JP6568348B2 - Fast reactor fuel assemblies and cores loaded with them - Google Patents

Description

  TECHNICAL FIELD The present invention relates to a fuel assembly for a fast reactor and a core for loading the same, and by making the Na void reactivity negative, the fuel assembly for a fast reactor capable of improving safety and the core for loading the same. About.

Generally, in a fast breeder reactor, a reactor core is disposed in a reactor vessel, and liquid sodium which is a coolant is filled in the reactor vessel. The fast reactor fuel assembly loaded in the core is composed of a plurality of fuel rods filled with depleted uranium ( ²³⁸ U) enriched in plutonium, a trumpet tube surrounding a plurality of bundled fuel rods, It has a lower end portion, an entrance nozzle for supporting a neutron shield located below the fuel rod, and a coolant outflow portion located above the fuel rod.

  A fast breeder reactor core has a core fuel region having an inner core region and an outer core region surrounding the inner core region, a blanket fuel region surrounding the core fuel region, and a neutron shield region surrounding the blanket fuel region. In the case of a standard homogeneous core, the Pu enrichment of the fuel assemblies loaded in the outer core region is higher than the Pu enrichment of the fuel assemblies loaded in the inner core region. As a result, the power distribution in the radial direction of the core is flattened.

  On the other hand, among high level radioactive waste (HLW) generated by reprocessing spent fuel in the nuclear reactor, minor actinide (MA) has long-term radioactivity, and this Research is being conducted to reduce the burden on the geological disposal site and reduce the environmental burden by reducing the harmfulness of HLW and increasing the decay rate by recycling MA and transmuting in a fast reactor. However, when MA is loaded on the reactor core, the void reactivity (reactivity applied by boiling of liquid Na as a coolant) tends to increase. Therefore, Non-Patent Document 1 has been proposed as one effective measure for suppressing the void reactivity.

  Non-Patent Document 1 describes the introduction of a fuel assembly in which a Na plenum region is provided on a fuel rod bundle. The fuel region of the large core has positive void reactivity, whereas the Na plenum region has negative void reactivity. Therefore, during transient events such as loss of coolant flow, the core due to voids in the Na plenum region There is a possibility that the overall effective void reactivity can be negative. The dependence of the void reactivity on the Na plenum region structural material ratio is shown. The smaller the structural material ratio, the smaller the void reactivity.

  Moreover, although it is related with a pressurized water reactor (PWR), the technique described in patent document 1 is known. Patent Document 1 discloses a configuration in which a plurality of disc springs are arranged between an upper end plug and an uppermost fuel pellet in a cladding tube of each fuel rod constituting a fuel assembly. A plurality of disc springs are used to preliminarily apply tension to the fuel pellets to prevent creep.

JP-A-62-3691

K. Kawashima, K. Fujimura, et al., “Advanced Design of the Axially Heterogeneous Core”, Proceedings of FR'91, pp. 3-1, Cypto. (1991).

  Non-Patent Document 1 discloses the dependence of the void reactivity on the proportion of the Na plenum region structural material, but does not specifically disclose how the fast reactor fuel assembly should be constructed. In particular, the fuel assemblies are transported for loading into the fast reactor core, but no mention is made of vibrations added to the fuel pellets in the fuel rods constituting the fuel assemblies during transportation.

  In Patent Document 1, a plurality of disc springs are disposed between the upper end plug and the uppermost fuel pellet. Therefore, the space formed between the upper end plug and the uppermost fuel pellet has a problem of expanding in the axial direction. Moreover, since patent document 1 is related to PWR, it does not include a Na plenum region disposed above the fuel rod in the fuel assembly.

  Accordingly, the present invention is to provide a fast reactor fuel assembly capable of reducing the void reactivity by reducing the volume ratio of the structural material in the Na plenum region, and a fast reactor core loaded with the fuel assembly. .

