JPH0570118B2 - - Google Patents

Description

[Detailed Description of the Invention] [Field of Application of the Invention] The present invention relates to a nuclear reactor structure, and in particular, to a nuclear reactor structure that can alleviate the occurrence of thermal stress caused by a temperature gradient in the axial direction of the reactor vessel near the liquid level of the reactor vessel, and , relates to a tank-type FBR reactor structure suitable for reducing the diameter of the reactor vessel.

[Background of the invention]

FIG. 6 is a longitudinal sectional view showing an example of a tank type FBR reactor structure. Typically, in a tank-type FBR reactor structure, a reactor vessel 1 is supported by welding to a roof slab 2, forming a closed space. A reactor core 4 is supported within the reactor vessel 1 by a core support structure member 3 . In addition, the roof slab 2 is equipped with an intermediate heat exchanger (hereinafter referred to as IHX) 5, which exchanges heat between the high temperature primary sodium heated in the reactor core 4 and secondary sodium, and an intermediate heat exchanger (hereinafter referred to as IHX) 5, which circulates the primary sodium in the reactor. circulation pump 6, control rod drive mechanism (not shown) that controls the output of the reactor core
The reactor is equipped with equipment such as a core upper mechanism 7 containing a core and a rotary plug 8 for positioning a fuel exchanger (not shown) at a predetermined position during fuel exchange.

The low-temperature primary sodium in the cold pool 9 is sucked into the circulation pump 6 suspended from the roof slab 2, pressurized, and led into the high-pressure pool 10. After that, it passes through the reactor core 4, reaches a high temperature, reaches the hot pool 11, flows into the IHX 5 through the opening in the hot pool 11 of the IHX 5 suspended from the roof slab 2, and flows into the IHX 5 through the heat transfer tube to the steam generator (not shown). ) exchanges heat with the circulating secondary sodium, returns to the cold pool 9 at a low temperature.
Above the high-temperature primary sodium in the hot pool 11, an alcon gas 12, which is an inert gas that does not react with sodium, is filled, filling the space between the sodium liquid level and the roof slab 2.

In such a reactor structure, the reactor vessel 1
Focusing on the temperature distribution, below the sodium liquid level, the temperature of the reactor vessel 1 is approximately the same as the sodium temperature, whereas above
For example, the temperature in the portion joined to the roof slab 2 is approximately room temperature. Therefore, a very large axial temperature gradient occurs in the reactor vessel 1 above the sodium liquid level, and a high thermal stress resulting from this occurs in the reactor vessel 1 near the sodium liquid level.

To prevent the occurrence of such high thermal stress,
Conventionally, for example, as shown in JP-A-53-67089 or JP-A-57-186193, reactor vessel 1 and high-temperature sodium are prevented from coming into contact with each other, and low-temperature sodium is brought into contact with each other. The high temperature level of the reactor vessel 1 is lowered to reduce the temperature gradient. In addition, in FIG. 6, 13 is a furnace wall cooling insulation structure.

Figure 7 is an explanatory diagram of a conventional furnace wall cooling insulation structure.
In FIG. 7, a double cylindrical container 19 is further provided inside the reactor vessel 1 to form a flow path for cooling the reactor vessel 1, and the space for this is This has prevented the reactor vessel 1 from becoming more compact.

Therefore, as shown in Figure 8, the hot pool 1
A vertical partition 14 serving as a heat insulating layer is provided between the high-temperature primary sodium 1 and the reactor vessel 1.
A bottom plate 17 having a sodium inlet/outlet hole 16 is provided between the reactor vessel 1 and the reactor vessel 1 to form an annulus space 18 in which low-temperature primary sodium accumulates, thereby alleviating the uniaxial temperature gradient of the reactor vessel 1. Ori,
A difference in liquid level between the liquid level of high temperature primary sodium in the hot pool 11 and the liquid level of low temperature primary sodium in the annulus space 18 is created by the circulation pump 6.

By the way, the sodium inlet/outlet hole 16 provided in the bottom plate 17 at the lower end of the vertical bulkhead 14 is normally shaped so that sodium is supplied into the annulus space 18, which is the space formed by the vertical bulkhead 14 and the reactor vessel 1. It was designed with the only constraint that it only needed to have an opening, and no other conditions were taken into account.

