-energy Monte Carlo method. Through parametric survey calculations by repeated use of MVP and a lattice burn-up calculation by MVP-BURN, a seed-blanket assembly configuration suitable for a concept of RMWR has been established, by evaluating precisely reactivity, a conversion ratio and a coolant void reactivity coefficient in a realistic computation time on a super computer. (orig.)

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.

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.

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

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

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.)

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

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

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

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

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.)

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)

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...

Experimental analysis of upward vertical two-phase flow in four-cusp channels simulating the conditions of a typical nuclear reactor channel, degraded by a loss of coolant accident

International Nuclear Information System (INIS)

Assad, A.C.A.

1984-01-01

The present work deals with an experimental analysis of upward vertical two-phase flow in channels with circular and four-cusp cross-sections. The latter simulates the conditions of a typical nuclear reactor channel, degraded by a loss of coolant accident. Simultaneous flow of air and water has been employed to simulate adiabatic steam-water flow. The installation of air-water separators helped eliminate instabilities during pressure-drop measurements. The gamma ray attenuation was utilized for the void fraction determination. For the four-cusp geommetry, new criteria for two-phase flow regime transitions have been determined, as well as new correlatins for pressure drop and void fraction, as function of the Lockhart-Martinelli factor and vapour mass-fraction, respectively. (Author) [pt

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

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

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)

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)

Application of noise analysis technique for monitoring the moderator temperature coefficient of reactivity in pressurized water reactors

International Nuclear Information System (INIS)

Shieh, D.J.; Upadhyaya, B.R.; Sweeney, F.J.

1987-01-01

A new technique, based on the noise analysis of neutron detector and core-exit coolant temperature signals, is developed for monitoring the moderator temperature coefficient of reactivity in pressurized water reactors (PWRs). A detailed multinodal model is developed and evaluated for the reactor core subsystem of the loss-of-fluid test (LOFT) reactor. This model is used to study the effect of changing the sign of the moderator temperature coefficient of reactivity on the low-frequency phase angle relationship between the neutron detector and the core-exit temperature noise signals. Results show that the phase angle near zero frequency approaches - 180 deg for negative coefficients and 0 deg for positive coefficients when the perturbation source for the noise signals is core coolant flow, inlet coolant temperature, or random heat transfer

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.

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

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...

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)

Study of characteristics of Th-U cycle in CANDU SCWR

International Nuclear Information System (INIS)

Shi, J.; Shi, G.

2010-01-01

The flexibility of CANDU technology allows the use of different fuel cycles including various uranium-driven thorium cycles. Direct self-recycle method and heterogeneous cycle modes with supercritical water as coolant were studied for (U,Th)O 2 CANFLEX fuel bundle. Lattice pitch and enrichment of driver fuel were treated as independent variables, taking account of coolant void reactivity, fuel burnup, and linear power uneven factor. In the end, appropriate cycle mode and parameters of bundle were chosen for (U,Th)O 2 cycle in CANDU SCWR. Calculations were processed by the two-dimensional multigroup neutron transport code WIMS-AECL release 3.1.2.1. (author)

The fuel string relocation effect - why the Bruce reactors were derated

Energy Technology Data Exchange (ETDEWEB)

Gold, M; Farooqui, M Z; Adebiyi, A S; Chu, R Y; Le, N T; Oliva, A F [Ontario Hydro, Toronto, ON (Canada); Balog, G; Qu, T; DeBuda, P G [Ontario Hydro, Tiverton, ON (Canada). Bruce Nuclear Generating Station-A

1996-12-31

In the CANDU Safety Analysis process, a series of design basis accidents are chosen and analyzed to confirm safety system effectiveness. Of all the postulated accidents, the Large Break Loss of Coolant Accident (LBLOCA) - a postulated break in the Heat Transport System piping near a component that services a large number of fuel channels - sets the most demanding requirements on the speed and reactivity depth of the shutdown system devices - shutoff rods and liquid poison injection. While the event is extremely improbable, it is reanalyzed periodically and its consequences examined to ensure continued shutdown system effectiveness. In March 1993, an additional effect was identified: if the break occurred in the piping on the inlet side of the core, this would cause sudden movement of the fuel bundles (so-called fuel string relocation) in a large number of channels. In Ontario Hydro`s Bruce NGS A, Bruce NGS B and Darlington reactors, each channel is fuelled against the flow. In this situation, the relocation of the fuel string results in a sudden positive reactivity increase. This reactivity increase is in addition to the reactivity due to the core coolant voiding. The combined reactivity effect could lead to power pulses much higher than those that would arise due to coolant voiding alone. To maintain safety margins in the event of such a postulated accident, the eight Bruce NGS A and Bruce NGS B units were initially derated to 60 percent power within 2 days of the identification and confirmation of this effect. This paper: describes the fuel string relocation phenomenon in detail; explains why the consequences differ at the various Ontario Hydro reactors; outlines the actions taken with respect to each of the Ontario Hydro reactors in the months following March 1993; describes the design solutions implemented to mitigate the problem and return the Bruce reactors to higher powers. 6 refs., 1 tab., 6 figs.

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

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.

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

Reactor core monitoring device

International Nuclear Information System (INIS)

Ishii, Takanobu; Handa, Hiroaki; Hayashi, Katsumi; Narita, Hitoshi; Shimozaki, Takaaki

1995-01-01

The device of the present invention reliably and conveniently detects an event of rapid increase of a coolant void coefficient at a portion of a channel by flow channel clogging event in a PWR-type reactor. Namely, upon flow channel clogging event, the coolant void coefficient is increased, an effective density is lowered, and a coolant shielding effect is lowered. Therefore, fast neutron fluxes at the periphery of a pressure tube are increased. The increase of the fast neutron fluxes is detected by a fast neutron flux detector disposed in a guide tube of an existent neutron flux detector. Based on the result, increase of coolant void coefficient can be detected. When an average void coefficient reaches from 30% to 100%, for example, the fast neutron fluxes are increased by about twice at a neutron permeation distance of coolants of about 10cm, thereby enabling to perform effective detection. (I.S.)

Study of IPR-R1 dynamics by reactivity random excitations

International Nuclear Information System (INIS)

Roedel, G.

1983-01-01

To demonstrate the viability of the utilization of analitical techniques of neutronic noise, a dynamic model for IPR-R1 reactor from CDTN was developed. This model allows reactivity feedback due to variations of temperature in fuel and coolant [pt

A Conceptual Study on a Supercritical CO_2-cooled Micro Modular Reactor

International Nuclear Information System (INIS)

Yu, Hwanyeal; Hartanto, Donny; Kim, Yonghee

2014-01-01

A Micro Modular Reactor (MMR) using Supercritical-CO_2 (S-CO_2) as coolant has been investigated from the neutronics perspective. The MMR is designed to be transportable so it can reach the remote areas. The thermal power of the reactor is 36.2 M Wth. The size of the active core is limited to 1.2 m length and 93.16 cm width. The size of whole core is 2.8 m length and 166.9 cm width. The reactor lifetime design target is 20 years. To maximize the fuel volume fraction in the core, high density uranium nitride UN"1"5 was used. The PbO/MgO reflector was also utilized to improve the neutron economy. The S-CO_2 is chosen as the coolant because it offers a higher thermal efficiency. In this study, neutronics calculations and depletion using McCARD Monte Carlo code has been done to determine the lifetime and behavior of the core. Several important safety parameters such as Control Rod worth, Doppler reactivity coefficients and coolant void reactivity coefficient have also been analyzed. (author)

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.

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.)

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.

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.

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

Development of demonstration advanced thermal reactor

Energy Technology Data Exchange (ETDEWEB)

Nishimura, Seiji; Oguchi, Isao; Touhei, Kazushige

1982-08-01

The design of the advanced thermal demonstration reactor with 600 MWe output was started in 1975. In order to make the compact core, 648 fuel assemblies, each comprising 36 fuel rods, were used, and the mean channel output was increased by 20% as compared with the prototype reactor. The heavy water dumping mechanism for the calandria was abolished. Advanced thermal reactors are suitable to burn plutonium, since the control rod worth does not change, the void reactivity coefficient of coolant shifts to the negative side, and the harmful influence of high order plutonium is small. The void reactivity coefficient is nearly zero, the fluctuation of output in relation to pressure disturbance is small, and the local output change of fuel by the operation of control rods is small, therefore, the operation following load change is relatively easy. The coolant recirculation system is of independent loop construction dividing the core into two, and steam and water are separated in respective steam drums. At present, the rationalizing design is in progress by the leadership of the Power Reactor and Nuclear Fuel Development Corp. The outline of the demonstration reactor, the reactor construction, the nuclear-thermal-hydraulic characteristics and the output control characteristics are reported.

Development of demonstration advanced thermal reactor

International Nuclear Information System (INIS)

Nishimura, Seiji; Oguchi, Isao; Touhei, Kazushige.

1982-01-01

The design of the advanced thermal demonstration reactor with 600 MWe output was started in 1975. In order to make the compact core, 648 fuel assemblies, each comprising 36 fuel rods, were used, and the mean channel output was increased by 20% as compared with the prototype reactor. The heavy water dumping mechanism for the calandria was abolished. Advanced thermal reactors are suitable to burn plutonium, since the control rod worth does not change, the void reactivity coefficient of coolant shifts to the negative side, and the harmful influence of high order plutonium is small. The void reactivity coefficient is nearly zero, the fluctuation of output in relation to pressure disturbance is small, and the local output change of fuel by the operation of control rods is small, therefore, the operation following load change is relatively easy. The coolant recirculation system is of independent loop construction dividing the core into two, and steam and water are separated in respective steam drums. At present, the rationalizing design is in progress by the leadership of the Power Reactor and Nuclear Fuel Development Corp. The outline of the demonstration reactor, the reactor construction, the nuclear-thermal-hydraulic characteristics and the output control characteristics are reported. (Kako, I.)

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

Isothermal temperature reactivity coefficient measurement in TRIGA reactor

International Nuclear Information System (INIS)

Zagar, T.; Ravnik, M.; Trkov, A.

2002-01-01

Direct measurement of an isothermal temperature reactivity coefficient at room temperatures in TRIGA Mark II research reactor at Jozef Stefan Institute in Ljubljana is presented. Temperature reactivity coefficient was measured in the temperature range between 15 o C and 25 o C. All reactivity measurements were performed at almost zero reactor power to reduce or completely eliminate nuclear heating. Slow and steady temperature decrease was controlled using the reactor tank cooling system. In this way the temperatures of fuel, of moderator and of coolant were kept in equilibrium throughout the measurements. It was found out that TRIGA reactor core loaded with standard fuel elements with stainless steel cladding has small positive isothermal temperature reactivity coefficient in this temperature range.(author)

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...

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)

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.

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.)

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

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.)

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.)

Hot spot formation and chemical reaction initiation in shocked HMX crystals with nanovoids: a large-scale reactive molecular dynamics study.

