JP2008139321A - Fuel assembly and core of nuclear reactor - Google Patents

Abstract

<P>PROBLEM TO BE SOLVED: To provide a core of a safe nuclear reactor of high conversion type or breeder type using light water or heavy water which does not use Na, such as light water reactors, and to provide a fuel assembly which can improve the utilization factor of nuclear fuels based on an improved method of the void coefficient with a new principle. <P>SOLUTION: The fuel assembly which employs light water or heavy water as the coolant and MOX fuel as the nuclear fuel material, and which is used for the nuclear reactor of high conversion type or breeder type of which moderator to fuel volume ratio is 1.0 or less, includes an upstream blanket 17 arranged at the lower end part, a downstream blanket 14 arranged at the upper end part, an upstream main exoergic part 16 and downstream main exoergic part 13 which are arranged between the upstream blanket 17 and downstream blanket 14 and have high Pu concentration, and a low concentration area 15 which is arranged between the upstream main exoergic part 16 and downstream main exoergic part 13 and has low Pu concentration. <P>COPYRIGHT: (C)2008,JPO&INPIT

Description

  The present invention relates to a fuel assembly and a reactor core, and more particularly to a fuel assembly and a reactor core capable of improving a void coefficient of a coolant in a reactor having a high conversion ratio using plutonium fuel. .

  The fast breeder reactor (LMFBR) using liquid sodium (Na) as a coolant is not yet ready for practical use. If a nuclear reactor that can achieve fuel growth or a high conversion ratio without using Na that is likely to cause a fire can be designed, safer nuclear power generation becomes possible.

  From this idea, measures to realize the excellent characteristics of LMFBR without using Na have been studied, but as in the case of LMFBR, the loss of the coolant increases the reactivity of the core, and the "void coefficient is positive" The phenomenon is likely to occur, and this phenomenon has become a major bottleneck, and a satisfactory core design has not yet been achieved.

  To summarize the cause of the positive void coefficient, the void generation hardens the neutron spectrum (shifts to higher energy) and the neutron "upper spectrum" (higher energy side) neutron generation rate vs. absorption ratio And the neutron absorption rate decreases at the neutron “lower side of spectrum” (lower energy side). By increasing the conversion ratio or breeding ratio, it is difficult to avoid the difficulty of safe operation of the reactor in the core in which the void coefficient of the core is positive.

  As a principle of a proposal for making the void coefficient negative, there is known a method for increasing leakage of neutrons accompanying generation of voids. As a configuration that realizes this principle, a method of increasing neutron leakage as an elongated core, a method of increasing neutron leakage by eliminating fuel in a fairly wide range, a flat core that leaks neutrons in the axial direction as a flat core, or an internal An axially inhomogeneous core with an internal blanket capable of neutron leakage towards is known.

  However, in each of the above configurations, since the core cannot be made large, there is a problem that the amount of fuel loading is small, the space utilization rate inside the reactor vessel is poor, and the economic causes of nuclear power generation are deteriorated.

  When examining the principle to solve the above-mentioned problems, the present inventors pay attention to the fact that the neutron absorption effect is reduced at the lower side of the spectrum, and there can be a “new principle” that can be realized. I found.

  In other words, if the reactor and fuel can be designed so that the neutron absorption effect does not decrease even if the spectrum is cured, and the characteristic that the moving distance of the neutron becomes longer (the area becomes wider) as the void ratio increases, A new principle can be realized.

  Such a “new principle” can be realized by dispersing and arranging neutron absorbers whose neutron absorption capability does not decrease even when the spectrum is cured. In other words, this “new principle” utilizes the property of better absorbing neutrons of neutron absorbers arranged in a dispersed manner as the distance of neutron movement increases with increasing void fraction.

  As a result of numerical calculations, this "new principle" is difficult to apply to small LMFBRs with very hard spectra and nuclear reactors with high plutonium enrichment. (Area) proved promising. Moreover, it became clear that combining this principle with the prior art can increase the fuel loading and contribute to the economic improvement of nuclear power generation.

