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Nuclear Science and Technology, Vol.8, No. 1 (2018), pp. 10-16 Thermal-hydraulics analysis for VVR-KN fuel lead test using PLTEMP code Ho Nguyen Thanh Vinh, Le Vinh Vinh, Nguyen Nhi Dien, Nguyen Kien Cuong, Huynh Ton Nghiem, Tran Quoc Duong and Nguyen Tien Vu Nuclear Research Institute, 01 Nguyen Tu Luc Street, Dalat, Vietnam E-mail: honguyenthanhvinh@gmail.com (Received 15 April 2018, accepted 18 September 2018) Abstract: VVR-KN is one of the low-enriched fuel types to be considered for a new research reactor (RR) of a Centre for Nuclear Energy Science and Technology (CNEST) of Vietnam. This fuel type was qualified by a lead test carried out with three fuel assemblies (FAs) in 6-MWt WWR-K research reactor at the Institute of Nuclear Physics, Kazakhstan. VVR-KN fuel was then used for conversion of the WWR-K reactor core from highly-enriched to low-enriched uranium fuel and the reactor was successfully commissioned in September 2016. PLTEMP is a thermal-hydraulic code with plate and coaxial tube models that seems to be suitable for VVR-KN fuel type. Before using PLTEMP code for thermal-hydraulics analysis of the new RR, a calculation for code validation was performed based on the data of the VVR-KN fuel lead test. First, MCNP5 code was used to calculate the power distribution of WWR-K reactor core with lead test fuel assemblies (LTAs) at the core center. Then, thermal-hydraulics parameters of the LTAs were obtained by using PLTEMP code together with calculated data of the power distribution and the lead test conditions. A comparison between the analytic results and the lead test data was made to confirm the suitability of PLTEMP code for thermal-hydraulics analysis of VVR-KN fuel under forced convection and downward flow conditions. Keywords:VVR-KN fuel type, MCNP5, PLTEMP, WWR-K reactor, VVR-KN fuel lead test. I. INTRODUCTION The CNEST project with a 10-MWt upgradable to 15-MWt research reactor (RR) has been in the pre-feasibility study phase. Meanwhile, the national research project for calculation of neutronic characteristics, thermal-hydraulics and safety analysis of the proposed RR loading with low enriched Russian fuel types has been deployed. One of selected fuel types intended for this RR is VVR-KN, which is composed of coaxial tubes. Additionally, PLTEMP code [1] with plate and tube models has been used at the Argonne National Laboratory for RRs using Russian coaxial tube fuel type. This code seems to be suitable for thermal-hydraulics calculation of the new RR at steady-state condition. The PLTEMP code validation with the coaxial fuel type and the natural convection mode was achieved by comparing with the experimental data in the Dalat research reactor [5]. On the other hand, in forced convection mode, the PLTEMP code was validated based on the data of the VVR-KN fuel lead test, which was performed in the 6-MWt WWR-K reactor of Kazakhstan [4]. Successful validation of the PLTEMP code was carried out in both forced and natural convection modes that confirms the suitability for thermalhydraulics analysis of the new RR not only in normal operation but also in cases the decay heat to be removed by natural convection of reactor pool water. This work includes neutronic and thermal-hydraulics calculations. A spatial ©2018 Vietnam Atomic Energy Society and Vietnam Atomic Energy Institute HO NGUYEN THANH VINH et al. thermal power distribution of WWR-K reactor core with experimental VVR-KN FAs is obtained by using MCNP5 code [2]. Thereafter, using PLTEMP code together with the results from the previous neutron calculations and the initial conditions of the fuel lead test