Development of a throttle and dampening device for the nuclear power plant of icebreaker LK-60

2021 ◽  
Vol 14 (2) ◽  
pp. 115-123
Author(s):  
E. A. Sukhorukova ◽  
S. P. Kolpakov

In the course of developing designs for mixing heat exchangers that operate on the principle of throttling the working medium on perforated grids, special attention is paid to ensuring the reliability of structures subject to erosive wear when subjected to dripping moisture and temperature stresses.JSC “NPO CKTI” has years of experience in the development of contact-type heat exchangers and was directly involved in the design of separate power plant equipment for LK-60, including a throttle and dampening device (TDD).It provides a description of the functional purpose of the TDD as part of the LK-60 nuclear power plant, the principle of operation and significant differences of the new design from those previously used. It is noted that while in previous designs the TDD included four columns connected in pairs, on LK-60 there are two columns located on top of the condenser. The TDD for LK-60 is designed to receive 132.5 t / h of steam of higher parameters than the previous generation designs intended to receive about 50 t / h of steam.The main technical solutions in the development of the design of the TDD are presented. The design provides access to the throttling lattices for diagnostics and their replacement if necessary which ensures a high degree of maintainability and reliability of the device. Perforation of the lattices arranged in series in the direction of the steam flow is made in such a way that the openings of the previous lattice, if possible, are not located opposite the openings of the subsequent lattice. The distance between the throttling lattices was taken from the conditions for ensuring the design course of the steam throttling process.Results are given of thermal and hydraulic calculations of the TDD. The calculation consists of two main parts. The first part includes thermal and hydraulic calculations with the determination of the degree of perforation of the lattices, the distribution of temperature and vapor pressure over the cross-sections of the TDD, etc. The second part contains the calculation of the cooling condensate injection nozzles.In the course of design studies, strength calculations were performed for all versions of TDD and individual parts. In addition, the nozzles underwent a full test cycle (determination of flow characteristics, water spray quality) in accordance with the test program.

1993 ◽  
Vol 640 (1-2) ◽  
pp. 371-378 ◽  
Author(s):  
Archava Siriraks ◽  
John Stillian ◽  
Dennis Bostic

2005 ◽  
Vol 113 (3) ◽  
pp. 308-313 ◽  
Author(s):  
Hyo-Joon Jeong ◽  
Eun-Han Kim ◽  
Kyung-Suk Suh ◽  
Won-Tae Hwang ◽  
Moon-Hee Han ◽  
...  

2016 ◽  
Vol 2016 ◽  
pp. 1-4 ◽  
Author(s):  
Mahdi Rezaeian ◽  
Jamshid Kamali

Due to high radioactivity and significant content of medium- and long-lived radionuclides, different operations with spent nuclear fuels (e.g., handling, transportation, and storage) shall be accompanied by suitable radiation protections. On the other hand, determination of radioactive source specification is the initial step for any radiation protection design. In this study, radioactive source specification of the spent fuels of Bushehr nuclear power plant, which is a VVER-1000 type pressurized water reactor, was determined. For the depletion and decay calculations, ORIGEN code was utilized. The results are presented for burnups of 30 to 49 GWd/MTHM and different cooling times up to 100 years. According to these results, total activity of a spent fuel assembly with initial enrichment of 3.92%, burnup of 49 GWd/MTHM, and cooling time of 3 years is 1.92 × 1016 Bq. The results can be utilized specifically in transportation/storage cask design for spent fuel management of Bushehr nuclear power plant.


2018 ◽  
Vol 1 (6) ◽  
pp. 177-184
Author(s):  
Son An Nguyen ◽  
Nguyen Trung Tran

In order to operate a nuclear power plant, ensuring safety is the most important factor. The function of safety rods are to shut down the reactor in case of emergency. The purpose of this paper to show the result of research and determine the value of safety rods SA, SB. Determination of the Boron concentration corresponding to each group of safety rods of OPR1000 nuclear reactor ensures the safely in the whole operation process. Experimental simulation is carried out in the system simulating core reactor OP1R1000 (CoSi OPR1000). The expermental result corresponds with the theoretic calculated result of Sa and Sb with 1500 pcm, 4000 pcm. The concentrations of Boron appropriately are 134 ppm and 284 ppm, respectively.


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