  The present invention has been made to solve the above-mentioned problems, and can improve the void coefficient adopting the above characteristics, and can safely use high-conversion or breeding-type light water or heavy water that does not use Na, such as a light water reactor. It is an object of the present invention to provide a fuel assembly and a reactor core capable of designing a nuclear reactor and improving the utilization rate of nuclear fuel.

  The fuel assembly according to the present invention is used in a high conversion type or breeding type nuclear reactor in which the coolant is light water or heavy water and the moderator-to-fuel volume ratio is 1.0 or less, and the nuclear fuel material is MOX fuel. In one fuel assembly, the fuel assembly is disposed between an upstream blanket disposed at a lower end portion thereof, a downstream blanket disposed at an upper end portion thereof, and the upstream blanket and the downstream blanket. An upstream main heat generating portion and a downstream main heat generating portion having a high Pu concentration, and a low concentration region having a low Pu concentration disposed between the upstream main heat generating portion and the downstream main heat generating portion. Features.

  According to the present invention, since a void coefficient improvement method based on a new principle can be obtained, a safe nuclear reactor using high conversion type or breeding type light water or heavy water that does not use Na, such as a light water reactor, can be designed. As a result, the utilization rate of nuclear fuel can be remarkably improved.

An embodiment of a fuel assembly according to the present invention will be described with reference to the drawings.
[First Embodiment]
FIGS. 1A to 1D are views for explaining a first embodiment of a fuel assembly according to the present invention. FIG. 1A is a longitudinal sectional view, and FIG. 1A is a cross-sectional view taken along line BB in FIG. 1A, and FIG. 1D is a cross-sectional view taken along line DD in FIG. 1A. .

  The fuel assembly 1 is configured by arranging a number of fuel rods 3 in a hexagonal lattice in a square channel box 2. Note that the upper and lower ends of each fuel rod 3 are fixed by upper and lower tie plates (not shown). Water having the functions of a coolant and a neutron moderator flows through the gap between the fuel rods 3 and 3 from the lower part (upstream side) to the upper part (downstream side) of the fuel assembly.

  In the fuel assembly 1 according to the present invention, the spacing between the fuel rods 3 and 3 is narrowed within an allowable range in order to increase the conversion ratio or proliferation rate of nuclear fuel. As a result, the moderator-to-fuel volume ratio ( (Hereinafter simply referred to as volume ratio) is small.

  This volume ratio was about 2 in a conventional nuclear reactor using water as a moderator, but in this embodiment using water, it is 1 or less and the target is 0.5 or less. As a nuclear fuel material, a pellet of mixed oxide fuel (hereinafter referred to as MOX) in which plutonium dioxide is mixed with uranium dioxide is filled in a cladding tube (not shown) to constitute a fuel rod 3.

  The basic configuration in the axial direction of the fuel assembly 1 shown in FIG. 1A is based on the example of a conventional fast reactor, and the upstream blanket 17 is upstream of the coolant at the lower end portion and the downstream side is at the upper end portion of the downstream side. There is a blanket 14, and the middle of both is a high concentration region with high Pu enrichment corresponding to the seed of the fast reactor.

  In the present embodiment, a low-concentration region 15 reminiscent of an internal blanket of a conventional fast reactor is arranged in the vicinity indicated by the DD arrow in FIG. In the low concentration region 15, the upstream side of the coolant is an upstream main heat generating portion 16, and the downstream side is a downstream main heat generating portion 13, and the fuel rod 3 is regularly arranged as shown in FIG. It is arranged.

  The low concentration region 15 is different from the conventional example in that the low concentration region 15 is not a blanket. As shown in FIG. 1 (d), the enriched boron with an increased concentration of boron 10 (B10) is used as a neutron absorber, and the absorption indicated by the mark ● The material segments 4 are arranged in a distributed manner. The Pu concentration in the low concentration region 15 is set to about 1/2 to 1/4 of the concentration in the upstream and downstream main heating portions (usually the Pu enrichment is 10 to 20%).

