JPS62273494A - Primary hydraulic testing method of boiling water type reactor - Google Patents

Abstract

(57) [Summary] This bulletin contains application data before electronic filing, so abstract data is not recorded.

Description

Detailed Description of the Invention 3. Detailed Description of the Invention [Objective of the Invention 1 <A Field of Application of Tojo] The present invention is directed to the reactor pressure vessel and reactor surroundings of a boiling water reactor (hereinafter referred to as BwR). Concerning primary water pressure test method for piping.

(Prior Art) A conventional example will be explained with reference to FIG. FIG. 6 is a system diagram illustrating the actual method for testing the hydraulic pressure of a BWR reactor pressure vessel, and reference numeral 1 in the figure indicates the reactor pressure vessel. A coolant and a reactor core (not shown) are housed within the reactor pressure vessel 1. The reactor pressure vessel 1 has reactor piping 2
However, a reactor piping valve 2a is also connected.

Generally, in the construction of a BWR plant, a primary water pressure test is performed after installing the reactor pressure vessel 1, the reactor piping 2 in its vicinity, the reactor piping valve 2a, etc. This is because a test run of the recirculation system (PLR system) etc. will be carried out afterwards, and it is necessary to conduct a pressure test on the reactor pressure vessel 1 etc. before that. During this secondary water pressure test, the reactor pressure vessel 1, the reactor piping 2, the reactor piping valve 28, etc. are heated to a predetermined temperature and then pressurized to a predetermined pressure. Here, the predetermined temperature is the minimum operating temperature of each structure, equipment, and piping with a certain margin (accounting for the temperature drop until the actual inspection). For example, when a test is conducted at t'C, 33°C is added to that t°C (t+33)'C.

The minimum operating temperature of the reactor pressure vessel 1 is 21°C, and the temperature of the R low side of the reactor piping 2 is 38°C. The equipment that performs this function will be explained below.

A temporary tank 3 is installed outside the reactor building (not shown), and steam for heating the reactor pressure vessel 1 is supplied to the temporary tank 3 via a steam line 4. The reference numeral 5 in the figure is a temporary circulation pump, and there are three VI-installed circulation pumps 5.
are installed in parallel.

This temporary circulation pump 5 connects the temporary circulation line 6 and PLR.
It is supplied into the reactor pressure vessel 1 via the discharge pipe 8 of the pump 7 . This reduces the water inside the reactor pressure vessel 1 to 15
Raise the temperature from ℃ (room temperature) to 60℃. This 60° C. is the predetermined temperature mentioned above. Then, it is returned to the temporary tank 3 via a temporary circulation line 10 in which a circulation line gate valve 9 is inserted in the reactor pressure vessel 1 . Such circulation raises the temperature of the reactor pressure vessel 1. Note that the reference numeral 6a in the figure is a temporary gate valve.

Further, the temperature of the reactor piping 2 connected to the reactor pressure vessel 1 is raised in the following manner. That is, the reactor piping valve 2a is opened, and the water in the reactor pressure vessel 1 is drained into the temporary tank 3 through the valve drain line 11 with the drain valve 11a inserted, the temporary line 12, and the temporary circulation line 10. By returning the reactor pressure vessel 1, the temperature is raised to 60° C. as in the reactor pressure vessel 1. As mentioned above, the minimum operating temperature of the reactor piping 2 is 38°C, and the temperature is raised to 60°C within a predetermined period (e.g., after several days), taking into consideration the large temperature drop at the end of the piping. (until the inspection date) to maintain humidity. In some cases, temporary heaters are also installed to raise the temperature.

Next, regarding pressurization, after the reactor pressure vessel 1 and the reactor piping 2 are heated, the circulation line gate valve 13 inserted in the temporary circulation line 6, the reactor piping valve 2a, and the reactor The pressure vessel drain line valve 14 (this reactor pressure vessel drain line valve 14 is inserted into the reactor pressure vessel drain line 15) is closed, and the temporary pressurizing pump 16 is operated.

