Design features of watercooled research reactors
10/02/2016 2016 - #03 Physics and technology of nuclear reactors
Chusov I.A. Shelegov A.S. Kochnov O.Yu.
https://doi.org/10.26583/npe.2016.3.12
UDC: 621.039
Brief review of the design, specific features of thermal hydraulics of reactor cores and circulation loops of pool-type research reactors is given. Main principal differences of research reactors as compared with industrial power reactor installations are outlined. Design of reactor installations is examined using the example of two research reactors VVR-M (Gatchina) and VVR-c (Obninsk). Direction of coolant circulation constitutes the feature of research reactor installations which is of key importance. In contrast to power installations propagation of coolant in research reactors is arranged in downwards direction, i.e. from core top to bottom. In connection with the above, particular design features of reactor support grids are discussed in the present study. A set of data is presented on the values of preset values of alarms and emergency protection triggering thresholds. The issue of modernization of the reactor core implemented by developing the family of fuel assemblies (FAs) of the new type is examined separately using the example of modernization of the VVR-M reactor. It is demonstrated that by changing the FA design it is possible to significantly increase the neutron flux density and per unit power of reactor facilities. Tables containing main technical characteristics of different FAs for nuclear reactors of the IRT type are presented. Certain circuit engineering solutions for coolant circulation loops and characteristic design of research loops aimed at the solution of different research tasks are discussed.
References
- Frenkel N.Z. Hydraulics. Moscow-Leningrad. Gosenergoizdat Publ., 1956. 456 p. (in Russian).
- Foks D.A. Hydrodynamic analysis of unsteady flow in pipelines. Moscow. Energoatomizdat Publ., 1981. 246 p. (in Russian)
- Kirsanov G.А., Konoplev K.A., Pikulik R.G., Shishkina Zh.A. Hydraulics of the active zone of the VVR-M reactor. Atomnaya energiya, 1975, v. 39, iss. 5, pp. 320-323 (in Russian).
- Kirsanov G.А., Konoplev K.A., Syasin А.N., Shishkina Zh.A. Determination of heat flux ultimate density for VVR-M reactor fuel assemblies. Leningrad. Preprint, LINP, no. 285, 1976. 17 p. (in Russian).
- Kirsanov G.А., Konoplev K.A., Shishkina Zh.A. To determination of critical heat fluxes and critical heat flux ratios in research pool-type reactors. Atomnaya energiya, 1986, v. 61, iss. 1, pp. 41-42 (in Russian).
- Kirsanov G.A., Konoplev K.A., Findaisen A., Shishkina Zh.A. Comparison of heat engineering potentialities of SFA for WWR-M reactor. Atomnaya energiya, 1989, v. 67, no. 2, pp. 97-100 (in Russian).
- Erykalov А.N., Zvezdkin V.S., Kirsanov G.А., Konoplev K.A., Lvov V.S., Petrov Yu.V., Ruzmanov А.P. WWR-M5 type thin-walled fuel elements for research reactors. Atomnaya energiya, 1986, v. 60, iss. 2, pp. 103-107 (in Russian).
- Zakharov А.S., Zvezdkin V.S., Konoplev K.A., Kirsanov G.A., Pikulik R.G., Orlov S.P., Lvov V.S., Saykov Yu.P. Finned fuel assemnlies of VVR-M reactor. Atomnaya energiya, 1993, v. 74, iss. 1, pp. 88-90. (in Russian).
- Kirsanov G.A., Konoplev K.A., Pikulik R.G., Saykov Yu.P., Tchmshkyan D.V., Tedoradze L.V. and Zakharov A.S. LEU WWR-M fuel assemblies burnable test. The RERTR-2000 International meeting on reduced enrichment for research and test reactors. October 1–6, 2000, Las Vegas, Nevada.
- Enin A.A., Erykalov A.N., Kirsanov G.A., Konoplev K.A., Lvov V.S., Petrov Yu.V., Saykov Yu.P., Zakharov A.S., Zvezdkin V.S. Design and Experience of HEU and LEU fuel for WWR-M reactors. In: Nuclear Engineering and Design, 1998, vol. 182, pp. 233–240.
- Kolesov V.V., Kochnov O.Yu., Volkov Yu.V., Ukraintsev V.F., Fomin R.V. Development of precision model of the VVR-c reactor for subsequent optimization of its design and breeding 99Mo and other radionuclides. Izvestiya vuzov. Yadernaya Energetika, 2011, no. 4, pp. 129-133 (in Russian).
- Zakharov A.S., Zvezdkin V.S., Kirsanov G.A., Konoplev K.A., Lvov V.S., Pikulik R.G., Saykov Yu.P. Development and tests of finned spent fuel assembly of WWR-M reactor. Preprint, PNPI, no. 1799, St. Petersburg, 1992. (in Russian).
- Kirillov P.L., Bobkov V.P., Zhukov A.V., Yur’ev Yu.S. Thermal-hydraulic Calculations in Nuclear Power. Handbook. Moscow. Izdat Publ., v. 1, 770 p. (in Russian).
- Petrov Yu.V., Erykalov A.N., Onegin M.S. A Neutronic Feasibility Study for Fuel Enrichment Reduction of the PNPI WWR-M reactor. Preprint, PNPI, no. 2401. Gatchina, 2000. 51 p. (in Russian).
- Konoplev K.A., Pikulik R. G., Saykov Yu.P. SFA tightness control at the WWR–M reactor in the USSR AS LNPI. In: LNPI methodical and applied works. Leningrad, 1988, pp. 129-130. (in Russian).
- Kirsanov G.A., Konoplev K.A., Saikov Yu.P., Zakharov A.S. The Test method and some results for WWR-M fuel. The 21st International Meeting on Reduced Enrichment for Research and Test Reactors (RERTR). October 18–23, 1998, San-Paulo, Brazil.
- Petuhov B.S., Genin L.G., Kovalev S.A., Solov’ev S.L. Heat Transfer in Nuclear Power Plants. Moscow. MEI Pibl., 2003, 548 p. (in Russian).
research reactor turbulence thermal hydraulics heat transfer in nuclear reactor fuel assembly reactor safety pooltype research reactor VVR-M2 VVR-M3 VVR-M5 VVR-c fuel assemblies
Link for citing the article: Chusov I.A., Shelegov A.S., Kochnov O.Yu. Design features of watercooled research reactors. Izvestiya vuzov. Yadernaya Energetika. 2016, no. 3, pp. 116-128; DOI: https://doi.org/10.26583/npe.2016.3.12 (in Russian).