  As shown in FIG. 1B, the BB cross section of FIG. 1A and the cross section of the downstream blanket 14 on the downstream side of the fuel assembly 1 are two rows of fuel along the inner surface of the channel box 2. There is a space (moderator / coolant space, hereinafter simply referred to as a water gap) 5 in which rods 3 and 3 are arranged, and about one row of fuel is removed inside.

  Further, the interior is provided with a low scattering space 6 formed by bundling gas plenum portions of a large number of fuel rods 3 formed by providing gas plenum-like spaces above the fuel rods 3. In addition, in FIG.1 (b), FIG.1 (c), and FIG.1 (d), the square member attached to the inner surface of the channel box 2 is a water exclusion material.

Next, the operation of the fuel assembly in the first embodiment shown in FIG. 1 will be described with reference to FIGS.
FIG. 2 shows the calculated value of the infinite multiplication factor, and FIG. 3 shows the void reactivity effect calculation result. In a fuel bundle composed of 37 fuel rods bundled in a hexagonal lattice, the water density of the moderator was changed and neutron spectrum calculation was performed. In the calculation, the temperature of water was fixed at 20 ° C. for simplicity of explanation.

  The volume ratio when the water density is 1 was calculated as 0.45. As the fuel in the fuel rod 3, a mixture of depleted uranium oxide and 6% or 13% Pu as an oxide was used.

Calculation values for the case where all 37 fuel rods 3 have the same composition are shown by solid lines, and calculation is made when one absorbent-filled rod filled with B ₄ C powder made of natural boron at 70% of the theoretical density is placed. Values are shown with dashed lines. Since the concentration of B10 can be increased to 5 times or more that of this calculation example, the operation of the present embodiment, which will be described in detail below, can be further enhanced.

Fig. 2 shows the change accompanying the change in water density at infinite multiplication factor _K∞ , and Fig. 3 shows the characteristic that the reactivity of the calculation system changes (void reactivity effect) as the water density decreases. It shows how it changes depending on the presence or absence of a stick.

  From FIG. 2, it can be seen that if a long absorbent filling rod is used as in the case of a normal fuel rod, the reactivity of the core is greatly suppressed due to a decrease in infinite multiplication factor.

FIG. 3 shows that when the Pu enrichment is high (13% enrichment), the void reactivity is increased because the water density decreases in the positive direction as shown by the curves b ₀ and b _1. The effect is positive and the improvement effect in the negative direction (shift from b ₀ to b ₁ ) is small, but when the Pu enrichment is low (6%), the solid line a decreases as the water density decreases. _{As shown} by the broken line a ₁ from _0, the opening of the curve increases, and it can be seen that the void coefficient is improved in the negative direction.

  The reason why the characteristics differ depending on the enrichment of Pu is that the neutron spectrum becomes harder as the enrichment of Pu increases, so that the neutron absorption characteristics of the absorber decrease.

In this calculation example, natural boron B ₄ C powder is used, but if the saturation characteristics of the absorbent segment are improved using concentrated boron, an improvement effect appears even if the Pu enrichment is increased to about 13%. Come. From both figures, it can be seen that it is important that the length of the absorbent-filled rod, i.e., the absorbent segment, be relatively short at the relatively low portion of the Pu enrichment.

  The low concentration region of the first embodiment shown in FIG. 1 is considered under such a background. When depleted uranium is used in the low-concentration region, such as an internal blanket of a fast reactor, Pu is generated with the progress of combustion in the core, so that the effect of improving the void effect decreases with combustion.

  When the enrichment of Pu is only reduced as in the present embodiment, the Pu concentration does not increase even if the combustion proceeds, and the neutron absorber concentration is also kept high. In addition, the void coefficient improvement effect does not deteriorate. Moreover, although Pu is sufficiently enriched, it causes sufficient fission, so that the power distribution of the entire core is not deteriorated as in the case of using an internal blanket.