Science.gov (United States)

Zhou, Tingting; Lou, Jianfeng; Zhang, Yangeng; Song, Huajie; Huang, Fenglei

2016-07-14

We report million-atom reactive molecular dynamic simulations of shock initiation of β-cyclotetramethylene tetranitramine (β-HMX) single crystals containing nanometer-scale spherical voids. Shock induced void collapse and subsequent hot spot formation as well as chemical reaction initiation are observed which depend on the void size and impact strength. For an impact velocity of 1 km s(-1) and a void radius of 4 nm, the void collapse process includes three stages; the dominant mechanism is the convergence of upstream molecules toward the centerline and the downstream surface of the void forming flowing molecules. Hot spot formation also undergoes three stages, and the principal mechanism is kinetic energy transforming to thermal energy due to the collision of flowing molecules on the downstream surface. The high temperature of the hot spot initiates a local chemical reaction, and the breakage of the N-NO2 bond plays the key role in the initial reaction mechanism. The impact strength and void size have noticeable effects on the shock dynamical process, resulting in a variation of the predominant mechanisms leading to void collapse and hot spot formation. Larger voids or stronger shocks result in more intense hot spots and, thus, more violent chemical reactions, promoting more reaction channels and generating more reaction products in a shorter duration. The reaction products are mainly concentrated in the developed hot spot, indicating that the chemical reactivity of the hmx crystal is greatly enhanced by void collapse. The detailed information derived from this study can aid a thorough understanding of the role of void collapse in hot spot formation and the chemical reaction initiation of explosives.

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

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...

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)

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)

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)

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)

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

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.

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

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

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

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.

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.

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

On-line prediction of BWR transients in support of plant operation and safety analyses

International Nuclear Information System (INIS)

Wulff, W.; Cheng, H.S.; Lekach, S.V.; Mallen, A.N.

1983-01-01

A combination of advanced modeling techniques and modern, special-purpose peripheral minicomputer technology is presented which affords realistic predictions of plant transient and severe off-normal events in LWR power plants through on-line simulations at a speed ten times greater than actual process speeds. Results are shown for a BWR plant simulation. The mathematical models account for nonequilibrium, nonhomogeneous two-phase flow effects in the coolant, for acoustical effects in the steam line and for the dynamics of the recirculation loop and feed-water train. Point kinetics incorporate reactivity feedback for void fraction, for fuel temperature, and for coolant temperature. Control systems and trip logic are simulated for the nuclear steam supply system

Safety-Related Optimization and Analyses of an Innovative Fast Reactor Concept

Directory of Open Access Journals (Sweden)

Dalin Zhang

2012-06-01

Full Text Available Since a fast reactor core with uranium-plutonium fuel is not in its most reactive configuration under operating conditions, redistribution of the core materials (fuel, steel, sodium during a core disruptive accident (CDA may lead to recriticalities and as a consequence to severe nuclear power excursions. The prevention, or at least the mitigation, of core disruption is therefore of the utmost importance. In the current paper, we analyze an innovative fast reactor concept developed within the CP-ESFR European project, focusing on the phenomena affecting the initiation and the transition phases of an unprotected loss of flow (ULOF accident. Key phenomena for the initiation phase are coolant boiling onset and further voiding of the core that lead to a reactivity increase in the case of a positive void reactivity effect. Therefore, the first level of optimization involves the reduction, by design, of the positive void effect in order to avoid entering a severe accident. If the core disruption cannot be avoided, the accident enters into the transition phase, characterized by the progression of core melting and recriticalities due to fuel compaction. Dedicated features that enhance and guarantee a sufficient and timely fuel discharge are considered for the optimization of this phase.

Reactivity feedbacks of a material test research reactor fueled with various low enriched uranium dispersion fuels

International Nuclear Information System (INIS)

Muhammad, Farhan; Majid, Asad

2009-01-01

The reactivity feedbacks of a material test research reactor using various low enriched uranium fuels, having same uranium density were calculated. For this purpose, the original aluminide fuel (UAl x -Al) containing 4.40 gU/cm 3 of an MTR was replaced with silicide (U 3 Si-Al and U 3 Si 2 -Al) and oxide (U 3 O 8 -Al) dispersion fuels having the same uranium density as of the original fuel. Calculations were carried out to find the fuel temperature reactivity feedback, moderator temperature reactivity feedback, moderator density reactivity feedback and moderator void reactivity feedback. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It was observed that the magnitudes all the respective reactivity feedbacks from 38 deg. C to 50 deg. C and 100 deg. C, at the beginning of life, of all the fuels were very close to each other. The fuel temperature reactivity feedback of the U 3 O 8 -Al was about 2% more than the original UAl x -Al fuel. The magnitudes of the moderator temperature, moderator density and moderator void reactivity feedbacks of all the fuels, showed very minor variations from the original aluminide fuel.

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

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.

Characterizing the effects of elevated temperature on the air void pore structure of advanced gas-cooled reactor pressure vessel concrete using x-ray computed tomography

Directory of Open Access Journals (Sweden)

Withers P.J.

2013-07-01

Full Text Available X-ray computed tomography (X-ray CT has been applied to nondestructively characterise changes in the microstructure of a concrete used in the pressure vessel structure of Advanced Gas-cooled Reactors (AGR in the UK. Concrete specimens were conditioned at temperatures of 105 °C and 250 °C, to simulate the maximum thermal load expected to occur during a loss of coolant accident (LOCA. Following thermal treatment, these specimens along with an unconditioned control sample were characterised using micro-focus X-ray CT with a spatial resolution of 14.6 microns. The results indicate that the air void pore structure of the specimens experienced significant volume changes as a result of the increasing temperature. The increase in the porous volume was more prevalent at 250 °C. Alterations in air void size distributions were characterized with respect to the unconditioned control specimen. These findings appear to correlate with changes in the uni-axial compressive strength of the conditioned concrete.

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

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....

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.)

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.)

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

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)

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

Uncertainty Evaluation of Reactivity Coefficients for a large advanced SFR Core Design

International Nuclear Information System (INIS)

Khamakhem, Wassim; Rimpault, Gerald

2008-01-01

Sodium Cooled Fast Reactors are currently being reshaped in order to meet Generation IV goals on economics, safety and reliability, sustainability and proliferation resistance. Recent studies have led to large SFR cores for a 3600 MWth power plants, cores which exhibit interesting features. The designs have had to balance between competing aspects such as sustainability and safety characteristics. Sustainability in neutronic terms is translated into positive breeding gain and safety into rather low Na void reactivity effects. The studies have been done on two SFR concepts using oxide and carbide fuels. The use of the sensitivity theory in the ERANOS determinist code system has been used. Calculations have been performed with different sodium evaluations: JEF2.2, ERALIB-1 and the most recent JEFF3.1 and ENDF/B-VII in order to make a broad comparison. Values for the Na void reactivity effect exhibit differences as large as 14% when using the different sodium libraries. Uncertainties due to nuclear data on the reactivity coefficients were performed with BOLNA variances-covariances data, the Na Void Effect uncertainties are near to 12% at 1σ. Since, the uncertainties are far beyond the target accuracy for a design achieving high performance, two directions are envisaged: the first one is to perform new differential measurements or in a second attempt use integral experiments to improve effectively the nuclear data set and its uncertainties such as performed in the past with ERALIB1. (authors)

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.)

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)

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)

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.

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)

Voids and superstructures: correlations and induced large-scale velocity flows

Science.gov (United States)

Lares, Marcelo; Luparello, Heliana E.; Maldonado, Victoria; Ruiz, Andrés N.; Paz, Dante J.; Ceccarelli, Laura; Garcia Lambas, Diego

2017-09-01

The expanding complex pattern of filaments, walls and voids build the evolving cosmic web with material flowing from underdense on to high density regions. Here, we explore the dynamical behaviour of voids and galaxies in void shells relative to neighbouring overdense superstructures, using the Millenium simulation and the main galaxy catalogue in Sloan Digital Sky Survey data. We define a correlation measure to estimate the tendency of voids to be located at a given distance from a superstructure. We find voids-in-clouds (S-types) preferentially located closer to superstructures than voids-in-voids (R-types) although we obtain that voids within ˜40 h-1 Mpc of superstructures are infalling in a similar fashion independently of void type. Galaxies residing in void shells show infall towards the closest superstructure, along with the void global motion, with a differential velocity component depending on their relative position in the shell with respect to the direction to the superstructure. This effect is produced by void expansion and therefore is stronger for R-types. We also find that galaxies in void shells facing the superstructure flow towards the overdensities faster than galaxies elsewhere at the same relative distance to the superstructure. The results obtained for the simulation are also reproduced for the Sky Survey Data Release data with a linearized velocity field implementation.

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.)

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

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.

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

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.)

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.

Development of quick-response area-averaged void fraction meter

International Nuclear Information System (INIS)

Watanabe, Hironori; Iguchi, Tadashi; Kimura, Mamoru; Anoda, Yoshinari

2000-11-01

Authors are performing experiments to investigate BWR thermal-hydraulic instability under coupling of neutronics and thermal-hydraulics. To perform the experiment, it is necessary to measure instantaneously area-averaged void fraction in rod bundle under high temperature/high pressure gas-liquid two-phase flow condition. Since there were no void fraction meters suitable for these requirements, we newly developed a practical void fraction meter. The principle of the meter is based on the electrical conductance changing with void fraction in gas-liquid two-phase flow. In this meter, metal flow channel wall is used as one electrode and a L-shaped line electrode installed at the center of flow channel is used as the other electrode. This electrode arrangement makes possible instantaneous measurement of area-averaged void fraction even under the metal flow channel. We performed experiments with air/water two-phase flow to clarify the void fraction meter performance. Experimental results indicated that void fraction was approximated by α=1-I/I o , where α and I are void fraction and current (I o is current at α=0). This relation holds in the wide range of void fraction of 0∼70%. The difference between α and 1-I/I o was approximately 10% at maximum. The major reasons of the difference are a void distribution over measurement area and an electrical insulation of the center electrode by bubbles. The principle and structure of this void fraction meter are very basic and simple. Therefore, the meter can be applied to various fields on gas-liquid two-phase flow studies. (author)

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.)

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.)

Recent IAEA Achievements in the Field of Fast Neutron Systems and Scope and Objectives of the Meeting

International Nuclear Information System (INIS)

Monti, S.

