Noble gas binary mixtures for gas-cooled reactor power plants

This paper examines the effects of using noble gases and binary mixtures as reactor coolants and direct closed Brayton cycle (CBC) working fluids on the performance of terrestrial nuclear power plants and the size of the turbo-machines. While pure helium has the best transport properties and lowest pumping power requirement of all noble gases and binary mixtures, its low molecular weight increases the number of stages of the turbo-machines. The heat transfer coefficient for a He–Xe binary mixture having a molecular weight of 15 g/mole is 7% higher than that of helium, and the number of stages in the turbo-machines is 24–30% of those for He working fluid. However, for the same piping and heat exchange components design, the loop pressure losses with He–Xe are 3 times those with He. Consequently, for the same reactor exit temperature and pressure losses in piping and heat exchange components, the higher pressure losses in the nuclear reactor decrease the net peak efficiency of the plant with He–Xe working fluid (15 g/mole) by a little more than 2% points, at higher cycle compression ratio than with He working fluid.     

Tribological performance of MoS2 coatings in liquid helium and at high loads

The main part of a narrow support element (NSE) of the W7-X superconducting coil system is an aluminium bronze pad, PVD coated on its spherical surface with MoS₂, which slides against the flat surface of the stainless steel coil housing, coated with MoS₂ spray. The operational requirements of the NSEs are: vacuum of p < 10⁻⁶ mbar, temperature T  4 K, maximum load P 1500 kN, typical displacement ≤5 mm, smooth sliding and no stick-slip events. The paper describes test results obtained with a downscaled NSE at T = 4.2 and 77 K. During the test the NSEs were submerged in liquid helium and nitrogen, respectively. Whereas the LN₂ test ran smoothly for up to 15,000 cycles, the test in LHe showed stick-slip from the very first cycle. The stick-slip disappeared after 50 cycles. Post mortem analysis of the tested parts revealed that in case of LHe the sprayed MoS₂ film was removed during the first 30–100 cycles by blistering and flaking. The reason for the loss of adhesion at LHe temperature is not known, several possible causes are under discussion. Further experiments under vacuum and at T 4 K are being prepared which are expected to help in clarifying the issue.     

Effects of working fluid and shaft rotation speed on the performance of HTR plants and the size of CBC turbo-machine

Investigated are the effects of the molecular weight of the working fluid, reactor exit temperature, and shaft rotation speed on the size and number of stages of the turbo-machine as well as the performance of high temperature reactor (HTR) plants with actively cooled reactor pressure vessel and direct or indirect Closed Brayton Cycles (CBCs). The present analyses for working fluids of helium (4 g/mol) and the 15 g/mol He–Xe and He–N₂ binary mixtures are performed for a reactor thermal power of 600 MW, shaft rotation speed of 3000–9000 rpm, and reactor exit temperature from 973 K to 1223 K. For the plants with indirect CBCs, the analyses assume a temperature pinch of 50 K in the IHX. Results show that the CBC compression ratio is relatively low (2.6 for He and He–Xe and 3.2 for He–N₂), increases very little with increasing the reactor exit temperature and near the maximum thermal efficiency of the plant. For the plants with a direct helium CBC, the thermal efficiency increases from 42% to 51% as the reactor exit temperature increases from 973 K to 1223 K, respectively, versus 37% to 47% for the plants with indirect He-CBC. The HTR plants with indirect He–Xe and He–N₂ CBCs and operating at a turbine inlet temperature of 1123 K have slightly higher thermal efficiencies (45.9% and 45.8%) than the He plant with indirect CBC (45.6%), generating 1.6 MWe more electrical power. The molecular weight of the working fluid has a very small effect on the plant thermal efficiency, but significantly reduces the size and number of stages of the CBC turbo-machine. Increasing the shaft rotation speed also decreases the size and number of stages of the CBC turbo-machine.     