In order to solve the above problems, a fuel assembly for a fast reactor according to the present invention includes a fuel rod cladding tube in which a plurality of fuel rods filled with fuel pellets containing nuclear fuel material are bundled and accommodated in a trumpet tube. The fuel assembly is loaded with a sodium plenum region defined by the trumpet tube above an upper end plug of the plurality of fuel rods, wherein each fuel rod is formed by the plurality of fuel pellets. A void reaction in the sodium plenum region during a transient event, including an upper gas plenum region defined by the fuel rod cladding, between the upper end of the fuel region and the upper end plug; The upper end of the disc spring or bamboo shoot spring or steel wool contacts the lower surface of the upper end plug and the lower end is the fuel rod cladding tube so that the degree is reduced. The upper gas plenum by shortening the axial length of the region, trumpet and the upper end plug defining said sodium plenum region and the upper gas plenum region by placing into contact with the upper surface of the uppermost fuel pellet the volume ratio of the constituent material of the sodium plenum region containing the fuel rod cladding that defines the is characterized in that so as to reduce.

The core of the fast reactor of the present invention includes an inner core region, an outer core region surrounding the inner core region, a radial blanket region surrounding the outer core region, and a neutron shield region surrounding the radial blanket region. A fast reactor core having a plurality of fuel rods loaded in the inner core region and the outer core region and filled with a plurality of fuel pellets containing nuclear fuel material in the fuel rod cladding tube and housed in a trumpet tube The fuel assembly includes a sodium plenum region defined by the trumpet tube above an upper end plug of the plurality of fuel rods, and each fuel rod has an upper end of a fuel region formed by the plurality of fuel pellets. between said upper end plug part has an upper gas plenum area defined by the fuel rod cladding tube, before during transient events including coolant loss of flow In order to reduce the void reactivity of the sodium plenum region, the upper end of the disc spring or bamboo shoot spring or steel wool contacts the lower surface of the upper end plug and the lower end is the uppermost portion in the fuel rod cladding tube so that the upper gas plenum region is reduced. The axial length of the upper gas plenum region is shortened by being disposed so as to contact the upper surface of the fuel pellet of the fuel pellet, and the trumpet tube and the upper end plug defining the sodium plenum region and the upper gas plenum region are defined. the volume ratio of the constituent material of the sodium plenum region containing the fuel rod cladding tube, characterized in that so as to reduce.

  According to the present invention, by reducing the volume ratio of the structural material in the Na plenum region, it is possible to provide a fast reactor fuel assembly capable of reducing void reactivity and a fast reactor core loaded with the fuel assembly.

  Problems, configurations, and effects other than those described above will be clarified by the following description of embodiments.

It is a longitudinal cross-sectional view of the fuel assembly for fast reactors of Example 1 which concerns on one Example of this invention. It is a longitudinal cross-sectional view of the fuel assembly for fast reactors of Example 1 shown in FIG. 1, and a cross-sectional view of the fast reactor core which loads it. It is a longitudinal cross-sectional view of the fast reactor core which loads the fuel assembly for fast reactors shown in FIG. 2, and is a figure which shows from a core center to an outer peripheral part. It is the longitudinal cross-sectional view of the fuel assembly for fast reactors of Example 2 which concerns on the other Example of this invention, and the cross-sectional view of the fast reactor core which loads it. FIG. 5 is a longitudinal cross-sectional view of a fast reactor core loaded with the fast reactor fuel assembly shown in FIG. 4, showing from the core center to the outer periphery. It is a figure which shows the relationship between the length of an upper gas plenum area | region, and Na void reactivity. It is a longitudinal cross-sectional view of the fuel rod which comprises the fuel assembly for fast reactors of Example 3 which concerns on the other Example of this invention.

  In the present specification, “the volume ratio of the structural material in the Na plenum region” means the Na plenum disposed above the upper end plug above the fuel region of the fuel rod constituting the fast reactor fuel assembly. The volume ratio of the structural material in the region including the region. That is, it means the volume ratio of the structural material in the region including the upper gas plenum region, the upper end plug and the Na plenum region. The structural material is a wrapper tube that defines a Na plenum region, a top end plug, a fuel rod cladding tube that defines an upper gas plenum region, and the like.

  Embodiments of the present invention will be described below with reference to the drawings.

