Passive core cooling system using a Steam Generator (SG) is implemented in case of a loss of reactor coolant accident (LOCA) to cool-down and depressurize the primary system using SGs. Figure IV-2 shows a scenario for small break LOCA.

To show the validity of this system, it is important to verify natural circulation flow characteristics of the primary side and heat transfer capability through SGs because supplied water from accumulators includes non-condensable gas (nitrogen). From this point of view, a simulation test had been performed in addition to the simple flow test in a single SG tube and numerical analyses using CANAC3-3D code. This test was performed by Kansai Electric Power, Kyushu Electric Power, The Japan Atomic Power, and Mitsubishi Heavy Industries.

III — 3. Conclusions

In the safety concept of APWR+ against small LOCA event, Steam Generators (SGs) are used for decay heat removal. Therefore, it is important to maintain natural circulation in the primary system during small LOCA events to transfer the decay heat from the core to the SG. Non-condensable gas, which dissolves in water from the accumulators and injected to primary loop on LOCA event, may accumulate in the top of U-bend tube and cause siphon brake for natural circulation at the U-bend region of SG.

Small LOCA test was performed using the ‘EOS’ simulation loop of a PWR to verify natural circulation and siphon break in SGs, and we concluded that natural circulation with non-condensable gas was maintained during LOCA event and heat removal by the SGs was continuously effective.

REFERENCES TO ANNEX IV

[1] INTERNATIONAL ATOMIC ENERGY AGENCY, Status of advanced light water reactor designs, p139-157, IAEA-TECDOC-1391, IAEA, Vienna (2004).

[2] SUZUTA, T., et al., Development of Advanced Computer Code CANAC3-3D for Next Generation PWR (Part2), Proc. of the 2nd Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety(NTHAS2), (2000).

[3] MAKIHARA, Y., et al., Development of Next Generation PWR (APWR+), Proc. of ICONE-9, Nice, France (2001).

[4] ARITA, S., et al., Safety Evaluation of Next Generation PWR (APWR+), Proc. of ICONE-10, Arlington, Virginia, USA (2002).

[5] TANAKA, T., et al., Examination of Natural Circulation and Heat Removal by Steam Generator, Proc. of the 6th International Conference on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-6),#N6P054, Nara, Japan (2004).