Tight Lattice Rod Bundle Research Articles

Study for Pressure Drop of Rod Bundle with Tight Lattice Array

The fuel assembly design with tightened fuel rod pitch gives a possibility of higher conversion ratio for Boiling Water Reactor. Pressure drop is one of the key design parameters to evaluate the possibility of BWR core with higher conversion ratio. The pressure drop tests for tight lattice rod bundle under steady conditions have been done under BWR fluid conditions. Two kind of a mini tight lattice bundle were used. One was a 7-rod test bundle with a hexagonal channel box. The other was a 14-rod test bundle with a rectangular channel box. The purpose of the 7-rod bundle tests was to survey the rod gap effect and the heating length effect on pressure drop. That of the 14-rod bundle test is to check the bundle shape and flow shape. The each pressure drop elements (head loss, friction loss, acceleration loss and spacer loss) were predicted based on the appropriate correlations for each item. The total pressure drop was obtained by sum of each pressure drop element. It was found from the compassion with data that the present prediction method has reasonable performance. The average and the standard deviation for the ratio of the predicted and the measured value is 1.05 and 9.7%.

OS8-2 二流体モデルを用いた稠密三角格子模擬ロッド流路内のサブチャンネルボイド率の解析(OS8 軽水炉・新型炉・核燃料サイクル,分散と集中の共存)

Void fraction data were obtained for air-water and air water with surfactant flows in a vertical channel with two subchannels simplifying a triangle tight lattice rod bundle. The data had the trends that the void fraction was lower in air-water system than air-water with surfactant one, and the void fractions for both systems were lower than those calculated by various correlations in literatures for circular pipe flow. In order to study the cause of the data trend, for annular flows as a first step, the void fraction was calculated by a two-fluid model subchannel code using several wall and interfacial friction correlations, and by assuming the liquid film to be uniform over the wall perimeter. The best agreement between the calculation and the experiment was obtained when NASCA correlation for wall friction force and modified RELAP5/MOD2 correlation incorporating reduced surface tension effects for interfacial friction force were used.

A turbulence model study for simulating flow inside tight lattice rod bundles

Performances of various turbulence models are evaluated for calculation of detailed coolant velocity distribution in a tight lattice fuel bundle. The individual models are briefly outlined and compared with respect to the prediction of wall shear stress and velocity field, for a fully developed flow inside a triangular lattice bundle. Comparisons clearly show the importance of proper modeling of the turbulence-driven secondary flows in subchannels. A quadratic k– ɛ model, which showed promising capability in this respect, is adjusted in its coefficients, and the adjusted model is applied to fully developed flow in an infinite triangular array, with various Reynolds numbers. The results show that the inclusion of adequate anisotropy modeling enables to accurately reproduce the wall shear stress distribution and velocity field in tight lattice fuel bundles.

Critical Power Correlation for Tight-Lattice Rod Bundles

Developing design correlation for the prediction of critical power in rod bundles is indispensable for R&D of Reduced-Moderation Water Reactor (RMWR) which adopts a triangular tight-lattice fuel rod configuration and axially double-humped-heated profile. In this research, critical power correlation for tight-lattice rod bundles is newly proposed using 7-rod axially uniform-heated data, 7-rod and 37-rod axially double-humped-heated data at Japan Atomic Energy Research Institute (JAERI). For comparatively high mass velocity region, the correlation is written in local critical heat flux-critical quality type. For low mass velocity region, it is written in critical quality-annular flow length type. The standard deviation of ECPR (Experimental Critical Power Ratio) to the whole JAERI data (694 data points) is 6%. The correlation is verified by Bettis Atomic Power Laboratory data (177 points, standard deviation: 7.7%). The correlation is confirmed being able to give good prediction for the effects of mass velocity, inlet temperature, pressure and heated equivalent diameter on critical power. The applicable range of the correlation is: Rod number lower than 37, rod gap from 1.0 to 2.29 mm, heated length from 1.26 to 1.8 m, mass velocity from 150 to 2,000 kg/m2·s and pressure from 2 to 11 MPa.

3526 稠密格子ロッドバンドルにおける限界出力相関式の改良と過渡限界出力試験の解析(S49-3 流動,原子炉用機器(2),S49 原子炉システムおよびその要素技術)

To develop the design correlation for the critical power in rod bundles is indispensable for R&D of Reduced-Moderation Water Reactor (RMWR) which adopts a triangular tight-lattice fuel rod configuration. In this research, critical power correlation developed for tight-lattice rod bundles is improved by using newly obtained 37-rod bundle data with 1.0mm gap between rods. Compared to the existing correlation, which based on the data obtained in 37-rod bundle data with 1.3mm gap, the improved one expands its adaptability to tighter rod bundles whose rod gap is 1mm, by revising the dependencies on mass velocity and equivalent heated diameter. The improved correlation is confirmed being able to give good critical power predictions to both its based datasets and Bettis Atomic Power Laboratory (BAPL) data. The standard deviation of ECPR (Experimental Critical Power Ratio) to JAERI 37-rod data and BAPL data are 7.3% and 8.4%, respectively. Moreover, the improved correlation is implemented into TRAC code, with which the analyses of transient processes postulated for the RMWR are carried out. For the postulated transients for the RMWR, it has been confirmed that BTs can be predicted within the accuracy of the implemented critical power correlation.