A suction pipe 17 of this temporary pressurizing pump 16 is branched and connected to the temporary circulation line 6. This temporary pressure pump 1
Temporary header 18, temporary line 19, PL
Pressurized steam is supplied into the reactor pressure vessel 1 via the R pump 7, the PLR discharge valve 20b, and the discharge pipe 8,
As a result, the inside of the reactor pressure vessel 1 is pressurized. Note that the reference numeral 20a in the figure is a suction valve. Further, whether or not the pressure has been increased to a predetermined pressure is monitored by a pressure-resistant pressure gauge device 21 attached to the upper part of the reactor pressure vessel 1. This pressure-resistant pressure gauge device 21 consists of two pressure gauges, of which the pressure gauge 218 is the official pressure gauge and the pressure gauge 21b is for confirmation. A predetermined pressure is applied through the above operations. Furthermore, all of the above equipment is temporary equipment, and before use, a pressure test will be conducted to confirm its soundness before use. Reference numeral 22 in the figure is an equipment sump, and this equipment sump 22 is provided with a sunbrine 23 from the temporary circulation line 6, as well as piping 24 branched and connected to the temporary circulation line 10. is installed. Also, all of the items indicated by the symbol @26 in the figure are on-off valves. It should be noted that all of the broken lines in FIG. 6 mean temporary equipment for this equipment.

According to the above configuration, there were the following problems.

(1) First, temporarily raise the temperature and pressurize the reactor pressure vessel 1, reactor piping 2, etc. Since it is configured to use the Q construction width, the temperature cannot be increased or increased until after the temporary equipment is installed.
Pressurization work could not be carried out, and at the same time, if electrical and instrumentation work around the reactor pressure vessel was to be carried out after the test was completed, it would only be possible after the above temporary equipment had been removed. This also applies to cases other than electrical and instrumentation work, where temporary piping for temporary equipment is installed from outdoors to indoors in the reactor pressure vessel 1, and if these are not removed, small piping or supports will be damaged. It is impossible to implement and the working conditions are poor.

In this way, the installation and removal of temporary equipment is BWR plant number one! The construction period was expected to be extended in the future.

(2) Next, when raising the temperature of the reactor pressure vessel 1 and the surroundings of the reactor, a large amount of steam is required to be used, which, combined with the above-mentioned temporary equipment, results in an increase in the cost required for plant construction.

(Problems to be Solved by the Invention) In the past, installation and removal of temporary equipment for heating and pressurizing the reactor pressure vessel and piping around the reactor was carried out during the plant construction period. With future extension,
The use of large amounts of steam as well as temporary equipment has led to an increase in costs, and the present invention was made based on this point, and its purpose is to shorten the construction period required for plant construction. It is an object of the present invention to provide a primary water pressure test method for a boiling water reactor, which can also reduce costs.

[Structure of the invention] (Means for solving the problem) That is, the primary water pressure test method for a boiling water reactor according to the invention of Ml is a method for testing the reactor pressure vessel and reactor surrounding piping of a boiling water reactor. The primary water pressure testing method for boiling water reactors, which involves pressure testing, is characterized by the fact that the pressure testing of the reactor pressure vessel and piping around the reactor is carried out using the equipment. be.

(Function) In other words, in the conventional hydraulic testing method, the temperature and pressure of the reactor pressure vessel and the piping around the reactor pressure vessel was raised and pressurized by installing special temporary equipment. This is done using this equipment.

(Embodiment) A first embodiment of the present invention will be described below with reference to FIGS. 1 and 2. It should be noted that the same parts as in the prior art are denoted by the same reference numerals and the explanation thereof will be omitted. In this embodiment, while the reactor pressure vessel 1 and the reactor piping 2 were previously raised and pressurized using temporary equipment, this was done using this equipment and some temporary equipment. The aim is to shorten the construction period and reduce costs. At this time, in this first example, the minimum operating temperature of the reactor piping 2 was lowered from the conventional 38°C to 10°C. This was achieved by lowering the material testing temperature that had been conventionally used, and by conducting a predetermined test at the lowered temperature, the desired characteristics were satisfied with Vf! It would be good if it were approved. In this example, as a result, the minimum operating temperature could be lowered to 10° C. as shown in (1) above. Therefore, the minimum operating temperature of the reactor pressure vessel 1 is 21°C as before, and that of the reactor piping 2 is 10°C. Therefore, in the temperature raising step, it was decided to raise the temperature to 26°C, which is 21°C plus 5°C.