Neutronic performance of HYLIFE-II fusion reactor using various thorium molten salts

HYLIFE-II is one of the major inertial fusion energy reactor design concepts in which a thick molten salt layer (Flibe = Li₂BeF₄) is injected between the reaction chamber walls and the explosions. Molten salt coolant eliminates the frequent replacement of solid first wall structure during reactors lifetime by decreasing intense neutron flux. This study presents the neutronic analysis of HYLIFE-II fusion reactor using various liquid wall coolants, namely, 75% LiF–25% ThF₄, 75% LiF–24% ThF₄–1% ²³³UF₄ or 75% LiF–23% ThF₄–2% ²³³UF₄. Neutron transport calculations for the evaluation of neutron spectra were conducted with the help of Scale 4.3 by solving the Boltzmann transport equation in S₈–P₃ approximation. The effects of flowing liquid wall thickness and type of coolant on the neutronic performance of the reactor were investigated. Furthermore, radiation damage calculations at the first wall structure with respect to type and thickness of liquid wall were carried out. Numerical results showed that using the flowing liquid wall containing the molten salt, 75% LiF–23% ThF₄–2% UF₄ with a thickness of 70 cm maintained tritium self-sufficiency of the (DT) fusion driver and extended the first wall lifetime to the reactors lifetime (30 full power years). In addition significant amount of high quality fissile fuel was bred through (n, γ) reaction of ²³²Th. Moreover, energy multiplication factor (M) was increased to 12 by high rate fission reactions of ²³³U occurring in the flowing wall. On the other hand, it was concluded that using the other two coolants, 75% LiF–25% ThF₄ or 75% LiF–24% ThF₄–1% ²³³UF₄, as liquid wall did not satisfy the radiation damage and the tritium sufficiency criteria together at any thickness, so that these two coolants were not suitable to improve neutronic performance of HYLIFE-II reactor.     

Simultaneous measurement of liquid velocity and interface profiles of horizontal duct wavy flow by ultrasonic velocity profile meter

A simultaneous measurement of the liquid velocity and interface profiles was performed for stratified-smooth and wavy flows in a horizontal duct using a ultrasonic velocity profile (UVP) meter. The influences of the reflections of ultrasonic pulses at the gas–liquid interface and channel bottom were reduced by using an absorbent for the ultrasonic pulses on the duct bottom wall and optimization of the liquid level and time interval between pulses. For a smooth–stratified flow, good comparison was obtained with a velocity profile obtained by particle tracking velocimetry (PTV) for video pictures taken simultaneously at the UVP measurement. Polystyrene beads were used as the reflector and tracers respectively, for the UVP and PTV measurements. The velocity profiles measured for a wavy flow with periodically-generated interfacial waves agreed well with the theoretical prediction for solitary waves. Turbulence component appeared in the velocity profiles of both the smooth–stratified and wavy flows.     

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.     

Very high swelling and embrittlement observed in a Fe–18Cr–10Ni–Ti hexagonal fuel wrapper irradiated in the BOR-60 fast reactor

The highest void swelling level ever observed in an operating fast reactor component has been found after irradiation in BOR-60 with swelling in Kh18H10T (Fe–18Cr–10Ni–Ti) austenitic steel exceeding 50%. At such high swelling levels the steel has reached a terminal swelling rate of 1%/dpa after a transient that depends on both dpa rate and irradiation temperature. The transient duration at the higher irradiation temperatures is as small as 10–13 dpa depending on which face was examined. When irradiated in a fast reactor such as BOR-60 with a rather low inlet temperature, most of the swelling occurs above the core center-plane and produces a highly asymmetric swelling loop when plotted vs. dpa. Voids initially harden the alloy but as the swelling level becomes significant the elastic moduli of the alloy decreases strongly with swelling, leading to the consequence that the steel actually softens with increasing swelling. This softening occurs even as the elongation decreases as a result of void linkage during deformation. Finally, the elongation decreases to zero with further increases of swelling. This very brittle failure is known to arise from segregation of nickel to void surfaces which induces a martensitic instability leading to a zero tearing modulus and zero deformation.     

The MEGAPIE-TEST project: Supporting research and lessons learned in first-of-a-kind spallation target technology

The megawatt pilot experiment (MEGAPIE) has been launched by six European institutions (PSI, FZK, CEA, SCK-CEN, ENEA and CNRS), JAEA (Japan), DOE (US) and KAERI (Korea) with the aim to carry out an experiment, in the SINQ target location at PSI (Switzerland), to demonstrate the safe operation of a liquid metal (lead–bismuth eutectic, LBE) spallation target hit by a 1 MW proton beam. The European Commission has joined the MEGAPIE project through the 5-year (2001–2006) project named MEGAPIE-TEST. This project has been formally concluded with an International Workshop, where the results and the lessons learned during the project have been summarised. This work presents a review of the outcome of that Workshop.     

