Thermal Hydraulic Analysis and Design of Nuclear Research Reactor - Power Uprating PROCEDURESBSEBSU, Farag Muftah Reactor Department, Tajoura Research Center P.O. Box 30878, Tajoura (Tripoli) Libya Fax: +218 21 360-4141, Phone +218 21 360 4142 Email:
[email protected] Abstract This report presents the outline of the core thermal hydraulic design and analysis (Operational Safety Analysis) of Budapest nuclear research reactor (WWR-M2 type), which is a tank type, light water-cooled nuclear research reactor with 36% enriched uranium coaxial annuli fuel. The Budapest nuclear research reactor is currently upgraded to 10 MWth of thermal power, while the cooling capacity of the reactor was designed and constructed for 20 MWth. This reserve in the cooling capacity serves redundancy today but can be used for future upgrading too. The core thermal hydraulic design was, therefore, done for the normal operation conditions so that fuel elements may have enough safety margins both against the onset of nucleate boiling (ONB) not to allow the nucleate boiling anywhere in the reactor core and against the departure from nucleate boiling (DNB). Thermal hydraulic performance was studied, and it is shown that the 36% enriched UAlx-Al fuels in WWR-SM fuel coolant channel 1 does not make possible to force up the reactor power to 20 MWth. The study was carried out for an equilibrium core, with compact load (223 fuel assemblies) under normal operation conditions only (steady state condition). 1. INTRODUCTION In this report we shall present the theoretical outline of the core power uprating thermal hydraulic design and analysis of WWRM2 research reactor, which it is a tank type, light water, cooled reactor with 36% enriched uranium coaxial annuli fuel. The WWR-M2 nuclear research reactor is currently uprated to 10 MWth of thermal power, while the cooling capacity of the reactor was designed and constructed for 20 MWth. This reserve in the cooling capacity serves redundancy today but can be used for future uprating too.[1] The reactor was first put into operation in 1959; its principal functions at that time were to serve as a facility for basic research experiments in the frameworks of research programs of the Academy of Science and industrial development projects. The reactor was first upgraded in 1967, a new type of fuel was introduced and beryllium reflector was applied, that allowed to increase the reactor thermal power from 2 MWth to 5 MWth, and after 27 years of operation a fullscale reconstruction and upgrading project was started. The reconstructed reactor was re-operated in 1992–1993. The design concept of the new reactor (upgrade one) is that it has great, flexibility of utilization and that it provides an adequate neutron flux for isotope 2 production, material testing, and neutron physics measurement. The performance of upgraded reactor has been investigated using the WWR-SM fuel type, with 10 to 20 MWth power level. [2] The core thermal hydraulic design was, therefore, done for the normal operation conditions so that fuel elements may have enough safety margins both against the Onset of Nucleate Boiling (ONB) not to allow the nucleate boiling anywhere in the reactor core and against the Departure from Nucleate Boiling (DNB). Thermal hydraulic performance was studied, and it is shown that the 36% enriched UAl xAl fuels in WWR-SM fuel coolant channel, dose not make possible to force up the reactor thermal power to 20 MWth. The study was carried out for an equilibrium core, with compact load (223 fuel assemblies) under normal operation conditions. 