Safety Analysis

시스템 코드와 열수력 실험을 이용한 원전 및 피동안전계통의 안전해석

New reactors must demonstrate that their safety systems can remove decay heat under design basis accidents and design extension conditions. ATHENA evaluates the accident mitigation capability of nuclear power plants and passive safety systems using thermal-hydraulic system codes such as MARS-KS, SPACE, MELCOR and GAMMA+, and supports the models with separate-effect experiments on condensation, boiling and radiative heat transfer.

Accident analysis of the innovative SMR (i-SMR)

i-SMR의 LOCA·비LOCA 사고와 피동계통 고장 조건에서 사고 완화 능력 평가

  • i-SMR
  • System code
  • LOCA

The Korean innovative SMR (i-SMR) removes decay heat with passive systems such as the passive auxiliary feedwater system (PAFS), the passive emergency core cooling system (PECCS) and the passive containment cooling system (PCCS). We build detailed plant input models and analyze representative LOCA and non-LOCA scenarios, multiple failures of passive safety valves, and heat-load sharing between PAFS and PCCS when PECCS fails.

(a) i-SMR reactor module and (b) passive safety systems (PAFS, PECCS, PCCS) connected to the emergency cooling tank.
Fig. 1. (a) i-SMR reactor module and (b) passive safety systems (PAFS, PECCS, PCCS) connected to the emergency cooling tank.
MELCOR nodalization of the i-SMR reactor building, containment and secondary system.
Fig. 2. MELCOR nodalization of the i-SMR reactor building, containment and secondary system.

Metal containment vessel and flooding safety system

금속격납용기의 열손실 저감과 침수안전계통 설계 변수 도출

  • SMR
  • CFD
  • TH experiment

A metal containment vessel is compact but loses heat by radiation during normal operation, while it must reject decay heat efficiently during an accident. We measure and simulate radiative and convective heat transfer across the MCV gap for different filling gases and emissivities, test radiation shielding, and determine design parameters of the flooding safety system such as the flooding valve size and MCV wall thickness.

Heat-loss mechanisms of the MCV: CFD for different filling gases, radiation shielding, and the experimental facility with its CFD model.
Fig. 3. Heat-loss mechanisms of the MCV: CFD for different filling gases, radiation shielding, and the experimental facility with its CFD model.

Condensation and boiling heat transfer

피동계통 성능을 좌우하는 응축·비등 열전달 실험과 코드 모델 검증

  • TH experiment
  • Model validation

Passive containment cooling relies on condensation, and core cooling depends on boiling and critical heat flux (CHF). We carry out experiments on dropwise and filmwise condensation, flow-boiling CHF and reflood heat transfer of accident-tolerant fuel cladding, develop mechanistic models from the data, and assess the condensation models implemented in the SPACE and MARS-KS codes.

Dropwise condensation on a vertical surface: (a) droplet and rivulet regions, (b, c) visualization.
Fig. 4. Dropwise condensation on a vertical surface: (a) droplet and rivulet regions, (b, c) visualization.
Visualization of pool boiling, flow boiling up to CHF, and safety injection experiments.
Fig. 5. Visualization of pool boiling, flow boiling up to CHF, and safety injection experiments.

Passive safety of molten salt reactors

상변화물질(PCM) 피동안전계통 등 용융염원자로의 계통 수준 안전성 분석

  • MSR
  • System code
  • Passive safety

For the PMFR, we develop a passive safety system that fills the space between the reactor vessel and the containment with a phase change material (PCM), absorbing decay heat as latent heat. System-level thermal-hydraulic behavior and safety of MSRs are analyzed with GAMMA+ and MELCOR, and freeze-valve release models are investigated for fuel salt drain.

PCM-based passive safety system of the PMFR: PCM cells between the reactor vessel and the containment vessel.
Fig. 6. PCM-based passive safety system of the PMFR: PCM cells between the reactor vessel and the containment vessel.