However, in the reactor wall protection device of such a tank-type FBR reactor structure, as shown in FIG. During a trip, the pressure difference between the liquid level of the low-temperature primary sodium in the space 18 disappears due to the stoppage of the circulation pump 6, and the liquid level of the low-temperature primary sodium rises in a short time. The axial temperature gradient becomes steeper at the position indicated by the circle in FIG. 9b during tripping than during normal operation. Therefore, there is a problem in that thermal stress is generated in the reactor core 4 and the reactor vessel 1 that supports the total sodium weight, which may threaten the structural integrity of the tank-type reactor structure.

[Purpose of the invention]

The present invention has been made in view of the above, and its purpose is to alleviate the thermal stress caused by the temperature gradient in the axial direction of the reactor vessel near the liquid surface of the reactor vessel, both during normal operation and during a trip. The object of the present invention is to provide a nuclear reactor structure in which the diameter of the reactor vessel can be reduced.

[Summary of the invention]

The present invention provides that the shape and dimensions of the low-temperature primary sodium inlet/outlet opening provided at the lower end of the bulkhead, which is a heat insulating layer of the reactor wall protection device provided near the liquid level of the reactor vessel in a tank-type FBR reactor structure, are In order to minimize the temperature gradient in the axial direction of the container, the low temperature primary sodium liquid level drops without much resistance, and
During tripping and shutdown, low temperature 1
This was done by focusing on the fact that it is desirable to have a shape and size that allows the liquid level of sodium to gradually rise.
A ₁ is the horizontal cross-sectional area of the annulus space, A ₂ is the sodium level difference between the hot pool and the above annulus space, H is the sodium level difference between the hot pool and the above annulus space, H is 1/2H is H ₀ , is the width of sodium temperature change during a reactor trip, ΔT, is the above annulus space The inflow velocity of sodium at the liquid level difference H ₀ to v ₀ ,
The density of sodium is ρ, the specific heat of sodium is C,
When the amount of heat released from the sodium in the annulus space to the hot pool and the atmosphere outside the reactor vessel during a reactor trip is Q ₁ and ₂ , respectively, A ₂ H ₀ /v ₀ A ₁ = ρCΔTHA ₂ /Q It is characterized by having a structure that satisfies ₁ + Q ₂ .

[Embodiments of the invention]

The embodiment of the present invention shown in FIG. 1 and the second embodiment will be described below.
This will be explained in detail using FIGS.

The outline of the reactor structure according to the present invention is the same as that shown in FIG. 6, but the reactor vessel 1 shown in FIG. The shape and size of the sodium inlet/outlet hole provided at the lower end of the vertical partition wall 14, which is a heat insulating layer provided between the hot primary sodium in the hot pool 11, has been devised.

FIG. 1 is a partial longitudinal sectional view showing an embodiment of the reactor wall protection device for a nuclear reactor structure according to the present invention. In Figure 1, 1 is a reactor vessel, 12 is argon gas as a cover gas, 11 is a hot pool containing high-temperature primary sodium, and 14 is a heat insulation provided between the high-temperature primary sodium and the reactor vessel 1. An annulus space 18 in which low-temperature primary sodium accumulates is formed in the space between the vertical partition wall 14 and the reactor vessel 1, and the bottom plate 17 at the lower end of the vertical partition wall 14 has a structure as shown in the figure. A sodium inlet/outlet hole (opening) 15 is provided.

Next, the principle of thermal stress reduction will be explained.

As previously explained with reference to FIG. 9, when the liquid level of low-temperature primary sodium in the annulus space 18 rises rapidly after a reactor trip, high thermal stress occurs in the reactor vessel 1 near the liquid level. cause However, as shown in Figure 2a, after the reactor trip, the liquid level of low-temperature primary sodium rises gradually, and the liquid level rises to maintain the temperature gradient in the reactor vessel 1 during rated operation. However, if the temperature in the annulus space 18 is lowered, high thermal stress will not occur after a reactor trip.