2. WWR-M REACTOR CORE OPERATION EXPERIENCE AND DESCRIPTION 2.1 WWR-M REACTOR CORE OPERATION EXPERIENCE The pool-type WWR-M reactors serve a wide range of scientific research and engineering purpose requiring a high neutron flux. The first reactor type was put in the operation on December 1959, in Leningrad, Russia as described in the International Atomic Energy Agency (IAEA) documents, (IAEA – Research Reactor – WWR-M Leningrad – Aug. 1960). The WWR-M reactor has various fuel coolant channels. The fuel assemblies for this reactor were first 3 produced in the late 1950; is initially fuel elements were composed of Al+UO2, cermets (1959-1963) but from beginning of 1963 the form is an Al+U alloy (WWR-M2). Their specific heat transfer surface was almost four times as large as the EK-10 rod elements used earlier in WWR-S pool reactors. In consequence, the designed power has been raised from 2 to 10 MWth. From 1967 together with WWR-M2 fuel elements, new fuel elements WWR-SM were in production with increased of active length (50 cm to 60 cm), which are still in use in reconstructed foreign research reactors. Many years of experience with these and other similar fuel elements in pool research reactors are availble (WWR-M in Gatchina and Kiev; WWR-Ts in Obninsk; WWR-K in Alma-Ata; Eva in Poland and WWR-SM in Germany and Hungary). The elaboration of new fuel elements type WWR-M was done in two stages. In the first stage, the optimization of the geometrical parameters of fuel elements was done leading to an increase of specific heat transfer surface by factor of 1.8. Some sets of these type assemblies (WWR-M3) were in service at the WWR-M reactor of the Petersburg Nuclear Institute of Physics (PNPI) “BP Konstantinov” in Gatchina between 1973 and 1980. In the second stage, the design concentrations of 235 U in the core was optimized (at 125 g/l), leading to a large increase in the power production of each assembly (WWRM5) and to raise the spare reactivity of the reactor. Table 1 shows the characteristic of the WWR-M fuel assemblies, and Table 2 shows the performance of WWR-M fuel assemblies in the WWR-M reactor of 4 PNPI. The base of the reactor core is a hexagonal grid plate with 397 identically formed holes. Fuel assemblies and beryllium displacers can be put into these holes, guide tubes of the control rods as well. The lattice pitch of 35 mm; the core positions are occupied by the fuel assemblies, control rods, beryllium displacers and isotope production channels. Stationary beryllium reflector of 20-cm average thickness surrounds the core. The cooling water is flowing down stream across the reactor core. The fuel of the reactor is of the WWR-M type (Russian product). It is an alloy of aluminum and uranium-aluminum eutectic with aluminum cladding. The uranium enrichment is 21%, 36%, and 80 %. The first fuel assembly contains two fuel tubes, with the outer tube (un fueled one) is of hexagonal shape, while the two inner ones are cylindrical. The reactor core horizontal cross sections is [10-13] shown in Figure 1. At the end of 1950 years the first Russian heat exchange assemblies TBC were developed, built from seamless tubing fuel elements WWR-M1, which were homogeneously filling up the active zone without occupying its volume for elements, only for constructional designations. The WWR-SM fuel coolant channel is three coaxial annuli (fuel elements) and the outer fuel element is a hexagonal shape with pitching is a 35 mm as shown in Figure 2. [310] 2.2 WWR-M REACTOR CORE DESCRIPTION The WWR-M is a cylindrical tank type reactor. The reactor core is placed 5.145 m below the surface of the reactor tank (in order to minimize the radiative exposure to the personel), which is open to 5 atmospheric pressure. The diameter of the tank is 2300 mm, and its height is 5685 mm. The heavy concrete reactor-shielding block is situated in a rectangular semi-hermetically sealed reactor hall. The base of the reactor core is a hexagonal grid plate as shown in Figure 1, with 397 identically formed holes. The fuel assemblies and the beryllium displacers can be put into these holes, as well as the guide tubes of the 18-absorber rods. The equilibrium core size (in this study) consists of 223 fuel