to obtain thermal-hydraulics parameters. Eventually, comparison of the calculated results with the experimental data confirms the suitability of the PLTEMP code for VVR-KN fuel under the experimental conditions. II. METHOD AND CALCULATION RESULTS A. VVR-KN fuel lead test VVR-KN fuel assembly (FA) (see Fig. 1) is a low-enriched fuel type that was used for the conversion of the WWR-K reactor core from highly-enriched uranium fuel of 36% (HEU) to low-enriched fuel of 19.75%of 235U (LEU) [3]. Also, this fuel is being considered for the new RR of the CNEST in Vietnam. Table I. reveals main characteristics of the VVR-KN FAs [3]. Table I. Characteristics of VVR-KN FAs. Fuel composition UO2 + Al Number of fuel elements per FA (standard FA/ FA with control rod inside) 8 tubes / 5 tubes Fuel cladding material Uranium enrichment, % Content of 235U per FA, g Al (SAV-1) 19.75 245 / 196 Content of 235U per a core volume unit, g/l Surface of heat removal per a core volume unit, cm2/cm3 Fuel element thickness, mm Thickness of gaps between fuel elements, mm Total weight of FA, kg Fuel meat thickness, mm Fuel meat length, mm Fuel meat uranium density, g/cm3 Volume fraction of water in FA Infinite-medium neutron multiplication factor, k∞ 104.4 / 83.5 5.46 / 4.33 1.6 2 4.66 / 4.21 0.7 600 3.0 0.54 1.648 There are two types of VVR-KN FAs, including the standard one with 8 fuel elements (FEs) and the other with 5 fuel elements for control rod placement inside FA-2 (Fig. 1). For the fuel lead test, the LEU VVR-KN FA’s width across flat is 66.3 mm, which makes it possible to be installed into the existing support plate using a triangular lattice with a spacing of 68.3 mm of the HEU VVR-C FA of WWR-K reactor. An increase in the FA robustness in case of beyond design basis accidents was achieved through the longitudinal spacer ribs introduced on the FE Fig. 1. Two types of VVR-KN FAs. 11 THERMAL-HYDRAULICS ANALYSIS FOR VVR-KN FUEL LEAD TEST … outside which ensure the guaranteed clearance between the FEs. reactor can reach at the permitted thermal power of up to 6 MWt. Its core consisted of 38 HEU VVR-C FAs, including 32 FAs with 5 FEs (FA-1) and 6 FAs with 3 FEs (FA-2) shown in Fig. 2. To ensure the required reactivity margin during the test, hexagonal beryllium blocks were installed into the core’s peripheral cells. A program was developed for the lead test fuel of the VVR-KN FAs in the WWR-K reactor. An experimental device of three experimental VVR-KN FAs (EFAs) with beryllium blocks around was placed in the reactor core center (Fig. 2). The WWR-K Fig. 2. WWR-K reactor core mapwith HEU VVR-C FAs and three LEU EFAs [4]. The purpose of the tests was to achieve a burn-up of 60% in which defined the life of the reactor FEs. For this purpose, three 8-element EFAs were manufactured with the following 235 U mass: 245.3 g in LTA-1, 244.7 g in LTA2, and 245.0 g in LTA-3. of 816 days, including 480 days at the reactor power rate of 6 MWt. The average 235U burnup of 59.7% (LTA-2) and 60.3% (LTA-3) was achieved. The safe operating mode for VVR-KN EFAs (no fuel cladding surface boiling) was maintained during the experiment. The parameters which were monitored during the fuel lead test include the coolant temperatures at the core inlet and outlet, the coolant temperature at the outlet of the experimental device with three VVR-KN EFAs inside, the coolant flow rate through the reactor core, the coolant level in the reactor pool, the coolant pressure at the EFAs outlet, the coolant and air activity beneath the reactor head. Table II presents thermal-hydraulics analysis results from the fuel lead test [4]. Table II. Thermal-hydraulics analysis results for VVR-KN EFAs. The VVR-KN fuel lead test was successfully performed with the total duration 12 Reactor power, MWt 5.65 Reactor pool water level, m 5.3 Total power of three EFAs, kWt 989 Power of most heat-rated EFA, kWt 333 HO NGUYEN THANH VINH et al. Core inlet coolant temperature, 0C 32 Maximum fuel cladding temperature, 0C 88 2 Maximum heat flux, kW/m Accordingly, thermal power of the three EFAs were of 325 kWt, 322 kWt, and 337 kWt respectively (Fig. 3). 