2013-01-01

Scope of this technical meeting: • Within FR development programmes, significant research and development (R&D) efforts are devoted to the design of innovative reactor cores ⇒ intrinsic safety features (enhanced negative reactivity feedbacks, reduced coolant void reactivity effects, etc.), ⇒ high performance (in terms of cycle length, high fuel burnup, breeding gain, etc.), ⇒ Minor Actinide transmutation capability. • The development of high performance in-core structural materials represents one of the most challenging aspects ⇒ high neutron flux, ⇒ liquid metal coolant, ⇒ high temperatures. Objectives and expected outcomes: • Present and discuss results of studies and on-going R&D and design activities in the field of innovative reactor core concepts; • Present and discuss results of studies and on-going R&D activities in the field of advanced reactor core structural materials; • Identification of research and technology gaps to be covered through new R&D initiatives to be carried out under the aegis of the IAEA

Analysis of multiple failure accident scenarios for development of probabilistic safety assessment model for KALIMER-600

International Nuclear Information System (INIS)

Kim, T.W.; Suk, S.D.; Chang, W.P.; Kwon, Y.M.; Jeong, H.Y.; Lee, Y.B.; Ha, K.S.; Kim, S.J.

2009-01-01

A sodium-cooled fast reactor (SFR), KALIMER-600, is under development at KAERI. Its fuel is the metal fuel of U-TRU-Zr and it uses sodium as coolant. Its advantages are found in the aspects of an excellent uranium resource utilization, inherent safety features, and nonproliferation. The probabilistic safety assessment (PSA) will be one of the initiating subjects for designing it from the aspects of a risk informed design (RID) as well as a technology-neutral licensing (TNL). The core damage is defined as coolant voiding, fuel melting, or cladding damage. Accident scenarios which lead to the core damage should be identified for the development of a Level-1 PSA model. The SSC-K computer code is used to identify the conditions which lead to core damage. KALIMER-600 has passive safety features such as passive shutdown functions, passive pump coast-down features, and passive decay heat removal systems. It has inherent reactivity feedback effects such as Doppler, sodium void, core axial expansion, control rod axial expansion, core radial expansion, etc. The accidents which are analyzed are the multiple failure accidents such as an unprotected transient overpower, a loss of flow, and a loss of heat sink events with degraded safety systems or functions. The safety functions to be considered here are a reactor trip, inherent reactivity feedback features, the pump coast-down, and the passive decay heat removal. (author)

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.)

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

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.

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

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...

Analytical evaluation on dynamical response characteristics of reduced-moderation water reactor with tight-lattice core under natural circulation core cooling

International Nuclear Information System (INIS)

Ishikawa, Nobuyuki; Okubo, Tsutomu

2009-01-01

The time-domain analyses with TRAC-BF1 code were performed for clarifying the dynamical response characteristics of the reduced-moderation water reactor (RMWR) with tight-lattice core configuration. The response characteristics were evaluated based on the step response basically utilized for dynamical system evaluation. As for the most fundamental dynamical characteristics, the channel flow response characteristics of single fuel assembly were evaluated. In the evaluation, the appropriate single-phase pressure drop setting at the inlet orifice was determined in terms of response stability from the design viewpoint. In addition, from the investigation on the relation of the response and transit time of coolant, it is confirmed that the channel flow response of RMWR is dominated by the transit time of vapor phase resulting from a high void fraction operation condition. As for a natural circulation flow response, it is clarified that the response is strongly influenced by the effect of two-phase pressure loss owing to a high void fraction condition. The reactor power response with reactivity feedback shows quite stable response characteristics on account of the small absolute value of void reactivity coefficient.

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.

Development of a computer code for Dalat research reactor transient analysis

International Nuclear Information System (INIS)

Le Vinh Vinh; Nguyen Thai Sinh; Huynh Ton Nghiem; Luong Ba Vien; Pham Van Lam; Nguyen Kien Cuong

2003-01-01

DRSIM (Dalat Reactor SIMulation) computer code has been developed for Dalat reactor transient analysis. It is basically a coupled neutronics-hydrodynamics-heat transfer code employing point kinetics, one dimensional hydrodynamics and one dimensional heat transfer. The work was financed by VAEC and DNRI in the framework of institutional R and D programme. Some transient problems related to reactivity and loss of coolant flow was carried out by DRSIM using temperature and void coefficients calculated by WIMS and HEXNOD2D codes. (author)

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

Thermal-hydraulics and neutronics studies on the FP7 CP-ESFR oxide and carbide cores

Energy Technology Data Exchange (ETDEWEB)

Ammirabile, L.; Tsige-Tamirat, H. [European Commission, JRC, Inst. for Energy, Petten (Netherlands)

2011-07-01

In the framework of the the Collaborative Project on European Sodium Fast Reactor (CP-ESFR) two core designs that are currently being proposed for the 3600 MWth sodium-cooled reactor concept: one is based on oxide fuel and the other on carbide fuel. Using the European Safety Assessment Platform (ESAP), JRC-IE has conducted static calculation on neutronics (incl. reactivity coefficients) and thermal-hydraulic characteristics for both oxide and carbide reference cores. The quantities evaluated include: keff, coolant heat-up, void, and Doppler reactivity coefficients, axial and radial expansion reactivity coefficients, pin-by-pin calculated power profiles, average and peak channel temperatures. This paper presents the ESAP models applied in the study together with the relevant results for the oxide and carbide core. (author)

Thermal-hydraulics and neutronics studies on the FP7 CP-ESFR oxide and carbide cores

International Nuclear Information System (INIS)

Ammirabile, L.; Tsige-Tamirat, H.

2011-01-01

In the framework of the the Collaborative Project on European Sodium Fast Reactor (CP-ESFR) two core designs that are currently being proposed for the 3600 MWth sodium-cooled reactor concept: one is based on oxide fuel and the other on carbide fuel. Using the European Safety Assessment Platform (ESAP), JRC-IE has conducted static calculation on neutronics (incl. reactivity coefficients) and thermal-hydraulic characteristics for both oxide and carbide reference cores. The quantities evaluated include: keff, coolant heat-up, void, and Doppler reactivity coefficients, axial and radial expansion reactivity coefficients, pin-by-pin calculated power profiles, average and peak channel temperatures. This paper presents the ESAP models applied in the study together with the relevant results for the oxide and carbide core. (author)

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

Nuclear power plant simulation on the AD10

International Nuclear Information System (INIS)

Wulff, W.; Cheng, H.S.; Mallen, A.N.; Stritar, A.

1985-01-01

A combination of advanced modeling techniques and the modern, special-purpose peripheral minicomputer AD10 is presented which affords realistic predictions of plant transient and severe off-normal events in LWR power plants through on-line simulations at a speed ten times greater than actual process speeds. Results are shown for a BWR plant simulation. The mathematical models account for nonequilibrium, nonhomogeneous two-phase flow effects in the coolant, for acoustical effects in the steam line and for the dynamics of the recirculation loop and feedwater train. Point kinetics incorporate reactivity feedback for void fraction, for fuel temperature, for coolant temperature, and for boron concentration. Control systems and trip logic are simulated for the nuclear steam supply system. 4 refs., 3 figs

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.

Bowing-reactivity trends in EBR-II assuming zero-swelling ducts

International Nuclear Information System (INIS)

Meneghetti, D.

1994-01-01

Predicted trends of duct-bowing reactivities for the Experimental Breeder Reactor II (EBR-II) are correlated with predicted row-wise duct deflections assuming use of idealized zero-void-swelling subassembly ducts. These assume no irradiation induced swellings of ducts but include estimates of the effects of irradiation-creep relaxation of thermally induced bowing stresses. The results illustrate the manners in which at-power creeps may affect subsequent duct deflections at zero power and thereby the trends of the bowing component of a subsequent power reactivity decrement

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.

RELAP5-3D code validation of RBMK-1500 reactor reactivity measurement transients

International Nuclear Information System (INIS)

Kaliatka, Algirdas; Bubelis, Evaldas; Uspuras, Eugenijus

2003-01-01

This paper deals with the modeling of transients taking place during the measurements of the void and fast power reactivity coefficients performed at Ignalina NPP. The simulation of these transients was performed using RELAP5-3D code model of RBMK-1500 reactor. At the Ignalina NPP void and fast power reactivity coefficients are measured on a regular basis and, based on the total reactor power, reactivity, control and protection system control rods positions and the main circulation circuit parameter changes during the experiments, the actual values of these reactivity coefficients are determined. Following the simulation of the two above mentioned transients with RELAP5-3D code, a conclusion was made that the obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviors of the separate MCC thermal-hydraulic parameters as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. The calculated reactivity and the total reactor core power behavior in time are also in reasonable agreement with the measured plant data. Despite of the small differences, RELAP5-3D code predicts reactivity and the total reactor core power behavior during the transients in a reasonable manner. Reasonable agreement of the measured and the calculated total reactor power change in time demonstrates the correct modeling of the neutronic processes taking place in RBMK-1500 reactor core

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

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

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)

Void formation and its impact on Cu−Sn intermetallic compound formation

International Nuclear Information System (INIS)

Ross, Glenn; Vuorinen, Vesa; Paulasto-Kröckel, Mervi

2016-01-01

Void formation in the Cu−Sn system has been identified as a major reliability issue with small volume electronic interconnects. Voids form during the interdiffusion of electrochemically deposited Cu and Sn, with varying magnitude and density. Electroplating parameters include the electrolytic chemistry composition and the electroplating current density, all of which appear to effect the voiding characteristics of the Cu−Sn system. In addition, interfacial voiding affects the growth kinetics of the Cu_3Sn and Cu_6Sn_5 intermetallic compounds of the Cu−Sn system. The aim here is to present voiding data as a function of electroplating chemistry and current density over a duration (up to 72 h) of isothermal annealing at 423 K (150 °C). Voiding data includes the average interfacial void size and average void density. Voids sizes grew proportionally as a function of thermal annealing time, whereas the void density grew initially very quickly but tended to saturate at a fixed density. A morphological evolution analysis called the physicochemical approach is utilised to understand the processes that occur when a voided Cu/Cu_3Sn interface causes changes to the IMC phase growth. The method is used to simulate the intermetallic thickness growths' response to interfacial voiding. The Cu/Cu_3Sn interface acts as a Cu diffusion barrier disrupting the diffusion of Cu. This resulted in a reduction in the Cu_3Sn thickness and an accelerated growth rate of Cu_6Sn_5. - Highlights: • Average void size is proportional linearly to thermal annealing time. • Average void density grows initially very rapidly followed by saturation. • Voids located close to the Cu/Cu_3Sn interface affect IMC growth rates. • Voids act as a diffusion barrier inhibiting Cu diffusion towards Sn. • Voids located at the interface cause Cu_3Sn to be consumed by Cu_6Sn_5.