assemblies, and the control rods, beryllium displacers and isotopes production channels, occupy the remaining core positions. A fixed beryllium reflector of 20-cm average thickness surrounds the core. The fuel assembly type is WWR-SM as shown in Figure 2 (consists of 3 coaxial fuel elements). The innermost is a tube, this is followed by a second fuel element with an annulus crosssection, and the third fuel element (outer) is a hexagonal shape. [1-10] 3. THMOD2 COMPUTER CODE DESCRIPTION The THMOD2 (Thermal Hydraulic Modeling version 2) [3,11-13] code is a one-dimensional computer program (axial direction) and it provides a capability for analysis of the steady state thermal hydraulic analysis of research reactors in which coaxial annuli and/or plates type fuel elements are adopted. In this code, subroutines to calculate temperature distribution in fuel elements. The THMOD2 code can calculate fuel temperatures under forced convection cooling mode with downflow direction. A heat transfer package is used for calculating heat transfer coefficient, DNB heat flux etc. 6 The heat transfer package was especially developed for research reactors, which operated under low pressure and low temperature conditions using coaxial annuli and/or plate-type fuel elements, just like the WWR-M2 reactor. 4. THERMAL HYDRAULIC CALCULATIONS Thermal hydraulic studies for the steady state conditions were made using the THMOD2 code with fuel coolant channel type WWRSM under low temperature and low pressure coolant conditions (in this study we consider only the WWR-M2 reactor operatinal safety analysis, i.e. the possibility of reactor power uprating using the same and/or exicting fuel assembly at normal operation of the reactor, and boiling occures or not. The cooling water flows downward through the reactor core, with inlet coolant temperature of 25-50 ºC, while the temperature difference between the core inlet and outlet is a round 5 º C with a volume flow rate calculated according to the following equation: P = V ρ c p ∆T (1) Where V is the reactor core total coolant volume flow rate, [m3/hr], P is the reactor core thermal power, [kW], cp is the average specific heat of coolant = 4.19 [kj/kg.ºC], ρ is the average reactor core coolant density = 988 [kg/m3], and ∆T is the temperature difference between reactor core outlet and inlet = Tout - Tin [ºC]. 7 The dependence Tsat (z) has been calculated by the well-known dependence of saturation temperature by the pressure depending on coordinate Tsat [P (z)] where: [7,10] P(z) = Po + ρ g ( Wd + z) ρ V2 z (ξ en + f + 1) 2 2b (2) where g, is the gravity acceleration, Po is the atmospheric pressure, Wd is the reactor’s pool depth, ξ en is the channel entrance friction coefficient, ƒ is the friction factor, b is the water spacing between fuel plates, V is the reactor core coolant velocity, and z is the channel axial distance. The reactor core inlet pressure is 1.512 [bar], and reactor core parameters are shown in Table 3, and coolant velocity is calculated by THMOD2 code for each volume flow rate and the reactor core configuration. The X59 [3, 14], and Dittus-Boelter’s [15] correlations were used for the calculation of the convection heat transfer coefficient. The H95 [3], and Bergles–Rohsenow’s [16] correlations for the Onset of Nucleate Boiling (ONB) temperature, and the X2000 [3] and existing international [17-21] correlations for DNB heat flux calculation. Boiling temperature and saturation temperature (i.e. the complete reactor core heat transfer package modeling) is described in THMOD2 code operation manual. [3, 13] 8 5. PROCEDURE OF THE REACTOR CORE UPRATING The THMOD2 code considers equal pressure drop for all channels of the reactor core, and calculates the velocity distribution for fuel coolant channels, using the dimensions of fuel elements as given in Table 4 for performing the upgrading calculations. The calculations were preformed