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Trivedi A.K.,Indian Institute of Technology Kanpur | Allison C.,Innovative Systems Software | Khanna A.,Indian Institute of Technology Kanpur | Munshi P.,Indian Institute of Technology Kanpur
Nuclear Engineering and Design | Year: 2016

The AP1000 is a Westinghouse 2-loop pressurized water reactor (PWR) with all emergency core cooling systems based on natural circulation. Its core design is very similar to a 3-loop PWR with 157 fuel assemblies. Westinghouse has reported their results of the safety analysis in its design control document (DCD) for a large break loss of coolant accident (LOCA) using WCOBRA/TRAC and for a small break LOCA using NOTRUMP. The current study involves the development of a representative RELAP5/SCDASIM model for AP1000 based on publically available data and its verification for a double ended cold leg (DECL) break in one of the cold legs in the loop containing core makeup tanks (CMT). The calculated RELAP5/SCDAPSIM results have been compared to publically available WCOBRA-TRAC and TRACE results of DECL break in AP1000. The objective of this study is to benchmark thermal hydraulic model for later severe accident analyses using the 2D SCDAP fuel rod component in place of the RELAP5 heat structures which currently represent the fuel rods. Results from this comparison provides sufficient confidence in the model which will be used for further studies such as a station blackout. The primary circuit pumps, pressurizer and steam generators (including the necessary secondary side) are modeled using RELAP5 components following all the necessary recommendations for nodalization. The core has been divided into 6 radial rings and 10 axial nodes. For the RELAP5 thermal hydraulic calculation, the six groups of fuel assemblies have been modeled as pipe components with equivalent flow areas. The fuel including the gap and cladding is modeled as a 1d heat structure. The final input deck achieved all steady state thermal hydraulic conditions as reported in the DCD. The analysis has been performed for the primary safety criteria, the peak clad temperature (PCT) as it is well established for a PWR that oxidation and hydrogen generation do not violate the safety criteria as long as PCT is under the safe limit. Results from this study show that the calculated value for the PCT is 1229 K well below the acceptance criteria of 1477 K, lower than the DCD value of 1311 K and higher than the TRACE value of 1186 K. © 2016 Elsevier B.V. All rights reserved. Source


Trivedi A.K.,Indian Institute of Technology Kanpur | Khanna A.,Indian Institute of Technology Kanpur | Munshi P.,Indian Institute of Technology Kanpur | Allison C.,Innovative Systems Software
Canadian Nuclear Society - 33rd Annual Conference of the Canadian Nuclear Society and 36th CNS/CNA Student Conference 2012: Building on Our Past... Building for the Future | Year: 2012

Station Black out (SBO) in Boiling Water Reactor of Laguna Verde Nuclear Power Plant (LVNPP) is analyzed. Each flow channel is modeled as a pipe divided into 14 nodes. Physical and thermodynamic properties of both recirculation loops are identical. There are four steam lines and they are considered separately in the model. SBO lead to loss of cooling in the core, severe damage of the fuel and hydrogen production. The maximum core surface temperature has gone up to 3000 K with total hydrogen accumulation about 430 kg. The maximum debris temperature in the lower plenum is 4233 K. Source


Trivedi A.K.,Indian Institute of Technology Kanpur | Sandeep K.T.,Indian Institute for Plasma Research | Allison C.,Innovative Systems Software | Khanna A.,Indian Institute of Technology Kanpur | And 3 more authors.
Fusion Engineering and Design | Year: 2014

The current work involves thermal hydraulic calculation of Lithium Lead Cooling System (LLCS) for the Indian test blanket module (TBM) for testing in International Thermonuclear Experimental reactor (ITER). It uses the RELAP portion of RELAP/SCDAPSIM/MOD4.0. Lithium-lead eutectic (LLE) has been used as multiplier, breeder and coolant in TBM. Thermodynamic and transport properties of the LLE have been incorporated into the code. The main focus of this study is to check the heat transfer capability of LLE as coolant for TBM system for steady state and the considered anticipated operational occurrences (AOO's), namely, loss of heat source, loss of primary flow and loss of secondary flow. The six heat transfer correlation (reported for liquid metals in the literature) has been tested for steady state analysis of LLCS loop and results are roughly same for all of them. A good agreement has been observed between the operating conditions of LLCS with those of RELAP5 calculations. Results from transient calculations show that a maximum temperature of 875 K is attained during a 300 s loss of primary flow (LLE). © 2014 Elsevier B.V. Source


Trivedi A.K.,Indian Institute of Technology Kanpur | Allison C.,Innovative Systems Software | Khanna A.,Indian Institute of Technology Kanpur | Munshi P.,Indian Institute of Technology Kanpur
Nuclear Engineering and Design | Year: 2014

This work analyses the influence of water addition in a boiling water reactor during a core isolation event in BWR. Injection of water is impacted by time as well as reactor vessel water level and it uses thermal hydraulic conditions representative of the reactor core isolation cooling (RCIC) system. A detailed RELAP/SCDAPSIM model of Laguna Verde BWR vessel and related reactor cooling system (provided by the Mexican Nuclear Regulatory Authority) has been used for this analysis. These calculations have been extended to the point of likely vessel failure or stable core cooling. They focus on initial heating and melting of the core where water addition is found to be most effective in limiting the extent of fuel melting. It also presents the results of a base case, a station blackout transient without water addition. These calculations have been carried out up to 5 h (after reactor scram) beyond the point of likely vessel failure. The maximum core surface temperature of 3042 K and hydrogen production of 367 kg is observed in this case. The importance of timing can be seen from 3500 s and 3700 s injection cases. One case leads to maximum core surface temperature of 1520 K with hydrogen production of 21 kg while the second case leads to temperature of 2940 K with hydrogen production of 193 kg. Temperatures (at the time of start of first injection) in both these cases are1371 K and 1590 K which explains this switching from stable core cooling to very high core surface temperature. © 2014 Elsevier B.V. Source


Allison C.,Innovative Systems Software | Hohorst J.,Innovative Systems Software | Venturi F.,University of Pisa | Forgione N.,University of Pisa
International Congress on Advances in Nuclear Power Plants, ICAPP 2014 | Year: 2014

Immediately after the accident at Fukushima Daiichi, Innovative Systems Software, ISS, and other members of the international SCDAP Development and Training Program started an assessment of the possible core/vessel damage states of representative LWRs during Fukushima-Daiichi-like Station Blackout transients. The initial 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. These initial calculations have been documented in the open literature. Since that initial assessment, additional calculations have been performed using the original Laguna Verde model as well as models with expanded nodalization to (a) better represent the turbine driven cooling system and (b) add a detailed 2D/3D thermal hydraulic nodalization of the containment. The expanded models have been used to evaluate (a) the heatup and potential failure of the RCS piping, and (b) the influence of containment heat up on the rate of core uncovery and heatup. This paper will describe the expanded models and the response of the vessel, RCS piping (including the turbine driven cooling systems), and containment for a range of Fukushima-Daiichi-like transients using the new models and then discuss the influence of the extended nodalization and calculations relative to the initial assessment performed in 2011. Source

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