  According to this embodiment, a new principle for improving the void coefficient to negative can be adopted, so that the axial length of the core can be increased, and more nuclear fuel can be loaded into the core. , The economic efficiency of nuclear power generation is improved.

  In addition, the water gap 5 and the low scattering space 6 are provided in the blanket part downstream from the BB cross section of FIG. Although this part is not an essential condition in the present embodiment, it has a function of adjusting leakage of neutrons downstream, and can have a property of improving the void coefficient in cooperation with the low concentration region 15. it can.

  When the void ratio increases, the water gap 5 part is greatly voided, and the ratio of neutrons leaking downstream through the low scattering space 6 increases, and the amount of neutrons contributing to the nuclear reaction decreases, which improves the void coefficient. Can contribute. However, in this case, since the nuclear fuel is partially excluded, the heat generating portion is reduced, so that this portion is preferably stopped on a small scale.

4A to 4F illustrate various axial characteristics of the fuel assembly in the first embodiment shown in FIG.
(A) is a fission material concentration ε distribution, and a lower end (upstream side) and an upper end (downstream part) are blankets in which depleted uranium is arranged to capture leaking neutrons and generate Pu. The low concentration region is MOX fuel, and the upper and lower portions are high concentration regions. (B) is the absorption cross section Σ _a of the neutron absorber. (C) is an axial void ratio distribution.

(D) is an infinite multiplication factor _K∞ distribution. The infinite multiplication factor of the upper and lower blanket portions is naturally low, but it is also somewhat lower in the low concentration region where the absorbent is added. The infinite multiplication factor when the void ratio increases is schematically shown by a broken line. In FIG. 4D, (A) shows when the void ratio is high, and (B) shows when the void coefficient is high because it is positive.

  In the low-concentration region and the surrounding area (b), the neutron absorption effect by the absorber is increased due to the increase in the neutron movement area, so that it is low. On the other hand, the upper and lower high density regions indicated by (b) away from the low density region are originally positive void coefficient portions, and therefore the infinite multiplication factor is high.

  The portion where the infinite multiplication factor on the lower side of the downstream blanket at the upper end portion is lowered corresponds to the BB cross section of FIG. 1 (a), and is reduced because the nuclear fuel material is largely eliminated. As described above, this part is not an essential part of the present embodiment, but has the effect of improving the void coefficient in the negative direction by leaking neutrons, as in the present embodiment.

  (E) is an axial output distribution. In FIG. 4 (e), (a) shows when the void ratio is high, and (b) shows when the void coefficient is high because it is positive. As the void ratio increases, the void ratio changes as indicated by the broken line.

  In the lower half of the core, the void ratio is usually not so high and the void distribution does not increase so much, so the power distribution does not change much, but the neutron absorption effect of the absorber in the low concentration region increases and spatially spreads. The output decreases in the range indicated by (), the void ratio becomes considerably high on the downstream side indicated by (B), and there is a region where the output slightly increases because the void coefficient is a positive region.

  (F) is an axial distribution of neutron importance. It is well known that in a standard system with a uniform composition in the axial direction, the distribution of the cosine (cos) squared is well known. It is slightly reduced in the low concentration region containing the absorbent. Since the low concentration region is disposed at a place where the importance is high as a whole, the axial lengths of the absorber and the low concentration region can be effectively set to be short.

[Second Embodiment]
Next, a second embodiment of the fuel assembly according to the present invention will be described with reference to FIGS.
FIG. 5 is a reference diagram for enlarging FIG. 1 (d) and comparing it with FIG. 6. FIG. 6 is a cross-sectional view showing the main part of the fuel assembly 1 a in the present embodiment. The same parts are denoted by the same reference numerals. This embodiment is different from the first embodiment in that a fuel reduction segment (V rod) in which half of the six fuel rods 3 surrounding the absorbent segment 4 is represented by a cavity is used. 7 is replaced by the arrangement.

