JPS6131994A - Fuel aggregate - 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] [Field of Application of the Invention] The present invention relates to a fuel assembly loaded into a nuclear reactor, and in particular, the present invention relates to a fuel assembly loaded in a nuclear reactor. The present invention relates to a fuel assembly suitable for increasing thermal margin and improving nuclear thermal-hydraulic stability.

[Background of the invention]

Nuclear reactor operation is subject to restrictions on linear power density, critical power ratio, and critical heat flux ratio. The former is imposed to prevent the fuel pellet from melting, and the latter two are imposed to prevent the fuel cladding from burning out. In nuclear reactor design, the core size is generally determined so that the linear power density satisfies a limit value when the total power is given.
Thereafter, the coolant flow rate and the core population enthalpy of the coolant are determined so that the critical power ratio and critical heat flux ratio are greater than or equal to the limiting values. Since nuclear reactors are generally operated at the above-mentioned total power, the linear power density limit can be complied with relatively easily. However, in recent years, there has been a growing demand for greater flexibility in reactor operation by extending the range of the design to reduce the coolant flow rate and change the core population enthalpy of the coolant. In response to this request, the limit output ratio and the limit heat flux ratio have small margins with respect to the limit values.

In addition, the operation of a nuclear reactor is carried out by a combination of nuclear phenomena and thermo-hydraulic phenomena, and the above-mentioned nuclear thermo-hydraulic stability refers to the current nuclear This is one indicator of whether or not the furnace is easy to maintain.
It is preferable for a nuclear reactor to always be operated in a nuclear-thermal-hydraulic stable state.During the operation of a nuclear reactor, if periodic fluctuations are forced due to disturbances, etc., the amplitude ratio of the output fluctuations is calculated as the reduction ratio (D). , R-, Decay Ratio), and the nuclear thermal-hydraulic stability in nuclear reactor operation is expressed by this decay ratio. When the width reduction ratio is smaller than 1.0, the output fluctuation gradually decreases and converges to a constant output, which means that the output is becoming nuclear-thermal-hydraulic stable. The closer it is to 0, the more stable it is.

A total of 49 fuel assemblies (7 x 7) are loaded into the reactor.
Or a total of 64 8 x 8 fuel rods (including one or two water rods that promote neutron debundling) and spacers that support the fuel rods with their longitudinal axes parallel to each other and space them apart. It consists of a channel box, etc. that surrounds the The coolant flows upward from the bottom of the fuel assembly as a single-phase flow, absorbs heat from the heat-generating parts of the fuel rods, causes boiling, and becomes a two-phase flow whose flow pattern is directed in the axial direction of the fuel rods. The flow changes sequentially into a bubble flow, a slag flow, and an annular flow, and flows out from the upper part of the fuel assembly. A total of 4 to 8 spacers are attached to the heat generating portion of the fuel rod at equal intervals.

As a prior art example of reducing the pressure loss of a spacer, the one described in Japanese Patent Application Laid-Open No. 55-107993 is known.

However, this technique does not recognize that the influence on the flow characteristics of the coolant, the thermal margin of the reactor, and the nuclear thermal hydraulic stability differs depending on the loading position of the spacer in the fuel assembly.

Therefore, in conventional fuel assemblies, nine identical spacers are attached in the direction of coolant flow. In such fuel assemblies, it is not taken into account that the distribution of flow rate and void fraction in the flow path cross section varies depending on the flow state of the coolant, such as single-phase flow, bubble flow, slug flow, and annular flow. Ta. Therefore, the thermal margin is small due to the distribution of the coolant flow rate, the distribution of the boiling start point, and the complicated flow of the gas phase and liquid phase caused by these. Furthermore, when the conventional fuel assemblies are loaded into a nuclear reactor, if the output changes due to a disturbance such as pressure, the position of the boiling start point in each assembly crosses the spacer, and the pressure loss changes. Due to this change in pressure loss,
The influence of disturbance becomes difficult to attenuate, as the coolant flow rate of each fuel assembly changes, and the accompanying change in the amount of voids affects core characteristics. The nuclear thermal-hydraulic stability margin of the Tamashi reactor was shrinking.