Experimental study on natural circulation and its stability in a heavy liquid metal loop

Motivated by the increasing interest in heavy liquid metal (HLM) cooled fast reactors and accelerator driven system (ADS), the TALL test facility was designed and constructed at KTH to investigate the thermal-hydraulic characteristics of HLM. In this paper, the HLM natural circulation characteristics in a HLM loop were investigated with experiments in the TALL test facility. The study includes measurements on (1) start-up of natural circulation from different initial conditions; (2) stability of natural circulation; (3) effects of influencing parameters and (4) capability of natural circulation. The experimental data are compared to predictions with a relevant code (RELAP5). Significant natural convection flow was observed in the experiments. It was found that the natural circulation was easily established and stabilized. It took only a few minutes to have a stable natural circulation prevailing from cold conditions. The natural circulation flowrate depends on the loop resistance, and the temperature difference between the hot leg and the cold leg, as determined by the power level and the heat sink capacity. The experiments show that the maximum flowrate for the natural circulation is 0.5 kg/s (corresponding to 0.5 m/s in the heat exchanger), resulting in heat removal of 15 kW from the core tank, which is comparable to the capacity of 100 W/cm of the electric heater elements. The preliminary analysis performed with the RELAP5 code is in reasonable agreement with the experimental data.     

Rapid-relocation model for describing high-fluence retention of rare gases implanted in solids

It has been known for a long time that the maximum areal density of inert gases that can be retained in solids after ion implantation is significantly lower than expected if sputter erosion were the only limiting factor. The difference can be explained in terms of the idea that the trapped gas atoms migrate towards the surface in a series of detrapping–trapping events so that reemission takes place well before the receding surface has advanced to the original depth of implantation. Here it is shown that the fluence dependent shift and shape of implantation profiles, previously determined by Rutherford backscattering spectrometry (RBS), can be reproduced surprisingly well by extending a simple retention model originally developed to account only for the effect of surface recession by sputtering (‘sputter approximation’). The additional migration of inert gas atoms is formally included by introducing an effective shift parameter Y_eff as the sum of the sputtering yield Y and a relocation efficiency Ψ_rel. The approach is discussed in detail for 145 keV Xe⁺ implanted in Si at normal incidence. Y_eff was found to increase with increasing fluence, to arrive at a maximum equivalent to about twice the sputtering yield. At the surface one needs to account for Xe depletion and the limited depth resolution of RBS. The (high-fluence) effect of implanted Xe on the range distributions is discussed on the basis of SRIM calculations for different definitions of the mean target density, including the case of volume expansion (swelling). To identify a ‘range shortening’ effect, the implanted gas atoms must be excluded from the definition of the depth scale. The impact-energy dependence of the relocation efficiency was derived from measured stationary Xe concentrations. Above some characteristic energy (20 keV for Ar, 200 keV for Xe), Y exceeds Ψ_rel. With decreasing energy, however, Ψ_rel increases rapidly. Below 2–3 keV more than 90% of the reemission of Ar and Xe is estimated to be due to bombardment induced relocation and reemission, only the remaining 10% (or less) can be attributed to sputter erosion. The relocation efficiency is interpreted as the ‘speed’ of radiation enhanced diffusion towards the surface. The directionality of diffusion is attributed to the gradient of the defect density on the large-depth side of the damage distribution where most of the implanted rare gas atoms come to rest. Based on SRIM calculations, two representative parameters are defined, the peak number of lattice displacements, N_d,m, and the spacing, z_r,d, between the peaks of the range and the damage distributions. Support in favour of rapid rare gas relocation by radiation enhanced diffusion is provided by the finding that the relocation efficiencies for Ar and Xe, which vary by up to one order of magnitude, scale as Ψ_rel=kN_d,m/Δz_r,d, independent to the implantation energy (10–80 keV Ar, 10–500 keV Xe), within an error margin of only ± 15%. The parameter k contains the properties of the implanted rare gas atoms. A recently described computer simulation model, which assumed that the pressure established by the implanted gas drives reemission, is shown to reproduce measured Xe profiles quite well, but only at that energy at which the fitting parameter of the model was determined (140 keV). Using the same parameter at other energies, deviations by up to a factor of four are observed.     