with the assumption that the three main primary pumps are operating at full load with a total flow rates as a function of the reactor core power according to Equation (1). Starting at 10 MWth the reactor core power level was gradually increased in steps of 1 MWth up to 20 MWth power level, and according to the maximum operating limits of the WWR-M2 research reactor for a fuel centerline temperature ≤ 150 ºC and the maximum cladding surface temperature ≤ 104 ºC. Using the old and new fuel element dimensions as shown in Table 4 as sample problems of THMOD2 code, we shall select the optimal fuel element dimensions suitable for WWR-M2 research reactor power uprating and also according to the reactor core design operating limits for fuel centerline temperature and fuel cladding surface temperature. Fuel elements should have enough safety margins both against the onset of nucleate boiling (ONB) not to allow the nucleate boiling anywhere in the reactor core and against the departure from nucleate boiling (DNB). [3, 22, 23] Table 5 shows the fuel centerline temperatures, fuel cladding surface temperatures, saturation temperatures, ONB temperatures, and boiling temperatures as a function of fuel coolant channel type as an 9 example for illustration, and from this table we shall consider only three types of fuel coolant channels for the reactor power uprating thermal hydraulic analysis. The maximum fuel centerline temperature and fuel cladding surface temperature as a function of reactor core power level and reactor coolant inlet temperature for three types of fuel coolant channels are shown in Table 6. The consequence of these results in Table 4 we will consider the WWR-SM1 fuel coolant channel for WWR-M2 research reactor core power uprating thermal hydraulic analysis. [3] The fuel cladding surface temperature, saturation temperature and ONB temperature as a function of reactor power level and reactor coolant inlet temperature and according to reactor core design conditions and operating limits are shown in Figure 3 and from this figure we shall select the maximum reactor operating power level and reactor coolant inlet temperature and other operating parameters as (P = 14 MWth, Tin = 40 ºC, TONB = 109 ºC and, and Tsat =104 ºC). 6. WWR-M2 REACTOR UPRATING THERMAL HYDRAULIC STUDIES In this section, we are planning to remodel the existing nuclear research reactor core of WWR-M2 at 10 MWth with 36 % enrichment uranium (Russian standard) fuel to investigate the thermal hydraulics and reactor core performance. 10 The temperature is shown as a function of coolant velocity because the coolant velocity is the only dominant variable to the fuel surface temperature. Both the ONB temperature and the saturation temperature become lower with an increase of coolant velocity because an increase in coolant velocity gives lower local pressure according to the increase of pressure loss. Figure 4 shows the calculated results of the fuel surface temperature, ONB temperature and saturation temperature where the difference between the ONB temperature and the fuel surface temperature is a minimum for sub-channel C, and it is equal to 5 ºC. The pressure at top and the bottom of WWR-SM1 fuel coolant channel are shown in Figure 5 with the coolant velocity as a parameter, to show the characteristics of pressure decrease due to the increase of coolant velocity. The increase of coolant velocity and decrease of pressure give lower temperature (TONB-Tsat). But in this case, the effects of an increase of coolant velocity and decrease of pressure on the increase of temperature difference (TONB-Tsat), due to the increase of coolant velocity is little in magnitude and only Tsat becomes lower according to the pressure decrease due to the increase of coolant velocity. Therefore, both of Tsat and TONB become lower with the increase of coolant velocity. On other hand, the fuel surface temperature