The RELAP/SCDAPSIM code, designed to predict the behaviour of reactor systems during normal and accident conditions, is being developed as part international SCDAP Development Training Program (SDTP). SDTP consists nearly 60 organizations in 28 countries supporting development technology, software, training materials for nuclear industry. program members licensed software users include universities, research organizations, regulatory vendors, utilities located Europe, Asia, Latin America,...

Immediately after the accident at Fukushima Daiichi, Innovative Systems Software and other members of international SCDAP Development Training Program started an assessment possible core/vessel damage states Daiichi Units 1–3. The included a brief review relevant severe experiments series detailed calculations using RELAP/SCDAPSIM. used RELAP/SCDAPSIM model Laguna Verde BWR vessel related reactor cooling systems. models were provided by Comision Nacional de Seguridad Nuclear y Salvaguardias,...

This work attempts to investigate the thermal hydraulic safety of lithium lead ceramic breeder (LLCB) test blanket system (TBS) in International Thermonuclear Experimental Reactor (ITER) with help modified code relap/scdapsim/mod4.0. The design basis accidents, in-vessel and ex-vessel loss coolant first wall (FW) module (TBM) are analyzed for this assessment. sequence accidents was started postulated initiating events (PIEs). A detailed modeling helium cooling (FWHCS) loop (LLCS) is...

Abstract The paper reports about the work performed in customizing RELAP/SCDAPSIM code features for an easier assessment of a conceptual design Small Modular Reactor (SMR) operating at supercritical pressure proposed by Schulenberg and Otic [13] frame EU ECC-SMART Project. First unsuccessful attempts to make use NRC version RELAP5 analysis thermal-hydraulic behaviour SCW SMR revealed numerical difficulties transition from subcritical pressures. This suggested activate cooperation customising...

Abstract The ASYST (Adaptive System Thermal-Hydraulic) code series is a recent development that integrates the functionalities of SCDAPSIM and SAMPSON. It can simulate analyze fluid reactor systems with Best Estimate Plus Uncertainty (BEPU) capability. proposed work involves conducting thorough validation verification process for model to predict condensation phenomena using condensation-separate effect test facility (present at Department Mechanical Nuclear Engineering, Khalifa University),...

The key objective of the test blanket module (TBM) program is to develop design technology for DEMO and future power-producing fusion reactors. proposed first wall system (TBS) a generalized concept testing in ITER, an experimental reactor under construction France presently. TBM (TBM FW) directly faces plasma cooled by helium cooling (FWHCS), which considered as critical component from ITER safety point view. scope this work comprises thermal hydraulic analysis FWHCS TBS assessment...

A comprehensive uncertainty analysis in the event of a severe accident two-loop pressurized water reactor is conducted using an package integrated ASYST code. The plant model based on nuclear power (NPP) Krško, Westinghouse-type plant. station blackout scenario with small break loss coolant analyzed, and all processes in-vessel phase are covered. best estimate plus (BEPU) methodology probabilistic propagation input used. uncertain parameters selected their impact safety criteria, operation...

In different stages of postulated severe accidents in CANDU reactors, the fuel channels may experience a series thermomechanical deformations, some which have significant impacts on accident progression; however, they not been mechanistically modeled by integrated codes such as MAAP-CANDU and SCDAP/RELAP5. This paper focuses development benchmarking mechanistic models for pressure tube (PT) ballooning sagging phenomena during channel heatup phase well assemblies core disassembly phase. These...

Great interest is given now to advanced nuclear reactors especially those using passive safety components. The Westinghouse AP1000 Advanced Passive pressurized water reactor (PWR) an 1117 MWe PWR designed achieve a high and performance record. system uses natural driving forces, such as gas, gravity flow, circulation convection. In this paper, the of during small break loss coolant accident (SBLOCA) investigated. This was done by modelling systems employed RELAP/SCDAPSIM code. describe...

The initial design of ITER incorporated the use carbon fiber composites in high heat flux regions and tungsten was used for low regions. current includes both these present work thermal hydraulic modeling analysis ex-vessel loss coolant accident (LOCA) divertor (DIV) cooling system. purpose this study is to show that new concept full able withstand scenarios. code RELAP/SCADAPSIM/MOD 4.0. A parametric also carried out with different in-vessel break sizes locations. discusses a number safety...