Therefore, as shown in FIG. 3, the inner and outer diameters of the vertical bulkhead 14 are D ₁ and D ₂ , the inner and outer diameters of the reactor vessel 1 are D ₃ and D ₄ , respectively, and the liquid level of high-temperature primary sodium in the hot pool 11 is Let H be the liquid level difference between the liquid level of low-temperature primary sodium in the annulus space 18 during rated operation, A ₁ be the opening cross-sectional area of the sodium inlet/outlet hole 15, and A ₂ be the horizontal cross-sectional area of the annulus space 18. . Furthermore, if the sodium flow rate at the sodium inlet/outlet hole 15 of low-temperature primary sodium flowing into the annulus space 18 immediately after a reactor trip is v, then the liquid level difference H is: H=ξv ₂ /2g...(1) Here , ξ: resistance coefficient at the sodium inlet/outlet hole 15 g: gravitational acceleration.

After the reactor trip, the liquid level difference H gradually decreases as time passes and eventually reaches zero. Therefore, as an average flow rate, when the liquid level difference _H0 is 1/2 of the liquid level difference during rated operation, the sodium inlet/outlet hole 15
When the sodium flow rate at is v ₀ , the following equation holds true, similar to equation (1): H ₀ =ξv ₀ ² /2g (2).

At this time, if the low temperature primary sodium flow rate flowing into the annulus space 18 is Q, then this flow rate Q
is expressed as the product of the inflow sodium flow rate v ₀ and the opening cross-sectional area A ₁ of the sodium inlet/outlet hole 15, and Q=v ₀ A ₁ (3).

Therefore, the time _t1 required for the sodium necessary to fill the annulus space 18 to flow into the annulus space 18 can be obtained by dividing the volume of the annulus space 18 by the flow rate of sodium flowing into the annulus space 18 per unit time. This can be expressed as t ₁ =A ₂ H ₀ /v ₀ A ₁ (4).

Next, let us consider the amount of heat escaping from the annulus space 18 into the inner hot pool 11 and to the outer side of the reactor vessel 1.

FIG. 4 is a diagram showing the flow of heat around the reactor vessel wall protection device. The heat transfer coefficients on the inner surface of the vertical bulkhead 14, inside the annulus space 18, and on the outer surface of the reactor vessel 1 are α ₁ , α ₂ , α ₃ , and the vertical bulkhead 14 , respectively.
and the thermal conductivity of the reactor vessel 1 are λ ₁ and
λ ₂ , the amount of heat escaping from the annulus space 18 to the inner hot pool 11 per unit time is Q ₁ , and the amount of heat escaping to the outside of the reactor vessel 1 is Q ₂ . In addition, the temperature difference between the low temperature primary sodium temperature in the annulus space 18 and the high temperature primary sodium temperature in the hot pool 11 is ΔT ₁ , and the temperature difference between the low temperature primary sodium temperature and the atmospheric gas temperature outside the reactor vessel 1 is defined as ΔT 1 . When ΔT _{2 is ΔT 2} , Q ₁ and Q ₂ can be respectively expressed by the following equations.

Q ₁ πHΔT ₁ /1/α ₁ D ₁ +1/α ₂ D ₂ +1/2λ ₁ lnD
₂ /D ₁ ...(5) Q ₂ =πHΔT ₂ /1/α ₂ D ₃ +1/α ₂ D ₄ +1/2λ ₂ l
nD ₄ /D ₃ ...(6) Therefore, the time required for the low temperature primary sodium temperature in the annulus space _{18 to decrease by ΔT is t 2 =} ρCΔTHA ₂ /Q ₁ +Q ₂ ... (7) Here, ρ is the density of sodium and C is the specific heat of sodium.

It is necessary to have a structure that satisfies this time t ₂ and the time t ₁ required for sodium to flow into and fill the annulus space 18, and in the present invention, the structure satisfies the following formula. .

A ₂ H ₀ /v ₀ A ₁ = ρCΔTHA ₂ /Q ₁ +Q ₂ ...(8) With this structure, while the sodium temperature decreases by ΔT, the liquid level rises by H, and after the reactor trip, annulus space 1 at each time of
The temperature gradient of the reactor vessel 1 near the sodium liquid level in the reactor vessel 8 does not change, and the thermal stress generated in the reactor vessel 1 can be minimized.

