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Determination of the Leakage Area of Radioactive Nitrogen 16N in Steam Generators in Reactors of KLT-40 Type

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Abstract

The paper studies a primary-to-secondary leak of nitrogen radionuclide 16N (T1/2 = 7.11 s, Eγ, max = 6.134 MeV, \({{{v}}_{{\gamma ,\max }}}\) = 69%) through a steam generator in a KLT-40 type reactor (used in ice breakers and floating power units, FPUs) with an ingress of water under pressure \({{P}_{{v}}}\) and temperature \({{T}_{{v}}}\) heated by a follow-up radioactive steam generation which is released under high pressure Pp through a helical steam duct of the steam generator. The content of the specified radionuclide in steam can be found and estimated by use of methods of gamma spectrometry and measurement of γ-activity concentration of steam and γ-radiation dose rate through the use of a simple physical and mathematical model making it possible to specify the reason and to determine the area of a leakage on a helical steam duct. The paper specifies the main areas in the structure of steam generators where radiation characteristics may be measured and their assessment techniques may be applied.

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Notes

  1. KLT-40—pressurized water nuclear reactor, developed at OKBM Afrikantov. Manufactured at the Nizhny Novgorod Machine-Building Plant. This type of reactor is widely used on icebreakers and floating power units (FPUs).

  2. This type of steam generator is typical for KLT-40 reactors and other pressurized water nuclear reactors [6].

  3. The speed of steam at its exit from the steam pipeline at the indicated values of temperature and pressure in Fig. 1, according to estimates, is ~600 m/s, and the average is 566 m/s. At such speeds of steam exit to the turbine rotor blades, a very high radial speed of the turbine would develop, which would entail a significant increase in centrifugal force, which would lead to a colossal increase in stresses in the turbine disk and especially in the rotor blades, which would lead to an excess of the allowable stresses of these structures and, as a result, to their destruction. These possible effects lead to the need to reduce the pressure and temperature of the steam by using appropriate technical solutions. https://tesiaes.ru/?p=8414 [in Russian].

  4. If the water-steam pipeline is made in the form of a spiral with radius Rsp with step hsp, then, with the length of the water–steam boundary region equal to ΔL0, the area of this region Sbpr is defined by the expression Sbor = (ΔL0/hsp) × 4π2Rsp(Rin + Rex)/2, where Rin and Rex are the inner and outer radii of the water-steam pipeline, respectively.

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Authors and Affiliations

  1. National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), 115409, Moscow, Russia

    A. P. Elokhin

  2. JSC Specialized Scientific and Research Institute of Instrumentation (SNIIP), 123060, Moscow, Russia

    S. N. Fedorchenko

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  1. A. P. Elokhin
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  2. S. N. Fedorchenko
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Correspondence to A. P. Elokhin.

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Elokhin, A.P., Fedorchenko, S.N. Determination of the Leakage Area of Radioactive Nitrogen 16N in Steam Generators in Reactors of KLT-40 Type. Phys. Atom. Nuclei 85 (Suppl 2), S42–S49 (2022). https://doi.org/10.1134/S106377882214006X

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  • DOI: https://doi.org/10.1134/S106377882214006X

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