Critical Power Correlation for Tight-Lattice Rod Bundles

Developing design correlation for the prediction of critical power in rod bundles is indispensable for R&D of Reduced-Moderation Water Reactor (RMWR) which adopts a triangular tight-lattice fuel rod configuration and axially double-humped-heated profile. In this research, critical power correlation for tight-lattice rod bundles is newly proposed using 7-rod axially uniform-heated data, 7-rod and 37-rod axially double-humped-heated data at Japan Atomic Energy Research Institute (JAERI). For comparatively high mass velocity region, the correlation is written in local critical heat flux-critical quality type. For low mass velocity region, it is written in critical quality-annular flow length type. The standard deviation of ECPR (Experimental Critical Power Ratio) to the whole JAERI data (694 data points) is 6%. The correlation is verified by Bettis Atomic Power Laboratory data (177 points, standard deviation: 7.7%). The correlation is confirmed being able to give good prediction for the effects of mass velocity, inlet temperature, pressure and heated equivalent diameter on critical power. The applicable range of the correlation is: Rod number lower than 37, rod gap from 1.0 to 2.29 mm, heated length from 1.26 to 1.8 m, mass velocity from 150 to 2,000 kg/m2·s and pressure from 2 to 11 MPa.

Analysis of Seven NEPTUN-III (Tight-Lattice) Bottom-Flooding Experiments with RELAP5/MOD3.3/BETA

Seven tight-lattice NEPTUN-III bottom-flooding experiments are analyzed by using the frozen version of RELAP5, RELAP5/MOD3.3/BETA. This work is part of the Paul Scherrer Institute (PSI) contribution to the High Performance Light Water Reactor (HPLWR) European Union project and aims at assessing the capabilities of the code to model the reflooding phenomena in a tight hexagonal lattice (which was one of the core geometries considered at the time for an HPLWR) following a hypothetical loss-of-coolant accident scenario. Even though the latest version of the code has as a default the new PSI reflood model developed by the author, which was tested and assessed against reflooding data obtained at standard light water reactor lattices, this work shows that for tight lattices, the code underpredicts the peak clad temperatures measured during a series of reflooding experiments performed at the NEPTUN-III tight-lattice heater rod bundle facility. The reasons for these differences are discussed, and the (possible) changes needed in the framework of RELAP5/MOD3.3 for improving the modeling of reflooding in tight lattices are investigated.

高稠密格子水冷却炉心の圧力損失(水冷却炉,革新型原子炉の開発および多目的利用技術,原子力要素技術開発)

In order to design Reduced-Moderation Water Reactor core from a thermal-hydraulic point of view, an evaluation method on a pressure drop in a tight-lattice rod bundle is required. In this study, the evaluation method by Martinelli-Nelson was investigated from a comparison with experimental data in a tight-lattice 37-rod bundle. The method tends to overestimate the pressure drop under low pressure condition (P_<2MPa), whereas the method can predict the pressure drop with high accuracy near the nominal operating condition (P=7.2MPa).

稠密三角格子サブチャンネル間の乱流混合(原子炉伝熱流動解析の革新(2),原子炉伝熱流動解析の革新)

A vertical test channel made up of two subchannels simplifying a triangle tight lattice rod bundle was newly constructed, and experimental data were obtained on turbulent mixing rate between the subchannels. Experiments were conducted in various single-phase and two-phase air-water flows under the hydrodynamic equilibrium flow conditions. The present mixing rate data were compared with those reported so far for triangle lattice rod bundle channels as well as square ones, having hydraulic diameters of about four times larger than the present tight lattice channel. It was found that the mixing rate was considerably smaller in the present channel than the previous triangle and square lattice channels, i.e., a channel size effect.

稠密バンドル用沸騰遷移相関式の改良(NP3 新型炉技術)

Developing design correlation for the prediction of critical power in rod bundles is indispensable for R&D of Reduced:Moderation Water Reactor (RMWR) which adopts a triangular tight-lattice fuel rod configuration with a gap of about 1 mm between rods. In this research, critical power correlation for tight-lattice rod bundles is newly proposed using 7-rod axially uniform-heated data, 7-rod and 37-rod axially double-humped-heated data at Japan Atomic Energy Developing design correlation for the prediction of critical power in rod bundles is indispensable for R&D of Reduced: Moderation Water Reactor (RMWR) which adopts a triangular tight-lattice fuel rod configuration with a gap of about 1 mm between rods. In this research, critical power correlation for tight-lattice rod bundles is newly proposed using 7-rod axially uniform-heated data, 7-rod and 37-rod axially double-humped-heated data at Japan Atomic Energy Research Institute (JAERI). For low mass velocity region (< 300 kg/m^2s), the correlation is written in critical quality-annular flow length type. For high mass velocity region (> 300 kg/m^2s), it is written in local critical heat flux-critical quality type. The standard deviation of ECPR (Experimental Critical Power Ratio) to the whole JAERI data (694 data points) is 6%. The correlation is verified by Bettis Atomic Power Laboratory data (177 points, standard deviation: 7.7%). The correlation is confirmed being able to give good prediction for the effects of mass velocity, inlet temperature, pressure and heated equivalent diameter on critical power. The applicable range of the correlation is: gap between rods from 1.0 to 2.29 mm, heated length from 1.26 to 1.8 m, mass velocity from 150 to 2000 kg/m^2s and pressure from 2 to 11 Mpa.