becomes lower with an increase of coolant velocity. It should be noticed in Figure 4 that the TONB is higher the fuel surface temperature at the coolant velocity of 4.5 – 9 m/sec. In this range of coolant velocity, no boiling occurs in the sub-channel and 11 on the other hand, two-phase flow occurs with nucleate boiling at the velocity less than 4 m/sec. Therefore, 4.75 m/sec should be adopted as design velocity for the WWR-M2 reactor core fuel coolant channel and with total volume flow rate of 2359 m3/hr. At the design velocity of 4.75 m/sec thus determined, the pressure drop between the core inlet and the bottom of WWR-SM1 fuel coolant channel is about 0.2323 bar as shown in Figure 5. The distribution of fuel centerline temperature, fuel cladding surface temperature and coolant temperature along the WWR-SM1 fuel coolant channel with the operating coolant velocity are shown in Figure 6. We tried to formulate new heat transfer coefficient correlation X59, and new critical heat flux correalation also using existing international experimental data and my correlations consider as better more limiting operation domain [3,14] as former correlations, therefore, the relationship of Nu vs. Re and heat transfer coefficient (The new one X59 and some of international correlations) applied for forced-convection single-phase flow in down flow direction, for WWR-SM1 fuel coolant channel with D = 4.71 – 5.47 mm with active length = 60 cm, is illustrated in Figure 7 with reactor core power = 14 MWth, Tin = 40 ºC. Figure 8 illustration the various DNB heat flux correlations (The new one X2000 and some of international correlations) described in the heat transfer package of the THMOD2 code. As for the core exist temperature of coolant, one should be careful of the following problem. If the coolant temperature is considerably high at 12 exist of the core, there is possibility that the coolant temperature should become the saturation temperature resulting in the two-phase flow at the location where the local pressure is the lowest in the primary cooling line. This situation should be avoided for a stable steady state operation condition. Figure 9 shows the calculation results of the average coolant temperature at exist of the fuel coolant channel and the saturation temperature where the local pressure is the lowest, as the function of coolant velocity in the fuel coolant channel. The results are shown for the core power of 14 MWth. In the condition of normal operation with the coolant velocity of 4.75 m/sec designed for the WWR-SM1 fuel coolant channel, the lowest pressure is about 1.167 bar with the saturation temperature of 95.76 ºC and the average bulk temperature of coolant at exist of WWR-SM1 fuel coolant channel is about 55.55 ºC as shown in Figure 9 and its consequently no boiling occurs in the primary cooling piping system. The maximum allowable fuel element cladding surface temperature is about 104 ºC as shown in Figure 10. The statistical comparison between the experimental data, X59 correlation and Dittus-Boelter correlation for calculation of Nu number as given in Table 7. Table 8 gives a statistical comparison summary between X2000 correlation and some of international DNB correlations, and also, Figure 11 shows the comparison between the X2000 correlation and some experimental data. [3, 24, 25] 13 Core thermal hydraulic characteristics [3] thus designed and analyzed for the forced-convection cooling mode at the reactor core power level of 10 and 14 MWth are summarized in Table 9. 