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

Evaluation of Melt Behavior with initial Melt Velocity under SFR Severe Accidents

Energy Technology Data Exchange (ETDEWEB)

Heo, Hyo; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of); Jerng, Dong Wook [Chung-Ang Univ, Seoul (Korea, Republic of)

2015-10-15

In the current Korean sodium-cooled fast reactor (SFR) program, early dispersion of the molten metallic fuel within a subchannel is suggested as one of the inherent safety strategies for the initiating phase of hypothetical core disruptive accident (HCDA). The safety strategy provides negative reactivity driven by the melt dispersal, so it could reduce the possibility of the recriticality event under a severe triple or more fault scenario for SFR. Since the behavior of the melt dispersion is unpredictable, it depends on the accident condition, particularly core region. While the voided coolant channel region is usually developed in the inner core, the unvoided coolant channel region is formed in the outer core. It is important to confirm the fuel dispersion with the core region, but there are not sufficient existing studies for them. From the existing studies, the coolant vapor pressure is considered as one of driving force to move the melt towards outside of the core. There is a complexity of the phenomena during intermixing of the melt with the coolant after the horizontal melt injections. It is too difficult to understand the several combined mechanisms related to the melt dispersion and the fragmentation. Thus, it could be worthwhile to study the horizontal melt injections at lower temperature as a preliminary study in order to identify the melt dispersion phenomena. For this reason, it is required to clarify whether the coolant vapor pressure is the driving force of the melt dispersion with the core region. The specific conditions to be well dispersed for the molten metallic fuel were discussed in the experiments with the simulant materials. The each melt behavior was compared to evaluate the melt dispersion under the coolant void condition and the boiling condition. As the results, the following results are remarked: 1. The upward melt dispersion did not occur for a given melt and coolant temperature in the nonboiling range. Over current range of conditions

Evaluation of Melt Behavior with initial Melt Velocity under SFR Severe Accidents

International Nuclear Information System (INIS)

Heo, Hyo; Bang, In Cheol; Jerng, Dong Wook

2015-01-01

In the current Korean sodium-cooled fast reactor (SFR) program, early dispersion of the molten metallic fuel within a subchannel is suggested as one of the inherent safety strategies for the initiating phase of hypothetical core disruptive accident (HCDA). The safety strategy provides negative reactivity driven by the melt dispersal, so it could reduce the possibility of the recriticality event under a severe triple or more fault scenario for SFR. Since the behavior of the melt dispersion is unpredictable, it depends on the accident condition, particularly core region. While the voided coolant channel region is usually developed in the inner core, the unvoided coolant channel region is formed in the outer core. It is important to confirm the fuel dispersion with the core region, but there are not sufficient existing studies for them. From the existing studies, the coolant vapor pressure is considered as one of driving force to move the melt towards outside of the core. There is a complexity of the phenomena during intermixing of the melt with the coolant after the horizontal melt injections. It is too difficult to understand the several combined mechanisms related to the melt dispersion and the fragmentation. Thus, it could be worthwhile to study the horizontal melt injections at lower temperature as a preliminary study in order to identify the melt dispersion phenomena. For this reason, it is required to clarify whether the coolant vapor pressure is the driving force of the melt dispersion with the core region. The specific conditions to be well dispersed for the molten metallic fuel were discussed in the experiments with the simulant materials. The each melt behavior was compared to evaluate the melt dispersion under the coolant void condition and the boiling condition. As the results, the following results are remarked: 1. The upward melt dispersion did not occur for a given melt and coolant temperature in the nonboiling range. Over current range of conditions

Measurements and calculations of reactivity for the IEA-R1 reactor

International Nuclear Information System (INIS)

Ferreira, P.S.B.; Maiorino, J.R.; Yamaguchi, M.

1988-01-01

This work shows a measurement of reactivity parameters, such as integral and diferential control rod worth, local void coefficient, and moderator temperature coefficient for the research reactor IEA-R1. The measured values were compared with those calculated through HAMMER-CITATION codes, having shown good agreement. (author) [pt

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

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

Learning from errors: analysis of medication order voiding in CPOE systems.

Science.gov (United States)

Kannampallil, Thomas G; Abraham, Joanna; Solotskaya, Anna; Philip, Sneha G; Lambert, Bruce L; Schiff, Gordon D; Wright, Adam; Galanter, William L

2017-07-01

Medication order voiding allows clinicians to indicate that an existing order was placed in error. We explored whether the order voiding function could be used to record and study medication ordering errors. We examined medication orders from an academic medical center for a 6-year period (2006-2011; n  = 5 804 150). We categorized orders based on status (void, not void) and clinician-provided reasons for voiding. We used multivariable logistic regression to investigate the association between order voiding and clinician, patient, and order characteristics. We conducted chart reviews on a random sample of voided orders ( n  = 198) to investigate the rate of medication ordering errors among voided orders, and the accuracy of clinician-provided reasons for voiding. We found that 0.49% of all orders were voided. Order voiding was associated with clinician type (physician, pharmacist, nurse, student, other) and order type (inpatient, prescription, home medications by history). An estimated 70 ± 10% of voided orders were due to medication ordering errors. Clinician-provided reasons for voiding were reasonably predictive of the actual cause of error for duplicate orders (72%), but not for other reasons. Medication safety initiatives require availability of error data to create repositories for learning and training. The voiding function is available in several electronic health record systems, so order voiding could provide a low-effort mechanism for self-reporting of medication ordering errors. Additional clinician training could help increase the quality of such reporting. © The Author 2017. Published by Oxford University Press on behalf of the American Medical Informatics Association. All rights reserved. For Permissions, please email: journals.permissions@oup.com

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.

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

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.)

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...

Void formation and its impact on Cu−Sn intermetallic compound formation

Energy Technology Data Exchange (ETDEWEB)

Ross, Glenn, E-mail: Glenn.Ross@aalto.fi; Vuorinen, Vesa; Paulasto-Kröckel, Mervi

2016-08-25

Void formation in the Cu−Sn system has been identified as a major reliability issue with small volume electronic interconnects. Voids form during the interdiffusion of electrochemically deposited Cu and Sn, with varying magnitude and density. Electroplating parameters include the electrolytic chemistry composition and the electroplating current density, all of which appear to effect the voiding characteristics of the Cu−Sn system. In addition, interfacial voiding affects the growth kinetics of the Cu{sub 3}Sn and Cu{sub 6}Sn{sub 5} intermetallic compounds of the Cu−Sn system. The aim here is to present voiding data as a function of electroplating chemistry and current density over a duration (up to 72 h) of isothermal annealing at 423 K (150 °C). Voiding data includes the average interfacial void size and average void density. Voids sizes grew proportionally as a function of thermal annealing time, whereas the void density grew initially very quickly but tended to saturate at a fixed density. A morphological evolution analysis called the physicochemical approach is utilised to understand the processes that occur when a voided Cu/Cu{sub 3}Sn interface causes changes to the IMC phase growth. The method is used to simulate the intermetallic thickness growths' response to interfacial voiding. The Cu/Cu{sub 3}Sn interface acts as a Cu diffusion barrier disrupting the diffusion of Cu. This resulted in a reduction in the Cu{sub 3}Sn thickness and an accelerated growth rate of Cu{sub 6}Sn{sub 5}. - Highlights: • Average void size is proportional linearly to thermal annealing time. • Average void density grows initially very rapidly followed by saturation. • Voids located close to the Cu/Cu{sub 3}Sn interface affect IMC growth rates. • Voids act as a diffusion barrier inhibiting Cu diffusion towards Sn. • Voids located at the interface cause Cu{sub 3}Sn to be consumed by Cu{sub 6}Sn{sub 5}.

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

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.

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

LOG-NORMAL DISTRIBUTION OF COSMIC VOIDS IN SIMULATIONS AND MOCKS

Energy Technology Data Exchange (ETDEWEB)

Russell, E.; Pycke, J.-R., E-mail: er111@nyu.edu, E-mail: jrp15@nyu.edu [Division of Science and Mathematics, New York University Abu Dhabi, P.O. Box 129188, Abu Dhabi (United Arab Emirates)

2017-01-20

Following up on previous studies, we complete here a full analysis of the void size distributions of the Cosmic Void Catalog based on three different simulation and mock catalogs: dark matter (DM), haloes, and galaxies. Based on this analysis, we attempt to answer two questions: Is a three-parameter log-normal distribution a good candidate to satisfy the void size distributions obtained from different types of environments? Is there a direct relation between the shape parameters of the void size distribution and the environmental effects? In an attempt to answer these questions, we find here that all void size distributions of these data samples satisfy the three-parameter log-normal distribution whether the environment is dominated by DM, haloes, or galaxies. In addition, the shape parameters of the three-parameter log-normal void size distribution seem highly affected by environment, particularly existing substructures. Therefore, we show two quantitative relations given by linear equations between the skewness and the maximum tree depth, and between the variance of the void size distribution and the maximum tree depth, directly from the simulated data. In addition to this, we find that the percentage of voids with nonzero central density in the data sets has a critical importance. If the number of voids with nonzero central density reaches ≥3.84% in a simulation/mock sample, then a second population is observed in the void size distributions. This second population emerges as a second peak in the log-normal void size distribution at larger radius.

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....

Investigation of circulating temperature fluctuations of the primary coolant in order to develop an enhanced MTC estimator for VVER-440 reactors

Energy Technology Data Exchange (ETDEWEB)

Kiss, Sandor; Lipcsei, Sandor [Hungarian Academy of Sciences, Budapest (Hungary). Centre for Energy Research - MTA

2017-09-15

Our aim was to develop a method based on noise diagnostics for the estimation of the moderator temperature coefficient of reactivity (MTC) for the Paks VVER-440 units in normal operation. The method requires determining core average neutron flux and temperature fluctuations. The circulation period of the primary coolant, transfer properties of the steam generators, as well as the source and the propagation of the temperature perturbations and the proportions of the perturbation components were investigated in order to estimate the feedback caused by the circulation of the primary coolant. Finally, after developing the new MTC estimator, determining its frequency range and optimal parameters, trends were produced based on an overall evaluation of measurements made with standard instrumentation during a one-year-long period at Paks NPP.

EBRPOCO - a program to calculate detailed contributions of power reactivity components of EBR-II

International Nuclear Information System (INIS)

Meneghetti, D.; Kucera, D.A.

1981-01-01

The EBRPOCO program has been developed to facilitate the calculations of the power coefficients of reactivity of EBR-II loadings. The program enables contributions of various components of the power coefficient to be delineated axially for every subassembly. The program computes the reactivity contributions of the power coefficients resulting from: density reduction of sodium coolant due to temperature; displacement of sodium coolant by thermal expansions of cladding, structural rods, subassembly cans, and lower and upper axial reflectors; density reductions of these steel components due to temperature; displacement of bond-sodium (if present) in gaps by differential thermal expansions of fuel and cladding; density reduction of bond-sodium (if present) in gaps due to temperature; free axial expansion of fuel if unrestricted by cladding or restricted axial expansion of fuel determined by axial expansion of cladding. Isotopic spatial contributions to the Doppler component my also be obtained. (orig.) [de

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.)

Real-time reactor coolant system pressure/temperature limit system

International Nuclear Information System (INIS)

Newton, D.G.; Schemmel, R.R.; Van Scooter, W.E. Jr.

