Conference Agenda

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Session Overview
6.01-2: Plant transient analysis - II
Tuesday, 17/Mar/2020:
11:15am - 12:45pm

Session Chair: DAVID PIALLA, EDF, France
Location: B-1048

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A comparative Simulation of Station Black Out Scenario Using MELCOR and CINEMA for APR1400

JinHo Song1, Dong-Gun Son1, JunHo Bae1, GwangSoon Ha1, Bub-Dong Chung2

1KAERI; 2Future and Challenge

Aa part of severe accident analysis code CINEMA (Code for INtegrated severe accidEnt Management Analysis) development, COMPASS (COre Meltdown Progression Accident Simulation Software) and SIRIUS (SImulation of Radioactive nuclides Interaction Under Severe accident) codes are being developed at Korea Atomic Energy Research Institute. Fundamental models of core melt progression including core heat up, core melting and relocation, formation of corium pool in the lower plenum, and vessel failure, and resultant fission product releases are discussed by comparing the accident progression during a station black out event for APR1400 simulated by CINEMA and MELCOR computer codes.

Experimental and computational research of the containment passive emergency pressure decrease system in the floating NPP with reactor KLT-40S

A. Khizbullin, O. Tyurikov


On the basis of the KLT-40S floating nuclear power plant (NPP), engineers of the JSC “Afrikantov OKBM” have developed the passive emergency pressure decrease system in the containment, which does not require the operation of supply and control systems or staff involvement. Floating power unit is the only source of energy (electric, thermal) in remote areas with decentralized power supply and its passive system provides a high level of safety in case of accidents with loss of coolant of the primary circuit.

The containment passive emergency pressure suppression system of the floating NPP with reactor KLT-40S was studied at the large-scale SPOT test facility. The containment of SPOT test facility has a volume of ~59 m3 and height of about 8 m. Passive system cooling circuit was modeled in full scale. Heat exchanger-condenser (HX-C) of the system is located in the containment model. Water evaporation tank has a volume of 25 m3. The relation of system power in range of 120 – 600 kW as a function of containment pressure for various partial air pressures in range of 40 – 170 kPa (abs.) was obtained. Various operation modes of the passive system in case of primary coolant leakage into the containment were studied.

As a result of experimental study a validation of KUPOL-MT v 1.0 code has been performed. KUPOL-MT is a lumped parameter code (LP-code) and is used for analysis of heat and mass transfer inside the containment volume of small reactor plants.

Numerical simulations were performed using the KUPOL-MT code to substantiate safety of the KLT 40S floating NPP in accidents of type LOCA.

Development and Preliminary Assessment of the new ASYST - ISA Integral Analysis BEPU Code using the PBF SFD-ST Bundle Heating and Melting Experiment, a Typical BWR Under Fukushima-Daiichi-Accident-Like Thermal Hydraulic Conditions and PWR for a Steam Line Break in the Containment

Chris Merrill Allison1, Judith Hohorst1, Alex Ezzidi2, Masanori Naitoh2, Raimon Pericas3

1Innovative Systems Software; 2The Institute of Applied Energy; 3Energy Software Services

ASYST (Adaptive SYStem Thermal-hydraulics) - ISA (Integral Simulation and Analysis) is a new code being developed jointly under the direction of the organizations noted above that combines the capabilities of SCDAPSIM and SAMPSON. The thermal hydraulic module, ASYST-THA, replaces the original US NRC-developed RELAP5 code used in RELAP/SCDAPSIM/MOD3.x and THA used in SAMPSON with new system level hydrodynamic options that include multi-dimensional, multi fluid models originally developed by ISS and IAE. The ASYST reactor-specific modeling options include modules describing the behavior of (a) the core/fuel assembly structures, (b) late phase debris/melt relocation, (c) the containment including melt spreading and molten core-concrete interactions, and (d) fission product release and transport. The core/fuel assembly behavior module uses derivatives of the SCDAPSIM/MOD3.x models and correlations while the late phase debris/melt relocation module uses a combination of SCDAPSIM/MOD3.x (2D based) models and SAMPSON MCRA, DCA, DSA (3D-based) models. The fission product release and transport module uses combinations of models from SCDAPSIM/MOD3.x and SAMPSON. The core-concrete interaction module uses a SCDAPSIM-based porous media model in combination with SAMPSON’s Debris-Concrete Interaction (DCRA) models and correlations. The reactor vessel, reactor coolant system and containment thermal hydraulic behavior is described by ASYST-THA in combination with the SAMPSON hydrogen combustion, hydrogen detonation and steam explosion modules, HYNA, DDOC and VESUVIUS, respectively. Funding and additional technical contributions for the development of ASYST comes from the contributors to the international SCDAP Development and Training Program (SDTP).

The paper provides a brief overview and the status of the development of the ASYST code along with a preliminary assessment of the code using the PBF SFD-ST in-pile bundle heating and melting experiment along with benchmark calculations for a representative station blackout transient calculation for a typical BWR under Fukushima-like thermal hydraulic conditions and a PWR for a steam line break in the containment.

Evaluation of Emergency Planning Zone for Small Modular Reactor, SMART

S.B. Kim, S.J. Han, R.J. Park, H.O. Kang

Korea Atomic Energy Research Institute

Emergency Preparedness and Response (EPR) is the last barrier in nuclear safety to protect the safety of the public and protect the environment. Emergency Planning Zone ( EPZ) refers to the area that has been set up around a nuclear power plant in advance to formulate contingency plan and make EPR to protect the public in a timely and effective manner in the event of an accident in nuclear power plant. IAEA issued the safety requirements GS-R-2 that require the specification of off-site emergency zones for which arrangement shall be made of taking urgent protective actions. In Korea, the NSSC(Nuclear Safety and Security Commission) issued emergency preparedness guideline in 1980 after TMI-2 accident. The EPZ for the nuclear power plant was specified as an 8-10 km from the site and protective actions are required to be implemented in this area. After the Fukushima, NSSC revised the guidelines and considered introducing the concept of PAZ (Precautionary Action Zone) and UPZ (Urgent Protective Action Planning Zone), based on the IAEA safety requirements and guidelines.

SMART (System-integrated Modular Advanced ReacTor) is a light water Small Modular Reactor (SMR) with a rated power of 360 MWt, developed in KAERI. and recently PPE (Pre-Project Engineering) has been finished with cooperation with K.A. CARE in Saudi-Arabia. For SMR designs employing novel features and technology, there is a need to consider a scalable EPZ based on more mechanistic methods, along with regulation, protection strategy, dose criteria, policy factors, and public acceptance.

EPZ need to be developed based on more conservative accident conditions, accounting for events of very low probability and events not considered in the design.

Abstract continued in remarks to committee.

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