CFD investigation of helical wire-wrapped 7-pin fuel bundle and the challenges in modeling full scale 217 pin bundle

Flow and temperature distributions of sodium in a heat generating fuel pin bundle with helically wound spacer wire have been predicted from basic principles by solving the three-dimensional conservation equations of mass, momentum and energy, for a wide range of Reynolds number. Turbulence has been modeled using the k– turbulence model. The geometry details of the bundle and heat flux from the fuel pin are similar to that of the Indian Prototype Fast Breeder Reactor (PFBR) that is currently under construction. The focus of the study is to assess the effect of transverse flow in promoting flow and temperature uniformity. It is seen that the ratio of maximum transverse velocity to the maximum axial velocity is nearly equal to the tangent of the rolling up angle of the spacer wire. Due to the wire wrap, the difference in bulk sodium temperature between the peripheral and central sub-channels is reduced to by a factor of 4 when compared to that without spacer wire. The film drop at the junction between wire and the pin is found to be only 70 °C. The predicted results are found to be in close agreement with that of the experimental results reported in literature. The present study considers a 7-pin bundle assembly of one helical pitch. The computational time and memory required for a 217 pin with 15 pitches assembly is ascertained to be 500 times that required for the current study. Hence, research activities have been directed towards developing a parallel CFD code and structural mesh generation software.     

CFD investigations of helical wire-wrap fuel pin bundle and its comparison with straight wire bundle

Computational Fluid Dynamics (CFD) investigations of a fast reactor fuel pin bundle wrapped with helical and straight spacer wires have been carried out and the advantages of using helical spacer wire have been assessed. The flow and temperature distributions in the fuel pin bundle are obtained by solving the statistically averaged 3-Dimensional conservation equations of mass, momentum and energy along with high Reynolds number k-ε turbulence model using a customized CFD code CFDEXPERT. It is seen that due to the helical wire-wrap spacer, the coolant sodium not only flows in axial direction in the fuel pin bundle but also in a transverse direction. This transverse flow enhances mixing of coolant among the sub channels and due to this, the friction factor and heat transfer coefficient of the coolant increase. Estimation of friction factor, Nusselt number, sodium temperature uniformity at the outlet of the bundle and clad hot spot factor which are measures of the extent of coolant mixing and non-homogeneity in heat transfer coefficient around fuel pin are paid critical attention. It is seen that the friction factor and Nusselt number are higher (by 25% and 15% respectively) for the helical wire wrap pin bundle compared to straight wire bundle. It is seen that for 217 fuel pin bundle the maximum clad temperature is 750 K for straight wire case and the same for helical wire is 720 K due to the presence of transverse flow. The maximum temperature occurs at the location of the gap between pin and wire. The ΔT between the bulk sodium in the central sub-channel and peripheral sub-channel is 30 K for straight wire and the same for helical wire is 18 K due to the presence of secondary transverse flow which makes the outlet temperature more uniform. The hotspot factor and the hot channel factors predicted by CFD simulation are 10% lower than that used in conventional safety analysis indicating the conservatism in the safety analysis.     

Distributed parameter analysis for the prediction of the fine structure of flow and temperature fields in wire-wrapped fuel pin bundle geometries

This paper describes the numerical method of a distributed parameter analysis code SPIRAL for the calculation of fluid flow and temperature in arbitrary channel geometries, discusses the numerical method in the modeling and solution of the problem, and presents some results, including comparison with experiments.

The derivation and solution of the finite element equations is discussed. In order to overcome difficulties arising from the geometry, the Galerkin finite element method using isoparametric elements was employed, and a procedure of finite element generation using curvilinear coordinate system was developed.

The SPIRAL code permits calculation of the fine structure of the multi-dimensional steady-state single-phase fluid flow and temperature fields in LMFBR fuel pin subassemblies in the presence of wire spacers. Calculated results are presented for crossflow velocity distributions and crossflow pressure drop characteristics in a tube bundle geometry with and without wire spacers, natural convection and heat transfer in horizontal annuli, flow in a wire-wrapped 7-pin bundle geometry and fully developed turbulent flow in a parallel 4-rod array contained in a rectangular duct.     

