RU2352005C1 - Method of steam and gas generators control tightness of ship nuclear engine installation with water heat transfer medium under pressure - Google Patents

Abstract

FIELD: physics; nuclear power.

SUBSTANCE: invention is intended for the control of tightness of steam and gas generators of ship nuclear engine installation on the stopped reactor, both at stationary pressure, and at carrying out of hydraulic trials. The method of tightness control of steam and gas generators of the ship nuclear engine installation with the water heat transfer medium under pressure includes sampling from each prestressly drained steam and gas generator on the stopped reactor and measuring of the content of tritium with the subsequent calculation of leaking quantity. Samplings of air moisture are performed from the pipe system of each steam and gas generator. Then tritium content in air of the pipe system of each steam and gas generator is determined. Before sampling, additional removal of water from pipelines of a contour of circulation of the drained steam and gas generators is performed. Sampling of moisture of air and determination of tritium content in air is performed with an interval of 24-48 hours.

EFFECT: increase of sensitivity of control method, possibility of earlier detection of steam and gas generator leakiness and possibility of determination quantity of leaking of the heat transfer medium of the first contour in specific steam and gas generator.

2 cl

Description

The invention relates to the field of nuclear engineering and can be used to control the tightness of the steam generators of a ship’s nuclear power plant (NPP) with water under pressure.

A known method of controlling the tightness of the primary circuit for controlling the parameters of the coolant, based on the registration of changes in flow or pressure of the coolant that occur when a leak occurs [Design of nuclear reactors. I.Ya. Emelyanov, V.I. Mikhan, V.I. Solonin. M .: Energoizdat, 1982. - 400 p.]. The advantage of this method is the unambiguity of identifying the leakage of the primary circuit with large leaks, the disadvantage is the low sensitivity with small leaks.

The closest way to control the tightness of steam generators is a method based on measuring the content of reference radionuclides in samples of the primary coolant and secondary water with a stopped reactor, followed by calculation of the leakage value. Water sampling of the second circuit is made from each steam generator after completion of hydraulic tests, while mixing water from the side of the second circuit of each steam generator. Moreover, the minimum detectable leakage is 14-28 g [RF patent No. 2300819, published June 10, 2007, bull. No. 16]. One of the disadvantages of this method is the choice of ¹³⁷ Cs as one of the reference radionuclides for monitoring the tightness of steam generators. As you know, ¹³⁷ Cs is a fission product of nuclear fuel, and an increase in its content in the primary coolant indicates the beginning of depressurization of the claddings of fuel elements [Engineering. Encyclopedia. M38 Nuclear engineering. T IV-25, book. 2, M.: Engineering. 2005, - 944 p., Ill.]. From the practice of operating nuclear power plants for transport purposes, it is known that most of the active zones generate their energy resource without depressurization of the claddings of fuel rods or with minor defects in them, at which the level of ¹³⁷ Cs in the primary coolant does not reach that indicated in the prototype, especially under conditions of constant operation of ion exchange filters primary coolant that sorb ¹³⁷ Cs. The use of a tritium second circuit as a reference radionuclide in water also has a number of certain limitations associated with the dynamics of its accumulation in the primary coolant of transport nuclear power plants. The level of tritium activity in the primary coolant indicated in the prototype is achieved only after an energy production of more than 10% [Control of the inter-circuit density of a transport nuclear power plant with an aqueous coolant under pressure. Bredikhin V.Ya., Zmitrodan A.A. Atomic energy, vol. 98, issue 3, March 2005, pp. 170-175]. Therefore, it becomes obvious that when the sealed cladding of the fuel rods and the energy production is less than 10%, the measurement of the content of the reference radionuclides indicated in the prototype in the secondary water is uninformative and does not have any significant advantages. In addition, to control the tightness of each steam generator separately, the procedure of mixing water from the side of the second circuit in each steam generator separately is required, which is necessary for taking a representative sample of water, and this is a difficult technical task in the conditions of transport nuclear power plants.

The objective of the invention is to create a method that allows for highly sensitive control of the tightness of each individual steam generator of a ship’s nuclear power plant with a water coolant under pressure in the standby mode both at stationary pressure in the reactor and during hydraulic tests, regardless of the state of tightness of the cladding of fuel rods and active energy production zones.

The technical result achieved by the implementation of the invention is to increase the sensitivity of the control method and, as a consequence, to detect leaks earlier, as well as to determine the rate of leakage of the primary coolant in a particular steam generator.

The specified technical result is achieved by the fact that in the method for monitoring the tightness of steam generators of a ship nuclear power plant with a water coolant, including sampling from each previously drained steam generator at a stopped reactor and measuring tritium activity with subsequent calculation of the leakage value, a new solution to the problem is that sampling of moisture in the air is carried out from the pipe system of each steam generator, while before sampling, hydraulic gates are removed and tkov circulation water from the piping circuit, and the selection of air moisture samples and measuring the tritium activity produce 24-48 hours at a time interval. The removal of hydraulic locks is necessary to ensure unimpeded sampling of air moisture from the tube system of the steam generators, and the removal of water is necessary to prevent dilution of the tritium activity received in the event of a defect in the tube system of the steam generator with the primary coolant, and, as a result, to increase the sensitivity of the control method.