AgInCd control rod failure in the QUENCH-13 bundle test

The QUENCH off-pile experiments performed at the Karlsruhe Research Center are to investigate the high-temperature behavior of Light Water Reactor (LWR) core materials under transient conditions and in particular the hydrogen source term resulting from the water injection into an uncovered LWR core. The typical LWR-type QUENCH test bundle, which is electrically heated, consists of 21 fuel rod simulators with a total length of approximately 2.5 m. The Zircaloy-4 rod claddings and the grid spacers are identical to those used in Pressurized Water Reactors (PWR) whereas the fuel is represented by ZrO₂ pellets. In the QUENCH-13 experiment the single unheated fuel rod simulator in the center of the test bundle was replaced by a PWR-type control rod. The QUENCH-13 experiment consisting of pre-oxidation, transient, and quench water injection at the bottom of the test section investigated the effect of an AgInCd/stainless steel/Zircaloy-4 control rod assembly on early-phase bundle degradation and on reflood behavior. Furthermore, in the frame of the EU 6th Framework Network of Excellence SARNET, release and transport of aerosols of a failed absorber rod were to be studied in QUENCH-13, which was accomplished with help of aerosol measurements performed by PSI–Switzerland and AEKI–Hungary.Control rod failure was initiated by eutectic interaction of steel cladding and Zircaloy-4 guide tube and was indicated at about 1415 K by axial peak absorber and bundle temperature responses and additionally by the on-line aerosol monitoring system. Significant releases of aerosols and melt relocation from the control rod were observed at an axial peak bundle temperature of 1650 K. At a maximum bundle temperature of 1820 K reflood from the bottom was initiated with cold water at a flooding rate of 52 g/s. There was no noticeable temperature escalation during quenching. This corresponds to the small amount of about 1 g in hydrogen production during the quench phase (compared to 42 g of H₂ during the pre-reflood phases). Posttest examinations of bundle structures revealed the presence of only little relocated AgInCd melt in the form of rivulets, mainly in the coolant channels surrounding the control rod simulator and at axial elevations between the third (0.55 m) and first spacer grids (−0.1 m).Results of QUENCH-13 on the onset of absorber rod failure are in agreement with CORA results of nine experiments each containing one or more AgInCd/stainless steel/Zircaloy-4 control rod assemblies. Bundle degradation triggered by early melt formation was, however, more pronounced in the CORA experiments with maximum bundle temperatures of 2300 K (compared to 1800 K in QUENCH-13). Consequently, QUENCH-13 allowed studying the initiation of absorber rod failure by eutectic reactions of SS-Zr, and later on of AgInCd-Zr, as well as the redistribution of the absorber material within the test bundle. Furthermore, input data for modeling of aerosol release during severe accidents are considered as benefits of the experiment.     

Activity level of gross α and gross β in airborne aerosol samples around the Qinshan NPP

The monitoring results of gross α and gross β activity from 2001 to 2005 for environmental airborne aerosol samples around the Qinshan NPP base are presented in this paper. A total of 170 aerosol samples were collected from monitoring sites of Caichenmen village, Qinlian village, Xiajiawan village and Yangliucun village around the Qinshan NPP base. The measured specific activity of gross α and gross β are in the range of 0.02 - 0.38 mBq/m^3 and 0.10 - 1.81 mBq/m^3, respectively, with an average of 0.11 mBq/m^3 and 0.45mBq/m^3, respectively. They are lower than the average of 0.15 mBq/m^3 and 0.52mBq/m^3, of reference site at Hangzhou City. It is indicated that the specific activity of gross α and gross β for environmental aerosol samples around the Qinshan NPP base had not been increased in normal operating conditions of the NPP.     

Experimental evaluation of the water ingression mechanism for corium cooling

Experiments were performed to assess the significance of water ingression cooling in the quenching of molten corium. Water ingression is a mechanism by which water penetrates into cracks and pores of solidified corium to enhance cooling that would otherwise be severely limited by the low thermal conductivity of the material. Quench tests were conducted with 2100 °C melts weighing 75 kg composed of UO₂, ZrO₂ and chemical constituents of concrete. The amount of concrete in the melts was varied between 4% and 23%. The melts were quenched with an overlying water layer; three tests were conducted at a system pressure of 1 bar and four tests at 4 bar. The measured cooling rates were found to decrease with increasing concrete content and, contrary to expectations, are essentially independent of system pressure. For the lower concrete content melts, cooling rates exceeded the conduction-limited rate with the difference being attributed to the water ingression mechanism. Measurements of the permeability of the corium “ingots” produced by the quench tests were used to obtain a second, independent set of dryout heat flux data, which exhibits the same trend as the quench test data. The data was used to validate an existing dryout heat flux model based on corium permeability associated with thermally induced cracking. The model uses the thermal and mechanical properties of the corium and coolant, and it reproduces the very particular data trend found for the dryout heat flux as a function of concrete content. The model predicts that water ingression cooling would be most effective for concrete-free corium mixtures such as in-vessel type melts. For such a melt the model predicts a dryout heat flux of 400 kW/m² at a pressure of 1 bar. The results of this study provide an experimental basis for a water ingression model that can be incorporated into computer codes used to assess accident management strategies.     