Development of a bundle correction method and its application to predicting CHF in rod bundles

A bundle correction method, based on the conservation laws of mass, energy, and momentum in an open subchannel, is proposed for the prediction of the critical heat flux (CHF) in rod bundles from round tube CHF correlations without detailed subchannel analysis. It takes into account the effects of the enthalpy and mass velocity distributions at subchannel level using the first derivatives of CHF with respect to the independent parameters. Three different CHF correlations for tubes (Groeneveld's CHF table, Katto correlation, and Biasi correlation) have been examined with uniformly heated bundle CHF data collected from various sources. A limited number of CHF data from a non-uniformly heated rod bundle are also evaluated with the aid of Tong's F-factor. The proposed method shows satisfactory CHF predictions for rod bundles both uniform and non-uniform power distributions.     

Development of the residual sodium quantification method for a fuel pin bundle of SFRs before and after dry cleaning

ABSTRACT

In a fuel handling system of sodium-cooled fast reactors (SFRs), it is necessary to remove the sodium remaining on spent fuel assemblies (FAs) before storing them in a spent fuel water pool (SFP). A next-generation SFR in Japan has adopted an advanced dry-cleaning system that consists of argon gas blowing to remove the metallic residual sodium on the FA, which increases economic competitiveness and reduces waste products thanks to a waterless process. In this R&D work, the performance of the dry cleaning process has been investigated.

This paper describes experimental and analytical studies focusing on the amount of residual sodium remaining on a fuel pin bundle before and after the argon gas blowing process. The experiments were conducted using a sodium test loop and a short (approximately 1 m) specimen consisting of a 7-pin bundle. The effects of the blowing gas velocity and the blowing time were quantitatively analyzed in the experiments. The blowing gas velocity was varied from 3.9 to 31.3 m/s, and 113 data-points of the residual sodium were collected during the experiment. On the basis of these experimental results, the residual sodium quantification method for the fuel pin bundle was constructed.     

CFD modelling of subcooled boiling—Concept, validation and application to fuel assembly design

The paper describes actual Computational Fluid Dynamics (CFD) approaches to subcooled boiling and investigates their capability to contribute to fuel assembly design. In a prototype version of the CFD code CFX a wall-boiling model is implemented based on a wall heat flux partition algorithm. It can be shown, that the wall boiling model is able to calculate the cross sectional averaged vapour volume fraction of vertical heated tubes tests with good agreement to published experimental data. The most sensitive parameters of the model are identified. Needs for more detailed experiments are established which are necessary to support further model development. The model is applied for investigation of the phenomena inside a hot channel of a fuel assembly. Here the essential phenomenon is the critical heat flux. Although subcooled boiling represents only a preliminary state towards the critical heat flux occurrence, essential parameters like swirl, cross flow between adjacent channels and concentration regions of bubbles can be determined. By calculating the temperature of the rod surface the critical regions can be identified which may later on lead to departure from nucleate boiling and possible damage of the fuel pin. The application of up-to-date CFD with a subcooled boiling model for the simulation of a hot channel enables the comparison and the evaluation of different geometrical designs of the spacer grids of a fuel rod bundle.     

A critical heat flux approach for square rod bundles using the 1995 Groeneveld CHF table and bundle data of heat transfer research facility

