Development of a Computer Code, ONCESG, for the Thermal-Hydraulic Design of a Once-Through Steam Generator

Due to their advantages, such as compactness and high efficiency in heat transfer, helically coiled heat exchangers have been widely used by different type of nuclear power plants, especially by small and medium size reactors (SMRs). In order to analyze the thermal-hydraulic characteristics of a helical coiled once through steam generator (OTSG) for a small integral pressurized water reactor, a computer code is developed in this paper. The code is based on two-fluid model. The constitutive correlations are recommended based on the assessments with the compiled databases from the reviewed literatures. NUSOL SG is validated and verified against heat transfer in helical coiled tubes, and the calculation results agree well with the experiment data. The present study could provide references for the investigators to perform further investigations on the thermal hydraulic characteristics of helical coiled OTSGs.

Download Full-text

Thermal–hydraulic design and analysis code development for steam generator of CFR600

Annals of Nuclear Energy ◽

10.1016/j.anucene.2015.12.007 ◽

2016 ◽

Vol 90 ◽

pp. 256-263 ◽

Cited By ~ 7

Author(s):

Peiwei Sun ◽

Zi Wang ◽

Jianmin Zhang ◽

Guanghui Su

Keyword(s):

Steam Generator ◽

Hydraulic Design ◽

Code Development

Download Full-text

Development of one-dimensional computer code DESOPT for thermal hydraulic design of sodium-heated once through steam generators

International Journal of Nuclear Energy Science and Technology ◽

10.1504/ijnest.2010.030556 ◽

2010 ◽

Vol 5 (2) ◽

pp. 143 ◽

Cited By ~ 10

Author(s):

G. Vaidyanathan ◽

A.L. Kothandaraman ◽

L.S. Siva Kumar ◽

V. Vinod ◽

I.B. Noushad ◽

...

Keyword(s):

Computer Code ◽

Steam Generators ◽

One Dimensional ◽

Hydraulic Design

Download Full-text

Computational-Theoretical Analysis of the Distribution of the Thermohydraulic Parameters and the Concentration of Dissolved Impurities in a PGV-1000M Horizontal Steam Generator Performed Using the STEG Computer Code

Atomic Energy ◽

10.1023/b:aten.0000041208.05809.4a ◽

2004 ◽

Vol 96 (6) ◽

pp. 407-410

Author(s):

N. B. Trunov ◽

V. I. Melikhov ◽

O. I. Melikhov ◽

Yu. V. Parfenov

Keyword(s):

Theoretical Analysis ◽

Steam Generator ◽

Computer Code ◽

Horizontal Steam Generator

Download Full-text

Calculation of the Dose Load During Dismantling of Large Components in the Process of Decommissioning of Nuclear Installations

Volume 2: Facility Decontamination and Decommissioning; Environmental Remediation; Environmental Management/Public Involvement/Crosscutting Issues/Global Partnering ◽

10.1115/icem2013-96141 ◽

2013 ◽

Cited By ~ 1

Author(s):

Martin Hornacek ◽

Vladimir Necas ◽

Peter Bezak

Keyword(s):

Steam Generator ◽

Decay Time ◽

Nuclear Power ◽

Power Plants ◽

Main Part ◽

Computer Code ◽

Point Of View ◽

External Exposure ◽

Dose Load ◽

Type Reactor

The paper is focused on the decommissioning process from the point of view of steam generator dismantling. A brief description of the steam generator (used in nuclear power plants with VVER 440 type reactor) and used computer code VISIPLAN 3D ALARA are given. The main part deals with the created model and dismantling strategy together with variable input parameters — decay time and decontamination. The obtained results — external exposure of workers and the influence of time and pre-dismantling decontamination — are studied. Also detailed analyses of every dismantling step considered are presented.

Download Full-text