Preliminary Assessment of the Possible BWR Core/Vessel Damage States for Fukushima Daiichi Station Blackout Scenarios Using RELAP/SCDAPSIM

Abstract

Immediately after the accident at Fukushima Daiichi, Innovative Systems Software and other members of the international SCDAP Development and Training Program started an assessment of the possible core/vessel damage states of the Fukushima Daiichi Units 1–3. The assessment included a brief review of relevant severe accident experiments and a series of detailed calculations using RELAP/SCDAPSIM. The calculations used a detailed RELAP/SCDAPSIM model of the Laguna Verde BWR vessel and related reactor cooling systems. The Laguna Verde models were provided by the Comision Nacional de Seguridad Nuclear y Salvaguardias, the Mexican nuclear regulatory authority. The initial assessment was originally presented to the International Atomic Energy Agency on March 21 to support their emergency response team and later to our Japanese members to support their Fukushima Daiichi specific analysis and model development.

Steam oxidation behavior of Y-bearing cladding tube with aluminizing coating

“…However, the reaction between Zr and steam at high temperature is accompanied by the release of large amounts of hydrogen gas [4][5][6]. According to the reports of the Fukushima Daiichi nuclear disaster, the oxidation of Zr cladding begins above 900°C and causes accelerated hydrogen generation above 1227°C [7][8][9]. For example, during the loss-of-coolant-accident (LOCA), cladding temperature will be increased up to 1200°C due to the decay heat from fuel accompanied by the evaporation of the cooling water, which results in the rapid steam-oxidation of Zr alloy and further produces significant and dangerous levels of hydrogen [10].…”

RELAP5/SCDAPSIM model development for AP1000 and verification for large break LOCA

“…This version used standard RELAP5 and SCDAP/RELAP5 input but included enhanced output options such as integrated 3D and time history plotting options. RELAP5/SCDAPSIM/MOD3.4 is the current production version while RELAP5/SCDAPSIM/MOD3.5 is an experimental version of the code with new QUENCH/PARAMETERexperiment driven SCDAP modeling improvements (Allison et al, 2012) and thermal hydraulic models of the production version. This version of the code will be referred as RS5/MOD3.5 throughout the current article.…”

Comparative study of the hydrogen generation during short term station blackout (STSBO) in a BWR

“…Immediately after the accident at Fukushima Daiichi several authors presented an assessment of the possible core/vessel damage states of the Fukushima Daiichi Units 1-3 station blackout, LOCA, among other scenarios using RELAP5/SCDAPSIM (Allison et al, 2012;Nuñez-Carrera et al, 2012;Trivedi et al, 2014). Allison et al (2012) presented an assessment of the possible core/vessel damage states of the Fukushima Daiichi Units 1-3 station blackout scenarios; the main conclusion is that the responses to the accident once the accident is underway can make a significant difference in the consequences of the accident.…”

“…Allison et al (2012) presented an assessment of the possible core/vessel damage states of the Fukushima Daiichi Units 1-3 station blackout scenarios; the main conclusion is that the responses to the accident once the accident is underway can make a significant difference in the consequences of the accident.…”

“…SCDAP/RELAP5 description is given by Allison et al (2012), which is a combination of the RELAP5 code for thermal hydraulics, SCDAP for the severe accident related phenomena and the COUPLE code for a finite element treatment of the vessel lower head (Siefken et al, 2001a,b). The code was developed by the Idaho National Engineering Laboratory (INEL) for LOCAs and transients within the primary system, reactor core phenomena and fission product release from the fuel.…”