Transient analysis following tripping of a primary circulating pump for 500-MWe PHWR power plant

Gaikwad, Avinash J.; Kumar, Rajesh; Vhora, S.F.; Chakraborty, Goutam; Raj, V. Venkat

doi:10.1109/tns.2003.809600

Cited by 9 publications

(4 citation statements)

References 3 publications

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“…It has been reported in recent literature that RELAP5/MOD3.2 is capable of simulating natural circulation phenomena [51]- [54]. The modeling philosophies were based on Dynna540 and Dyna220 codes which were developed in house [55]- [58]. RELAP5 has been extensively used to analyze accident scenarios and operational transients for research reactors, power reactors like PHWR, PWRs and BWRs.…”

Section: Main Heat Transport (Mht) System Of Ahwrmentioning

confidence: 99%

Effect of Loop Configuration on Steam Drum Level Control for a Multiple Drum Interconnected Loops Pressure Tube Type Boiling Water Reactor

Gaikwad

Vijayan

Iyer

et al. 2009

IEEE Trans. Nucl. Sci.

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Section: Main Heat Transport (Mht) System Of Ahwrmentioning

confidence: 99%

Effect of Loop Configuration on Steam Drum Level Control for a Multiple Drum Interconnected Loops Pressure Tube Type Boiling Water Reactor

Gaikwad

Vijayan

Iyer

et al. 2009

IEEE Trans. Nucl. Sci.

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“…Such study aims at analysing all the worst possible SBO scenarios and design verification to avoid severe accident conditions [8]. The thermal hydraulics modeling methodology and simulation philosophy of 700MWe PHWR for the present study are based on [9][10][11][12][13][14][15][16], though RELAP5/MOD3.2 code has been used here. Sensitivity studies were carried out to finalize the present nodalisation, which are not presented due to space limitation.…”

Section: Passive Decay Heat Removal System (Pdhrs) For 700 Mwe Phwrmentioning

confidence: 99%

Effect of Coolant Inventories and Parallel Loop Interconnections on the Natural Circulation in Various Heat Transport Systems of a Nuclear Power Plant during Station Blackout

Gaikwad

Vijayan

Bhartya³

et al. 2008

Science and Technology of Nuclear Installations

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Provision of passive means to reactor core decay heat removal enhances the nuclear power plant (NPP) safety and availability. In the earlier Indian pressurised heavy water reactors (IPHWRs), like the 220 MWe and the 540 MWe, crash cooldown from the steam generators (SGs) is resorted to mitigate consequences of station blackout (SBO). In the 700 MWe PHWR currently being designed an additional passive decay heat removal (PDHR) system is also incorporated to condense the steam generated in the boilers during a SBO. The sustainability of natural circulation in the various heat transport systems (i.e., primary heat transport (PHT), SGs, and PDHRs) under station blackout depends on the corresponding system's coolant inventories and the coolant circuit configurations (i.e., parallel paths and interconnections). On the primary side, the interconnection between the two primary loops plays an important role to sustain the natural circulation heat removal. On the secondary side, the steam lines interconnections and the initial inventory in the SGs prior to cooldown, that is, hooking up of the PDHRs are very important. This paper attempts to open up discussions on the concept and the core issues associated with passive systems which can provide continued heat sink during such accident scenarios. The discussions would include the criteria for design, and performance of such concepts already implemented and proposes schemes to be implemented in the proposed 700 MWe IPHWR. The designer feedbacks generated, and critical examination of performance analysis results for the added passive system to the existing generation II & III reactors will help ascertaining that these safety systems/inventories in fact perform in sustaining decay heat removal and augmenting safety.

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“…The detailed mathematical models for the PHT system and other major systems [1][2][3][4][5][6][7][8][9][10][11][12][13][14][15] are not described here because of space limitation. The pressurizer and the SG models are described in this section.…”

Section: Mathematical Modelmentioning

confidence: 99%

Selection of a steam generator pressure control program for a 500 MWe pressurized heavy water reactor power plant through transient analysis

Gaikwad

Kumar

Vhora³

et al. 2003

Proceedings of the Institution of Mechanical Engineers, Part A:

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A personal computer based 500 MWe pressurized heavy water reactor (PHWR) power plant transient analysis code has been developed for analysing plant operational transients, for sorting out design issues and for taking decisions on new operating procedures and safety-related transient events. This code is based on the coupled solution of the conservation equations of mass, momentum and energy and the equation of the fluid state for simulation of process transients/dynamics. It includes mathematical models for all the major equipment/subsystems, i.e. the reactor core heat transfer, steam generators (SGs), surge tank (pressurizer), bleed condenser, etc., of the 500 MWe PHWR power plant. All the associated control systems on the primary and secondary sides are also incorporated. These codes are also validated against the available plant data. This code was also used to resolve the issues related to SG pressure control program selection. Two fundamental types of SG pressure control program are available for nuclear power plants, i.e. the constant boiler pressure program (CBPP) and the variable boiler pressure program (VBPP) for power manoeuvring from zero to 100 per cent of full power (FP), and vice versa. With CBPP the secondary steam cycle design and operation are simplified, at the cost of more load on the primary heat transport (PHT) pressure control system, whereas VBPP reduces the load on the PHT pressure controller. Parametric studies were carried out by taking into account the reactor trip, turbine trip and the IRV sizing design basis transient for comparison of these SG pressure control programs, i.e. (1) the earlier old design VBPP with constant PHT temperature 25 per cent of FP onwards, (2) CBPP and (3) VBPP similar to the 220 MWe PHWR SGPC program. After examining the two extremes in cases 1 and 2, the third scheme similar to the 220 MWe boiler pressure control (BPC) program was found to be most suitable. Schemes 2 and 3 put more load on the PHT pressure control and make use of the capacities enhanced by the pressurizer. The dip in the pressurizer level is acceptable in the CBPP, i.e. in the second case also. All the pros and cons of the selection of a BPC program were clearly quantified and resolved by considering a wide range of transients. All the related transient studies are discussed in this paper.

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Transient analysis following tripping of a primary circulating pump for 500-MWe PHWR power plant

Cited by 9 publications

References 3 publications

Effect of Loop Configuration on Steam Drum Level Control for a Multiple Drum Interconnected Loops Pressure Tube Type Boiling Water Reactor

Effect of Loop Configuration on Steam Drum Level Control for a Multiple Drum Interconnected Loops Pressure Tube Type Boiling Water Reactor

Effect of Coolant Inventories and Parallel Loop Interconnections on the Natural Circulation in Various Heat Transport Systems of a Nuclear Power Plant during Station Blackout

Selection of a steam generator pressure control program for a 500 MWe pressurized heavy water reactor power plant through transient analysis

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