2237 高稠密格子水冷却炉心の除熱技術の開発 (2) : 大型熱特性試験とモデル実験

In this manuscript, the key point of large-scale thermal-hydraulic tests and model experiments which simulate the Reduced-Moderation Water Reactor (RMWR) core are reported. The aim of the large-scale thermal-hydraulic tests using 37-rod bundle test facility is to make clear the rod number effect, rod gap effect etc. for investigating the feasibility of RMWR. Thermal-hydraulic model experiments will be performed in order to verify an advanced 3-D two-phase flow simulation method. Void fraction distribution and its fluctuation and basic thermal characteristics in tight lattice rod bundles will be measured by high-frame-rate neutron radiography technique etc. We will develop the database by 2007 that can resolve the fundamental feasibility subjects for the RMWR.

Critical Power in Axially Uniformly-Heated Tight-Lattice Rod Bundles

Critical power experiments using the test section with an axially uniformly-heated tight-lattice rod bundle have been carried out. One of the aims of the experiments is to propose the critical power correlation. Critical power in tight-lattice rod bundles is very important on the thermal-hydraulic design of the Reduced-Moderation Water Reactor (RMWR). Many parameter effects on critical power such as flow and configuration parameters, axial and radial power distributions, bundle size etc. have to be considered. In this investigation, we focused on the fundamental parameter effects such as flow and configuration parameter and radial power distribution on critical power as the first step. Test section was an axially uniformly-heated 7-rod bundle with gap between rods of 1.0 mm, rod diameter of 12.3mm and heated length of 1.8m. Critical power correlation for the axially uniformly-heated tight rod bundles was derived based on the present and BAPL databases. An excellent agreement was obtained between the developed correlation and data (371 points) within an error of ±4.6%.

G204 稠密バンドル内の流動特性に関する大規模直接解析

G204 稠密バンドル内の流動特性に関する大規模直接解析

D108 中性子トモグラフィによる稠密バンドル内二相流の可視化と計測

Neutron tomography thermal-hydraulic measurement technique is originally developed based on the neutron radiography, computed tomography and two-phase flow measurement techniques. The purpose of the developing is to measure the void fraction distribution in the Reduced-Moderation Water Reactor which is a water-cooled breeder reactor designed by the JAERI as a future reactor. We have visualized and measured the void distribution of air/water two-phase flow and boiling flow in tight-lattice rod bundles. We used the research reactor JRR-3 as a neutron source. Three-dimensional data can be obtained in order to evaluate the numerical analysis codes. In this manuscript, the neutron tomography system, comparison between the reconstruction methods of computed tomography and examples of the measured two-phase flow data which was taken in the 7 rod bundle with a gap between rods of 1.0mm.

1914 稠密燃料バンドルの伝熱特性評価

A new fuel assembly design with tightened fuel rod has been proposed to achieve high conversion ratio about 1 on Boiling Water Reactor. The steady and transient critical power performance of the tight lattice rod bundle had not been known sufficiently in spite of important items for evaluating the possibility of BWR core with high conversion ratio. Therefore, the critical power measurement tests for tight lattice rod bundle have been performed under BWR steady and transient conditions. As that result, it was found that Arai CHF correlation had enough ability for predicting the steady-state critical power and transient dryout time for tight lattice BWR rod bundle.

DNB experiments for high-conversion PWR core design

It is very important to clarify the departure from nucleate boiling (DNB) performance of core fuel assemblies for the high conversion pressurized water reactors (PWR) [1,2]. To investigate this, DNB experiments were performed in tight lattice rod bundles, using the model fluid Freon 12 and water under the actual operating conditions. In addition, DNB heat flux measurements in an annular-flow channel were carried out for the design of fertile rods, which are installed in thimble tubes.

Numerical Simulation of Reflooding Behavior in Tight-Lattice Rod Bundles

To evaluate the applicability of the reflood analysis code REFLA for ordinal pressurized water reactors to the analysis of reflooding phenomena in light water high conversion reactors (LWHCRs) with tight-lattice cores, a numerical simulation of the NEPTUN LWHCR test was performed with the REFLA code. The NEPTUN LWHCR test was performed at the Swiss Federal Insitute for Reactor Research with a test section simulating the tight-lattice core of an LWHCR. The results indicate no potential problems in the use of REFLA for the simulation of reflooding behavior in tight-lattice rod bundles. To improve the code, however, it is recommended to modify models of core heat transfer at a high flooding rate and core water distribution (integration of droplet flow) in the axial direction, and to investigate core pressure drop and horizontal cross flow.