Safety Injection Tank Research Articles

Integral Thermal-Hydraulics Tests for the Safety Evaluation of VVER-440/213 Nuclear Reactors and Safety Code Validation

The Paks nuclear power plant is equipped with pressurized water reactors of the VVER-440/213 type. These plants have a number of special features, namely, six-loop primary circuit, horizontal steam generators, loop seal in both hot and cold legs, safety injection tank setpoint pressure higher than secondary pressure, etc. As a consequence of the special design solutions, the transient behavior of such a reactor system is different from the usual pressurized water reactor system behavior. To study the transient behavior of these plants, the PMK-2 integral-type facility, a thermal-hydraulic model of the Paks nuclear power plant, was designed and constructed.A short description of the specific design solutions of the VVER-440/213-type plants is given with the modeling aspects and similarity criteria applied to the design of the PMK-2 facility. Since the startup of the facility in 1985, 48 experiments have been performed primarily in an international framework with the participation of several experts from European and overseas countries to study one- and two-phase natural circulation, loss-of-coolant accidents, special plant transients, and experiments in support of the accident management measures. The results of several experiments illustrate the system effects of special design solutions and the effectiveness of bleed-and-feed accident management measures. A brief commentary on the thermal-hydraulic system code validation is provided, and conclusions are offered.

The development of passive design features for the Korean Next Generation Reactor

Four passive design features, such as a fluidic device, a passive secondary condensing system (PSCS), a passive cavity flooding system, and a passive hydrogen ignitor, are under development as part of the Korean Next Generation Reactor (KNGR), an advanced PWR, to increase its safety. The fluidic device, which is located at the discharge of the safety injection tank, is a system to inject the borated water into the reactor coolant system in a passively regulating way to elongate the allowable start-up time of emergency diesel generators and to enhance performance against the loss of coolant accidents. The PSCS, which supplements the auxiliary feedwater system, secures the heat removal through steam generators in the case of the loss of feedwater event. The PSCS takes inlet flow from the steam line and returns condensate into the feedwater line after condensation through condenser tubes. Fusible plugs have been adopted for passive cavity flooding. If the ambient temperature is high enough, the plugs between the in-containment refueling water storage tank (IRWST) and the reactor cavity melt so that the IRWST water starts to flow into the reactor cavity by gravity. Passive hydrogen ignitors (i.e. catalytic ignitors) have been adopted in addition to the active hydrogen ignitors to maintain containment hydrogen concentrations below a detectable limit of 10 volume percent. Applicability of these passive features on KNGR has been studied following the three-step approach of (1) preliminary analyses using computer codes and small-scale experimental facilities, (2) detailed analyses through large-scale tests, code running, and uncertainty validation, and (3) quality assurance by design verification and analyses on interfaces with other systems.

Design options for the safety injection system of Korean next generation reactor

Loss of coolant accident (LOCA) analyses for various configurations of safety injection system (SIS) are performed to optimize the emergency core cooling system (ECCS) performance for the Korean next generation reactor (KNGR). The KNGR is an advanced light water reactor (ALWR) adopting the advanced design feature of a direct vessel injection (DVI) configuration and passive fluidic device in the discharge line of the safety injection tank (SIT). To determine the feasible SIS configuration and the optimum capacities of the SIT and high pressure safety injection pump (HPSIP), licensing design basis and best estimate LOCA analyses are performed for the limiting large break and small break spectrum, respectively. The analyses results show that the four-train DVI injection with the current system design is a more feasible configuration than the other ones considered and the adoption of a fluidic device SIT enhances the ECCS performance for large break LOCA. For small break LOCA, in the case of cold leg break, the DVI4 configuration is better than other configurations and also meets the EPRI ALWR requirement of no core uncovery for up to a 15.24 cm (6 in) diameter small break. However, in the case of DVI line break, slight core uncovery is predicted and also the system behavior is significantly affected by reactor vessel (RV) downcomer modeling. Therefore, the DVI4 configuration is more feasible for KNGR ECCS performance, but further investigations are required to resolve the ECCS bypass issues for large break LOCA and to develop a proper RV downcomer model for analysis of DVI line break in small break LOCA.