  According to the present embodiment, the effect of the absorbent can be further enhanced. In this example, three V rods 7 are arranged around one absorbent material segment 4, but it is not necessary to limit to three. Although it can be reduced to one or six, too much will lead to a decrease in nuclear fuel material, so a design that takes this into consideration is necessary.

[Third Embodiment]
Next, a third embodiment of the fuel assembly according to the present invention will be described with reference to FIG. The fuel assembly 1b in the present embodiment is a modification of the fuel assembly 1a in the second embodiment, and includes three V rods 7 and three W rods 8 around one absorbent material segment 4. It is in surrounding.

The W rod 8 is a moderator segment, and water, solid moderator ZrH ₂ , ZrD ₂ , C, Be, BeO, Al ₂ O ₃ , CeO _2, etc. are used as neutron moderators. In the present embodiment, an example is shown in which the same number of V bars 7 and W bars 8 are arranged. However, it is not necessary to arrange these bars in the same number.

[Fourth Embodiment]
Next, a fourth embodiment of the fuel assembly according to the present invention will be described with reference to FIG. FIG. 8 corresponds to FIG. 1 (d), and the fuel assembly 1c in the present embodiment is different from the first embodiment in that a larger number of absorber segments 4 are arranged regularly and dispersively. It is to have done. According to the present embodiment, the absorption effect is increased by increasing the number of the 23 absorbent segments 4 that are regularly distributed between the fuel rods 3 and 3 in the channel box 2, thereby reducing the concentration. The length of the region can be reduced.

[Fifth Embodiment]
Next, a fifth embodiment of the fuel assembly according to the present invention will be described with reference to FIGS.
FIG. 9 is a reference diagram shown for enlarging FIG. 1 (b) and comparing it with FIG. 10, and FIG. 10 shows a main part of the fuel assembly 1d of the present embodiment. This embodiment is different from the first embodiment in that a large number of large hollow tubes 9 larger than the diameter of the fuel rod 3 are bundled, and extra gaps are filled with a small tube 10 to form a low scattering space 6. It is to have done. According to the present embodiment, since the amount of the structural material that forms the space 6 can be reduced, the characteristics of the low scattering space 6 can be improved.

[Sixth Embodiment]
Next, a sixth embodiment of the fuel assembly according to the present invention will be described with reference to FIG. The present embodiment is different from the fifth embodiment in that the low scattering space 6 is constituted by a container 11 having a substantially square cross section and an internal reinforcing material 12 provided in the container 11. According to the present embodiment, since the low scattering space 6 is configured using fewer structural materials, a better low scattering space 6 can be configured.

[Seventh Embodiment]
Next, a seventh embodiment of the fuel assembly according to the present invention will be described with reference to FIG. FIG. 12 corresponds to FIG. 1A, and this embodiment is different from the first embodiment in that the downstream main heating region 13 shown by the BB cross section in FIG. The upper part is that all of the fuel rods 3 are filled as shown in FIG. The downstream blanket 14 has the same configuration.

  According to the present embodiment, if only this portion is exactly the same as the first embodiment, it may be insufficient for improving the void coefficient. Since the amount of nuclear fuel loaded can be improved and the amount of blanket fuel loaded can be increased, it is possible to contribute to the flattening of power distribution and Pu multiplication.

[Eighth Embodiment]
Next, an eighth embodiment of the fuel assembly according to the present invention will be described with reference to FIG. FIG. 13 corresponds to FIG. 1A, and this embodiment is different from the first embodiment in that a plurality of main heat generating regions 18 are arranged so as to be divided into a number of regions 18. , 18 is provided with a low concentration region 15. According to the present embodiment, since the axial length of the fuel assembly can be significantly increased while keeping the void coefficient negative, the amount of nuclear fuel loaded into the core can be greatly increased, and the economics of nuclear power generation can be improved. Can be improved.