1991-01-01

This patent describes an system, used in controlling the operating of a nuclear reactor coolant system, which automatically calculates and displays allowable reactor coolant system pressure/temperature limits within the nuclear reactor coolant system based upon real-time inputs. It comprises: means for producing signals representative of real-time operating parameters of the nuclear reactor cooling system; means for developing pressure and temperature limits relating the real-time operating parameters of the nuclear reactor coolant system, for normal and emergency operation thereof; means for processing the signals representative of real-time operating parameters of the nuclear reactor coolant system to perform calculations of a best estimate of signals, check manual inputs against permissible valves and test data acquisition hardware for validity and over/under range; and means for comparing the representative signals with limits for the real-time operating parameters to produce a signal for a real-time display of the pressure and temperature limits and of the real-time operating parameters use an operator in controlling the operation of the nuclear reactor coolant system

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

Influence of void ratio on thermal performance of heat pipe receiver

International Nuclear Information System (INIS)

Gui Xiaohong; Tang Dawei; Liang Shiqiang; Lin Bin; Yuan Xiugan

2012-01-01

Highlights: ► The temperature gradient increases significantly and the utility ratio of PCM decreases obviously as void ratio increases. ► Void cavity influences the process of phase change greatly. ► PCM melts slowly during sunlight periods and freezes slowly during eclipse periods as void ratio increases. ► The temperature gradient of PCM zone is very significant with the effect of void cavity. - Abstract: In this paper, influence of void ratio on thermal performance of heat pipe receiver under microgravity is numerically simulated. Accordingly, mathematical model is set up. Numerical method is offered. The temperature field of Phase Change Material (PCM) canister is shown. Numerical results are compared with numerical ones of National Aeronautics and Space Administration (NASA). Numerical results show that the temperature gradient increases significantly and the utility ratio of PCM decreases obviously as void ratio increases. Void cavity influences the process of phase change greatly. PCM melts slowly during sunlight periods and freezes slowly during eclipse periods as void ratio increases. The thermal resistance of void cavity is much bigger than that of PCM canister wall. Void cavity prevents the heat transfer between PCM zone and canister wall. The temperature gradient of PCM zone is very significant with the effect of void cavity. So the thermal stress of heat pipe receiver may increase, and the lifetime may decrease as void ratio increases.

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

In-pile TREAT Test L04: simulating a lead sub-assembly in an unprotected LMFBR loss-of-coolant accident

International Nuclear Information System (INIS)

Tylka, J.P.; Bauer, T.H.; Wright, A.E.; Davies, A.L.; Herbert, R.; Woods, W.J.

1983-01-01

Test L04 in the PFR/TREAT series is the first multi-pin, in-pile simulation of a LMFBR transient undercooling/overpower (TUCOP) accident using full length prototypic fuel irradiated in a fast reactor. L04 is a gridded 7-pin bundle test performed in the ANL Mk-III integral loop in a flowing sodium environment and uses prototypic, bottom plenum, UK reactor fuel, preirradiated in the PFR to an axial peak burn-up of 4.2 a/o. The objective of L04 was the study, by simulation, of coolant voiding and fuel motion during the initiating phase of a hypothetical TUCOP accident in a large LMFBR. Test L04 is intended to study the behavior of a centrally located, lead subassembly with the highest power-to-flow ratio

Burnup effects on criticality, breeding and safety of 1,000 MWe gas-cooled fast breeder reactor

International Nuclear Information System (INIS)

Yoshida, Hiroyuki; Ohta, Fumio

1977-12-01

Burnup characteristics of 1,000 MWe, PuO 2 - UO 2 fuelled helium-cooled fast breeder reactor have been studied concerning criticality, breeding and safety. A 26-energy group cross-section set produced from ENDF/B-3 was used. Criticality and breeding were studied with two-dimensional burnup code APOLLO and 4-energy group cross-section set generated by collapsing the mentioned cross-section set. Safety aspects such as Doppler reactivity effect, coolant-depressurisation and steam-ingression reactivity effect were studied with multi-dimensional diffusion theory code CITATION and perturbation theory code PERKY, as well as the 26-energy group cross-section set. The following were revealed: (1) The reactivity swing over a year's irradiation is merely 1.5% ΔK/K. This small swing may permit relatively long fuel dwelling in GCFR and , thus, the frequency of outages for refuelling can be minimised. (2) The surplus fissile plutonium over a year's irradiation is about 360 Kg, and the system doubling time is about 9 years. The GCFR studied has excellent breeding, compared with those in PuO 2 -UO 2 fuelled LMFBR and other GCFRs. (3) The coolant-depressurisation reactivity effect becomes more positive with burnup. This is not so serious as the sodium-void reactivity effect of LMFBR. (4) In the start-up core, the steam-ingression reactivity effect due to steam ingression to the core and blanket from the secondary coolant system becomes positive at certain steam density (0.02gr/cc) and this positive effect increases with steam density. With advance of burnup, however, the effect becomes negative, this increasing with steam density. After all, the steam ingression is no hazard in operation of GCFR since the reactivity effect is negative in the equilibrium state. (auth.)

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.)

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)

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....

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

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

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....

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

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

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

Closure behavior of spherical void in slab during hot rolling process

Science.gov (United States)

Cheng, Rong; Zhang, Jiongming; Wang, Bo

2018-04-01

The mechanical properties of steels are heavily deteriorated by voids. The influence of voids on the product quality should be eliminated through rolling processes. The study on the void closure during hot rolling processes is necessary. In present work, the closure behavior of voids at the center of a slab at 800 °C during hot rolling processes has been simulated with a 3D finite element model. The shape of the void and the plastic strain distribution of the slab are obtained by this model. The void decreases along the slab thickness direction and spreads along the rolling direction but hardly changes along the strip width direction. The relationship between closure behavior of voids and the plastic strain at the center of the slab is analyzed. The effects of rolling reduction, slab thickness and roller diameter on the closure behavior of voids are discussed. The larger reduction, thinner slab and larger roller diameter all improve the closure of voids during hot rolling processes. Experimental results of the closure behavior of a void in the slab during hot rolling process mostly agree with the simulation results..

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

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.

Development of neutronics and thermal hydraulics coupled code – SAC-RIT for plate type fuel and its application to reactivity initiated transient analysis

International Nuclear Information System (INIS)

Singh, Tej; Kumar, Jainendra; Mazumdar, Tanay; Raina, V.K.

2013-01-01

Highlights: • A point reactor kinetics code coupled with thermal hydraulics of plate type fuel is developed. • This code is applicable for two phase flow of coolant. • Safety analysis of IAEA benchmark reactor core is carried out. • Results agree well with the results available in literature. - Abstract: A point reactor kinetics code SAC-RIT, acronym of Safety Analysis Code for Reactivity Initiated Transient, coupled with thermal hydraulics of two phase coolant flow for plate type fuel, is developed to calculate reactivity initiated transient analysis of nuclear research and test reactors. Point kinetics equations are solved by fourth order Runge Kutta method. Reactivity feedback effect is included into the code. Solution of kinetics equations gives neutronic power and it is then fed into a thermal hydraulic code where mass, momentum and thermal energy conservation equations are solved by explicit finite difference method to find out fuel, clad and coolant temperatures during transients. In this code, all possible flow regimes including laminar flow, transient flow and turbulent flow have been covered. Various heat transfer coefficients suitable for single liquid, sub-cooled boiling, saturation boiling, film boiling and single vapor phases are incorporated in the thermal hydraulic code

Zero waste machine coolant management strategy at Los Alamos National Laboratory

International Nuclear Information System (INIS)

Carlson, B.; Algarra, F.; Wilburn, D.

1998-01-01

Machine coolants are used in machining equipment including lathes, grinders, saws and drills. The purpose of coolants is to wash away machinery debris in the form of metal fines, lubricate, and disperse heat between the part and the machine tool. An effective coolant prolongs tool life and protects against part rejection, commonly due to scoring or scorching. Traditionally, coolants have a very short effective life in the machine, often times being disposed of as frequently as once per week. The cause of coolant degradation is primarily due to the effects of bacteria, which thrive in the organic rich coolant environment. Bacteria in this environment reproduce at a logarithmic rate, destroying the coolant desirable aspects and causing potential worker health risks associated with the use of biocides to control the bacteria. The strategy described in this paper has effectively controlled bacterial activity without the use of biocides, avoided disposal of a hazardous waste, and has extended coolant life indefinitely. The Machine Coolant Management Strategy employed a combination of filtration, heavy lubricating oil removal, and aeration, which maintained the coolant peak performance without the use of biocides. In FY96, the Laboratory generated and disposed of 19,880 kg of coolants from 9 separate sites at a cost of $145K. The single largest generator was the main machine shop producing an average 14,000 kg annually. However, in FY97, the waste generation for the main machine shop dropped to 4,000 kg after the implementation of the zero waste strategy. It is expected that this value will be further reduced in FY98

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

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.)

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

Is abdominal wall contraction important for normal voiding in the female rat?

Directory of Open Access Journals (Sweden)

Boone Timothy B

2007-03-01

Full Text Available Abstract Background Normal voiding behavior in urethane-anesthetized rats includes contraction of the abdominal wall striated muscle, similar to the visceromotor response (VMR to noxious bladder distension. Normal rat voiding requires pulsatile release of urine from a pressurized bladder. The abdominal wall contraction accompanying urine flow may provide a necessary pressure increment for normal efficient pulsatile voiding. This study aimed to evaluate the occurrence and necessity of the voiding-associated abdominal wall activity in urethane-anesthetized female rats Methods A free-voiding model was designed to allow assessment of abdominal wall activity during voiding resulting from physiologic bladder filling, in the absence of bladder or urethral instrumentation. Physiologic diuresis was promoted by rapid intravascular hydration. Intercontraction interval (ICI, voided volumes and EMG activity of the rectus abdominis were quantified. The contribution of abdominal wall contraction to voiding was eliminated in a second group of rats by injecting botulinum-A (BTX, 5 U into each rectus abdominis to induce local paralysis. Uroflow parameters were compared between intact free-voiding and BTX-prepared animals. Results Abdominal wall response is present in free voiding. BTX preparation eliminated the voiding-associated EMG activity. Average per-void volume decreased from 1.8 ml to 1.1 ml (p Conclusion The voiding-associated abdominal wall response is a necessary component of normal voiding in urethane anesthetized female rats. As the proximal urethra may be the origin of the afferent signaling which results in the abdominal wall response, the importance of the bladder pressure increment due to this response may be in maintaining a normal duration intermittent pulsatile high frequency oscillatory (IPHFO/flow phase and thus efficient voiding. We propose the term Voiding-associated Abdominal Response (VAR for the physiologic voiding-associated EMG

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

Molecular dynamics modeling and simulation of void growth in two dimensions

Science.gov (United States)

Chang, H.-J.; Segurado, J.; Rodríguez de la Fuente, O.; Pabón, B. M.; LLorca, J.