7. CALCULATION RESULTS On the bases of the results obtained using THMOD2 code thermal hydraulic calculation for WWR-M2 Nuclear Research Reactor core power uprating we can conclude that theoretically it is possible to increase the reactor core thermal power level up to 14 MWth safely and without any operational problems of the reactor using the existing WWR-SM1 fuel coolant channels (3 coaxial fuel elements). Acknowledgment The authors would like to express their appreciation to the Prof. Dr. L. Rádonyi, head of Department for Energy, Budapest University of Technology and Economics, Hungary for his continuous encouragement, valuable suggestions and supporting this work. Also, their thanks are forwarded to Prof. Dr. Tamás Jászay and Prof. Dr. Tamás Kornyi for thier suggetions and discussions. References 1. KFKI, (Central Research Institute for Physics): The Budapest Research Reactor Safety Report Analysis, Budapest, Hungary, (1994). 2. Hargitai, T.: Refueling strategy at the Budapest research Reactor, 2nd International Topical Meeting on Research Reactor 14 fuel management, Organized by the European Nuclear Society (ENS), Binges, Belgium, Nov. 29-31 (1998). 3. BSEBSU, F. M.: Thermal Hydraulic Analysis of Water-Cooled Nuclear Research Reactors, PhD. Thesis, Budapest University of Technology and Economics, Budapest-Hungary, (2001 Oct. 17). 4. Eryekalov, A. N., et. al.: Thin-Walled Fuel Elements WWR-M5 for Research Reactors, Atomic Energy. Vol. 60/2, (1986), pp. 103 - 106, (in Russian), 5. Verkhovyekh, P. M., et. al.: Remarks to the Reconstruction of Active Zone in the Nuclear Reactor Type WWR-M, Atomic Energy, Vol. 41/3, (1976), pp. 201-203, (in Russian). 6. Enin, A. A., et al.: Design and experience of HEU and LEU fuel for WWR-M reactors, Nucl. Eng. and Des., Vol. 182, (1998), pp. 233-240. 7. Eryekalov, A. N., et. al.: Reduction of the Enrichment in Fuel Elements for WWR-M Reactors, IAEA-SM-310/113P, (1984), pp. 710-726. 8. IAEA – Research Reactors Documents, WWR-M - Leningrad, (1962), pp. 165-169. 9. Zakharov, A. S., et. al.: Control Rods of A WWR-M Reactor Fitted with Finned External Fuel Elements, Atomic Energy, (1993), 74/1, pp. 88 - 90. 15 10. Eryekalov, A. N. and Petrov, Yu. V.: Parameters Characterizing Reactor for Physical Experiments, Atomic Energy, (1968), 25/1, pp. 82 - 84. 11. BSEBSU, F. M., and BEDE, G.: A Simple Computer Program for the Calculations of Reactor Channel Temperature Distribution, Periodic Polytechnica Series Mech. Eng., Budapest University of Technology and Economics, Budapest, Hungary, Vol. 41/2, (1997), pp. 133. 12. BSEBSU, F. M., and BEDE, G.: Nuclear Reactor Channel Modelling Using THMOD2 code, KERNTECHNIK, Vol. 64/5-6, (1999), pp. 269-273. 13. BSEBSU, F. M.: THMOD2 Code Operation Manual, Internal Report, Department for Energy, Budapest University of Technology and Economics, Budapest, Hungary. (1998). 14. BSEBSU, F. M., and BEDE G.: Theoretical study in SinglePhase Forced-Convection Heat Transfer Characteristics for Narrow Annuli Fuel Coolant Channels, Periodic Polytechnica Series Mech. Eng., Budapest University of Technology and Economics, Budapest-Hungary. Under Press, (2002). 15. Todreas N. E., and Kasimi, M. S.: Thermal Hydraulic Fundamental, Vol. I, & II Hemisphere Publishing Corporation, NY, USA. (1990). 16. Fenech H., and Rohsenow W. M.: Heat Transfer, Ch.16, pp. 335-418, The Technology of Nuclear Reactor Safety, Vol. 2. 16 Reactor Material and Engineering, edited by T. J. Thompson and J. G. Beckerly, MIT Press, (1973). 17. Sudo Y., and Kaminaga M.: A New CHF Correlation Scheme Proposed for Vertical Rectangular Channels Heated from Both Sides in Nuclear Research Reactors, Transactions of the ASME, Vol. 115, (May 1993), pp. 426-434. 18. Tong L. S.: Prediction of Departure from Nucleate Boiling for an Axially Non-Uniform Heat Flux Distribution, Journal of Nuclear Energy, Vol. 21, (1967), pp.241-248. 19. Mishima K., and Nishihara H.: The Effect of Flow Direction and Magnitude on CHF for Low Pressure Water in Thin Rectangular Channels, Nucl. Eng. and Des., Vol. 86, (1985), pp. 165-181. 20. Mishima K., Nishihara H., and Shibata T.: CHF Correlations Related to the Core Cooling of a Research Reactor, JAERI– M84–073, (1983), pp. 312 – 320. 