[Ninth Embodiment]
Next, a ninth embodiment of the fuel assembly according to the present invention will be described with reference to FIG. FIG. 14 corresponds to FIG. 1A, and this embodiment is different from the first embodiment in that a main heat generating region 18 is mainly provided between the downstream blanket 14 and the upstream blanket 17. The low-concentration region 15 is provided between the upper end of the main heat generating region 18 and the downstream blanket 14.

  This embodiment is particularly suitable when the Pu enrichment degree of the main heat generation region is slightly low and the degree of improvement of the required void coefficient is not so large, and is absorbed between the downstream blanket and the main heat generation region. By disposing the low concentration region including the material segment, the reactivity loss of the fuel assembly can be reduced.

  If necessary, the present embodiment is possible even if the absorbent segments are dispersedly arranged in the cross-sectional direction with a length of 10 to 30 cm, for example, around ± 50 cm in the axial central portion where the Pu concentration in the main heating region 18 is slightly low. The effect of the form is obtained.

Although the embodiments of the fuel assembly according to the present invention have been described over the first to ninth examples, there are many other embodiments, and these will be supplementarily described.
In each of the above embodiments, the case of water (light water) as the coolant has been described. However, heavy water or a mixture of heavy water and light water may be used. In particular, when a mixture of light water and heavy water is used, the density of light water can be reduced by heavy water to suppress the neutron moderation characteristics.

  Further, the thickness of the cladding tube of the fuel rod of each fuel assembly described above can be increased to 1 to 2 mm from the conventional case (about 0.8 mm). Thereby, the atomic ratio of water to fuel can be reduced, and the neutron moderation characteristic can be suppressed.

In addition, the void coefficient has been described assuming a boiling water reactor, but even in the case of a pressurized water reactor, when the cooling water boils, a state of a coolant void is generated, which is the same as the problem of the void coefficient described above. A situation is assumed. As the strong neutron absorbing material has been described by taking only the _{B 4 C, EuB 6, Eu} 2 O 3 -Gd 2 O 3, HfH 2 , etc. can be used.

  Although it is assumed that the absorbent material segment 4 is disposed over the entire low concentration region 15, it may be shorter or longer. In the case where the low concentration region 15 is relatively long, there may be a case where it is better to dispose the absorbent segment 4 on one side of the low concentration region 15 as in the case where the absorber segment 4 is adjacent to the blanket.

  The most representative of the fuel reduction segment (V-bar) 7 is a gas, but in the present invention, most ceramics that can be used in a nuclear reactor can be used. It does not matter if fuel material or parent material is included. However, it is desirable that neutron scattering and absorption are not so large, but this is not a requirement.

The moderator of the moderator segment (W bar) 8 includes water, ZrH ₂ , ZrD ₂ , HfH ₂ , C, Be, BeO, Al ₂ O ₃ , CeO _2, etc., but the axial direction inside the fuel assembly of the segment It is preferable to dispose a small number of positions on the downstream side of the coolant flow where the void ratio increases. If the number is large, the hardened neutron spectrum will be too soft and the conversion ratio and growth ratio will be reduced. It should be noted that a small number of low density regions may be arranged.

Next, an embodiment of the fuel rod to be incorporated into the fuel assembly according to the present invention will be described.
FIG. 15 shows an embodiment of the fuel rod used in the fuel assembly of the first embodiment shown in FIG. FIG. 15A shows the fuel rod 3 disposed in the vicinity of the channel box 2.

That is, a depleted uranium oxide pellet (lower DUO ₂ ) 21, which is a lower blanket, and an upstream main heating region are formed in the cladding tube 19 from the lower end plug 20 side, that is, the lower side (upstream side of the coolant flow). The upstream MOX pellet 22, the MOX low concentration pellet 23 constituting the low concentration region formed by reducing the Pu concentration, the downstream MOX pellet 24 constituting the downstream main heating region, and the depleted uranium oxide pellet constituting the downstream blanket (Upper DUO ₂ ) 25. Further, the downstream side is a gas plenum 26 (not shown in FIG. 1A), and the upper end is sealed with an upper end plug 27.

