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The primary side of the steam generator contains vertical primary collectors and horizontal heat exchange tubes. The PACTEL testįacility consists of three primary loops while the reference reactor Reactor after modernization is now 1500 MW).

Scaled down nominal power (nominal thermal power of reference The maximum heating power in the core is 1000 kW, which is roughly 20% of the Of 144 electrically heated fuel rod simulators. The maximum primary and secondary side pressures areĨ.0 MPa and 5.0 MPa, respectively (see Table 1). Have been kept to preserve the natural circulation pressure heads. Besides the emergency coreĬooling systems are simulated in PACTEL. Included in PACTEL: a pressure vessel, main circulation loops, Plants located at Loviisa and managed by Fortum Power and Heat.Īll the main parts of the reference reactor primary loop are (1:305) out-of-pile model of the VVER-440 type nuclear power Parallel Channel Test Loop, PACTEL, is a volumetrically scaled Pipes were prepared and calculations with these models were carried out. Nodalization cases four, ﬁve and eight layers of heat exchange (Riikonen, 1994 Tuunanen and Nakada, 1993). Guidelines of previous RELAP5 model of PACTEL steam generator TRACE/SNAP modeling of the steam generator was based on the

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To test the modeling capabilities of TRACE code for VVERs.

The LOF-10 experiment (Kouhia and Puustinen, 1998) was chosen The many loss-of-feedwater experiments carried out with PACTEL, Symbolic Nuclear Analysis Package (SNAP) model editor. Pare a model for horizontal steam generator of PACTEL using the The way towards whole model of the PACTEL facility was to pre* Corresponding author.

The ﬁfth Finnish reactor unit of EPR type is being constructed inĪ new TRACE V5.0 thermal hydraulic code has been recently At the moment there are two VVER-440 unitsĪnd two boiling water reactor (BWR) units operating in Finland. Water reactor (EPR) type PWR features can be carried out (Rantakaulio et al., 2010). Thus, also research activities on European pressurized Modiﬁed in 2009 to include two new loops with vertical steam Type pressurized water reactors (PWR) and over 200 experiments It wasĭesigned to model the thermal hydraulic behavior of the VVER-440 The Parallel Channel Test Loop (PACTEL) facility (Tuunanen etĪl., 1998), constructed in 1990 at Lappeenranta University of Technology (LUT) in Finland, is the largest facility for VVER-440. Simulated natural circulation phase was found also in the TRACE code calculations. The expected ﬂow reversal in the lowest tube layers during the Modeling of the pipe layers increased the accuracy of the results. Only after the uppermost cell on the secondary side had voided thoroughly. The steam superheating in the calculations was possible When the uppermost tube layer had uncovered. In the experiment the steam started to superheat immediately Ove restimated slightly this heat transfer. Secondary side degraded gradually during the uncovery of the heat exchange tubes. The heat transfer from the primary to the However, at the ﬁnal state the calculated secondary sideĬollapsed level had decreased more than in the experiment. In rather good agreement with the experiment. The simulation of PACTEL loss-of-feedwater experiment LOF-10, the main parameters of the calculations Different nodalization options were introduced. Lowest heat exchange tubes were studied in detail. The phenomenon of heat transfer from primary to secondary side, steam superheating and ﬂow reversal in the The calculation results showed that TRACE is capable in simulating horizontal steam generator behavior both in steady state and during loss-of-feedwater transient. Thermal hydraulic code and assess different modeling options of the code. Main objectives were to prepare a simulation model for its horizontal steam generator with the TRACE Parallel Channel Test Loop (PACTEL) is an integral test facility for a VVER-440 type nuclear reactor. Results of a loss-of-feedwater (LOF-10) experiment at the PACTEL facility. This paper describes the modeling of horizontal steam generator with the TRACE code and calculation Laboratory of Nuclear Engineering, Lappeenranta University of Technology (LUT), Faculty of Technology, LUT Energy, PO Box 20, FIN-53851 Lappeenranta, Finland Journal homepage: TRACE code modeling of the horizontal steam generator of the PACTEL facilityĪnd calculation of a loss-of-feedwater experiment No.Annals of Nuclear Energy 37 (2010) 1494–1501Ĭontents lists available at ScienceDirect If you want a new game key just click on the “ Get It Free” button and follow the simple instructions.

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