508 2 Hot-spot heat flux, kW/m 374 0 Hot-spot saturation temperature, C 107.4 0 Surface boiling onset temperature, C 114.7 Surface boiling onset temperature margin 1.43 Flux instability safety margin 2.1 B. Calculation of thermal power distribution MCNP5 code was used to calculate thermal power of three EFAs placed in the WWR-K RR, as well as their spatial thermal power distribution in the reactor core. Fig. 3. Thermal power of the three EFAs. The axial thermal power distribution was obtained as well and shown in Fig. 4. As it is evident from the Fig. 4 that the power peaking factor of EFAs were of 1.287, 1.286, and 1.290 each one. MCNP5 is a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron, or coupled neutron/photon/ electron transport, including the capability to calculate eigenvalues for critical systems [2]. The calculation of the thermalhydraulics parameters, as well as safe margins, required attention to thermal power at hottest positions. So that, the hot channel factors of each fuel element of EFAs were calculated by MCNP5 code and presented in Table III. Geometrical model of WWR-K RR including three LEU VVR-KN EFAs, all HEU VVR-C FAs, beryllium reflector and core configuration was imported into MCNP5 input file. After executing the code, results of the EFAs thermal power distribution were gained. Factor of thermal power 1.400 1.300 1.200 1.100 1.000 0.900 0.800 0.700 0.600 45 75 105 135 165 195 225 255 285 315 345 375 405 435 465 495 525 555 Distance from the top of EFAs Fig. 4. Axial thermal power distribution of EFAs. 13 THERMAL-HYDRAULICS ANALYSIS FOR VVR-KN FUEL LEAD TEST … Table III. Factor of thermal power at the hottest position of EFAs. 1st FE 2nd FE 3rd FE 4th FE 5th FE 6th FE 7th FE 8th FE 1st EFA 1.455 1.417 1.404 1.415 1.445 1.478 1.534 1.613 2nd EFA 1.458 1.420 1.408 1.419 1.448 1.482 1.538 1.617 3rd EFA 1.363 1.327 1.316 1.326 1.353 1.385 1.437 1.511  The coolant flow rate through an EFA is 13 m3/h;  Total thermal power of the EFAs is 983 kWt; C. Thermal-hydraulics analysis The calculation of cladding surface temperature, coolant temperature and safety margins of EFAs was performed by using PLTEMP V3.8 code. All these conditions together with geometry of EFAs, coolant flow and previous calculated results of the EFAs thermal power distribution were imported into input file of PLTEMP code. Additionally, the system error, such as uncertainty of power and flow measurement, heat transfer coefficient, must be considered to obtain the desired results. The obtained thermal-hydraulics parameters are presented in Table IV and Fig. 5. PLTEMP code is a thermal-hydraulics code for plate and concentric-tube geometry with capability of calculating natural circulation flow. It was also used to analyze thermal-hydraulics parameters for core conversion of the Dalat RR from HEU to LEU fuel [5]. The initial conditions of the VVR-KN fuel lead test were as follows: Eventually, the computing results were compared with experimental ones obtained from VVR-KN fuel lead test (Table V).  