The critical heat flux (CHF) approach using CHF look-up tables has become a widely accepted CHF prediction technique. In these approaches, the CHF tables are developed based mostly on the data bank for flow in circular tubes. A set of correction factors was proposed by Groeneveld et al. [Groeneveld, D.C., Cheng, S.C., Doan, T., 1986. 1986 AECL-UO Critical Heat Flux lookup table. Heat Transf. Eng. 7(1–2), 46] to extend the application of the CHF table to other flow situations including flow in rod bundles. The proposed correction factors are based on a limited amount of data not specified in the original paper. The CHF approach of Groeneveld and co-workers is extensively used in the thermal hydraulic analysis of nuclear reactors. In 1996, Groeneveld et al. proposed a new CHF table to predict CHF in circular tubes [Groeneveld, D.C., et al., 1996. The 1995 look-up table for Critical Heat Flux. Nucl. Eng. Des. 163(1), 23]. In the present study, a set of correction factors is developed to extend the applicability of the new CHF table to flow in rod bundles of square array. The correction factors are developed by minimizing the statistical parameters of the ratio of the measured and predicted bundle CHF data from the Heat Transfer Research Facility. The proposed correction factors include: the hydraulic diameter factor (K_hy), the bundle factor (K_bf), the heated length factor (K_hl), the grid spacer factor (K_sp), the axial flux distribution factors (K_nu), the cold wall factor (K_cw) and the radial power distribution factor (K_rp). The value of constants in these correction factors is different when the heat balance method (HBM) and direct substitution method (DSM) are adopted to predict the experimental results of HTRF. With the 1995 Groeneveld CHF Table and the proposed correction factors, the average relative error is 0.1 and 0.0% for HBM and DSM, respectively, and the root mean square (RMS) error is 31.7% in DSM and 17.7% in HBM for 9852 square array data points of HTRF.     

A modified equilibrium void distribution model applicable to subchannel-scale vapor–liquid cross flow model for conventional square and tight lattice BWR fuel bundles

As a part of the advanced subchannel code development project sponsored by Ministry of Economy, Trade and Industry, Japan, this paper describes improvement of the equilibrium void distribution model that is a main part of the vapor–liquid cross flow model.The three-component cross flow (TCCF) model is defined as the present framework that separates contributions of diversion, turbulent mixing and void drift. The Lahey's void settling model is introduced to express the latter two components. Based on the high-resolution air–water database and other published data of steam-water tests, general trends of vapor–liquid cross flow processes are examined. It can be assumed that subchannel void distributions are dominated by the three major effects, i.e. the fluid dynamic effect, the geometrical effect and the narrow gap effect.The equilibrium void distribution model is modified to include the above-mentioned three effects. Three characteristic parameters are assigned for each of the three effects and they are identified experimentally as functions of the void fraction. Multi-dimensional lattice geometries are incorporated based on the two-dimensional flow network model. The network equation is constructed by mapping the equilibrium void balance problem into the force-deflection problem. The resultant models are verified based on equilibrium void distribution data obtained by Sadatomi and Kawahara.     

Modification of thermal hydraulic code,THEATRe, with fuel rod behavior models and its application to reactivity-initiated accident (RIA)

During reactor operation, many complex changes occur in fuel rod which affects its thermal, mechanical and material properties. These changes also affect the reactor response to the transient and accident situations. Realistic simulation of fuel rod behavior under transients such as reactivity-initiated accident (RIA) is of great significance. In this study, thermal hydraulic analysis code THEATRe (Thermal Hydraulic Engineering Analysis Tool in Real-time) has been modified by addition of fuel rod behavior models for dynamic simulation of nuclear reactor. Transient changes in gas-gap parameters were taken into account by modeling the gas-gap behavior. Thermo-mechanical behavior of fuel rod is modeled to take into account the thermal, elastic and plastic deformation. To simulate RIA, point reactor kinetics model is also incorporated in the THEATRe code. To demonstrate the transient fuel rod behavior, AP1000 reactor is modeled and three hypothetical RIA cases are simulated. The RIA is considered at three different reactor power levels, i.e. 100, 50 and 1% of nominal power. The investigated parameters are fuel temperature, cladding stress and strain, fuel and cladding thermal conductivity and heat transfer coefficient in gas-gap. Modified code calculates the fuel rod temperatures according to updated fuel, clad and gas-gap parameters at the onset of steady-state operation and during the transient. The modified code provides lower steady-state fuel temperature as compared to the original code. Stress and strain analyses indicate that the hoop and radial strain is higher at high power locations of the fuel rod; therefore, gap closure process will initially occur in the central portion of the fuel rod and it should be given more emphasis in the safety analysis of the fuel rod and nuclear reactor during accidents and transients.     

