Maolong Liu

Estimation of the Depressurization Process of Fukushima Daiichi NPP Unit 1 with SAMPSON

Abstract As Units 1, 2, and 3 of the Fukushima Daiichi nuclear power plant (NPP) entered the phase of long-term station blackout following the huge tsunami, the decay heat could not be effectively removed from the reactor vessel and resulted in high in-vessel pressure and temperature. The Tokyo Electric Power Company announced that the safety relief valves of Fukushima Daiichi NPP Unit 1 (1F1) were never manually opened. However, the measured reactor pressure was decreased to ~1 MPa at 2:43 on March 12, 2011. Such unanticipated depressurization might accelerate core uncovery and on the other hand delay containment failure caused by direct containment heating. In addition, the failure time and the failure path of the boiling water reactor pressure boundary before manual depressurization have a huge impact on the resulting source term. The authors modeled the creep failure of the stainless steel guide tubes of the source range monitor in the core and the main steam line and estimated the possible depressurization mechanism of 1F1 using the SAMPSON (Severe Accident Analysis Code with Mechanistic, Parallelized Simulations Oriented towards Nuclear Field) severe accident analysis code.

Development of simple success criteria regarding alternative water injection for emergency response to long-term station blackout of BWR

This paper discusses the conditions regarding the alternative water injection which prevents core damage under long-term station blackout of a Boiling Water Reactor (BWR). A BWR-5 model plant was analyzed by RELAP5/SCDAP mod3.5 by changing six parameters regarding the accident scenario and alternative water injection. It was found that preventing core damage was almost equivalent to preventing cladding rupture. When the temperature of fuel rod reaches the region where cladding creep rupture occurs, oxidation and heatup of Zircaloy begin to accelerate each other as a result of increased surface area by that cladding rupture. By summarizing the results of sensitivity analyses, the discriminants for calculating an index of core condition and their threshold values for preventing core damage were proposed. Such information prepared for each BWR plant would be useful in the actual emergent condition where the measurements of core parameters are unavailable.

Improvement of SAMPSON Code for BWR Severe Accident Analysis

The SAMPSON code has been developed in the IMPACT project in Japan to investigate severe accident phenomena for light water reactors. It integrates various analysis modules into a single code. The authors improved the fuel rod heat-up module of SAMPSON code by modeling the oxidation reaction of various core structures, including Zircaloy, stainless steel and B4C. And the creep failures of the Zircaloy fuel cladding and stainless steel monitoring guide tubes of the source range monitor (SRM) in the reactor core was also modeled for severe accident analysis.Copyright