2013-10-01

The mechanisms of growth of a circular void by plastic deformation were studied by means of molecular dynamics in two dimensions (2D). While previous molecular dynamics (MD) simulations in three dimensions (3D) have been limited to small voids (up to ≈10 nm in radius), this strategy allows us to study the behavior of voids of up to 100 nm in radius. MD simulations showed that plastic deformation was triggered by the nucleation of dislocations at the atomic steps of the void surface in the whole range of void sizes studied. The yield stress, defined as stress necessary to nucleate stable dislocations, decreased with temperature, but the void growth rate was not very sensitive to this parameter. Simulations under uniaxial tension, uniaxial deformation and biaxial deformation showed that the void growth rate increased very rapidly with multiaxiality but it did not depend on the initial void radius. These results were compared with previous 3D MD and 2D dislocation dynamics simulations to establish a map of mechanisms and size effects for plastic void growth in crystalline solids.

Molecular dynamics modeling and simulation of void growth in two dimensions

International Nuclear Information System (INIS)

Chang, H-J; Segurado, J; LLorca, J; Rodríguez de la Fuente, O; Pabón, B M

2013-01-01

The mechanisms of growth of a circular void by plastic deformation were studied by means of molecular dynamics in two dimensions (2D). While previous molecular dynamics (MD) simulations in three dimensions (3D) have been limited to small voids (up to ≈10 nm in radius), this strategy allows us to study the behavior of voids of up to 100 nm in radius. MD simulations showed that plastic deformation was triggered by the nucleation of dislocations at the atomic steps of the void surface in the whole range of void sizes studied. The yield stress, defined as stress necessary to nucleate stable dislocations, decreased with temperature, but the void growth rate was not very sensitive to this parameter. Simulations under uniaxial tension, uniaxial deformation and biaxial deformation showed that the void growth rate increased very rapidly with multiaxiality but it did not depend on the initial void radius. These results were compared with previous 3D MD and 2D dislocation dynamics simulations to establish a map of mechanisms and size effects for plastic void growth in crystalline solids. (paper)

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

Structure-dependent behavior of stress-induced voiding in Cu interconnects

International Nuclear Information System (INIS)

Wu Zhenyu; Yang Yintang; Chai Changchun; Li Yuejin; Wang Jiayou; Li Bin; Liu Jing

2010-01-01

Stress modeling and cross-section failure analysis by focused-ion-beam have been used to investigate stress-induced voiding phenomena in Cu interconnects. The voiding mechanism and the effect of the interconnect structure on the stress migration have been studied. The results show that the most concentrated tensile stress appears and voids form at corners of vias on top surfaces of Cu M1 lines. A simple model of stress induced voiding in which vacancies arise due to the increase of the chemical potential under tensile stress and diffuse under the force of stress gradient along the main diffusing path indicates that stress gradient rather than stress itself determines the voiding rate. Cu interconnects with larger vias show less resistance to stress-induced voiding due to larger stress gradient at corners of vias.

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.

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

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.

Large-Scale Reactive Atomistic Simulation of Shock-induced Initiation Processes in Energetic Materials

Science.gov (United States)

Thompson, Aidan

2013-06-01

Initiation in energetic materials is fundamentally dependent on the interaction between a host of complex chemical and mechanical processes, occurring on scales ranging from intramolecular vibrations through molecular crystal plasticity up to hydrodynamic phenomena at the mesoscale. A variety of methods (e.g. quantum electronic structure methods (QM), non-reactive classical molecular dynamics (MD), mesoscopic continuum mechanics) exist to study processes occurring on each of these scales in isolation, but cannot describe how these processes interact with each other. In contrast, the ReaxFF reactive force field, implemented in the LAMMPS parallel MD code, allows us to routinely perform multimillion-atom reactive MD simulations of shock-induced initiation in a variety of energetic materials. This is done either by explicitly driving a shock-wave through the structure (NEMD) or by imposing thermodynamic constraints on the collective dynamics of the simulation cell e.g. using the Multiscale Shock Technique (MSST). These MD simulations allow us to directly observe how energy is transferred from the shockwave into other processes, including intramolecular vibrational modes, plastic deformation of the crystal, and hydrodynamic jetting at interfaces. These processes in turn cause thermal excitation of chemical bonds leading to initial chemical reactions, and ultimately to exothermic formation of product species. Results will be presented on the application of this approach to several important energetic materials, including pentaerythritol tetranitrate (PETN) and ammonium nitrate/fuel oil (ANFO). In both cases, we validate the ReaxFF parameterizations against QM and experimental data. For PETN, we observe initiation occurring via different chemical pathways, depending on the shock direction. For PETN containing spherical voids, we observe enhanced sensitivity due to jetting, void collapse, and hotspot formation, with sensitivity increasing with void size. For ANFO, we

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

Sodium-cooled fast reactor core designs for transmutation of MHR spent fuel

International Nuclear Information System (INIS)

Hong, S. G.; Kim, Y. H.; Venneri, F.

2010-01-01

In this paper, the core design analyses of sodium cooled fast reactors (SFR) are performed for the effective transmutation of the DB (Deep Burn)-MHR (Modular Helium Reactor). In this concept, the spent fuels of DB-MHR are transmuted in SFRs with a closed fuel cycle after TRUs from LWR are first incinerated in a DB-MHR. We introduced two different type SFR core designs for this purpose, and evaluated their core performance parameters including the safety-related parameters. In particular, the cores are designed to have lower transmutation rate relatively to our previous work so as to make the fuel characteristics more feasible. The first type cores which consist of two enrichment regions are typical homogeneous annular cores and they rate 900 MWt power. On the other hand, the second type cores which consist of a central non-fuel region and a single enrichment fuel region rate relatively higher power of 1500 MWt. For these cores, the moderator rods (YH 1.8 ) are used to achieve less positive sodium void worth and the more negative Doppler coefficient because the loading of DB-MHR spent fuel leads to the degradation of these safety parameters. The analysis results show that these cores have low sodium void worth and negative reactivity coefficients except for the one related with the coolant expansion but the coolant expansion reactivity coefficient is within the typical range of the typical SFR cores. (authors)

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.

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.

Mechanism of Void Prediction in Flip Chip Packages with Molded Underfill

Science.gov (United States)

Wu, Kuo-Tsai; Hwang, Sheng-Jye; Lee, Huei-Huang

2017-08-01

Voids have always been present using the molded underfill (MUF) package process, which is a problem that needs further investigation. In this study, the process was studied using the Moldex3D numerical analysis software. The effects of gas (air vent effect) on the overall melt front were also considered. In this isothermal process containing two fluids, the gas and melt colloid interact in the mold cavity. Simulation enabled an appropriate understanding of the actual situation to be gained, and, through analysis, the void region and exact location of voids were predicted. First, the global flow end area was observed to predict the void movement trend, and then the local flow ends were observed to predict the location and size of voids. In the MUF 518 case study, simulations predicted the void region as well as the location and size of the voids. The void phenomenon in a flip chip ball grid array underfill is discussed as part of the study.

Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

Energy Technology Data Exchange (ETDEWEB)

Armstrong, J.; Hamilton, H.; Hyland, B. [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada)

2013-07-01

A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

International Nuclear Information System (INIS)

Armstrong, J.; Hamilton, H.; Hyland, B.

2013-01-01

A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

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

Thermal hydraulic And RSG-Gas Core Reactivity Characteristics Due To Cold Water Insertion Accident

International Nuclear Information System (INIS)

Hastuti, Endiah Puji; Suparlina, Lily; Tukiran

2000-01-01

Under normal operating condition,the primary coolant is circulated by 2 out of the 3 primary coolant pumps. Unnecessary operation of the reserve pump would result in a temperatur decrease of the primary coolant by less than 5 o C. the corresponding increase of reactivity amounts to Δρ ≤0,1 %. The analysis was done using silicide core configuration data with 3.55 gU /cm 3 fuel loading. The calculation model was done with and without automatic control rod. The calculation results for the worst case condition, shows that reactor reached the maximum power 28.52 MW at 81.1 seconds, after the accident occurred. The maximal fuel element, cladding and outlet coolant temperatures are 148.3 o C,142.1 o C, and 75.7 o C, respectively. Safety margins for DNBR and flow instability reached 1.25 and 4.20, respectively. Comparing to the RSG-GAS safety margin at transient condition reguirement >1.48, RSG-GAS has enough safety margin if the power trip executed at 114% of 25 MW

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

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

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

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...

Void fraction fluctuations in two-phase gas-liquid flow

International Nuclear Information System (INIS)

Ulbrich, R.

1987-01-01

Designs of the apparatus in which two-phase gas-liquid flow occurs are usually based on the mean value of parameters such as pressure drop and void fraction. The flow of two-phase mixtures generally presents a very complicated flow structure, both in terms of the unsteady formation on the interfacial area and in terms of the fluctuations of the velocity, pressure and other variables within the flow. When the gas void fraction is near 0 or 1 / bubble or dispersed flow regimes / then oscillations of void fraction are very small. The intermittent flow such as plug and slug/ froth is characterized by alternately flow portions of liquid and gas. It influences the change of void fractions in time. The results of experimental research of gas void fraction fluctuations in two-phase adiabatic gas-liquid flow in a vertical pipe are presented

Coupled 3D-neutronics / thermal-hydraulics analysis of an unprotected loss-of-flow accident for a 3600 MWth SFR core

International Nuclear Information System (INIS)

Sun, K.; Chenu, A.; Mikityuk, K.; Krepel, J.; Chawla, R.

2012-01-01

The core behaviour of a large (3600 MWth) sodium-cooled fast reactor (SFR) is investigated in this paper with the use of a coupled TRACE/PARCS model. The SFR neutron spectrum is characterized by several performance advantages, but also leads to one dominating neutronics drawback - a positive sodium void reactivity. This implies a positive reactivity effect when sodium coolant is removed from the core. In order to evaluate such feedback in terms of the dynamics, a representative unprotected loss-of-flow (ULOF) transient, i.e. flow run-down without SCRAM in which sodium boiling occurs, is analyzed. Although analysis of a single transient cannot allow general conclusions to be drawn, it does allow better understanding of the underlying physics and can lead to proposals for improving the core response during such an accident. The starting point of this study is the reference core design considered in the framework of the Collaborative Project on the European Sodium Fast Reactor (CP-ESFR). To reduce the void effect, the core has been modified by introducing an upper sodium plenum (along with a boron layer) and by reducing the core height-to-diameter ratio. For the ULOF considered, a sharp increase in core power results in melting of the fuel in the case of the reference core. In the modified core, a large dryout leads to melting of the clad. It seems that, for the hypothetical event considered, fuel failure cannot be avoided with just improvement of the neutronics design; therefore, thermal-hydraulics optimization has been considered. An innovative assembly design is proposed to prevent sodium vapour blocking the fuel channel. This results in preventing a downward propagation of the sodium boiling to the core center, thus limiting it to the upper region. Such a void map introduces a negative coolant density reactivity feedback, which dominates the total reactivity change. As a result, the power level and the fuel temperature are effectively reduced, and a large dryout

Coupled 3D-neutronics / thermal-hydraulics analysis of an unprotected loss-of-flow accident for a 3600 MWth SFR core

