RU2300819C2 - Method for checking shipboard nuclear power plant for tightness - Google Patents

Abstract

FIELD: nuclear power engineering; checking steam generators of shipboard pressurized-water reactor for tightness during hydraulic tests.

SUBSTANCE: proposed method for checking steam generator of shipboard power plant for tightness includes measurement of reference radionuclide content in primary and secondary coolant samples with reactor shut down followed by calculation of amount of leakage. Primary and secondary coolant samples are taken from each steam generator upon completion of hydraulic tests. Prior to do so, primary-circuit water is stirred. This method provides for unbiased check of steam generator for tightness at low estimation threshold and identifying leaks of any amount.

EFFECT: ability of monitoring each steam generator for condition.

3 cl, 1 tbl

Description

The invention relates to nuclear engineering and can be used to control the tightness of the steam generators (GH) of marine nuclear power plants (NPPs) with a pressurized water-water reactor during hydraulic tests.

A known method of controlling the density of ship equipment at a stopped reactor by holding at a primary coolant pressure greater than the working one. At the same time, “omissions” of the test medium (leakage of the primary coolant) are controlled. The degree of tightness or density of the GHG is determined by the amount of leakage of the coolant of the primary circuit [1, 2]. In practice, the presence and magnitude of coolant leaks is determined by the pressure drop of the primary circuit medium during hydraulic tests. However, in the presence of micro-leaks, the pressure drop is insignificant, and it can be noticed only with prolonged exposure, which in practice is sometimes difficult to implement. On ship nuclear power plants, the effect of a pressure drop in the presence of leaks may not be recorded due to an increase in the temperature of the primary circuit medium due to the residual heat of the reactor core. To lower the detection threshold, a more sensitive leak detection method is needed. In addition, the method of monitoring GHG leaks during hydraulic tests should allow a comparison of the degree of tightness of GHGs at various stages of the nuclear power plant life cycle.

Closest to the technical nature of the proposed method is a method for monitoring the inter-circuit tightness of a ship nuclear power plant, including measuring the content of the reference radionuclide in samples of the primary coolant and in the secondary water on a stopped reactor [3].

The disadvantage of this method is that the total leakage of all steam generators is determined, and not each. In addition, due to the relatively large mass of water in the secondary circuit of the nuclear power plant (which is more than an order of magnitude greater than the mass of water in each GHG), it is not possible to quickly determine the leakage rate (kg / h) and, as a result, the lack of initial data for comparing the tightness state each GHG at all stages of the nuclear power plant life cycle.

The objective of the invention is to provide a method that allows you to quickly and objectively control the tightness (density) of each GHG after completion of hydraulic tests.

The technical result achieved during the implementation of the invention is the possibility of identifying the inter-circuit leakage of each GHG, including micro-leaks, which allows early diagnostics of the state of each GHG as part of the general monitoring of the technical condition of ship nuclear power plants.

To achieve the specified technical result, it is proposed to use a method based on measuring the content of the reference radionuclide in the samples of the primary coolant and in the water of the second circuit in a stopped reactor, followed by calculation of the leakage value.

Distinctive features of the proposed method is that the measurement of the reference radionuclide content in the primary coolant and in the water of the second circuit of each steam generator is carried out after hydraulic testing, leakage is calculated using a different formula, and long-lived radionuclides that are weakly adsorbed on surfaces are used as reference radionuclides loop equipment, for example, tritium, cesium-137, etc. The use of a long-lived and weakly sorbed radionuclide as a reference is necessary to exclude the introduction of corrections for decay and sorption losses when determining the leakage of a steam generator [3]. The choice of a reference radionuclide is determined by the radionuclide composition of the primary coolant before carrying out hydraulic tests.

The method is as follows. To determine the tightness of the steam generator after completion of hydraulic tests, samples of the coolant of the first circuit and water samples are taken from the second circuit of each SG after mixing. The mixing operation can be carried out by bubbling. If the GHG was drained during hydraulic tests, then the GHG in the second circuit is filled with distillate, which is thoroughly mixed. The samples determine the content of the reference radionuclide. The amount of leakage is determined by the formula:

Leak rate is calculated using the formula

Where:

Q is the amount of leakage, kg;

a ₁ - specific activity of the reference radionuclide in the primary coolant, Bq / kg;

and ₂ - specific activity of the reference radionuclide in the water of the second GHG loop, Bq / kg;

M ₂ - the mass of water in the second circuit of the GHG, kg (indicated in the technical documentation for the GHG);

g is the leakage rate, kg / h;

τ is the duration of hydraulic tests, h

In the proposed method, the lower threshold for determining the leakage rate is interconnected with the detection limit of the reference radionuclide in water samples of the second GHG circuit, which can be significantly reduced due to the use of more sensitive recording equipment or methods with preliminary concentration of the reference radionuclide in water samples of the second GHG circuit [4].