  FIG. 15 (b) shows a short fuel rod 28 which is disposed inside the fuel rod shown in FIG. 15 (a) and is shortened to form a moderator / coolant space (water gap) on the downstream side. Thus, the downstream portion is the burnishing portion 29. The downstream MOX section is slightly shortened, the downstream depleted uranium region is removed, and the gas plenum 26 is moved to the lower end plug 20 side. The plenum 26 includes a ring-shaped metal tube 30 that supports the MOX pellet 22 and a heat insulating pellet 31.

FIG. 15C corresponds to the first fuel rod of the present invention, but a long gas plenum 32 is provided to provide the low scattering space 6 on the downstream side. A stack of B ₄ C pellets 33 is arranged substantially in line with the axial height of the stack of MOX low-concentration pellets 23 in FIG. This B ₄ C pellet 33 acts as an absorber segment using concentrated boron as a neutron absorber.

The top and bottom of a stack of B ₄ C pellet 33 is disposed a metal wool 34, and prevents any chance of mixing during powder occurrence of MOX pellets 22, 24 and B ₄ C pellet 33. By using zirconium for the metal wool 34, it is possible to adsorb tritium slightly generated by the nuclear reaction of B ₄ C.

  FIG. 15 (d) corresponds to the second fuel rod of the present invention, but a stack of MOX low concentration pellets 23 is provided to enhance the features of the present invention. On the downstream side, a long low-scattering space of the gas plenum 32 is provided for the same purpose as in FIG.

  FIG. 16 schematically illustrates a standard configuration of various fuel rods used in the present invention. FIG. 16 (a) is a first fuel rod of the present invention corresponding to FIG. 15 (c). FIG. 16B shows the second fuel rod of the present invention, which is mostly occupied by the full length main heating type MOX fuel 35.

FIG. 16C shows the third fuel rod of the present invention, which corresponds to the V rod 7 shown in FIG. That is, the fuel reduction segment 36 is configured in the same manner as the gas plenum 26 provided at the lower end of FIG. 15 (b) with substantially the same height and length as the absorbent segment made of the B ₄ C pellet 33 in FIG. 16 (a). And constitutes an intermediate plenum 37.

  FIG. 16D shows the fourth fuel rod of the present invention, which corresponds to the W rod 8 shown in FIG. That is, the moderator segment 38 is configured over substantially the same length as the absorbent material segment in FIG. Such a configuration can be used when a non-hydrogen-containing moderator is used. In the case of a moderator containing hydrogen, it is natural that the moderator must be isolated from the fuel filling space.

  FIG. 16 (e) shows the fifth fuel rod of the present invention, and the configuration of the moderator segment arrangement portion is the same as in FIG. 16 (d). A moderator segment 38 is provided on the downstream side of the stack of the MOX low-concentration pellets 23 so as to contribute to improvement of the void coefficient in the downstream seed heat generation region. The configuration shown in FIG. 16 is a very typical configuration, and various modifications are naturally expected to realize the idea of the present invention.

[First Embodiment of Reactor Core]
Next, a first embodiment of a reactor core according to the present invention will be described with reference to FIGS.
FIG. 17 (a) is a cross-sectional view schematically showing the reactor core in the present embodiment, and FIG. 17 (b) is a longitudinal direction of the fuel assembly loaded at the position “1” in FIG. 17 (a). FIG. 17C is a cross-sectional view showing a conventional fuel assembly loaded at a position other than “1” in the core of FIG.

  The fuel assembly shown in FIG. 17 (b) is used to improve the “core coolant void coefficient” that is most necessary for the safe and reliable operation of the reactor. Depending on the design, the void coefficient can be greatly improved, or it can be improved on a small scale.