The mark @101 in the figure is a condensate tank, and this condensate tank 1
01 is connected to a steam line 102. The condensate at a water temperature of 15°C stored in the condensate tank 101 is
By the MUWC (t! [water make-up water system) pump 103, M
UWC line 104, RHR (residual heat removal system) line 1
05, water is supplied into the reactor pressure vessel 1 via the PLR piping 8 and filled with water. Next, the RHR pump 106 is operated. In other words, the temperature of the condensate supplied into the reactor pressure vessel 1 is 15° C. as described above, and this temperature is 25° C. using the heat generated by the shaft power of the RHR pump 106.
The temperature is raised to 6°C. At that time, considering the heat radiation from the reactor pressure vessel 1, the concentration increase rate is about 1°C/hr, and therefore approximately 11 hours of operation is required to raise the concentration by 11°C (26°C - 15°C). Become. Further, the temperature drop after heating is about 0.1°C/40hr, and based on the data of the atmospheric temperature of about 10°C, it will not drop below 21°C if the period until the inspection mill is 3 days. The above is the 4-temperature operation.

In the figure, reference numeral 106a is an RHR heat exchanger, and reference numeral 106b is an RHR isolation valve.

Next is the pressurization work, and the water in the SLC tank 108 is distributed to the SLC using the 5LC (water-resistant injection system) pump 107.
109 into the reactor pressure vessel 1 from the bottom. As a result, the inside of the reactor pressure vessel 1 is pressurized. The above SLC piping 109 has a 5LC1llilli valve 109.
a inserts t'L T t, N le. Reactor pressure vessel 1
The maximum working pressure of is 87.9Kg/cd, and the pressure test is 109.875 K3/c, which is 1.25 times that.
tA, approximately 110 Kg/cd is required. However, the height of the reactor pressure vessel 1 is about 22 cm, so there is a pressure difference of about 2.2'J/cd between its upper and lower parts. Therefore, at the top, we further add 9/cd to 5 to make 1
The test should be carried out at a pressure of 15 to 9/cd. However, the SLC Bonn 7107 Teha is about 80 to 9/
The pressure can only be raised to crN\a, so a temporary pressurizing pump 110 (indicated by a broken line in the figure) is used.

With this temporary pressure pump 110, 80Kg/ci to 1
15 to 9/ci. In the figure, reference numeral 111 is a temporary header, and 112 is a temporary pressure line.

In this way, in the case of this embodiment, instead of using temporary equipment as in the past to heat up and pressurize the reactor pressure vessel 1, this equipment is used to perform the work. Temporary equipment is used for only part of the pressurization.

In the figure, reference numeral 112 is an RCIC (reactor isolation cooling system), and reference numeral 113 is a drain line. Further, reference numeral 114 is an on-off valve.

According to this embodiment, the following effects can be achieved.

(1) First, the work to raise the temperature and pressurize the reactor pressure vessel 1 is performed using a permanently installed surface, rather than using only temporary equipment as in the past. Therefore, installation and removal work of temporary equipment is significantly reduced, and as a result, the number of man-hours required for plant construction is significantly reduced.

In addition, pond work ('I
Whereas the start of 4-gas instrumentation work, installation of small piping, etc.) was delayed, this does not occur in this embodiment, and the plant construction period can be significantly shortened. Looking at FIG. 2, which compares the conventional case and the case of this embodiment, it is clearly understood that the construction period is shortened.

In Figure 2, AT is pneumatic, HT is pressure resistant (water is used)
, Inspection indicates the inspection respectively. Furthermore, ECC8 indicates the emergency core cooling system, CR indicates the control ring, and CRD indicates the IJ 10 rod drive system. As is clear from Figure 2, the construction period has been significantly shortened (specifically, about 14 days).