The advantages of the proposed method are illustrated by the following examples. According to published data, the minimum value of the leakage of the medium, fixed by the method according to the pressure drop, is 1.3 · 10 ^-2 cm ³ / s [1, 2, 6]. When conducting hydraulic tests (within 24 hours), the minimum controlled leakage will be 1.3 · 10 ^-3 · 3.6 · 10 ³ · 24 = 1.12 · 10 ³ cm ³ . For water, this corresponds to 1120 g.

The minimum controlled leakage, determined by the content of the reference radionuclide (tritium) in the secondary circuit [3], is: the specific activity of the reference radionuclide in the primary coolant

a ₁ = 3.7 · 10 ⁷ Bq / kg;

specific activity of the reference radionuclide in the water of the second GHG loop

a ₂ = 7.4 · 10 ² Bq / kg;

mass of water in the secondary circuit of a nuclear power plant

M ₂ = 18000 kg;

accept

M _pr / M ₂ = 0 Q = 7.4 · 10 ² · 18000 / 3.7 · 10 ⁷ = 3.6 · 10 ^-1 kg = 360 g.

In the calculations, a ₂ is the lower limit for determining the content of the reference radionuclide in the secondary media without prior concentration.

Minimum-controlled leakage during hydraulic tests, determined by the proposed method, is

- when using cesium-137 as a reference radionuclide

a ₁ = 1.85 · 10 ⁵ Bq / kg,

a ₂ = 7.4 Bq / kg;

take a lot of water in the GHG

M _2ng = 700 kg Q = 7.4 · 700 / 1.85 · 10 ⁵ = 28 · 10 ^-3 kg = 28 g.

- when using tritium as a reference radionuclide

a ₁ = 3.7 · 10 ⁷ Bq / kg,

a ₂ = 7.4 · 10 ² Bq / kg;

take a lot of water in the GHG

M _2ng = 700 kg Q = 7.4 · 10 ² · 700 / 3.7 · 10 ⁷ = 14 · 10 ^-3 kg = 14 g.

Table. The value of the minimally controlled leak, g Analogue Prototype The proposed method 1120 360 14-28

Sampling of media from the primary circuit and each steam generator is carried out according to the accepted sampling technology for a particular nuclear power plant. The measurement of the content of reference radionuclides in the selected samples is performed on commercially available instruments. To measure the tritium content, a radiometer is used, for example, RZHS-05. In order to lower the threshold for the determination of tritium in PG water, known methods of its concentration are used [7]. When using the TRIATHLER multifunction radiometer, preconcentration is not required. Without preconcentration, the cesium-137 content in the water sample from GHG (sample weight 0.5 kg) is measured on a serial spectrometer with a detector of the DGDK-50A type. When using the KRK-1 type radiometer as a measuring instrument, known methods with preliminary concentration are used [4].

Thus, in comparison with known methods, the proposed method provides operational and objective control of the tightness of each GHG during hydraulic tests of ship nuclear power plants and the identification of any leaks, including micro-leaks. This allows you to monitor the technical condition of the steam generator to exclude its sudden failure [5].

Information sources

1. Kuznetsov V.A. and other marine nuclear power plants. M .: Atomizdat, 1976, p. 323, 346-351.

2. Machines, mechanisms, steam boilers, vessels and ship apparatuses. Norms and rules of hydraulic and air tests. GOST 22161-76.

3. Bredikhin V.Ya., Rakov V.T., Zmitrodan A.A. A method for controlling the inter-circuit tightness of a ship’s nuclear power plant with a water coolant. RF patent No. 2203510, 2003, bull. No. 12.

4. Moskvin L.N., Gumerov M.F., Efimov A.A. and others. Methods of chemical and radiochemical control in nuclear energy. - M .: Energoatomizdat, 1989 .-- 264 p.

5. Shitilov S.A. The destruction of materials induced by environmental influences as a cause of engineering failure. - Nuclear technology abroad, 1997, No. 8, pp. 11-22.

6. Babkin V.T. and others. Tightness of motionless connections of hydraulic systems. - M.: Mechanical Engineering, 1977, p. 104-110.

7. Chebotina M.Ya., Nikolin O.A. The supply of tritium from the Beloyarsk NPP to the Olkhov swamp-river ecosystem and the Pyshma river // Questions of radiation safety. - Ozersk: PO "Mayak", 2004, No. 2, p. 42-48.

Claims (3)

1. A method for monitoring the tightness of the steam generator of a ship’s nuclear power plant, including measuring the content of the reference radionuclide in the samples of the coolant of the first and secondary circuits with a stopped reactor, followed by calculation of the leakage rate, characterized in that the samples of the coolant of the primary circuit and the secondary water are produced from each steam generator after completion of hydraulic tests, while pre-mixing the water in the second circuit of the steam generator. 2. The method according to claim 1, characterized in that the amount of leakage is determined by the formula

where Q is the amount of leakage, kg; and ₁ - specific activity of the reference radionuclide in the primary coolant, Bq / kg; and ₂ - specific activity of the reference radionuclide in the water of the second GHG loop, Bq / kg; M ₂ - the mass of water in the second circuit of the steam generator, kg; 3. The method according to claim 1, characterized in that as reference radionuclides use radionuclides with a long half-life and weakly sorbed on the surfaces of the secondary circuit.