  FIG. 1 is a longitudinal sectional view of a fast reactor fuel assembly of Example 1 according to one embodiment of the present invention. As shown in FIG. 1, a fast reactor fuel assembly 1 is configured by bundling and housing a plurality of fuel rods 2 in a trumpet tube 7 having a hexagonal cross section. The fast reactor fuel assembly 1 has a Na plenum region 12 in a stainless steel (SUS) trumpet tube 7 above the upper end of the fuel rod 2. The Na plenum region 12 is a space defined by the trumpet tube 7 and contains only fluid Na (liquid Na) as a coolant in this space. Further, the fast reactor fuel assembly 1 has an entrance nozzle (introducing liquid Na) in order to allow liquid Na, which is a coolant, to flow between the fuel rods 2 below the trumpet tube 7 shown in FIG. A handling head (not shown) and the like are provided above the trumpet tube 7 (not shown). Moreover, the space | interval of adjacent fuel rods 2 is hold | maintained by the wire spacer (not shown) wound around the fuel rod surface.

The fuel rod 2 is a mixed oxide fuel (hereinafter referred to as MOX (Mixed Oxide) fuel) in which a plutonium oxide (PuO ₂ ) and a deteriorated uranium oxide (UO ₂ ) are mixed in a fuel rod cladding tube 4 made of stainless steel. A plurality of fuel pellets 10 sintered) are filled and sealed in the axial direction. The fuel pellet 10 is not limited to the MOX fuel, but may be one containing nuclear fuel such as Pu. The fuel rod 2 is provided with an upper end plug 3 and a lower end plug 9 at the upper and lower ends, respectively, to seal the inside of the fuel rod cladding tube 4. The fuel rod 2 includes a lower gas plenum region 6 between a lower end plug 9 and a lower end portion of a fuel region 11 formed by a plurality of fuel pellets 10. The lower gas plenum region 6 is a space defined by the fuel rod cladding tube 4 and holds fission products (FP) gas generated by fission of the fuel pellet 10 in the space. The length of the lower gas plenum region 6 is determined so that the stress of the fuel rod cladding tube 4 due to the internal pressure by the FP gas is less than the design limit while holding all the FP gas generated during the fuel life. Yes.

  As shown in the enlarged view of region A in FIG. 1, an upper gas plenum region 5 is provided between the upper end plug 3 and the fuel region 11 on the upper end side of the fuel rod 2. The upper gas plenum region 5 is a space defined by the fuel rod cladding tube 4, and the disc spring 8 is arranged inside. The disc spring 8 has an upper end in contact with the lower surface of the upper end plug 3 and a lower end in contact with the upper surface of the uppermost fuel pellet 10 forming the fuel region 11 (the upper end portion of the fuel region 11). Thus, the position of the fuel pellet 10 forming the fuel region 11 is held by the biasing force of the disc spring 8 when the fast reactor fuel assembly 1 is transported or transported.

FIG. 2 shows a longitudinal sectional view of the fast reactor fuel assembly 1, 1 ′ and a transverse sectional view of the fast reactor core 21 in which it is loaded. As shown in the cross-sectional view of FIG. 2, the fast reactor core 21 includes an inner core region 22, a core fuel region having an outer core region 23 surrounding the inner core region 22, a radial blanket region 24, and a neutron shield region 25. Have In the radial direction of the fast reactor core 21, the radial blanket region 24 surrounds the core fuel region and is adjacent to the core fuel region, and the neutron shield region 25 surrounds the radial blanket region 24. The control rod 26 has a plurality of neutron absorption rods in which boron carbide (B ₄ C) pellets are enclosed in a stainless steel cladding tube, and these neutron absorption rods are accommodated in a trumpet tube having a hexagonal cross section. The The control rod 26 has two independent systems of a main furnace stop system and a post-furnace stop system, but these are shown without distinction in FIG.

  In FIG. 2, the Pu enrichment of the fast reactor fuel assembly 1 ′ loaded in the outer core region 23 is made higher than the Pu enrichment of the fast reactor fuel assembly 1 loaded in the inner core region 22. Yes. The fast reactor fuel assemblies 1 and 1 'have the same shape as the fast reactor fuel assembly 1 shown in FIG. 1, but are filled in the fuel rods 2' in the fast reactor fuel assembly 1 '. By increasing the Pu enrichment of the fuel pellet 10 ', the power distribution in the radial direction of the fast reactor core 21 is flattened.

