Analysis of mass transfer processes in a reactor during a loss-of-coolant accident
The purpose of the work was to optimize the parameters of the spillage system equipped with a gas pressure hydroaccumulator for a ship water-to-water reactor in a loss-of-coolant accident. The water-gas ratio in the hydroaccumulator and the hydraulic resistance of the path between the hydroaccumulator and the reactor were optimized at the designed hydroaccumulator geometric volume.
The main dynamic processes were described using a mathematical model and a computational analysis. A series of numerical calculations were realized to simulate the behavior dynamics of the coolant level in the reactor during the accident. Estimates of the minimum and maximum values of the coolant level were obtained: depending on the initial water-gas ratio in the hydroaccumulator at different diameters of the flow restrictor on the path between the hydroaccumulator and the reactor. These results were restricted by the conditions that, during spillage, the coolant level should remain above the core and below the blowdown nozzle. The first condition implies that the core is in a safe state, the second excludes the coolant water blowdown. The optimization goal was to achieve the maximum time interval in which these conditions would be satisfied simultaneously.
The authors propose methods for selecting the optimal parameters of the spillage system; these methods provide the maximum time for the core to be in a safe state during a loss-of-coolant accident at the designed hydroaccumulator volume. Using these methods, it is also possible to make assessments from the early stages of designing reactor plants.
- Innovative Small and Medium Sized Reactors: Design Features, Safety Approaches and R&D Trends. Final report of a technical meeting held in Vienna, 7-11 June 2004. IAEA-TECDOC-1451. IAEA, Vienna, 2005.
- Zverev D.L., Pahomov A.N., Polunichev V.I., Veshnyakov K.B., Kabin S.V. New generation reactor plant RITM-200 for the perspective nuclear icebreaker. Atomnaya energiya, 2012, v. 113, iss. 6, pp. 323-328 (in Russian).
- Fadeev Yu.P., Belyaev V.M., Pahomov A.N., Polunichev V.I., Veshnyakov K.B., TurusovA.Yu., Vorob’yov V.M. NPP of the raised power for icebreakers. IV International scientific and technical conference «Innovative projects and technologies of nuclear power engineering» (ISTC NIKIET-2016): the collection of reports. Moscow. JSC «NIKIET» Publ., 2016, v 1, pp. 577-581 (in Russian).
- Kuul V.S., Samoilov O.B., Falkov A.A. VPBER-600 Integral reactor thermohydraulic study during LOCAs. NURETH-5.Proc.Int.Top.Meet. on Nuclear Reactor Thermal Hydraulics, Salt Lake City, Utah, USA, September 21-24, 1992.
- Vorob’yova M.V., Gusev A.S., Lepyohin A.N., Morozov O.N. The calculated analysis of reactor plantsafety for floating atomic station in a Severodvinsk with use of a code of improved estimation RELAP5/mod3.2. Proc. of the VII-th International youth scientific and technical conference «Future of technic science» (Nizhny Novgorod, May 16, 2008). Nizhny Novgorod. NSTU n.a. R.E. Alekseev Publ., 2008, pp. 189-190 (in Russian).
- Lepyohin A.N., Vorob’yova M.V., Gusev A.S. Reactor plant KLT-40S safety analysis for floating small-power atomic thermal-electric power station in loss-of-coolant accidents: report theses. Molodyozh’ v nauke [Youth in science]: the collection of reports. VIII scientific and technical conference (Sarov, November 10-12, 2009).Sarov: FSUE «RFYaC-VNIIEF», 2010. - С. 16-24.
- Vorob’yova M.V., Fakeev A.A., Sadovnikov A.V., Lepyohin A.N. LOCA safety analysis forthe atomic lighter carrier «Sevmorput’» reactor in support of its life extension capability. Technologies of supply of life cycle of nuclear power plants: reviewed scientific and technical collection. FNO FSUE «NITI n.a. A.P. Aleksandrov» (Sosnovy Bor). St.-Petersburg. OC «Izdatel’stvo VVM» Publ., 2016, no. 2 (4), pp. 8-23 (in Russian).
- Lepyohin A.N., Gusev A.S., ShvetsovYu.K., Sokolov A.N. Analyze of loss-of-coolant accidents in RITM-200 power plant at passive safety systems connection at KORSAR/BR and RELAP/SCDAPSIM/MOD3.4 codes. Technologies of supply of life cycle of nuclear power plants: reviewed scientific and technical collection. FNO FSUE «NITI n.a. A.P. Aleksandrov» (Sosnovy Bor). St.-Petersburg. OC «Izdatel’stvo VVM» Publ., 2017, no.1 (7), pp. 29-40 (in Russian).
- Idel’chik I.E. The reference book on hydraulic resistances. Moscow. Mashinostroenie Publ., 1990, 672 p. (in Russian).
- Chirkin V.S. Thermоphysical properties of materials of nuclear engineering. Moscow. Atomizdat Publ., 1968, 484 p. (in Russian).
- Rivkin S.L., Aleksandrov A.A. Thermоphysical properties of water and water steam. Moscow. Energiya Publ., 1980, 424 p. (in Russian).
- Thermophysical properties database of material for Light Water Reactors and Heavy Water Reactors. Final report of a coordinated research project 1999–2005.IAEA-TECDOC-1496. IAEA, Vienna, 2006.
- RELAP5/mod3. Сode Manual. Idaho National Engineering Laboratory, June 1995, NUREG/CR 5535 V1-V5.
- Migrov Yu.A., Volkova S.N., Yudov Yu.V., Danilov I.G., Korotaev V.G., Kut’in V.V., Bondarchik B.R., Benediktov D.V. KORSAR – thermohydraulic code of new generation for safety justification of the atomic power station with VVER. Teploenergetika, 2001, no. 9, pp. 36-43 (in Russian).