So far, we have discussed sodium liquid level fluctuations during reactor tripping, but next we will discuss startup time. At startup, there is a relationship in which the sodium temperature in the annulus space 18 decreases while the sodium temperature gradually increases, so as shown in FIG. 5, the temperature gradient in the reactor vessel 1 does not become very large. Therefore, there is essentially no need to control the liquid level, and no structural improvements are required.

Therefore, in the present invention, as shown in FIG. 1, the sodium inlet/outlet hole 15 has an inverted eight-shaped longitudinal cross-sectional structure, and according to this structure, low temperature primary sodium flows into the annulus space 18 during a reactor trip. The inflow flow rate of is kept low (because the contraction phenomenon occurs), and formula (8) can be satisfied,
The axial temperature gradient of the reactor vessel 1 in the vicinity of the liquid level of the reactor vessel 1 is suppressed, and the generation of thermal stress can be alleviated.

〔Effect of the invention〕

As explained above, according to the present invention, it is possible to suppress the rate of rise in the liquid level of low-temperature primary sodium in the annulus space between the circulation pump and the heat-insulating partition installed in the reactor vessel when the reactor trips. This has the effect that thermal stress caused by a temperature gradient in the axial direction of the reactor vessel near the liquid surface of the reactor vessel can be alleviated, and the diameter of the reactor vessel can be reduced.

[Brief explanation of the drawing]

FIG. 1 is a partial vertical sectional view showing an embodiment of the reactor wall protection device of the reactor structure of the present invention, and FIG. 2 is the temperature in the axial direction near the reactor vessel liquid level after a reactor trip in the case of the present invention. A diagram for explaining the changes, FIG. 3 is a dimensional explanatory diagram of the furnace wall protection device in FIG. 1, and FIG. 4 is an explanatory diagram of the flow of heat around the furnace wall protection device in FIG. 1.
Figure 5 is a diagram to explain the temperature change in the axial direction near the reactor vessel liquid level during startup, and Figure 6 is a diagram for tank type
A vertical cross-sectional view of the FBR reactor structure, Figure 7 is an explanatory diagram of a conventional reactor wall cooling and insulation structure, Figure 8 is a partial cross-sectional view of another conventional reactor wall protection device, and Figure 9 is a case according to Figure 8. FIG. 2 is a diagram for explaining temperature changes in the axial direction near the reactor vessel liquid level after a reactor trip. 1... Reactor vessel, 2... Roof slab, 4...
... Core, 5 ... Intermediate heat exchanger, 6 ... Circulation pump, 7 ... Core upper mechanism, 8 ... Rotating plug, 9
... Cold pool, 11 ... Hot pool, 1
4... Vertical partition wall, 15... Sodium inlet/outlet hole,
17...Bottom plate, 18...Annual space.

Claims (1)

[Scope of Claims] 1. A reactor vessel that houses a reactor core, primary cooling system equipment, and in-reactor equipment, and a heat insulating layer that thermally protects the reactor vessel wall disposed inside the reactor vessel. In a nuclear reactor equipped with a partition and an opening for supplying sodium in a cold pool from below to an annulus space formed between the partition and the reactor vessel, the opening is The cross-sectional area of the opening is A ₁ , and the horizontal cross-sectional area of the annulus space is
A ₂ , H is the sodium liquid level difference between the hot pool and the annulus space, H ₀ is 1/2H, ΔT is the sodium temperature change width when the reactor is tripped, and the liquid level difference in the annulus space is H ₀ . The inflow velocity of sodium is v ₀ , the density of sodium is δ, the specific heat of sodium is C, the amount of heat released from the sodium in the annulus space to the hot pool and the atmosphere outside the reactor vessel at the time of reactor trip is Q ₁ , respectively. A nuclear reactor structure characterized by having a structure satisfying A ₂ H ₀ /v ₀ A ₁ =ρCΔTHA ₂ /Q ₁ +Q ₂ when Q ₂ . 2. The nuclear reactor structure according to claim 1, wherein the vertical cross-sectional shape of the opening is an inverted figure-eight shape.