Here, FIG. 3 shows a longitudinal sectional view from the core center to the outer periphery of the fast reactor core 21 loaded with the fast reactor fuel assemblies 1, 1 ′ shown in FIG. In Figure 3, the center axis _{O 21} of the high-speed reactor core 21 depicts 1/2 core to axial symmetry. As shown in FIG. 3, the fuel region 11 of the fast reactor fuel assembly 1 loaded in the inner core region 22 and the fuel region 11 ′ of the fast reactor fuel assembly 1 ′ loaded in the outer core region 23. The axial (O ₂₁ ) direction length is equal. The enrichment of the fissile material Pu (weight ratio of Pu with respect to the weight of the total heavy metal HM (Heavy metal)) is higher than the Pu enrichment of the fast reactor fuel assembly 1, and the fast reactor fuel assembly 1 ′. The Pu enrichment is set high, and the power in the radial direction of the core is flattened.

  As described above, in the fast reactor core 21 loaded with the fast reactor fuel assemblies 1 and 1 ′ having the Na plenum region 12 of the present embodiment, the flow rate of the liquid Na that is the coolant decreases due to a pump failure or the like. Assuming the case, the temperature of the liquid Na in the vicinity of the upper ends of the fuel rods 2 and 2 ′ rises most. As a result, a decrease in the density of liquid Na, which is a coolant in the Na plenum region 12, or generation of voids makes it easier for neutrons generated in the fuel regions 11 and 11 'to leak upward, thereby greatly reducing void reactivity. Have

  The conventional fast reactor has a gas plenum region in the lower part of the fuel rod in the fuel assembly loaded in the core, like the fast reactor fuel assemblies 1, 1 'of the present embodiment. Each fuel rod is provided with a coil spring between the upper end of the fuel pellet and the upper end plug to hold the position of the fuel pellet when the fuel assembly is transported or transported. The axial length of the region where the coil spring is disposed (upper gas plenum region) is about 10 cm. The function of the coil spring is to hold the fuel pellets during fuel transfer at room temperature before fuel irradiation.

  On the other hand, in the fast reactor fuel assembly 1, 1 ′ of the present embodiment, the stainless steel disc spring 8 is arranged in the upper gas plenum region 5 as described above. This is because the disc spring 8 is provided only to hold the position of the fuel pellet 10 during the transport or transport of the fast reactor fuel assemblies 1, 1 ', and is particularly required during operation of the fast reactor. By focusing on not being. Thus, the axial length of the upper gas plenum region 5 can be reduced by adopting a configuration in which the urging force is applied to the fuel pellet 10 by the disc spring 8. The axial length of the upper gas plenum region 5 in this embodiment is about 2 cm.

  As described above, the void reactivity of the fast reactor loaded with the fast reactor fuel assemblies 1, 1 ′ having the Na plenum region decreases as the volume ratio of the structural material in the Na plenum region decreases. As described above, the volume ratio of the structural material in the Na plenum region is the region B surrounded by the one-dot chain line, that is, the region including the upper gas plenum region 5, the upper end plug 3 and the Na plenum region 12 as shown in FIG. It means the volume ratio of the structural material. Therefore, in this embodiment, the axial length of the upper gas plenum region 5 of the Na plenum region 12 is shortened from about 10 cm to about 2 cm (about 1/5 or less) to reduce the volume ratio of the structural material. Yes. As a result, the void reactivity is reduced, and it is possible to improve the safety at the time of flow loss transient (Unprotected Loss of Flow: ULOF), assuming a scram failure.

  According to the present embodiment, the void reactivity can be reduced by reducing the volume ratio of the structural material in the Na plenum region. In addition, since the void reactivity can be reduced, the safety of the fast reactor can be improved even if an event such as a scram failure occurs.

  FIG. 4 shows a vertical cross-sectional view of a fast reactor fuel assembly of Example 2 according to another embodiment of the present invention, and a cross-sectional view of a fast reactor core loaded with the fuel assembly. The same components as those shown in FIGS. 1 and 2 are denoted by the same reference numerals. In the first embodiment, the fast reactor fuel assemblies 1 and 1 ′ having the same shape are loaded in the inner core region 22 and the outer core region 23 of the fast reactor core, respectively. In contrast, the present embodiment is different in that the fast reactor fuel assembly 1a loaded in the inner core region 22 and the fast reactor fuel assembly 1b loaded in the outer core region 23 have different shapes. Below, the description which overlaps with Example 1 is abbreviate | omitted.

