• Nuclear Engineering and Technology
• /
• v.52 no.4
• /
• pp.742-754
• /
• 2020

Zircaloy cladding oxidation is an important phenomenon for both design basis accident and severe accidents, because it results in cladding embrittlement and rapid fuel temperature escalation. For this reason during the last decade, many experts have been conducting experiments to identify the oxidation phenomena that occur under design basis accidents and to develop mathematical analysis models. However, since the study of design extension conditions (DEC) is relatively insufficient, it is essential to develop and validate a physical and mathematical model simulating the oxidation of the cladding material at high temperatures. In this study, the QUENCH-05 and -06 experiments were utilized to develop the best-fitted oxidation model and to validate the SPACE code modified with it under the design extension condition. It is found out that the cladding temperature and oxidation thickness predicted by the Cathcart-Pawel oxidation model at low temperature (T < 1853 K) and Urbanic-Heidrick at high temperature (T > 1853 K) were in excellent agreement with the data of the QUENCH experiments. For 'LOCA without SI' (Safety Injection) accidents, which should be considered in design extension conditions, it has been performed the evaluation of the operator action time to prevent core melting for the APR1400 plant using the modified SPACE. For the 'LBLOCA without SI' and 'SBLOCA without SI' accidents, it has been performed that sensitivity analysis for the operator action time in terms of the number of SIT (Safety Injection Tank), the recovery number of the SIP (Safety Injection Pump), and the break sizes for the SBLOCA. Also, with the extended acceptance criteria, it has been evaluated the available operator action time margin and the power margin. It is confirmed that the power can be enabled to uprate about 12% through best-estimate calculations.

• Nuclear Engineering and Technology
• /
• v.52 no.9
• /
• pp.1955-1962
• /
• 2020

In a sudden shutdown of primary pump or coolant loss accident in a marine nuclear power plant, the primary flow decreases rapidly in a transition process from forced circulation (FC) to natural circulation (NC), and the lower flow enters the steam generator (SG) causing reverse flow in the U-tube. This can significantly compromise the safety of nuclear power plants. Based on the marine natural circulation steam generator (NCSG), an experimental loop is constructed to study the characteristics of reverse flow under middle-temperature and middle-pressure conditions. The transition from FC to NC is simulated experimentally, and the characteristics of SG reverse flow are studied. On this basis, the experimental loop is numerically modeled using RELAP5/MOD3.3 code for system analysis, and the accuracy of the model is verified according to the experimental data. The influence of the flow variation rate on the reverse flow phenomenon and flow distribution is investigated. The experimental and numerical results show that in comparison with the case of adjusting the mass flow discontinuously, the number of reverse flow tubes increases significantly during the transition from FC to NC, and the reverse flow has a more severe impact on the operating characteristics of the SG. With the increase of flow variation rate, the reverse flow is less likely to occur. The mass flow in the reverse flow U-tubes increases at first and then decreases. When the system is approximately stable, the reverse flow is slightly lower than obverse flow in the same U-tube, while the flow in the obverse flow U-tube increases.