Authors & Affiliations
Semchenkov A.A.1, Kustova I.N.1, Nikel O.A.1, Kabanov Y.A.2
1 N.A. Dollezhal Research and Development Institute of Power Engineering, Moscow, Russia
2 JSC “State Scientific Center – Research Institute of Atomic Reactors”, Dimitrovgrad, Russia
Kustova I.N.1 – Engineer of the 2nd Category.
Nikel O.A.1 – Design Engineer of the 3rd Category.
Kabanov Y.A.2 – Chief Specialist of MBIR Project Office.
The stability of the circulation in the steam-water circuit, that is, the preservation of the thermohydraulic parameters within acceptable limits for perturbations characteristic of the operation, largely determines the reliability of the steam generator itself. Under certain conditions, pulsating pressure, flow and temperature pulsations may appear in boiling apparatus. For the first time, ripples were encountered during the creation of once-through boilers, and the harmful effect of flow instability on the apparatus design led to the need for a detailed study of the problem. In this work, a computational analysis of the thermal-hydraulic stability of the reverse steam generator (OPG) of the MBIR nuclear research facility (NRF) is carried out. The need for such a study is due to the fact that the OPG consists of three parallel connected steam generating modules. With such a design, water flow rate pulsations may appear, including with overturning flow rates in separate steam-generating channels. To avoid this, it is necessary to conduct throttling at the inlet to each OPG module, therefore, in this study, the hydraulic resistance of the chokes is determined. The research methodology is based on the use of a numerical thermohydraulic model of a reverse steam generator, made by means of a certified thermohydraulic design code HYDRA-IBRAE/LM/V1. The OPG RNU MBIR at the economizer and superheating sections has twisted heat exchange intensifiers that direct water and steam both in the longitudinal and transverse directions relative to the bundle of heat exchange tubes. This circumstance required a special approach in calculating the hydraulic resistance and heat transfer coefficients along the path of the OPG modules. After the creation of the computational model, it had to be verified, and since the design of the OPG of the NRF MBIR is similar to the design of the OPG of the BOR-60 reactor, the experimental data obtained on the OPG BOR-60 were used for verification. The calculations of the normal operation modes of the OPG NRF MBIR after the throttling of the modules confirmed the absence of fluctuations in the flow rate in these modes.
reverse steam generator, heat and hydraulic stability, model, calculation, analysis, nuclear research facility, experiment, verification
1. Sroyelov V.S., Privalov Yu.V., Shtynda Yu.Ye., Bocharin P.P., Shtrutek Y., Manek O., Malasek V., Lokhman K. Teplogidravlicheskiye ispytaniya modeli obratnogo parogeneratora natriy-voda [Thermohydraulic tests of the sodium-water reverse steam generator model]. Atomnaya energiya – Atomic Energy, 1986, vol. 61, no. 2, pp. 138–139.
2. Afremov D.A., Zhuravleva Yu.V., Kudryavtsev A.V., Safronov D.V., Semchenkov A.A. Verifikatsiya koda HYDRA-IBRAE/LM/V1 i yego ispol'zovaniye dlya modelirovaniya RU BREST-OD-300. Godovoy otchet NIKIET-2016 [Verification of the HYDRA-IBRAE/LM/V1 code and its use for modeling RU BREST-OD-300. Annual Report NIKIET]. Moscow, NIKIET JSC Publ., 2016. Pp. 138–140.
3. Alipchenkov V.M., Anfimov A.M., Afremov D.A., Gorbunov V.S., Zeigarnik Yu.A., Kudryavtsev A.V., Osipov S.L., Mosunova N.A., Strizhov V.F., Usov E.V. Bazovyye polozheniya, tekushcheye sostoyaniye razrabotki i perspektivy dal'neyshego razvitiya teplogidravlicheskogo raschetnogo koda novogo pokoleniya HYDRA_IBRAE/LM dlya modelirovaniya reaktornykh ustanovok na bystrykh neytronakh [Basic provisions, current development status and prospects for further development of the new generation HYDRA_IBRAE/LM thermo-hydraulic calculation code for modeling fast neutron reactor plants]. Teploenergetika – Thermal Engineering, 2016, no. 2, pp. 54–64.
4. Mosunova N.A., Alipchenkov V.M., Pribaturin N.A., Strizhov V.F., Usov E.V., Lobanov P.D., Afremov D.A., Semchenkov A.A., Larin I.A. Lead coolant modeling in system thermal-hydraulic code HYDRA-IBRAE/LM and some validation results. Nuclear Engineering and Design, 2020, vol. 359, article 110463.
5. Idelchik I.E. Spravochnik po gidravlicheskim soprotivleniyam [Handbook of hydraulic resistance]. Moscow, Mashinostroyeniye Publ., 1992.
6. Kirillov P.L., Bobkov V.P., Zhukov A.V., Yuriev Yu.S. Handbook of thermal hydraulic calculations in nuclear energy. Moscow, Publishing House of Atomic Energy, 2010, vol. 1.
7. Fast reactor steam generators with sodium on the tube side. Design and operational parameters. IAEATECDOC- 730, ISSN 1011-4289. IAEA, VIENNA, 1994.
8. Petrov P.A. Gidrodinamika pryamotochnogo kotla [Hydrodynamics of a once-through boiler]. Moscow, Gosenergoizdat Publ., 1960. 169 p.
9. Morozov I.I., Gerliga V.A. Ustoychivost' kipyashchikh apparatov [The stability of boiling apparatus]. Moscow, Atomizdat Publ., 1969. 280 p.