  As shown in FIG. 4, the fast reactor fuel assembly 1a loaded in the inner core region 22 of the fast reactor core 31 includes a plurality of fuel pellets 10 filled in the fuel rod cladding tube 4 of the fuel rod 2a. An internal blanket pellet 32 using deteriorated uranium as a nuclear fuel is disposed below the center in the axial direction. Thus, the fuel region is divided into the first fuel region 11a above and the second fuel region 11a 'below, with the internal bracket pellet 32 as a boundary. The total axial length of the first fuel region 11a, the inner blanket pellet 32, and the second fuel region 11a 'is shorter than the axial length of the fuel region 11 in the first embodiment. As described above, the arrangement position of the internal blanket pellets 32 is preferably lower than the approximate center in the axial direction of the plurality of fuel pellets 10, but is not necessarily limited thereto, and the axial direction of the plurality of fuel pellets 10 is not necessarily limited. It is good also as a structure arrange | positioned in the approximate center.

  As shown in FIG. 4, the fast reactor fuel assembly 1b loaded in the outer core region 23 of the fast reactor core 31 has a fuel region longer than the axial length of the fuel region 11 in the first embodiment. 11b. That is, the fast reactor fuel assembly 1b accommodates the fuel rods 2b in which the number of fuel pellets 10 filled in the fuel rod cladding tube 4 is increased as compared with the fuel rods 2 shown in the first embodiment.

FIG. 5 is a longitudinal sectional view from the core center to the outer periphery of the fast reactor core 31 loaded with the fast reactor fuel assemblies 1a and 1b shown in FIG. In Figure 5, the center axis _{O 31} of the high-speed reactor core 31 illustrates 1/2 the core in the axial object. As shown in FIG. 5, deteriorated uranium is formed between the first fuel region 11a and the second fuel region 11b of the fuel rod 2a accommodated in the fast reactor fuel assembly 1a loaded in the fast reactor core 31. An internal blanket pellet 32 is provided as a fuel material. The axial length (O ₃₁ ) of the first fuel region 11a is longer than the axial length of the second fuel region 11a ′. The total of the axial lengths of the first fuel region 11a, the inner blanket pellet 32, and the second fuel region 11a ′ is determined in the fast reactor fuel assembly 1 loaded in the fast reactor core 21 of the first embodiment. The axial length of the fuel region 11 of the fuel rod 2 to be accommodated is shorter. The axial length of the fuel region 11b of the fuel rod 2b accommodated in the fast reactor fuel assembly 1b loaded in the outer core region 23 is the same as that of the first fuel region 11a, the inner blanket pellet 32 and the first fuel region 2b. Longer than the sum of the axial lengths of the two fuel regions 11b.

  The inner core region 22 of the fast reactor core 31 shown in FIG. 5 has a low output by having a region (internal blanket pellet 32) that does not contain Pu or the like, which is a fissile material, below the approximate center in the axial direction. Therefore, the neutron flux near the upper end of the first fuel region 11a containing Pu and the like is high. Therefore, when it is assumed that the liquid Na serving as the coolant boils, the neutrons leaking from the fuel pellets 10 in the first fuel region 11a to the Na plenum region 12 increase, and the result is shown in FIG. Compared with the fast reactor core 21, the void reactivity can be further reduced.

  FIG. 6 is a diagram showing the relationship between the length of the upper gas plenum region and the Na void reactivity. FIG. 6 is a diagram in which the effect of reducing the void reactivity according to this example is evaluated by analysis. The analysis method used for the evaluation is as follows.

In general, nuclear reactions in nuclear reactors are governed by Boltzmann transport equations that govern the behavior of neutrons in space and energy. In a normal fast reactor core calculation, the above transport equation is based on a diffusion approximation that approximates the flow of neutron flux (unit: number of neutrons per 1 cm ² ) proportional to the gradient of neutron flux distribution. In many cases, the analysis code is used. Furthermore, in the design calculation of fast reactors in Japan, nuclear isotopes such as Pu ( ²³⁹ Pu, ²⁴⁰ Pu, ²³⁸ U, ²³⁷ Np, ²⁴¹ Am, etc.), reactor structural materials (Fe etc.), coolants Among the cross-section libraries that systematize data such as neutron cross-section (unit: 1 / cm ² ) that is the reaction probability of ( ²³ Na) etc. with neutrons, the latest library JENDL-4.0 prepared in Japan Based on the above, the 70-group nuclear data set for the fast reactor, which is discretized into 70 energy groups using the energy distribution of the standard fast reactor neutron flux, is often used. The above analysis method for the fast reactor has been verified by critical experiment equipment for fast reactors in Japan and overseas and critical experiment analysis of actual fast reactors.