Stopping power of polymeric foils for swift heavy ions

The stopping power of polypropylene PP(C₃H₆) and Polyethylene naphthalate PEN (C₇H₅O₂) polymeric foils has been measured, using transmission technique, for Si, Cl and Ti ions covering the energy range 1.0–4.5 MeV/u. These measured stopping power values have been compared with the corresponding values generated from the widely used semi-empirical formulations and standard data tables. The applicability of these formulations and data tables, in the light of the experimental values, is highlighted.     

Neutron irradiation effects on optical absorption of KU1 and KS-4V quartz glasses and Infrasil 301

In this work we have compared the effects of neutron (10²¹–10²² n/m² fluences) and gamma irradiation (23.8 MGy dose) on the IR–vis–UV optical absorption spectra of high purity silica with different OH content: KU1 (800 ppm), KS-4V (<0.2 ppm), and commercial silica Infrasil 301 (<8 ppm). The results show that the UV–vis optical degradation of the silica, after neutron irradiation at the highest fluence is similar for the three grades studied, while gamma-induced optical absorption depends on the material grade (KS-4V shows the lowest optical absorption). The effects of both types of radiation on the IR band related with the hydroxyl group (3650 cm⁻¹) depend on the silica grade. For KU1, the shape of this band changes with neutron fluence. For Infrasil 301 gamma and neutron irradiated, this band height increases, possibly due to free molecular or hydrogen atoms. The shift to lower energies observed for the 2260 cm⁻¹ band in the three neutron irradiated silica grades, reflects the changes induced by neutrons in the lattice bonding angle distribution.     

CFD analysis of thermal–hydraulic behavior of heavy liquid metals in sub-channels

CFD analysis was carried out for thermal–hydraulic behavior of heavy liquid metal flows, especially lead–bismuth eutectic, in sub-channels of both triangular and square lattices. Effect of various parameters, e.g. turbulence models and pitch-to-diameter ratio, on the thermal–hydraulic behavior was investigated. Among the turbulence models selected, only the second order closure turbulence models reproduce the secondary flow. For the entire parameter range studied in this paper, the amplitude of the secondary flow is less than 1% of the mean flow. A strong anisotropic behavior of turbulence is observed. The turbulence behavior is similar in both triangular and square lattices. The average amplitude of the turbulent velocity fluctuation across the gap is about half of the shear velocity. It is only weakly dependent on Reynolds number and pitch-to-diameter ratio. A strong circumferential non-uniformity of heat transfer is observed in tight rod bundles, especially in square lattices. Related to the overall average Nusselt number, CFD codes give similar results for both triangular and square rod bundles. Comparison of the CFD results with bundle test data in mercury indicates that the turbulent Prandtl number for HLM flows in rod bundles is close to 1.0 at high Peclet number conditions, and increases by decreasing Peclet number. Based on the present results, the SSG Reynolds stress model with semi-fine mesh structures is recommended for the application of HLM flows in rod bundle geometries.     

Observation of Radioactive Aerosol Emissions from the Sarcophagus at the Chernobyl Nuclear Power Plant

It is shown that high wind velocity outdoors results in a higher concentration of radioactive aerosols in the ventillation exhaust from the sarcophagus. Aerosol samples from manholes in the roof of the sarcophagus have been collected, for the first time ever, in January–December 2002 on trilayer filter packets. The ¹³⁷Cs concentration in outgoing flows is 0.7–2.3 Bq/m³. The activity median aerodynamic diameter of the aerosol carrier particles is 0.7–1.8 m. At the same time, the ¹³⁷Cs concentration in the atomspheric layer at the ground dear the sarcophagus was 1000 times lower and the carrier-particle sizes 2–4 times larger.     