The coolant temperature at the core inlet is 320C; Table IV. Cladding temperature and coolant temperature of the EFAs with system error. Distance from the top (mm) 45 75 105 135 165 195 225 255 285 315 345 375 405 1st EFA Cladding Coolant temp. temp. (0C) (0C) 56.072 60.471 64.942 68.992 73.012 76.444 79.210 81.649 83.535 84.721 85.662 85.938 85.496 33.269 34.208 35.301 36.540 37.914 39.405 40.983 42.624 44.305 45.994 47.668 49.307 50.877 2nd EFA Cladding Coolant temp. temp. (0C) (0C) 56.379 60.833 65.361 69.463 73.535 77.013 79.817 82.291 84.206 85.414 86.374 86.662 86.225 14 33.293 34.250 35.365 36.628 38.027 39.547 41.154 42.825 44.537 46.257 47.963 49.632 51.232 3rd EFA Cladding Coolant temp. temp. (0C) (0C) 55.158 59.403 63.721 67.632 71.515 74.829 77.502 79.858 81.682 82.830 83.741 84.009 83.584 33.223 34.128 35.181 36.374 37.697 39.134 40.653 42.232 43.850 45.476 47.088 48.666 50.179 HO NGUYEN THANH VINH et al. 435 465 495 525 555 585 84.554 83.253 81.232 78.900 76.608 77.175 52.353 53.718 54.949 56.030 56.960 57.812 85.282 83.975 81.941 79.591 77.282 77.861 52.735 54.124 55.379 56.480 57.428 58.296 82.675 81.417 79.466 77.213 74.999 75.547 51.600 52.914 54.100 55.141 56.037 56.858 Temperature, 0C 90 80 Cladding temperature of EFAs 70 60 50 Coolant temperature of EFAs 40 30 20 45 105 165 225 285 345 405 465 525 Distance from the top of EFAs, mm Fig. 5. Cladding and coolant temperature of the EFAs. Table V. Comparison of the experimental with calculated results. Parameters Lead Test PLTEMP Reactor power, MWt 5.65 5.65 Power of three EFAs, kWt 989 983 333 337 Core inlet coolant temperature, C 32 32 Maximum fuel cladding temperature, 0C 88 86.7 374 387 Hot-spot saturation temperature, 0C 107.4 107.4 0 Surface boiling onset temperature, C 114.7 115.4 Surface boiling onset temperature margin 1.43 1.5 Flux instability safety margin 2.1 2.2 Power of the most heat-rated EFA, kWt 0 2 Hot-spot heat flux, kW/m surface boiling onset temperature margin (ONBR) is 1.5. III. CONCLUSION According to the calculated results shown in Table IV, the highest cladding temperature of VVR-KN EFAs is 86.6620C, the highest coolant temperature is 58.2960C, and in Table V, the minimum flux instability safety margin (FIR) is 2.2 and the minimum Because the highest temperature at the outlet of the EFAs is far from saturated temperature (107.40C) so much, belong with obtained FIR and ONBR, it can be concluded that the EFAs operated safely during the test. 15 THERMAL-HYDRAULICS ANALYSIS FOR VVR-KN FUEL LEAD TEST … [4] F. M. Arinkin, L. V. Chekushina, P. V. Chakrov, Sh. Kh. Gizatulin, S. N. Koltochnik, D, Nakipov, A. A. Shaimerdenov, N. Hanan, P. Garner and J. Roglands-Ribas, “Results of the trial of lead test assemblies in the WWR-K reactor”. RRFM 2014, Slovenia, 2014. As it can be also seen in Table V, that the thermal-hydraulics parameters published from the VVR-KN fuel lead test and those computed by PLTEMP code are similar. Therefore, the PLTEMP code can be used for the thermal-hydraulics design of the new nuclear research reactor loaded with LEU VVR-KN FAs of the CNEST project. [5] “Report on neutronics and steady-state thermalhydraulics analyses for DNRR full LEU core configurations”, Reactor Center, Dalat Nuclear Research Institute, VINATOM, Dalat, 2012. REFERENCES [1] Arne P. Olson, M. Kalimullah, “A users guide to the PLTEMP/ANL V3.8 Code”, ANL/RERTR, Argonne National Laboratory, June 2009. [2] J. F. Briesmeister, Ed., “MCNP - A General Monte Carlo N-Particle Transport Code, Version 5”, LA-13709-M, April 2000. [3] Yu. S. Cherepnin, S. A. Sokolov, S. Yu. Bulkin, V. A. Lukichev, O. V. Kravtsova, A. I. Radaev, “Conversion of the WWR-K research reactor to low-enriched fuel as the basis for the development and introduction of the VVR-KN fuel assemblies in existing and advanced pooltype research reactors”, NIKIET, Moscow, Russia, 2014. 16