  For the core system shown in FIGS. 4 and 5, the neutron diffusion equation, which is spatially discretized by the difference method and discretized into 70 groups corresponding to the above nuclear data set, is numerically solved, The neutron flux distribution is obtained, and the cross-sectional area of many isotope nuclides constituting the core is averaged using the number density of each nuclide, and the product of the neutron flux obtained above is spatially and energy-wise. To calculate the reaction rates of various nuclear reactions. In FIG. 6, with the axial length of the upper gas plenum region 5 of the fast reactor core 31 as a parameter, the case where the liquid Na, which is the coolant, is not boiling during normal operation and the transient loss of coolant flow rate (ULOF) Assuming the event, when the liquid Na in the trumpet tube 7 boiled, solve the above neutron diffusion equation to determine the effective neutron multiplication factor before and after boiling, and calculate the void reactivity from the difference Yes.

  The length (cm) of the upper gas plenum region on the horizontal axis indicates the upper surface of the fuel pellet 10 located at the uppermost portion (the upper end portion of the first fuel region 11a) and the upper end plug 3 in the fast reactor fuel assembly 1a. The axial direction distance between the lower surface of each is shown. Further, the Na void reactivity ($) on the vertical axis indicates the fuel region (the first fuel region 11a, the inner blanket pellet 32, and the second fuel region) including the inner blanket pellet 32 in the fast reactor core 31 of the present embodiment. 11b shows the Na void reactivity (unit is $ (dollar)) when the liquid Na, which is the coolant inside the trumpet tube 7, in the Na plenum region 12 boils (voids).

Here, the unit ($) of Na void reactivity will be described. There are two types of neutrons generated by fission: prompt neutrons with a lifetime of the order of 10 ⁻⁴ [seconds] and delayed neutrons of several tens of seconds depending on the nuclide. If the nuclear reactor becomes critical only by prompt neutrons, the reactivity of the order of 10 ⁻⁴ [second], that is, the output fluctuates, so that it becomes difficult to control the reactivity with a mechanical device such as the control rod 26. Actually, there are delayed neutrons, so if the change in reactivity is less than the ratio of the number of delayed neutrons to the total number of neutrons (this is called the effective delayed neutron ratio, expressed as βeff), it is mechanical. Control by a simple device is sufficiently possible. The change in reactivity, which is just the effective delayed neutron ratio (βeff), is called 1 dollar ($) and is used as a scale of reactivity. On the other hand, if the insertion reactivity due to control rod misdrawing exceeds 1.0 $, the criticality is exceeded only by prompt neutrons, the mechanical control cannot follow, and fuel may break.

  Returning to FIG. 6, as described above, when the length of the conventional upper gas plenum region is about 10 cm, the Na void reactivity is about 1.3 $. However, in the fast reactor fuel assembly 1a of the present embodiment, the length of the upper gas plenum region is about 2 cm, and the Na void reactivity is about 0.3 $. Therefore, as shown in FIG. 6, by loading the fast reactor fuel assembly 1a of the present embodiment into the fast reactor core 31, the Na void reactivity can be reduced by about 1.0 $ compared to the conventional case.

  According to this example, compared with Example 1, it becomes possible to further reduce the void reactivity.

  Further, according to the present embodiment, as described above, the total length of the fuel rod 2a is shortened by about 8 cm compared to the conventional case (a reduction in the axial length of the upper gas plenum region 5), and the pressure loss of the core portion is reduced. A secondary effect of decreasing can also be achieved.

  FIG. 7 shows a longitudinal sectional view of fuel rods constituting a fast reactor fuel assembly of Example 3 according to another embodiment of the present invention. In the first and second embodiments, the disc spring 8 is arranged in the upper gas plenum region 5, whereas in the present embodiment, the upper gas plenum is not arranged in the upper gas plenum region 5. The difference is that the axial length of the region 5 is further shortened. In FIG. 7, the same components as those in the first and second embodiments are denoted by the same reference numerals.