Heterogeneous cores for improved safety performance: A case study: The supercritical water fast reactor

Light water reactor (LWR) technology is nowadays the most successful commercial application of fission reactors for the production of electricity. However, in the next few years, nuclear industry will have to face new and demanding challenges: the need for sustainable and cheap sources of energy, the need for public acceptance, the need for even higher safety standards, the need to minimize the waste production are only a few examples. It is for these very reasons that a few next generation nuclear reactor concepts were selected for extensive research and development; super critical water reactors are among them. The use of a supercritical coolant would allow for higher thermal efficiencies and a more compact plant design, since steam generators, or steam separators and driers would not be needed, hence achieving a better economy. Moreover, because of the high heat capacity of supercritical water, relatively less coolant would be needed to refrigerate the reactor, therefore the feasibility to design a water cooled fast reactor: the supercritical water fast reactor (SCFR). This system presents unique features combining well-known fast and light water reactor characteristics in one design (e.g. a tendency to a positive void reactivity coefficient together with loss of coolant accident – LOCAs as a design basis accident). The core is in fact loaded with highly enriched MOX fuel (average plutonium content of 23%), and presents a peculiar and significant geometrical and material heterogeneity (use of radial and axial blankets, solid moderator layers, 12 different enrichment zones). The safety analysis of this very complex core layout, together with the optimization of the void reactivity effect through core design, is the main objective of this work.     

Two phase flow 1D turbulence model for poly-disperse upward flow in a vertical pipe

A 1D test-solver was developed in recent years for modeling of two phase bubbly flows in pipe geometry. The solver considers a number of bubble classes and calculates bubble-size resolved void fraction profiles in the radial direction. A successful implementation was achieved regarding bubble forces models (non-drag forces). Discrepancies appeared when coalescence and breakup rates were significant. These rates depend upon local turbulence quantities, which are possible reason for discrepancies. Originally the test-solver is equipped by Sato model (Sato, Y., Sadatomi, M., Sekoguchi, K., 1981. Momentum and heat transfer in two-phase bubble flow. I. International Journal Multiphase Flow 7, 167–177 .) which accounts for turbulence via shear- and bubble-induced viscosities calculated out of empirical correlations. One equation for the turbulent kinetic energy was solved, while the dissipation rate was calculated out of a correlation. In order to improve calculation of the local turbulence parameters, a two-phase k– turbulence model was adopted instead. The account for the bubble-induced turbulence was made via a source term taken out of literature. Comparisons between new and old turbulence modeling against experimental data showed better agreement for the new model. The experiments covered a wide range of water and air superficial velocities for upward bubbly flow in two pipe's diameters: 50 and 200 mm. The main feature of the new model is providing more reliable values of turbulence parameters for application in coalescence and breakup models. A comparison with CFX 5.7 calculations in a 50 mm pipe showed better calculation results when the source term was considered in the k– equations. An implementation into CFX is planned.     

A comparative CFD investigation of helical wire-wrapped 7, 19 and 37 fuel pin bundles and its extendibility to 217 pin bundle

Preliminary investigations of sodium flow and temperature distributions in heat generating fuel pin bundles with helical spacer wires have been carried out. Towards this, the 3D conservation equations of mass, momentum and energy have been solved using a commercial computational fluid dynamics (CFD) code. Turbulence has been accounted through the use of high Reynolds number version of standard k– model, with uniform mesh density respecting wall function requirements. The geometric details of the bundle and the heat flux in are similar to that of the Indian Prototype Fast Breeder Reactor (PFBR) that is currently under construction. The mixing characteristics of the flow among the peripheral and central zones are compared for 7, 19 and 37 fuel pin bundles and the characteristics are extended to a 217 pin bundle. The friction factors of the pin bundles obtained from the present study is seen to agree well with the values derived from experimental correlations. It is found that the normalized outlet velocities in the peripheral and central zones are nearly equal to 1.1–0.9, respectively which is in good agreement with the published hydraulic experimental measurements of 1.1–0.85 for a 91 pin bundle. The axial velocity is the maximum in the peripheral zone where spacer wires are located and minimum in the zones which are diametrically opposite to the respective zone of maximum velocity. The sodium temperature is higher in the zones where the flow area and mass flow rates are less due to the presence of the spacer wires though the axial velocity is higher there. It is the minimum in the peripheral zones where the circumferential flow is larger. Based on the flow and temperature distributions obtained for 19 and 37 pin bundles, a preliminary extrapolation procedure has been established for estimating the temperatures of peripheral and central zones of 217 pin bundle.