The method is as follows. The steam generators are drained, then the hydraulic gates and residual water are removed from the pipelines of the circulation circuit connecting the pipe system of the steam generators with the air space of the compartment by blowing with air or nitrogen. Moisture samples of air are taken from the pipe system of each steam generator. First, a control sample is taken to determine the initial tritium content in the air, then after 24-48 hours the air moisture is re-sampled and the tritium content in the air is determined. Sampling of air moisture from the pipe system of the steam generator is carried out by passing air through a desiccant (indicator silica gel) using an aspirator with an adjustable air flow rate. The aspirator is connected to the pipe fitting of each steam generator alternately. In order to prevent the creation of additional vacuum in the steam generator, its pipe system communicates with the air space of the compartment through the process fitting. To exclude the possibility of tritium-containing moisture entering from the compartment into the steam generator pipe system, a barrier filter with a desiccant is installed on the air connection. To determine the tritium content in the air of the pipe systems of the steam generator, the adsorption method of sampling air moisture is used. As a moisture adsorbent use indicator silica gel, placed in a transparent Plexiglas column. After sampling, the moisture from silica gel is isolated by thermal vacuum desorption. In the obtained moisture, the tritium activity is measured on a TRIATHLER radiometer, then according to the method [Belovodsky LF, Gaevoy VK, Grishmanovsky V.I. Tritium. M .: Energoizdat, 1985, 248 p.] Determine the tritium content in the air. The results of the control sample and the sample taken after 24-48 hours from each steam generator are compared. If an increase in the tritium content in the air exceeds the error of the measurement procedure, a conclusion is made about the leakage of the pipe system of the steam generator. The amount of leakage is estimated by the formula:

,

,

where G is the magnitude of the leakage of the primary coolant into the air of the pipe system of the steam generator, kg / h;

And ¹ _air - tritium activity in the air of the pipe system of the steam generator, control sample, Bq / m ³ ;

And ² _air - the activity of tritium in the air of the pipe system of the steam generator, a second test, Bq / m ³ ;

And _{I con} - tritium activity in the primary coolant, Bq / kg;

V is the air volume of the pipe system of the steam generator, m ³ (indicated in the technical documentation of the steam generator);

t is the time interval between sampling, h

For example, with tritium activity in the primary coolant at a level of 3.7 × 10 ⁴ Bq / kg, which is achieved with an energy production of less than 1%, the minimum leakage recorded will be about 0.2 g / h.

To confirm the possibility of monitoring the tightness of steam generators, full-scale tests were conducted at an experimental research nuclear power plant with a dosed "input" of the primary coolant into the pipe system of one of the steam generators. During the tests, a container filled with the aqueous phase of the primary coolant was connected to the technological connection of the air cavity of the pipe system of the steam generators with air. The results are presented in the table.

Sampling point The content of tritium in the air, Bq / m ³ Notes PG-4 - After "exposure" of each GHG
within ~ 48 hours
tritium activity in the coolant
I circuit was 10 ⁷ Bq / kg PG-2 - PG-3 - PG-1 - PG-4 27 The dosage of the coolant I circuit air in the compartment - Note: the sign (-) in the table. means that the measurement result is below the detection limit (10 Bq / m ³ ) methodology for determining the tritium content in the air.

Thus, in comparison with the known methods, the proposed method provides an increase in the control sensitivity and earlier detection of leakage of steam generators at a stopped reactor both at stationary pressure and during hydraulic tests, regardless of the state of tightness of the cladding of fuel rods and energy production of the core.

Claims (2)

1. A method of monitoring the tightness of the steam generators of a ship’s nuclear power plant with a pressurized water coolant, including sampling from each previously drained steam generator at a stopped reactor and measuring tritium content, followed by calculating the leakage rate, characterized in that air moisture samples are taken from each pipe system steam generator with subsequent determination of the tritium content in the air of the pipe system of each steam generator, and before sampling, an additional itelnoe removing water from the circulation loop piping drained steam generator, and the selection of air moisture samples and determining tritium content of the air is carried out at intervals of 24-48 hours. 2. The method according to claim 1, characterized in that the amount of leakage is calculated by the formula:

where G is the magnitude of the leakage of the primary coolant into the air of the pipe system of the steam generator, kg / h;
And ¹ _air - tritium activity in the air of the pipe system of the steam generator, control sample, Bq / m ³ ;
And ² _air - the activity of tritium in the air of the pipe system of the steam generator, a second test, Bq / m ³ ;
A _{I con} - tritium activity in the primary coolant, Bq / kg;
V is the air volume of the pipe system of the steam generator, m ³ ;
t is the time interval between sampling, h