  As shown in FIG. 7, a plurality of fuel rods 2 a accommodated in the fast reactor fuel assembly are composed of a lower surface of the upper end plug 3 and an upper surface of the uppermost fuel pellet 10 forming the fuel region 11 (the fuel region 11 No structural material is disposed in the upper gas plenum region 5 which is a space defined by the fuel rod cladding tube 4.

  In this embodiment, the axial length L of the upper gas plenum region 5 is set in consideration of the difference in thermal expansion coefficient between the fuel rod cladding tube 4 made of stainless steel and the fuel pellet 10 containing nuclear fuel such as Pu. Hereinafter, the thermal expansion of the fuel pellet 10 and the thermal expansion of the fuel rod cladding tube 4 will be specifically described.

When the length of the fuel region 11 formed by a plurality of fuel pellets 10 containing nuclear fuel such as Pu is 100 cm as standard, the fuel expansion is considered in consideration of the thermal expansion due to the temperature change from the normal temperature to the rated output state. The amount of thermal expansion of the pellet 10 is
10 ^{−5 [} 1 / K] × 100 [cm] × 1000 [K] = 1 [cm],
The amount of thermal expansion of the fuel rod cladding tube 4 is
1.75 × 10 ^{−5 [} 1 / K] × 100 [cm] × 450 [K] = 0.8 [cm]. Therefore, if the axial length L of the upper gas plenum region 5 is 0.2 [cm], the difference in the amount of expansion (thermal expansion amount) associated with the thermal expansion of the fuel pellet 10 and the fuel rod cladding tube 4 can be absorbed. .

  If the length between the upper surface of the uppermost fuel pellet 10 forming the fuel region 11 (the upper end of the fuel region 11) and the upper end plug 3 is about 0.2 [cm], the inside of the trumpet 7 When transporting or transporting a fuel assembly for a fast reactor containing a plurality of fuel rods 2a, the fuel pellet 10 moves to the lower surface side of the upper end plug 3 and contacts in the fuel rod cladding tube 4. Even in such a case, the moving stroke is limited to 0.2 [cm]. Therefore, even if the fuel pellet 10 collides with the upper end plug 3, the soundness of the fuel pellet 10 is maintained.

  Therefore, in this embodiment, the axial length of the upper gas plenum region 5 is shortened from about 2 cm to 0.2 cm (1/10) as compared with the first and second embodiments, and the upper gas plenum region is also reduced. The structure 5 does not include any structural material. Thereby, the volume ratio of the structural material in the Na plenum region can be further reduced.

  According to the present embodiment, as compared with Embodiments 1 and 2, the volume ratio of the structural material in the Na plenum region can be further reduced, and the void reactivity can be further reduced.

  The fuel rod 2a of the present embodiment is used for the fast reactor fuel assembly 1, 1 'loaded in the fast reactor core 21 shown in FIG. 2 or in the fast reactor core 31 shown in FIG. By applying to the fuel assemblies 1a and 1b, a fast reactor core capable of reducing the void reactivity can be realized.

  Moreover, in the above-mentioned Example 1 and Example 2, although it was set as the structure which distributes the stainless steel disc spring 8 to the upper gas plenum area | region 5, if it replaces with this and has the same function, it will be a bamboo shoot You may use a spring, steel wool, etc.

  In addition, this invention is not limited to an above-described Example, Various modifications are included. For example, the above-described embodiments have been described in detail for easy understanding of the present invention, and are not necessarily limited to those having all the configurations described. Further, a part of the configuration of one embodiment can be replaced with the configuration of another embodiment, and the configuration of another embodiment can be added to the configuration of one embodiment. Further, it is possible to add, delete, and replace the configurations of other embodiments with respect to a part of the configurations of the embodiments.

