Development of Texas-V code surrogate model for assessment of steam explosion impact in Nordic BWR
2015 (English)In: International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, American Nuclear Society, 2015, Vol. 9, 7222-7235 p.Conference paper (Refereed)Text
Severe accident mitigation strategies in Nordic boiling water reactors (BWRs) employ core melt cooling in a deep pool of water under the reactor pressure vessel. Corium melt released from the vessel is expected to fragment, solidify and form a porous debris bed coolable by natural circulation. However, steam explosion can occur upon melt release threatening containment integrity and potentially leading to large early release of radioactive products to the environment. Significant aleatory and epistemic uncertainties exist in accident scenarios, melt release conditions, and modeling of steam explosion phenomena. Assessment of the risk of ex-vessel steam explosion requires application of the Integrated Deterministic Probabilistic Safety Analysis (IDPSA). IDPSA is a computationally demanding task which makes unfeasible direct application of Fuel-Coolant Interaction codes. The goal of the current work is to develop a Surrogate Model (SM) of the Texas-V code and demonstrate its application to the analysis of explosion impact in the Nordic BWR. The SM should be computationally affordable for IDPSA analysis. We focus on prediction of the steam explosion loads in a reference Nordic BWR design assuming a scenario of coherent corium jet release into a deep water pool. We start with the review of the Texas-V sub-models in order to identify a list of parameters to be considered in implementation of the SM. We demonstrate that Texas-V exhibits chaotic response in terms of the explosion impulse as a function of the triggering time and introduce a statistical representation of the explosion impulse for given melt release conditions and arbitrary triggering time. We demonstrate that characteristics of the distribution are well-posed. We then separate out the essential portion of modelling uncertainty by identification of the most influential uncertain parameters using sensitivity analysis. Both aleatory uncertainty in characteristics of melt release scenarios and water pool conditions, and epistemic uncertainty in FCI modeling are considered. Ranges of the uncertain parameters are selected based on the available information about prototypic severe accident conditions in a Nordic BWR. A database of Texas-V solutions is generated and used for the development of the SM. Performance, predictive capability and application of the SM to risk analysis are discussed in detail.
Place, publisher, year, edition, pages
American Nuclear Society, 2015. Vol. 9, 7222-7235 p.
Ex-vessel steam explosion, IDPSA, ROAAM+, Sensitivity study, Severe accident, Surrogate model
Other Engineering and Technologies not elsewhere specified
IdentifiersURN: urn:nbn:se:kth:diva-187421ScopusID: 2-s2.0-84963998006ISBN: 978-151081184-3OAI: oai:DiVA.org:kth-187421DiVA: diva2:930436
16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2015; Chicago
QC 201605242016-05-242016-05-232016-05-24Bibliographically approved