A comparative study of ZPR-6/7 with MCNP/5 and MC-2/REBUS

This work models the INL ZPR-6/7 assembly employing two different approaches: a probabilistic approach using MCNP/5 and a deterministic one using MC²-2/REBUS. With MCNP/5, each drawer of the assembly is modeled in detail with regard to geometry and fuel loading. In the deterministic approach, the MC²-2 collapses cross-sections in energy and space into a 15 few group structure homogenized spatially over each drawer for the REBUS 3D model. Various reactivity coefficients and reaction rates at different locations inside the core were evaluated and compared for both approaches and contrasted to published experimental data and were found to be in good agreement.     

A mass spectrometry method for quantitative and kinetic analysis of gas release from nuclear materials and its application to helium desorption from UO2 and fission gas release from irradiated fuel

A new system has been developed to determine absolute quantities of gas (mainly noble gases) released during thermal desorption in the range from 10^?12 to 10^?5 mol with a precision of few percent. The system is actually designed for simultaneous measurement of gaseous elements like He, Xe, Kr, thermally released from nuclear fuel samples and also allows the determination of the release kinetics as a function of time. This system, called Quantitative GAs MEasurement System (Q-GAMES), is based on the principle of collecting, purifying and spiking the sample gas in a “high-pressure” chamber, and continuous sampling of the gas for mass spectrometric analysis without sample depletion during the experiment. It is equipped with its own spike generator and with different gas purification systems. It is shown that this system fulfills the requirement to work with two existing very high-temperature gas desorption facilities for nuclear materials. This paper describes the Q-GAMES principle, the spiking system, its calibration, its operative mode, the different quantification techniques, as well as its technical data, in combination with some examples of typical application.     

NF-86、NF-87与FWT-60三种辐射变色染料尼龙薄膜剂量计性能的比较研究

NF-86和NF-87两种剂量计均是以尼龙为基质材料,分别以六羟乙基付品红氰化物和付品红氰化物为辐射变色染料的薄膜剂量计。本文给出这两种剂量计与国外同类产品——美国远西公司的FWT-60辐射变色尼龙薄膜剂量计主要剂量学性能比较研究的结果。结果表明NF-86、NF-87剂量计的性能已接近和达到美国FWT-60剂量计的水平,能够为国内辐射加工中γ剂量测量(10~3—10~5Gy)提供一种简便快速而又可靠的监测手段。     

Application of a Zircaloy/Steam Oxidation Model to a CFD Code and its Validation against a CANDU Fuel Channel Experiment: CS28-2

Oxidation of a Zircaloy cladding exposed to high-temperature steam is an important phenomenon in the safety analysis of CANDU reactors during a postulated loss-of-coolant accident (LOCA), since a Zircaloy/steam reaction is highly exothermic and results in hydrogen production. As part of a computational fluid dynamics (CFD) simulation of the CS28-2 high-temperature experiment for this accident analysis, two Zircaloy/steam reaction models based on a parabolic rate law are implemented in a commercial CFD code (CFX-10) through a user FORTRAN. It is confirmed that the present oxidation models for the CFX-10 reproduce the results of each empirical correlation in the verification tests well. Then the CFX-10 predictions of a temperature rise and hydrogen production due to Zircaloy/steam oxidation are compared with the results of the CS28-2 experiment. From these validation processes, it is shown that the Urbanic-Heidrick model, which is widely used in CANDU fuel channel codes, is also applicable to a CFX-10 simulation of Zircaloy/steam oxidation in a CANDU fuel channel.     

Stress corrosion fracture of zircaloy cladding in fuel rods subjected to power increases: A model for crack propagation and the failure threshold stress

A crack may form and propagate by a stress corrosion mechanism, in the zircaloy cladding of a water cooled fuel rod, if it is subjected to a sufficiently severe power increase (ramp), the likely responsible chemical species being iodine produced by fissioning of the fuel. By formulating and analysing a model, and relating the theoretical predictions to observations on the size of plastic zones associated with a propagating crack, valuable information is obtained concerning the micro-mechanics of stress corrosion fracture in zircaloy; this information is then used as input for a theoretical analysis of crack formation. A model describing the crack formation process is developed and it is shown that the threshold stress can be low in relation to the yield stress of Irradiated zircaloy, a prediction that accords with some recent experimental observations. Implications of the results with respect to both fuel failure predictions and possible cladding improvements by means of texture changes are discussed.