1, 1 ', 1a, 1b, 1c ... Fast reactor fuel assemblies 2, 2', 2a ... Fuel rod 3 ... Upper end plug 4 ... Fuel rod cladding tube 5 ... Upper gas plenum region 6 ... Lower gas plenum region 7 ... Wrapper Tube 8 ... Belleville spring 9 ... Lower end plugs 10, 10 ', ... Fuel pellets 11, 11', 11b, 11c ... Fuel region 11a ... First fuel region 11a '... Second fuel region 12 ... Na plenum region 21 , 31 ... Fast core 22 ... Inner core region 23 ... Outer core region 24 ... Radial blanket region 25 ... Neutron shield region 26 ... Control rod 32 ... Internal blanket pellet

Claims (8)

A fuel assembly in which a plurality of fuel rods filled with fuel pellets containing nuclear fuel material are bundled in a fuel rod cladding tube and bundled into a trumpet tube and loaded into a fast reactor core,
A sodium plenum region defined by the trumpet tube above an upper end plug of the plurality of fuel rods;
Each fuel rod has an upper gas plenum region defined by the fuel rod cladding tube between an upper end of the fuel region formed by the plurality of fuel pellets and the upper end plug .
In order to reduce the void reactivity of the sodium plenum region during transient events including loss of coolant flow, the upper end of the upper gas plenum region is contacted with the upper end of a disc spring or bamboo shoot spring or steel wool against the lower surface of the upper end plug. A trumpet tube defining the sodium plenum region by shortening the axial length of the upper gas plenum region by disposing the contact lower end so as to contact the upper surface of the uppermost fuel pellet in the fuel rod cladding tube; fast reactor fuel assembly, characterized in that the volume ratio of the upper end plug and construction material of sodium plenum region containing the fuel rod cladding that defines the upper gas plenum region is so reduced. The fuel assembly for a fast reactor according to claim 1,
The fuel assembly for a fast reactor, wherein each of the fuel rods is provided with an internal blanket pellet having a deteriorated uranium as a nuclear fuel below a substantial center of the fuel region. The fuel assembly for a fast reactor according to claim 2,
Each fuel rod has a lower gas plenum region defined by the fuel rod cladding tube between a lower end portion of the fuel region and a lower end plug. A core of a fast reactor having an inner core region, an outer core region surrounding the inner core region, a radial blanket region surrounding the outer core region, and a neutron shield region surrounding the radial blanket region,
A fuel assembly that is loaded in the trumpet tube by bundling a plurality of fuel rods loaded in the inner core region and the outer core region and filling a plurality of fuel pellets containing nuclear fuel material in the fuel rod cladding tube,
A sodium plenum region defined by the trumpet tube above an upper end plug of the plurality of fuel rods;
Each fuel rod has an upper gas plenum region defined by the fuel rod cladding tube between an upper end of the fuel region formed by the plurality of fuel pellets and the upper end plug .
In order to reduce the void reactivity of the sodium plenum region during transient events including loss of coolant flow, the upper end of the upper gas plenum region is contacted with the upper end of a disc spring or bamboo shoot spring or steel wool against the lower surface of the upper end plug. A trumpet tube defining the sodium plenum region by shortening the axial length of the upper gas plenum region by disposing the contact lower end so as to contact the upper surface of the uppermost fuel pellet in the fuel rod cladding tube; core of a fast reactor, wherein the volume ratio of the upper end plug and construction material of sodium plenum region containing the fuel rod cladding that defines the upper gas plenum region is so reduced. In the core of the fast reactor according to claim 4,
The axial length of the fuel region of each fuel rod housed in the fuel assembly loaded in the inner core region and the fuel region of each fuel rod housed in the fuel assembly loaded in the outer core region A fast reactor core characterized in that the axial lengths of the cores are equal. In the core of the fast reactor according to claim 5,
A fast reactor core characterized in that a Pu enrichment of a fuel assembly loaded in the outer core region is higher than a Pu enrichment of a fuel assembly loaded in the inner core region. In the core of the fast reactor according to claim 4,
The axial length of the fuel region of each fuel rod housed in the fuel assembly loaded in the outer core region is the fuel of each fuel rod housed in the fuel assembly loaded in the inner core region. A fast reactor core characterized by being longer than the axial length of the region. In the core of the fast reactor according to claim 7,
Each of the fuel rods accommodated in the fuel assembly loaded in the inner core region is provided with an internal blanket pellet using depleted uranium as a nuclear fuel below a substantial center of the fuel region. The core of the furnace.

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