In order to improve nuclear thermal-hydraulic stability with increased thermal margin in the conventional fuel assembly described above, the arrangement spacing and enrichment distribution of fuel rods must be changed to reduce the coolant flow rate and voids. It is necessary to change the distribution of the amount of coolant, to reduce changes in the flow rate of coolant in each fuel assembly even when disturbances are applied, and to reduce changes in the nuclear characteristics of the reactor core. However, it is economically undesirable to drastically change the design of such a fuel assembly.

[Purpose of the invention]

In view of the above-mentioned circumstances, the present invention increases the thermal margin of a nuclear reactor without significantly changing the design of the fuel assembly.
The object is to provide a fuel assembly suitable for improving stability.

[Summary of the invention]

In order to achieve the above object, in the fuel assembly of the present invention loaded in a nuclear reactor where heat is removed by the boiling heat transfer phenomenon of the coolant, (1) 3/4 to 4/4 of the total length from the lower end of the fuel rod is The spacer installed in the region where the two-phase flow between the two-phase flow becomes an annular flow is configured so that the pressure loss is large at the center of the coolant flow path cross section, and the pressure loss is small at the periphery. ) The spacer installed in the non-boiling region between 0/6 and 1/6 of the total length from the bottom end of the fuel rod causes a large pressure loss in the center of the coolant flow path cross section, and a small pressure loss in the periphery. Configure it like this.

(3) The spacer installed in the region where the flow pattern of the two-phase flow is bubble flow or slug flow between 1/6 and 3/4 of the total length from the bottom end of the fuel rod is different from the spacer installed in other regions. Constructed to reduce pressure loss.

In addition, in order to reduce the pressure loss around the cross section of the coolant flow path in the spacer described in (1) and (2), it is necessary to make the lower or upper end of the structural material blade-shaped, or to It is sufficient to make the projected area of the structural material on the periphery smaller than that at the center.

The reason for configuring the fuel assembly in this way is as follows.

First, in (1) above, when the flow mode of a two-phase flow is an annular flow, the gas phase and liquid phase exist separately, so compared to bubble flow or slug flow, where the gas phase and liquid phase exist as a mixture. , the effect of the gas phase disrupting the flow is as small as about 1/4 or less, and the lateral flow of the coolant is reduced. As a result, the coolant flow rate decreases in the periphery of the flow path cross section, and the thermal margin becomes smaller. Therefore, by reducing the pressure loss in this peripheral area, we intended to increase the coolant flow rate and increase the thermal margin. It is something.

Next, in the single-phase flow region near the fuel assembly population, there is no gas phase, so compared to two-phase flows such as bubble flow and slug flow,
Since the lateral flow is small, the coolant flow rate distribution within the channel cross section strongly depends on the pressure drop distribution.

Since a channel box exists at the periphery of the flow path cross section, pressure loss increases and the coolant flow rate decreases. Therefore, by adopting the configuration (2) above, the pressure loss in the peripheral area is reduced, the coolant flow rate is increased, and the complicated movement of the gas phase and liquid phase in the downstream boiling region is reduced, thereby improving thermal efficiency. Increase your margin.

Furthermore, the reason for the configuration as described in (3) above is to improve the nuclear thermal hydraulic stability of the reactor core. '19 By making the pressure drop coefficient smaller in the area including the boiling start point of the coolant than in other areas, even if the boiling start position changes due to disturbance and crosses the spacer, the change in pressure loss due to the spacer will remain unchanged. Since it is smaller than the pressure loss in the area, the pressure loss from the fuel assembly to the outlet can be prevented from changing drastically. Therefore, changes in the coolant flow rate of each fuel assembly are reduced, and changes in the amount of voids and core characteristics are also reduced, and the effects of disturbances are quickly attenuated.
The nuclear thermal hydraulic stability of the reactor core can be improved. From 176 to 3/3 of the total length from the bottom end of the fuel rod shown in (3) above
In the region where the flow pattern of the two-phase flow between 4 and 4 is bubble flow or slug flow, the gas phase has a large effect of disrupting the flow, and the coolant flow rate tends to be uniformly distributed within the cross section of the flow path. Even if the pressure loss distribution in the cross section of the coolant flow path is changed using a spacer, the thermal margin does not become much larger.