(2) Furthermore, workability has been improved as the number of temporary facilities has been significantly reduced, making it possible to improve safety.

(3) Unlike conventional methods, this method does not require large-scale temporary facilities or a large amount of steam, so it is extremely effective in reducing costs.

Next, a second embodiment will be described with reference to FIG.

This example uses SLC pump 17 at 115 Kfl/
Since it can be used up to cli, the use of the temporary pressure pump 109 in the first embodiment is unnecessary. The other configurations are the same as those of the first embodiment, and the explanation thereof will be omitted.

Therefore, the first fruit I7! ! In addition to producing the same effects as in the example, since the temporary equipment is not used at all, the plant construction period can be shortened more than in the first example, and other effects are great. This is shown in FIG. Like the above-mentioned Fig. 2, this Fig. 4 is a comparison between this embodiment and the conventional example. According to this, the plant construction period can be shortened by about 20 days, and the effect is greater than that of the above-mentioned first embodiment. It can be seen that it was the one that ruled the Qin Dynasty.

Fig. 5 shows a detailed comparison of the reactor pressure vessel hydraulic schedule for the first embodiment, the second embodiment, and the conventional case, and shows that the first embodiment has a shorter construction period than the conventional case. It can be seen that the construction period is significantly shortened, and that the second embodiment shortens the construction period even more.

Incidentally, in the first and second embodiments, pressurization is performed by an SLC system, but the present invention is not limited to this, and a CRD system may also be used. In this case, pressurization will be performed using driving water of the CRD system.

It is also being considered to further lower the minimum operating temperature of the reactor pressure vessel and the piping surrounding the reactor, which could significantly reduce or even eliminate the temperature raising process.

[Effect of the invention 1] As detailed above, according to the primary water pressure testing method for a 'faIli water reactor according to the present invention, temporary equipment is required only in some steps or not at all during the water pressure test. So, plant l! Significantly shorten the construction period,
Moreover, it is possible to reduce the number of man-hours required for construction, which is extremely effective in reducing costs and improving safety.

[Brief explanation of the drawing]

Fig. 1 is a system diagram showing the implementation of the first practical example of the present invention;
The figure shows a comparison of the process of the conventional example and the first embodiment, Fig. 3 is a system diagram showing the implementation of the second embodiment, and Fig. 4 shows a comparison of the process of the second embodiment and the conventional one. 5 is a diagram showing a comparison of the steps of the first embodiment and the second embodiment and the conventional method, and FIG. 6 is a system diagram showing the conventional example. 1... Reactor pressure vessel, 2... Nuclear reactor piping, 2
a... Nuclear reactor piping valve, 106... RHR system pump, 107... SLC system pump, 110... Temporary pressurizing pump.

Claims (3)

[Claims] (1) In the primary water pressure test method for a boiling water reactor, the pressure test of the reactor pressure vessel and the piping around the reactor is carried out. A primary water pressure test method for a boiling water nuclear reactor, characterized in that it is carried out using the equipment provided. (2) A heating process in which the residual heat removal system pump is operated to raise the temperature of the reactor pressure vessel and reactor piping to a predetermined temperature using the heat generated by its shaft power, and the boric acid water injection system or control rod drive a first pressurization step of pressurizing the reactor pressure vessel and the reactor surrounding piping to just below a predetermined pressure using a mechanical system;
The method according to claim 1, further comprising a second pressurizing step of pressurizing the reactor pressure vessel and the piping around the reactor from before a predetermined pressure to a predetermined pressure using a temporary pressurizing pump. Primary water pressure test method for water reactors. (3) A heating process in which the residual heat removal system pump is operated to raise the temperature of the reactor pressure vessel and reactor piping to a predetermined temperature using the heat generated by its shaft power, and the boric acid water injection system or control rod drive mechanism 2. The primary water pressure test method for a boiling water reactor according to claim 1, further comprising a step of pressurizing a reactor pressure vessel and piping around the reactor to a predetermined pressure by means of a system.