Thermohydraulics Department

The Thermohydraulics Department has acquired important experience in the thermohydraulic analysis of VVER-type reactors, both in experimental and in analytical fields. The basis of the first one is the PMK-2, an integral-type scaled-down model of the primary and partly the secondary circuit of Paks NPP. Experiments and computer codes (RELAP5, ATHLET and CATHARE) have been used: to study the transient and accident behaviour of VVER-440/V213 plants to study special problems (e.g. the effect of non-condensable gases)to develop possible accident management strategies for various accident scenarios (e.g. secondary side bleed and feed).The department participates in the RELAP Users' Club CAMP and in the VVER-specific development of the ATHLET and CATHARE codes, including the coupling of the 3D neutron kinetics code KIKO-3D with ATHLET. A number of the NPP's safety enhancement projects have been supported by computer analysis with the above codes. To support the development of symptom based emergency operating procedures for the Paks NPP the analysis of a wide range of accident scenarios has been performed. The department has acquired important experience in the thermohydraulic analysis of VVER-type reactors, both in experimental and in analytical fields. The basis of the first one is the PMK-2, an integral-type scaled-down model of the primary and partly the secondary circuit of Paks NPP. Experiments and computer codes (RELAP5, ATHLET and CATHARE) have been used: to study the transient and accident behaviour of VVER-440/V213 plants to study special problems (e.g. the effect of non-condensable gases) to develop possible accident management strategies for various accident scenarios (e.g. secondary side bleed and feed).The department participates in the RELAP Users' Club CAMP and in the VVER-specific development of the ATHLET and CATHARE codes, including the coupling of the 3D neutron kinetics code KIKO-3D with ATHLET. Extended design basis accident analysis was performed for the safety reassessment of the Paks NPP in the framework of the AGNES project. A number of the NPP's safety enhancement projects have been supported by computer analysis with the above codes. To support the development of symptom based emergency operating procedures for the Paks NPP the analysis of a wide range of accident scenarios has been performed. The PMK-2 is a scaled-down model of the Paks Nuclear Power Plant equipped with VVER-440/213-type reactors of Soviet design. It is a full pressure model of the plant with a volume and power scaling of 1:2070. Due to the importance of gravitational forces in both single- and two-phase flow the elevation ratio is 1:1 except for the lower plenum and pressuriser. The six loops of the plant are modeled by a single active loop. The coolant is water under the same operating conditions as in the plant, so transients can be started from nominal operating conditions. The core model consists of 19 electrically heated rods with uniform power distribution. In the core the heated length, spacer type and elevations, as well as the channel flow area are the same as in the Paks NPP. The main circulating pump of the PMK-2 serves to produce the nominal operating conditions and to simulate the flow coast-down following pump trip. The pump cannot be applied to two-phase conditions; therefore it is accommodated in a by-pass line. The flow coast-down is modeled by closing a control valve. For natural circulation the by-passed cold leg part is opened. The horizontal design of the VVER-440 steam generator is modeled by horizontal heat transfer tubes between hot and cold vertical collectors in the primary side. In the secondary side of the steam generator the steam/water volume ratio is maintained. From the emergency core cooling systems the four hydroaccumulators of the Paks NPP are modeled by two vessels. They are connected to the downcomer and upper plenum similar to those of the reference system. The high and low-pressure injection systems are modeled by the use of piston pumps. In the first design of the PMK-NVH facility only the primary circuit of plant was modeled. This version was used until 1990. The PMK-2 facility is an upgraded version - first of all by addition of a controlled secondary heat removal system - extending the capability of the test loop to transient processes initiated by events in the secondary circuit. A research program has been going on at the Institute to give scientific support to the in-vessel corium retention in the Paks Nuclear Power Plant of VVER-440/213 type. The concept to be implemented in the plant is based on the external reactor pressure vessel cooling (ERVC) to preserve the reactor pressure vessel (RPV) integrity in severe accident sequences leading to core melt. The research program includes integral type experimental and computer code modeling of the severe accident sequence. For the experimental modeling of the cooling loop to be implemented in the plant, an integral type model, the CERES (Cooling Effectiveness on Reactor External Surface) facility was designed and constructed, containing a section of the reactor vessel wall, with a scaling ratio of 1:40 for the vessel external surface and 1:1 for the elevations. For the computer code model of the CERES facility, the RELAP5/mod3.3 code, fully validated by PMK-2 test results, have been used.

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