To improve nuclear thermal and hydraulic stability of nuclear reactors.

One method is to reduce the pressure drop within the fuel assembly to ensure sufficient flow even if the pump driving the flow of coolant stops. For this purpose, it is effective to reduce the pressure loss of the spacer as much as possible, and see (3) above.
The configuration of the fuel assembly is consistent with this method. Therefore, it is considered that it is not good to increase the pressure loss due to the population of the fuel assembly as described in (2) above. However, the above (2
), the pressure loss is slightly larger in the single-phase flow region, and even when a disturbance is applied, the thermal-hydraulic response in the single-phase flow region and the two-phase flow region is small, and the nuclear heat of the reactor is reduced. Hydraulic stability is improved.

After the coolant flows into the fuel assembly as a single-phase flow, it boils along the flow direction, and the flow style changes to bubble flow, slug flow, and annular flow. This flow pattern is determined by the coolant flow rate and void ratio, but in normal operation of a boiling water reactor, a bubble flow or slug flow occurs between 1/6 and 374 of the total length of the fuel assembly. It can be considered that the flow becomes a circular flow between 3/4 and 474.

[Embodiments of the invention]

A first embodiment of the present invention will be described in detail below with reference to FIG. This example is a BW with an electrical output of 1100 MWe class.
The present invention is applied to a fuel assembly 1 in which approximately 800 fuel assemblies are loaded into a fuel assembly R. The fuel rods 2 are supported with their longitudinal axes parallel and spaced apart by spacers 3;
It is surrounded by a channel box 4. The coolant flows from the artificial orifice 5, generates heat from the fuel rod heat generating part 6, causes boiling, and flows out from the upper part as a two-phase flow. In this fuel assembly l, eight spacers are installed at intervals of about 0.4 m over the total length of the fuel rod heat generating part 6 of about 3.7 m, and from the lower end 0/6 to 1/6 of the total length. One spacer 301 is installed in the non-boiling region between the two. This spacer 301 is divided into two areas, a center part 8 away from the channel box 4 and a peripheral part 9 near the channel box, by a dashed line 7, and in the center part 8, the upper and lower ends of the spacer structural material are not blade-shaped. The -A' cross section is rectangular lO, and the structural members of the peripheral portion 9 have blade-like upper and lower ends so that the B-B' cross section has a streamlined shape 11 (an example of the configuration of (2) above). In the spacer 301 configured in this way, the pressure loss is greater at the center 8 depending on whether the upper and lower ends of the structural material are blade-shaped or not.
and peripheral part 9: l: 0.66, which is 34% smaller in the peripheral part. In a fuel assembly in which this spacer 301 is installed, the increase in pressure loss at the periphery due to the channel box 4 is offset by making the upper and lower ends of the structural material blade-shaped, so the pressure loss at the center and periphery of the flow path cross section is offset. are approximately equal, and the coolant flow rate is approximately uniform in the non-boiling part. As a result, the boiling start points of many fuel rods in the fuel assembly are approximately the same, and the movement of the gas phase and liquid phase in the boiling portion is reduced. In connection with this, the coolant liquid film on the fuel cladding tube is difficult to peel off, and the liquid film sufficiently removes heat, increasing the thermal margin. Five spacers 302 are installed in a region where the flow pattern of the two-phase flow is bubble flow or slug flow between 1/6 and 374 of the total length from the lower end of the fuel rod. This spacer 30
In No. 2, the upper and lower ends of the structure are shaped like blades so that the c-c' cross section is streamlined (an example of the construction in (3) above). In the spacer 302 configured in this manner, the structural material has a streamlined shape, so that the pressure loss coefficient is reduced. This spacer 3
In the fuel gathering point body in which 02 is installed, the output changes due to disturbance, and even if the boiling noise device crosses the spacer 302, the change in pressure loss can be reduced. Therefore, the change in pressure loss from the population of the fuel assembly to the outlet is small, and the change in the flow rate of each fuel assembly in the core is also small, and the influence of the disturbance quickly attenuates, improving the nuclear thermal hydraulic stability of the core. can be improved. Spacers 3
Install two 03. This spacer 303 is divided into two areas by a dashed line 15: a center part 16 far away from the channel box, and a peripheral part 17 near the channel box. The cross section is rectangular 18, and the structural members of the peripheral portion 17 have blade-like upper and lower ends, and are configured as shown in the BE' cross section. The spacer 303 is the spacer 3 installed in the non-boiling region.
Similar to No. 01, both of No. 9 and No. 9 have a rectangular cross section at the center of the spacer structural material, and the upper and lower ends at the periphery of No. 1 are shaped like blades. However, compared to the spacer 303, the spacer 301 has a structure in which the peripheral section 11 is more streamlined and has a smaller pressure loss. This is because in the annular flow region where the spacer 303 is installed, the gas phase has the effect of disturbing the flow, and the coolant flow rate tends to be uniform even in the non-boiling region, whereas in the non-boiling region where the spacer 301 is installed. In this case, there is no gas phase and the coolant flow rate tends to be small in the peripheral area, so it is necessary to make the cross section streamlined to further reduce pressure loss and make the coolant flow rate uniform. In the spacer 303 in which the flow mode of the two-phase flow is an annular flow, the pressure loss between the center portion 16 and the peripheral portion 17 is 1:0.89, which is 11% smaller at the peripheral portion. In the fuel assembly in which this spacer 303 is installed, the flow rate of coolant in the peripheral area tends to decrease due to the presence of the channel box, but this is due to the effect of installing the spacer with less pressure loss in the peripheral area and the effect of the gas phase disrupting the flow. The flow rate of the coolant becomes almost uniform. As a result.

