EDN: LMLULB
Authors & Affiliations
Zubkov A.G.1, Oleksyuk D.A.1, Vertikov E.A.1, Noskov A.S.1, Scherbinin A.A.1, Morozkin O.N.2, Shishkin A.A.2
1 National Research Centre “Kurchatov Institute”, Moscow, Russia
2 Join Stock Company TVEL, Moscow, Russia
Zubkov A.G.1 – Research Fellow. Contacts: 1, pl. Akademika Kurchatova, Moscow, Russia, 123182. Tel.: +7 (926) 282-56-58; e-mail: This email address is being protected from spambots. You need JavaScript enabled to view it..
Oleksyuk D.A.1 – Head Of Department, Cand. Sci. (Tech.).
Vertikov E.A.1 – Engineer.
Noskov A.S.1 – Head Of Laboratory.
Shcherbinin A.A.1 – Deputy Head Of Department.
Morozkin O.N.2 – Project Manager.
Shishkin A.A.2 – Director Of Department.
Abstract
The paper presents the results of experimental studies of local parameters of coolant (temperature and velocity), as well as pressure drops, obtained on two multi-rod fuel assembly models. The experiments were performed at the thermal physics test facility KS in the National Research Center “Kurchatov Institute” in order to obtain data for the validation of software tools for thermohydraulic analysis of the core. The paper describes the methods of measuring local parameters. Experimental studies were carried out in a wide range of operating parameters of the coolant, in particular, at nominal operating parameters of cores of pressurized water reactors. Local temperature studies were performed using thermocouples installed in the flow, and local velocity studies were performed using Pitot tubes. The paper presents a methodology for the primary processing of experimental data. The computational analysis of the experimental data was performed using the SC-INT subchannel code. The article presents a step-by-step configuration of the input parameters of the code responsible for the characteristics of experimental fuel assembly models. A comparison of the calculated and experimental data on the magnitude of the pressure drop, local temperatures of the coolant and local speeds was performed, and a good agreement between the calculation and experiment was shown.
Keywords
thermophysical studies, local coolant parameters, fuel assemblies, fuel assembly models, pressurized water reactors, turbulent mixing, validation of subchannel codes, temperature measurement, velocity measurement
Article Text (PDF, in Russian)
References
- Klemin A.I., Polyanin L.N., Strigulin M.M. Teplogidravlichekii raschet i teplotekhnicheskaya nadezhnost' yadernykh reaktorov [Thermal hydraulic calculation and thermal reliability of nuclear reactors]. Moscow, Atomizdat Publ., 1980. 261 p.
- Tong L.S. Boiling heat transfer and two-phase flow. Routledge, 2018.
- Lisenkov E.A., Bezrukov Yu.A., Seleznev A.V., Vasilchenko I.N., Bogdanov A.S., Khripachev Yu.B. Eksperimental'nyye issledovaniya kriticheskikh teplovykh potokov na modelyakh TVS-2M s peremeshivayushchimi reshetkami [Experimental studies of critical heat fluxes on TVS-2M models with mixing grids]. Tyazheloye mashinostroyeniye – Heavy Engineering, 2017, no. 9, pp. 10–15.
- Samoilov O.B., Kupriyanov A.V., Falkov A.A., Shipov D.L., Molodtsov A.A., Lukyanov V.E. Eksperimental'nyye issledovaniya kriticheskikh teplovykh potokov i razrabotka korrelyatsii dlya TVSA s peremeshivayushchimi reshetkami-intensifikatorami [Experimental inwestigations of critical heat flux and development of critical heat flux correlation for TVSA with mixing grids]. Voprosy atomnoy nauki i tekhniki. Seriya: Yaderno-reaktornyye konstanty – Problems of Atomic Science and Technology. Series: Nuclear Reactor Constants, 2016, no. 3, pp. 47–53.
- Dmitriev S.M., Lukyanov V.E., Samoilov O.B. Obosnovaniye korrelyatsii dlya rascheta kriticheskogo teplovogo potoka v teplovydelyayushchikh sborkakh al'ternativnoy konstruktsii s peremeshivayushchimi reshetkami-intensifikatorami dlya VVER-1000 [The Substantiation of the Correlation for Critical Heat Flux Calculation for Alternative Design Fuel Assemblies with Mixing Spacer Grids in VVER-1000]. Izvestiya vuzov. Yadernaya Energetika, 2012, no. 1, pp. 99–108. DOI: https://doi.org/10.26583/
npe.2012.1.12.