21. Koweri Y., et. al.: Experimental Study on DNB Heat Flux Correlations for JMTR Safety Analysis, Int. MTG. on Reduced Enrichment for Research Reactor and Test Reactor, New Port RI, 23 – 27 Sept. (1990). 22. BSEBSU, F. M., et al.: Tajoura Reactor Power Uprating– Thermal Hydraulic Analysis, International Multidisciplinary Conference on Environmental and Economical Development in Libya and Hungary, Godollo, Hungary, April 27-28, (1998). 17 23. Sudo Y., Ando H., Ikawa H., and Ohnishi N.: Core Thermohydraulic Design with 20% LEU Fuel for Upgraded Research Reactor JRR-3, Journal of Nucl. Sci. and Tech., 22/7, (July 1985), pp. 551-564. 24. Teyssedou A., et al.: Critical Heat Flux Data in a Vertical Tube at Low and Medium Pressures, Nucl. Eng. and Des., Vol. 149, (1994), pp. 185-194. 25. Groeneveld D. C., et. al.: The 1995 look-up Table for Critical Heat Flux in Tubes, Nucl. Eng. and Des., Vol. 163, (1996), pp. 1-23. 18 Figure 1. WWR-M2 research reactor core horizontal cross-section. 19 Figure 2. WWR-SM Fuel Coolant Channel 20 Figure 3. The maximum cladding surface temperature, saturation temperature, and ONB temperature as a function of reactor core power level and reactor coolant inlet temperature for fuel elements of WWR-SM fuel coolant channel dimensions. 21 Figure 4. Maximum cladding surface temperature, saturation temperature, and ONB temperature as a function of reactor coolant velocity of sub-channel C. 22 Figure 5. The pressure at reactor top and bottom as a function of reactor coolant velocity of sub-channel C. 23 Figure 6. The axial distribution of fuel centerline temperature, fuel surface temperature, and coolant temperature along the coolant subchannel D of WWR-SM1 fuel coolant channel. 24 Figure 7. Illustration of heat transfer correlation applied for forcedconvection single-phase flow for down flow. 25 1200 1100 1000 900 800 700 QDNB[W/cm ] 2 WWR-M Sub-channel D Tin= 40 C Pin= 1.512 bar De= 6 mm Power= 14 MWth o 600 500 400 300 200 100 0 1 2 3 4 5 6 Labuntsov Mirshak Bernath Biasi Bsebsu, X2000 Tong, W3 7 8 9 10 11 Coolant Velocity [m/sec] 26 Figure 8. Illustration of DNB critical heat flux correlation used for sub-channel D of WWR-SM1 fuel coolant channel. 27 Figure 9. Calculated results of average core exit coolant temperature and saturation temperature at lowest pressure in primary coolaing line vs. core coolant velocity. 28 Figure 10. Calculated results of maximum cladding surfaces of the fuel element 3 of WWR-SM1 fuel coolant channel vs. core coolant velocity. 29 Figure 11. Comparison of X2000 correlation, CHF data with predictions of correlations and look-up CHF table (L=1.4 m, P = 4.9 bar) 30 Table 1. Characteristics of WWR-M assemblies. Fuel element Wall (Meat) Thickness [mm] 2.3(0.9) 2.5(0.7) 1.25(0.53) 1.25(0.39) 2.5(0.9) 1.25(0.43) 1.25(0.43) Specific heat Transfer Surface [cm /cm ] 3.67 3.67 6.6 6.6 3.67 6.6 6.6 2 3 235 Assembly Type WWR-M1 WWR-M2 WWR-SM WWR-M5 WWRM2E WWRM5E U Uraniu Compositio n UO2+Al U+Al U+Al UO2+Al UO2+Al UO2+Al UO2+Al m Density [g/cm ] 1.5 1.33 0.77 1.2 2 2 3 3 235 U [% ] 20 36 90 90 36 36 21 Conc. in core [g/l] 50 61.2 125 125 122 102 83 31 Table 2. Performance of WWR-M assemblies on the WWR-M reactor of PNPI. Characteristic Operating period Reactor power [MW] Mean (max.) burnup in unloaded assemblies [%] Number of used single assemblies Mean power production per assembly [MWday/Ass.] Total power production [GW-day] WWR-M1 WWR-M2 WWR-M3 WWR-M5 1959-63 10 47(76) 184 9.7 1.8 1963-79 16 41(91) 2765 10 28 1973-80 18 28(73) 638 7.7 5 1980-97 18 29(59) 2235 14.7 32.8 Table 3. Core Design Description Parameters Reactor type Power level, MW Vertical positions Fuel positions Irradiation position Beryllium displacers Horizontal beam Tank type 10 397 223 51 123 10 32 Radial Tangential Fuel Type Meat Material Clad Material Active Length, mm Lattice Pitch, mm Moderator, coolant Reflector Control Rod Absorber Safety Rod Automatic Rod Manual Rod Coolant inlet