Since the coolant liquid film on the fuel cladding tube is difficult to peel off, heat is removed sufficiently by the liquid film, and the thermal margin increases.

In this fuel assembly, a spacer 301 is installed in an area where the coolant is non-boiling, a spacer 302 is installed in an area where the flow pattern of the two-phase flow is a bubble flow or a slug flow, and a spacer 303 is installed in an annular flow area. The ratio of the three pressure loss coefficients is 0.90:0.69:1.00.

By installing the spacer 302 with low pressure loss in the region of bubble flow or slug flow in this way, even when the output changes due to disturbance, the change in the coolant flow rate of each fuel assembly can be reduced, and the influence of the disturbance is reduced. It can dampen quickly and improve the nuclear thermal-hydraulic stability of the reactor core.

The width reduction ratio in the fuel assembly according to the first embodiment of the present invention is 0.
．． 58 and 69. In a conventional fuel assembly, a fuel assembly in which a spacer whose structural material has a rectangular cross section is installed over the entire length of the fuel rod, the width reduction ratio can be calculated as 0.68, so it is 15%.
An improvement in the attenuation ratio can be seen. Thermal margin also increases by 9%.

Next, a second embodiment of the present invention will be explained in detail using FIG. This embodiment, like the first embodiment, has an electrical output of 1
The target is approximately 800 fuel assemblies l loaded into a 100MWei (7) BWR.

In this fuel assembly, there is a non-boiling region between 0/6 and 1/6 of the total length from the lower end of the fuel rod, and a two-phase flow between 3/4 and 4/4 of the total length from the lower end of the fuel rod. As in the first embodiment, one spacer 301 is provided in each region where the flow pattern is annular.

Spacer 30: Install 12 pieces. The difference from the first embodiment is that the central part 2
0 and the peripheral portion 21, the cross section of the structural material is rectangular 22, and five spacers 304t with a large comparison point pressure loss are installed (an example of using the configurations (1) and (2) above). In the fuel assembly constructed in this way, the effects of the configurations (1) and (2) above can be expected. In other words, compared to the conventional fuel assembly, both the width reduction ratio and the thermal margin are 7.
% improved.

A third embodiment of the present invention will be described in detail with reference to FIG. This embodiment targets the same fuel assembly as the first and second embodiments.