- Kuzina Ju.A., Arnoldov M.N., Orlov Yu.I., Sorokin A.P. Teplofizicheskiye issledovaniya: ot pervogo stenda k krupnomasshtabnoy atomnoy energetike [Thermophysical investigations: from the first to stand large-scale nuclear energy]. Voprosy atomnoy nauki i tekhniki. Seriya: Yaderno-reaktornyye konstanty – Problems of Atomic Science and Technology. Series: Nuclear Reactor Constants, 2021, no. 2, pp. 236–255.
- Gidrodinamicheskiye osobennosti techeniya teplonositelya za peremeshivayushchey distantsioniruyushchey reshetkoy TVS-Kvadrat reaktora PWR [Hydrodynamic features of the coolant flow behind the mixing spacer grid TVS-Kvadrat of the PWR reactor]. Teploenergetika – Thermal Engineering, 2019, no. 4, pp. 32–38.
- Kobzar L.L., Oleksyuk D.A., Semchenkov Y.M. Experimental and computational investigations of heat and mass transfer of intensifier grids. Kerntechnik, 2015, vol. 80, no. 4, pp. 349–358. DOI: 10.3139/124.110508.
- Oleksyuk D.A. Razrabotka i eksperimental'noye obosnovaniye programmy dlya poyacheykovogo teplogidravlicheskogo rascheta aktivnykh zon reaktorov tipa VVER. Diss. kand. tekh. nauk [Development and experimental validation of the program for cell-by-cell thermal-hydraulic calculation of WWER reactor cores. Diss. cand. tech. sci.] Moscow, 2002. 194 p.
- SC-INT. Certification passport of the program for electronic computers No. 578 dated 31.03.2023.
- Kobzar L.L., Oleksyuk D.A. Experimental Studies of the Efficiency of Heat-and-Mass Transfer Intensifier Spacer-Grids. At Energy, 2019, vol. 125, issue 5, pp. 290–296. DOI: https://doi.org/10.1007/s10512-019-00483-8.
- GOST 8.586.2-2005 (ISO 5167-2:2003). Izmereniye raskhoda i kolichestva zhidkostei i gazov s pomoshch'yu standartnykh suzhayushchikh ustroistv. Chast' 2. Diafragmy. Tekhnicheskiye trebovaniya. Vved. 2007.01.01 [GOST 8.586.2-2005 (ISO 5167-2:2003). Flow and quantity measurement of liquids and gases using standard orifices. Part 2. Diaphragms. Specifications. Introduction. 2007.01.01]. Moscow, Standartinform Publ., 2007.
- Vertikov E.A., Oleksyuk D.A., Zubkov A.G., Malyutin M.A. K voprosu o validatsii poyacheykovykh kodov dlya rascheta aktivnykh zon reaktorov tipa VVER [On the validation of cell-by-cell codes for calculating WWER reactor cores]. Sbornik dokladov XXIV Mezhdunarodnoy konferentsii molodykh spetsialistov po yadernym energeticheskim ustanovkam [Proc. of the XXIV International Conference of Young Specialists on Nuclear Power Plants]. Podolsk, 2024, pp. 195–203.
- Altshul A.D., Voitinskaya Yu.A., Kazennov V.V. et al. Gidravlicheskiye poteri v vodovodakh elektrostantsiy [Hydraulic losses in power plant water conduits]. Moscow, Energoatomizdat Publ., 1985. 104 p.
- Borisov V.D. Poperechnoye peremeshivaniye teplonositelya v puchkakh sterzhney [Transverse mixing of coolant in rod bundles]. Preprint IAE-3269/5, Moscow, IAE Publ., 1980. 28 p.
- Oleksyuk D.A., Kireeva D.R. Validation of the SC-INT code using experimental data on coolant mixing in a 37-rod fuel assembly with heat exchange intensifying spacer grids. Proc. of the M&C 2017 – International Conference on Mathematics & Computational Methods Applied to Nuclear Science & Engineering. Jeju, Korea, 2017.
- Subcooled Boiling Data from Rod Bundles: technical report. EPRI. United States, 2002. No. 1003383.
UDC 532.57, 536.24, 621.039.5
Problems of Atomic Science and Technology. Series: Nuclear and Reactor Constants, 2025, no. 2, 2:18