Energy Technology Data Exchange (ETDEWEB)

Sun, K. [Paul Scherrer Institut PSI, 5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne EPFL, 1015 Lausanne (Switzerland); Chenu, A. [Ecole Polytechnique Federale de Lausanne EPFL, 1015 Lausanne (Switzerland); Mikityuk, K.; Krepel, J. [Paul Scherrer Institut PSI, 5232 Villigen PSI (Switzerland); Chawla, R. [Paul Scherrer Institut PSI, 5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne EPFL, 1015 Lausanne (Switzerland)

2012-07-01

The core behaviour of a large (3600 MWth) sodium-cooled fast reactor (SFR) is investigated in this paper with the use of a coupled TRACE/PARCS model. The SFR neutron spectrum is characterized by several performance advantages, but also leads to one dominating neutronics drawback - a positive sodium void reactivity. This implies a positive reactivity effect when sodium coolant is removed from the core. In order to evaluate such feedback in terms of the dynamics, a representative unprotected loss-of-flow (ULOF) transient, i.e. flow run-down without SCRAM in which sodium boiling occurs, is analyzed. Although analysis of a single transient cannot allow general conclusions to be drawn, it does allow better understanding of the underlying physics and can lead to proposals for improving the core response during such an accident. The starting point of this study is the reference core design considered in the framework of the Collaborative Project on the European Sodium Fast Reactor (CP-ESFR). To reduce the void effect, the core has been modified by introducing an upper sodium plenum (along with a boron layer) and by reducing the core height-to-diameter ratio. For the ULOF considered, a sharp increase in core power results in melting of the fuel in the case of the reference core. In the modified core, a large dryout leads to melting of the clad. It seems that, for the hypothetical event considered, fuel failure cannot be avoided with just improvement of the neutronics design; therefore, thermal-hydraulics optimization has been considered. An innovative assembly design is proposed to prevent sodium vapour blocking the fuel channel. This results in preventing a downward propagation of the sodium boiling to the core center, thus limiting it to the upper region. Such a void map introduces a negative coolant density reactivity feedback, which dominates the total reactivity change. As a result, the power level and the fuel temperature are effectively reduced, and a large dryout

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.

Atomistic simulations of void migration under thermal gradient in UO2

International Nuclear Information System (INIS)

Desai, Tapan G.; Millett, Paul; Tonks, Michael; Wolf, Dieter

2010-01-01

It is well known that within a few hours after startup of a nuclear reactor, the temperature gradient within a fuel element causes migration of voids/bubbles radially inwards to form a central hole. To understand the atomic processes that control this migration of voids, we performed molecular dynamics (MD) simulations on single crystal UO 2 with voids of diameter 2.2 nm. An external temperature gradient was applied across the simulation cell. At the end of the simulation run, it was observed that the voids had moved towards the hot end of the simulation cell. The void migration velocity obtained from the simulations was compared with the available phenomenological equations for void migration due to different transport mechanisms. Surface diffusion of the slowest moving specie, i.e. uranium, was found to be the dominant mechanism for void migration. The contribution from lattice diffusion and the thermal stress gradient to the void migration was analyzed and found to be negligible. By extrapolation, a crossover from the surface-diffusion-controlled mechanism to the lattice-diffusion-controlled mechanism was found to occur for voids with sizes in the μm range.

Close correlation of herpes zoster-induced voiding dysfunction with severity of zoster-related pain: A single faculty retrospective study.

Science.gov (United States)

Fujii, Mizue; Takahashi, Ichiro; Honma, Masaru; Ishida-Yamamoto, Akemi

2015-11-01

Herpes zoster (HZ), a common vesiculo-erythematous skin disease associated with reactivation of varicella zoster virus in the cranial nerve, dorsal root, and autonomic ganglia, is accompanied by several related symptoms represented by postherpetic neuralgia. Among them, involvement of vesicorectal dysfunction is relatively rare. The vesicorectal symptom can usually be recovered in transient course, but is quite important in terms of impaired quality of life. Male individuals affected with HZ and skin lesions on sacral dermatome have been reported as independent risk factors of zoster-related voiding dysfunction. In this study, urinary symptoms were focused upon and six patients with zoster-related voiding dysfunction at a single faculty of dermatology in Japan from 2009 to 2014 were retrospectively analyzed. All patients showed HZ lesions on the sacral area and the urinary symptom recovered in approximately 2 months (14 days to 7 months). The term of treatment for zoster-associated urinary dysfunction was positively correlated with that for zoster-related pain without significance (r = 0.661, P = 0.153). Average treatment term for pain relief of sacral HZ accompanied by voiding dysfunction (91.3 ± 76.44 days) was significantly longer than that of sacral HZ without urinary symptom (18.9 ± 20.42 days) (P = 0.032). These results suggested that zoster-related voiding dysfunction would mainly be involved in sacral HZ and closely associated with severity of zoster-related pain. Dermatologists should be aware that severe zoster-related pain accompanied by sacral HZ, which is related to prolonged treatment of pain relief, can be a predictive factor of voiding dysfunction. © 2015 Japanese Dermatological Association.

Transient analysis and thermal hydraulic margins of GHARR-1 using the PARET/NAL code

International Nuclear Information System (INIS)

Adoo, N.A.

2009-06-01

The PARET code has been adapted by the IAEA for testing transient behaviour in research reactors. The PARET code provides a coupled thermal hydrodynamic and point kinetics capability with a continuous reactivity feedback and an optional voiding model that estimates the voiding produced by the subcooled boiling. The present version of the PARET/ANL 73 code provides a convenient means of assessing the various models and correlations proposed for the use in the analysis of research reactor behaviour. The Monte Carlo N-Particle code (MCNP) has been used to obtain power peaking profile for a two channel PARET/ANL model. A PARET model with the corresponding neutronics and thermal hydraulic characteristics for the miniature neutron source reactor (MNSR) has been used to simulate reactivity accidents for the Ghana Research Reactor - 1(GHARR-1) under the MNSR operation conditions of natural circulation, normal operation and reactivity insertion accidents. The simulation results via the insertion of large reactivity demonstrated the high inherent safety features of the MNSR for which the high negative reactivity feedback of moderator temperature limits power excursion and avoids consequently the escalation of clad temperature to the level of onset of sub-cooled void formation. The hot channel peaking factors for both radial and axial were found to be 1.17 and 1.44 respectively. Thermal hydraulic performance characteristics were investigated and the safety margins determined. The peak clad and coolant temperatures ranged from 59.18 0 C to 106.75 0 C and 42.95 0 C to 178.44 0 C respectively at which nucleate boiling will occur within the flow channels of the core. (au)

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)

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

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

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

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)

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.

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.

Analysis of an Advanced Test Reactor Small-Break Loss-of-Coolant Accident with an Engineered Safety Feature to Automatically Trip the Primary Coolant Pumps

International Nuclear Information System (INIS)

Polkinghorne, Steven T.; Davis, Cliff B.; McCracken, Richard T.

2000-01-01

A new engineered safety feature that automatically trips the primary coolant pumps following a low-pressure reactor scram was recently installed in the Advanced Test Reactor (ATR). The purpose of this engineered safety feature is to prevent the ATR's surge tank, which contains compressed air, from emptying during a small-break loss-of-coolant accident (SBLOCA). If the surge tank were to empty, the air introduced into the primary coolant loop could potentially cause the performance of the primary and/or emergency coolant pumps to degrade, thereby reducing core thermal margins. Safety analysis performed with the RELAP5 thermal-hydraulic code and the SINDA thermal analyzer shows that adequate thermal margins are maintained during an SBLOCA with the new engineered safety feature installed. The analysis also shows that the surge tank will not empty during an SBLOCA even if one of the primary coolant pumps fails to trip

Reactor core of light water-cooled reactor

International Nuclear Information System (INIS)

Miwa, Jun-ichi; Aoyama, Motoo; Mochida, Takaaki.

1996-01-01

In a reactor core of a light water cooled reactor, the center of the fuel rods or moderating rods situated at the outermost circumference among control rods or moderating rods are connected to divide a lattice region into an inner fuel region and an outer moderator region. In this case, the area ratio of the moderating region to the fuel region is determined to greater than 0.81 for every cross section of the fuel region. The moderating region at the outer side is increased relative to the fuel rod region at the inner side while keeping the lattice pitch of the fuel assembly constant, thereby suppressing the increase of an absolute value of a void reactivity coefficient which tends to be caused when using MOX fuels as a fuel material, by utilizing neutron moderation due to a large quantity of coolants at the outer side of the fuel region. The void reactivity coefficient can be made substantially equal with that of uranium fuel assembly without greatly reducing a plutonium loading amount or without greatly increasing linear power density. (N.H.)

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

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

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)

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

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.

Performance investigation of an automotive car radiator operated with nanofluid-based coolants (nanofluid as a coolant in a radiator)

International Nuclear Information System (INIS)

Leong, K.Y.; Saidur, R.; Kazi, S.N.; Mamun, A.H.

2010-01-01

Water and ethylene glycol as conventional coolants have been widely used in an automotive car radiator for many years. These heat transfer fluids offer low thermal conductivity. With the advancement of nanotechnology, the new generation of heat transfer fluids called, 'nanofluids' have been developed and researchers found that these fluids offer higher thermal conductivity compared to that of conventional coolants. This study focused on the application of ethylene glycol based copper nanofluids in an automotive cooling system. Relevant input data, nanofluid properties and empirical correlations were obtained from literatures to investigate the heat transfer enhancement of an automotive car radiator operated with nanofluid-based coolants. It was observed that, overall heat transfer coefficient and heat transfer rate in engine cooling system increased with the usage of nanofluids (with ethylene glycol the basefluid) compared to ethylene glycol (i.e. basefluid) alone. It is observed that, about 3.8% of heat transfer enhancement could be achieved with the addition of 2% copper particles in a basefluid at the Reynolds number of 6000 and 5000 for air and coolant respectively. In addition, the reduction of air frontal area was estimated.

On the formation of voids in internal tin Nb$_{3}$Sn superconductors

CERN Document Server

Scheuerlein, C; Haibel, A

2007-01-01

In this article we describe three void growth mechanisms in Nb$_{3}$Sn strands of the internal tin design on the basis of combined synchrotron micro-tomography and x-ray diffraction measurements during in-situ heating cycles. Initially void growth is driven by a reduction of void surface area by void agglomeration. The main void volume increase is caused by density changes during the formation of Cu3Sn in the strand. Subsequent transformation of Cu-Sn intermetallics into the lower density a-bronze reduces the void volume again. Long lasting temperature ramps and isothermal holding steps can neither reduce the void volume nor improve the chemical strand homogeneity prior to the superconducting A15 phase nucleation and growth.

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)

Analysis of actual status of works on technology of heavy liquid metal coolants

International Nuclear Information System (INIS)

Martynov, P.N.; Askhadullin, R.Sh.; Orlov, Yu.I.; Storozhenko, A.N.