In the fuel assembly of the third embodiment, two spacers 303 similar to those of the first embodiment are installed in the annular flow region between 3/4 and 474 of the total length from the lower end of the fuel rod. The difference from the first embodiment is that from 0/6 to 1/2 of the total length from the lower end of the fuel rod.
A spacer 305 in which the cross-sectional shape of the structural material is rectangular 25 in both the center part 23 and the peripheral part 24 in the non-boiling region of /6
1/4 and 3 of the total length from the bottom end of the fuel rod.
In order to reduce pressure loss in the region where the bubble flow or slug flow occurs between (
An example of using configuration 3) in combination). In a fuel assembly constructed in this way, the effects of having the construction (1) (a) described above can be expected. In other words, compared to the conventional fuel assembly,
The width reduction ratio is improved by 8% and the thermal margin is improved by 6%.

The same electrical power 110 as mentioned in the embodiment 1.2.3
A fourth embodiment of the present invention, which targets approximately 800 fuel assemblies loaded in a 0MWe class BWR, will be described in detail with reference to FIG.

In the fuel assembly of this embodiment, two spacers 303 are installed between 374 and 4/4 of the total length from the lower ends of the fuel rods, as in the first embodiment. In other areas, six spacers 305 are installed in which the structural material has a rectangular cross section 25 at the center 23 and peripheral portion 24 of the cross section of the coolant flow path (as described above).
An example of the configuration of 1).

In the fuel assembly constructed in this way, the effect of the configuration (1) can be expected, and the thermal margin is improved by 4% compared to the previous conventional fuel assembly.

As a method of changing the pressure loss between the center and peripheral parts of the cross section of the coolant flow path of the fuel assembly, in the embodiments 1, 2, 3 and 4, the upper and lower ends of the spacer structural material are made into blade shapes or not. Yes.

In the third embodiment, a method of changing the projected area of the spacer structural material onto the flow path was also used. In the former method, as shown in Fig. 5, the pressure loss ratio can be reduced to 1.0 depending on the shape of the spacer structural material.
It can be changed arbitrarily within the range of 0 to 0.66. Furthermore, in the latter method, it is known that the pressure loss is approximately proportional to the projected area of the spacer structural material. Spacers installed in the region where the flow pattern of the two-phase flow is bubble flow or slug flow between 1/6 and 3/4 of the total length from the bottom end of the fuel rod have a lower pressure loss than spacers installed in other regions. In order to reduce the pressure loss or to change the pressure loss of the spacer within the cross section of the flow path, any of the above three methods may be used.

In the detailed description of the present invention, the coolant flow path of the fuel assembly was divided into two regions, a center portion and a week edge portion, by the dashed dotted line 7 in FIG. 1, and the pressure loss was varied in the two regions. Generally, pressure loss varies greatly depending on flow resistance such as in a channel box, so it is good to divide the area into an area including the fuel rods adjacent to the channel box and an area other than the area.

〔Effect of the invention〕

As explained above, according to the present invention, thermal margin can be increased and nuclear thermal hydraulic stability can be improved simply by replacing the spacer of the fuel assembly.

Therefore, there is no need for major design changes to the fuel assembly.
It can be applied to nuclear reactors currently in operation, and can be useful for downsizing nuclear reactors or increasing their power density. Therefore, the safety and economic effects of implementing the present invention are significant.

[Brief explanation of the drawing]

FIG. 1 is a block diagram of a fuel assembly according to a first embodiment of the present invention, FIG. 2 is a block diagram of a fuel assembly according to a second embodiment of the present invention, and FIG. 3 is a block diagram of a fuel assembly according to a second embodiment of the present invention. FIG. 4 is a configuration diagram of a fuel assembly according to the fourth embodiment of the present invention, and FIG. 5 is an explanatory diagram of the relationship between the spacer structural material and pressure loss. Out. l\...Fuel assembly, 2...Fuel rod, 301-30
6...S.S.T0 Mochi 1st (Tai) 2nd Kogyo3 mouth

Claims (1)

[Claims] 1. Consists of a large number of fuel rods and spacers that support their longitudinal axes in parallel and are spaced apart. Coolant flows from the bottom and flows out from the top, and is removed by the boiling heat transfer phenomenon of the coolant. In heated nuclear reactor fuel assemblies,
The spacer is installed in a region where the flow pattern of the two-phase flow becomes an annular flow between 3/4 and 4/4 of the total length from the lower end of the fuel rod. , a fuel assembly characterized in that it is configured to reduce pressure loss in the peripheral area.