Temperature. ΟC Coolant inlet Pressure, bar 8 2 WWR-SM UAlx-Al Al (SAV-I) 600 35 H2O Beryllium B4C (18) 3 1 14 35 1.52 Table 4. WWR-SM fuel coolant channels, fuel meat and clad dimensions [mm]. CHANNEL TYPE WWRSM0 WWRSM1 WWRSM2 WWRSM3 Fuel Element I CTH FTH CTH 0.90 1.09 2 1.02 6 0.90 0.70 1.09 8 0.84 7 0.70 0.90 0.95 6 1.02 6 0.90 Fuel Element II CTH FTH CTH 0.90 1.09 2 1.02 6 0.90 0.70 1.09 8 0.84 7 0.70 0.90 0.95 6 1.02 6 0.90 Fuel Element III CTH FTH CTH 0.94 1.09 2 1.02 6 0.90 0.74 1.09 8 0.84 7 0.70 094 0.956 1.026 0.90 33 WWRSM4 WWR-M51 WWR-M52 WWR-M53 0.80 0.36 0.43 0.41 0.90 0.53 0.39 0.43 0.80 0.36 0.43 0.41 0.80 0.36 0.43 0.41 0.90 0.53 0.39 0.43 0.80 0.36 0.43 0.41 0.80 0.36 0.43 0.41 0.90 0.53 0.39 0.43 0.80 0.36 0.43 0.41 CTH = Clad Thickness, and FTH= Fuel meat Thickness Table 5. The comparison between the centerline temperatures, fuel cladding surface temperatures, saturation temperatures, ONB temperatures, and boiling temperatures at P = 10 MWth, and Tin = 50 º C for the fuel coolant channels. channel type P, [MWth] 10 10 TF, [ºC] 155.0 3 139.5 9 TCl, [ºC] 109.2 2 102.3 0 Tsat, [ºC] 109.04 106.60 TONB, [ºC] 111.39 109.01 TBLG, [ºC] 138.80 138.58 WWR-SM0 WWR-SM1 34 WWR-SM2 WWR-SM3 WWR-SM4 WWR-M51 WWR-M52 WWR-M53 10 10 10 10 10 10 144.5 0 153.6 2 154.8 4 184.0 8 190.8 5 191.1 3 103.9 1 108.2 4 109.4 6 123.0 3 125.6 1 125.8 8 107.94 109.20 109.20 110.83 111.10 111.10 110.31 111.52 111.52 113.17 113.40 113.40 136.79 139.04 139.04 142.01 142.54 142.54 Table 6. The fuel centerline temperature and fuel cladding surface temperature as a function of reactor core power level, coolant inlet temperature and fuel coolant channel type. Fuel channel Type Tin =35 º P MWth 10 13 15 Tin =40 º Tin =50 º Tin =35 º Tin =40 º Tin =50 C C C Fuel Centerline Temperature [ºC] º C C C Clad Surface Temperature [ºC] WWR- 152.74 160.27 164.72 153.20 160.47 164.77 155.03 161.17 165.89 100.65 105.29 108.11 103.42 107.89 110.61 109.22 113.63 116.21 35 SM0 18 20 10 13 WWRSM1 15 18 20 10 13 WWRSM2 15 18 20 170.78 174.51 135.10 142.29 146.61 170.78 174.51 139.28 148.01 152.21 157.99 161.57 170.64 174.26 136.34 143.31 147.51 170.64 174.26 141.84 148.69 152.76 158.36 161.83 171.41 174.82 139.59 146.17 150.16 171.41 174.82 144.50 150.93 154.77 160.06 163.34 112.04 114.50 92.63 97.49 100.49 112.04 114.50 93.97 99.24 102.00 105.89 108.34 114.44 116.85 95.73 100.47 103.41 114.44 116.85 97.63 102.05 104.75 108.54 110.93 119.81 122.08 102.30 106.84 109.66 119.81 122.08 103.91 108.11 110.68 114.31 116.61 Table 7. The statistical comparison between the experimental data, X59 correlation and Dittus-Boelter correlation for calculation of Nusselt Number. Correlation X59 Mean 114.37 Standard Deviation (±) 46.03 S. Error (±) 7.1 Data No. 42 36 DittusBoelter Experimental 128.5 2 121.6 2 55.43 55.76 8.6 8.6 42 42 Table 8. Statistical comparison summary between X2000 correlation and some of international DNB correlations. QDNB Mean, Standard Deviation S. Error No. 37 [W/cm2] Labuntso v Mirshak Biasi X2000 227.69 315.38 224.05 260.64 (±) (±) 18.75 31.68 59.15 40.56 1.50 2.54 4.74 3.25 156 156 156 156 Table 9. Summary of core thermal hydraulic analysis and design for WWR-M2 research reactor core Parameter Primary system total volume flow rate, [m /hr] Flow ratio in active core region, [%] Coolant velocity in WWR-SM1 sub-channels, [m/sec] Core inlet coolant temperature, [oC] Average temperature through primary circuit system, [oC] Core inlet pressure, [bar] Pressure loss through active reactor core, [bar] Minimum temperature margin to ONB, [oC] Minimum DNB ratio, [--] Maximum cladding surface temperature (upper limit), [oC] 3 10 MWth 1750 78 3 50 5 1.512 0.173 7 2.41 104 14 MWth 2359 78 4.75 40 5 1.512 0.232 5 1.86 104 38 Core exit coolant temperature, [oC] Onset Nucleate Boiling temperature, TONB, [oC] Saturation temperature, Tsat, [oC] q′′ , [W/cm2] ONB 64 111 108.5 108 55.55 109 104 108.8 39