2014-01-01

Principle duties in heavy liquid metal coolant technology (HLMC) are provision of the purity of coolant and surfaces of circulation loop for maintenance of design thermohydraulic characteristics, prevention of structural materials corrosion and erosion during long service life and present-day safety precautions on different stages of reactor facility operation. For this reason, current HLMC (Pb-Bi, Pb) technology must include coolant pre-operation and charging; monitoring and regulating of coolant oxygen potential; hydrogen purification of coolant and surfaces of circulation loop from lead oxides-based slags; coolant filtration; reactor cover gas purification from coolant aerosols. The current topical problem is personnel training on the questions of HLMC technology [ru

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

Interfacial area, velocity and void fraction in two-phase slug flow

International Nuclear Information System (INIS)

Kojasoy, G.; Riznic, J.R.

1997-01-01

The internal flow structure of air-water plug/slug flow in a 50.3 mm dia transparent pipeline has been experimentally investigated by using a four-sensor resistivity probe. Liquid and gas volumetric superficial velocities ranged from 0.55 to 2.20 m/s and 0.27 to 2.20 m/s, respectively, and area-averaged void fractions ranged from about 10 to 70%. The local distributions of void fractions, interfacial area concentration and interface velocity were measured. Contributions from small spherical bubbles and large elongated slug bubbles toward the total void fraction and interfacial area concentration were differentiated. It was observed that the small bubble void contribution to the overall void fraction was small indicating that the large slug bubble void fraction was a dominant factor in determining the total void fraction. However, the small bubble interfacial area contribution was significant in the lower and upper portions of the pipe cross sections

On hydrogen-induced plastic flow localization during void growth and coalescence

Energy Technology Data Exchange (ETDEWEB)

Ahn, D.C.; Sofronis, P. [Department of Mechanical Science and Engineering, University of Illinois at Urbana-Champaign, 1206 West Green Street, Urbana, IL 61801 (United States); Dodds, R.H. Jr. [Department of Civil and Environmental Engineering, University of Illinois at Urbana-Champaign, 205 North Mathews Avenue, Urbana, IL 61801 (United States)

2007-11-15

Hydrogen-enhanced localized plasticity (HELP) is recognized as a viable mechanism of hydrogen embrittlement. A possible way by which the HELP mechanism can bring about macroscopic material failure is through hydrogen-induced accelerated void growth and coalescence. Assuming a periodic array of spherical voids loaded axisymmetrically, we investigate the hydrogen effect on the occurrence of plastic flow localization upon void growth and its dependence on macroscopic stress triaxiality. Under a macroscopic stress triaxiality equal to 1 and prior to void coalescence, the finite element calculation results obtained with material data relevant to A533B steel indicate that a hydrogen-induced localized shear band forms at an angle of about 45 {sup circle} from the axis of symmetry. At triaxiality equal to 3, void coalescence takes place by accelerated hydrogen-induced localization of plasticity mainly in the ligament between the voids. Lastly, we discuss the numerical results within the context of experimental observations on void growth and coalescence in the presence of hydrogen. (author)

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.)

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.)

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

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

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)

Method of simulating spherical voids for use as a radiographic standard

International Nuclear Information System (INIS)

Foster, B.E.

1977-01-01

A method of simulating small spherical voids in metal is provided. The method entails drilling or etching a hemispherical depression of the desired diameter in each of two sections of metal, the sections being flat plates or different diameter cylinders. A carbon bead is placed in one of the hemispherical voids and is used as a guide to align the second hemispherical void with that in the other plate. The plates are then bonded together with epoxy, tape or similar material and the two aligned hemispheres form a sphere within the material; thus a void of a known size has been created. This type of void can be used to simulate a pore in the development of radiographic techniques of actual voids (porosity) in welds and serve as a radiographic standard

Size-effects at a crack-tip interacting with a number of voids

DEFF Research Database (Denmark)

Tvergaard, Viggo; Niordson, Christian Frithiof

2008-01-01

A strain gradient plasticity theory is used to analyse the growth of discretely represented voids in front of a blunting crack tip, in order to study the influence of size effects on two competing mechanisms of crack growth. For a very small void volume fraction the crack tip tends to interact...... of the characteristic material length relative to the initial void radius. For a case showing the multiple void mechanism, it is found that the effect of the material length can change the behaviour towards the void by void mechanism. A material model with three characteristic length scales is compared with a one...

Systems design of direct-cycle supercritical-water-cooled fast reactors

International Nuclear Information System (INIS)

Oka, Yoshiaki; Koshizuka, Seiichi; Jevremovic, Tatjana; Okano, Yashushi

1995-01-01

The system design of a direct-cycle supercritical-water-cooled fast reactor is presented. The supercritical water does not exhibit a change of phase. the recirculation system, steam separator, and dryer of a boiling water reactor (BWR) are unnecessary. Roughly speaking, the reactor pressure vessel and control rods are similar to those of a pressurized water reactor, the containment and emergency core cooling system are similar to a BWR, and the balance of plant is similar to a supercritical-pressure fossil-fired power plant (FPP). the electric power of the fast converter is 1,508 MW(electric). The number of coolant loops is only two because of the high coolant enthalpy. Containment volume is much reduced. The thermal efficiency is improved 24% over a BWR. The coolant void reactivity is negative by placing thin zirconium-hydride layers between seeds and blankets. The power costs would be much reduced compared with those of a light water reactor (LWR) and a liquid-metal fast breeder reactor. The concept is based on the huge amount of experience with the water coolant technology of LWRs and FPPs. The oxidation of stainless steel cladding is avoided by adopting a much lower coolant temperature than that of the FPP

BWR fuel assembly bottom nozzle with one-way coolant flow valve

International Nuclear Information System (INIS)

Taleyarkhan, R.P.

1987-01-01

In a nuclear reactor having a flow of coolant/moderator fluid therein, at least one fuel assembly installed in the fluid flow, the fuel assembly is described comprising in combination: a bundle of elongated fuel rods disposed in side-by-side relationship so as to form an array of spaced fuel rods; an outer tubular flow channel surrounding the fuel rods so as to direct the flow of coolant/moderator fluid along the fuel rods; bottom and top nozzles mounted at opposite ends of the flow channel and having an inlet and outlet respectively for allowing entry and exit of the flow of coolant/moderator fluid into and from the flow channel and along the fuel rods therein; and a coolant flow direction control device operatively disposed in the bottom nozzle so as to open the inlet thereof to the flow of coolant/moderator fluid in an inflow direction into the flow channel through the bottom nozzle inlet but close the inlet to the flow of coolant/moderator fluid from the flow channel through the bottom nozzle inlet upon reversal of coolant/moderator fluid flow from the inflow direction

Safety design/analysis and scenario for prevention of CDA with ECCS in lead-bismuth-cooled fast reactor

International Nuclear Information System (INIS)

Minoru, Takahashi; Vaclav, Dostal; Abu Khalid, Rivai; Novitrian; Yumi, Yamada

2007-01-01

Safety design has been developed to show safety feature of Pb-Bi-cooled direct contact boiling water small fast reactor (PBWFR). The core is designed to have negative void reactivity even if the entire core and upper plenum are voided by steam intrusion from above. In-vessel type control rod driving mechanisms are used to prevent control rods from accidental ejection due to high pressure in the reactor vessel. In cases of coolant leakage from reactor vessel and feed water pipes, Pb-Bi coolant level in the reactor vessel is kept at the required level for decay heat removal by means of closed type guard vessel. Dual pipes are adopted to avoid leak of water in the feedwater system. Pump trip in feedwater systems initiates loss of coolant flow (LOF) event, although there is no concern of loss of flow accident due to primary pump trip. Injection of high pressure water slows down the flow-coast-down of feedwater at the LOF event. It has been evaluated that the fuel temperature is kept lower than safety limits at the unprotected loss of flow and heat sink (ATWS). A scenario for prevention of the core disruptive accident (CDA) with the emergency core cooling system (ECCS) is examined. The reactor becomes super-critical when the reactor vessel is filled with water. It is necessary to use water with boric acid for the ECC system, and additional backup rods for sub-critical core in water injection. (authors)

New insight on bubble-void transition effects in irradiated materials

International Nuclear Information System (INIS)

Dubinko, V.I.

1993-01-01

An account of elastic interaction between cavities and point defects is shown to result in new critical quantities for bubblevoid transition effects in irradiated cubic crystals. In contrast to previous theories, the present one gives not only critical quantities which determine the onset of bias-driven void swelling but the maximum stationary number density and the corresponding mean radius of voids as well as the duration of the bimodal regime. The void density and swelling rate are shown to be independent from the gas level. In the region of low temperatures/high dose rates, the void density appears to be independent from irradiation parameters as well. The relationships among material constants are found at which the stabilization of gas bubbles occurs via the dislocation loop punching mechanism resulting in a drastic change in the cavity behaviour under irradiation such as the saturation (or even suppression) of void swelling and void lattice formation. The theoretical results are compared with experimental data and further experimental tests are proposed. (author). 38 refs., 1 tab., 11 figs

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....

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.

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

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.

Stress Voiding in IC Interconnects - Rules of Evidence for Failure Analysts

Energy Technology Data Exchange (ETDEWEB)

FILTER, WILLIAM F.

1999-09-17

Mention the words ''stress voiding'', and everyone from technology engineer to manager to customer is likely to cringe. This IC failure mechanism elicits fear because it is insidious, capricious, and difficult to identify and arrest. There are reasons to believe that a damascene-copper future might be void-free. Nevertheless, engineers who continue to produce ICs with Al-alloy interconnects, or who assess the reliability of legacy ICs with long service life, need up-to-date insights and techniques to deal with stress voiding problems. Stress voiding need not be fearful. Not always predictable, neither is it inevitable. On the contrary, stress voids are caused by specific, avoidable processing errors. Analytical work, though often painful, can identify these errors when stress voiding occurs, and vigilance in monitoring the improved process can keep it from recurring. In this article, they show that a methodical, forensics approach to failure analysis can solve suspected cases of stress voiding. This approach uses new techniques, and patiently applies familiar ones, to develop evidence meeting strict standards of proof.

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

"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.

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

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

Nucleation of voids in materials supersaturated with mobile interstitials, vacancies and divacancies

International Nuclear Information System (INIS)

Wolfer, W.G.; Si-Ahmed, A.

1982-01-01

In previous void nucleation theories, the void size has been allowed to change only by one atomic volume through vacancy or interstitial absorption or through vacancy emission. To include the absorption of divacancies, the classical nucleation theory is here extended to include double-step transitions between clusters. The new nucleation theory is applied to study the effect of divacancies on void formation. It is found that the steady-state void nucleation rate is enhanced by several orders of magnitude as compared to results with previous void nucleation theories. However, to obtain void nucleation rates comparable to measured ones, the effect of impurities, segregation and insoluble gases must still be invoked. (author)

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

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.