Small break LOCA analysis for optimization of safety injection system of Korean standard nuclear power plant

The Ulchin Nuclear Power Plant Units 3 and 4 (UCN 3&4) design was based on ABB-Combustion Engineering's (ABB-CE's) standardized 3800 MWt Nuclear Steam Supply System (System 80). Despite the 25% core power difference between the two plants (3800 MWt for System 80, 2815 MWt for UCN 3&4), several design features have not been changed, especially the capacity of Safety Injection System (SIS). Focusing on this fact, the sensitivity study for UCN 3&4 SIS capacity on the behavior of the postulated small break Loss of Coolant Accident (LOCA) was performed to optimize the SIS capacity. Firstly, the sensitivity study for Safety Injection Tank (SIT) capacity on the postulated small break LOCA was performed by changing the size of SIT from 100% to 60% (100%, 85%, 75%, 60%, respectively) for the 0.5 ft 2 (465 cm 2) pump discharge leg break. Secondly, the sensitivity study of High Pressure Safety Injection (HPST) pump capacity on the behavior of small break LOCA was investigated by reducing its discharge flow rate of UCN 3&4 to 50% (100%, 85%, 75%, 70%, 60% and 50%, respectively) for the 0.05 ft 2 (46.5 cm 2 pump discharge leg break. Finally, the spectrum analysis was performed using the acceptable results of these two studies, i.e., the reduced SIT size and HPSI pump capacity to 70% of UCN 3&4 to retain sufficient margin for the small break LOCA. The results of this spectrum analysis show that the peak cladding temperature (PCT) and peak local cladding oxidation (PLO) for the limiting break size of 0.05 ft 2 (46.5 cm 2 pump discharge leg break were 1459°F (793°C) and 0.486% respectively, which are in compliance with the Emergency Core Cooling System (ECCS) acceptance criteria of 2200°F (1204°C) and 17% specified in 10 CFR 50.46. The resuts of this sensitivity study can be adapted to the optimal design of SIS for the future 2815 MWt Korean Standard Nuclear Power Plant and be contributed to the increase of economical gain.

Photomask array placement on slices in LSI processing : A. F. Tasch, Jr. and V. R. Porter. Microelectron & Reliab.12, 539 (1973)

The effects of coolant injection into a reactor vessel by the passive actuation of SITs (Safety Injection Tanks) on the in-vessel core melt progression under a severe accident have been evaluated in an APR (Advanced Power Reactor) 1400. A best estimate simulation from initiating events of 2-in. and 3-in. break SBLOCAs (Small Break Loss Coolant Accidents) without SI (Safety Injection) to a reactor vessel failure has been carried out using the SCDAP/RELAP5 computer code in conditions with and without the passive actuation of the SITs. The SCDAP/RELAP5 results have shown that the passive coolant injection into the reactor vessel by the actuation of the SITs leads to postpone the reactor vessel failure time by 4–5 h in the SBLOCA without SI. At the time of the reactor vessel failure, 44.5% and 42.8% of the fuel rod cladding were oxidized in the 2-in. and 3-in. SBLOCAs with the actuation of the SITs, respectively. However, 37.5% and 34.6% of the fuel rod cladding were oxidized in the 2-in. and the 3-in. SBLOCAs without the actuation of the SITs, respectively. Even though the SITs are designed for a large break LOCA, their actuation is very effective on the delay of a reactor vessel failure in spite of more hydrogen generation in the SBLOCA without SI.

Sealing action of packing rings

The effects of coolant injection into a reactor vessel by the passive actuation of SITs (Safety Injection Tanks) on the in-vessel core melt progression under a severe accident have been evaluated in an APR (Advanced Power Reactor) 1400. A best estimate simulation from initiating events of 2-in. and 3-in. break SBLOCAs (Small Break Loss Coolant Accidents) without SI (Safety Injection) to a reactor vessel failure has been carried out using the SCDAP/RELAP5 computer code in conditions with and without the passive actuation of the SITs. The SCDAP/RELAP5 results have shown that the passive coolant injection into the reactor vessel by the actuation of the SITs leads to postpone the reactor vessel failure time by 4–5 h in the SBLOCA without SI. At the time of the reactor vessel failure, 44.5% and 42.8% of the fuel rod cladding were oxidized in the 2-in. and 3-in. SBLOCAs with the actuation of the SITs, respectively. However, 37.5% and 34.6% of the fuel rod cladding were oxidized in the 2-in. and the 3-in. SBLOCAs without the actuation of the SITs, respectively. Even though the SITs are designed for a large break LOCA, their actuation is very effective on the delay of a reactor vessel failure in spite of more hydrogen generation in the SBLOCA without SI.