  The feature of the new fuel assembly configuration to improve the void coefficient is that the plutonium concentration is locally reduced and neutron absorbers are distributed in the vicinity of this. However, there is a drawback that the degree of burnup of the fuel is lowered. Therefore, it is necessary to perform an optimization design. The core shown in FIG. 17A is configured based on such a background.

  The basic structure in the axial direction of the fuel assembly shown in FIG. 17 (b) is that there is an upstream blanket 17 upstream of the coolant at the lower end and a downstream blanket 14 at the upper end of the downstream. This fuel assembly is an internal blanket that forms a high concentration fuel region with high Pu enrichment corresponding to the seed of the fast reactor, but unlike that, a low concentration region 15 with reduced plutonium concentration is arranged. . The fuel assembly schematically shown in FIG. 17 (c) is a conventional fuel assembly without a low concentration region.

  18 (a) to 18 (c) are examples of a cross-sectional view of the fuel assembly of FIG. 17 (b), and FIG. 18 (a) is a cross-sectional view taken along arrow AA in FIG. 18 (b) is a cross-sectional view taken along the line BB in FIG. 17 (b), FIG. 18 (c) is a cross-sectional view taken along the line CC in FIG. 17 (b), and FIGS. ) Is an example of the fuel rod used in the fuel assembly of the present invention, and FIG. 19A shows the fuel rod arranged along the inner surface of the channel box.

In the cladding tube 19, from the lower end (upstream side of the coolant), a stack of uranium oxide pellets (lower DUO ₂ ) 21 using depleted uranium, a stack of upstream MOX pellets 22 constituting the upstream main heat generation region, the present invention The stack of MOX low concentration pellets 23 with reduced Pu concentration, the stack of downstream MOX pellets 24 constituting the downstream main heat generating region, and the upper DUO ₂ play a major role in ensuring the action of the absorbent segment at 25 and gas plenum 26.

In FIG. 19B, a moderator / coolant space (hereinafter referred to as a water gap) is formed inside the fuel rod arrangement shown in FIG. 19A, or a downstream water removal container is arranged. This is an example of a fuel rod shortened for this purpose, in which the downstream upper DUO ₂ 25 region is removed.

FIG. 19C is a fuel rod particularly important for generating the operation of the present invention, and is shortened for the same purpose as in FIG. 19B. The difference from the fuel rod in FIG. 19B is that the stack of B ₄ C pellets 33 is sandwiched between the metal wool 34 on the upper and lower sides at almost the same height as the stack of the MOX low concentration pellets 23 in FIG. It is the point which has been arrange | positioned and comprised the "absorbent material segment".

When wool-like zirconium is used as the metal wool 34, a small amount of tritium generated by the reaction of boron with neutrons (which may corrode the cladding tube 19) is selectively adsorbed. Soundness is ensured. Another purpose of the metal wool 34 is to prevent mixing of the fuel material with a small amount of B ₄ C which may occur.

  By the way, in the fuel assembly according to the present invention, in order to increase the conversion ratio or proliferation ratio of the nuclear fuel, the interval between the fuel rods is narrowed within an allowable range. Called volume ratio).

  This volume ratio was about 2 in a conventional nuclear reactor using water as a moderator, but in this embodiment using water, it is 1 or less and the target is 0.5 or less. As a nuclear fuel material, MOX pellets are filled in a cladding tube.

  In this way, even a light water cooled nuclear reactor allows high conversion or proliferation of nuclear fuel, but the void coefficient of water is positive. Therefore, if the volume ratio is reduced with the conventional nuclear reactor configuration, it has been difficult to avoid the fact that safe operation of the nuclear reactor becomes difficult.

  On the other hand, the greatest feature of the present invention is that a low-concentration fuel region reminiscent of an internal blanket of a conventional fast reactor is arranged in the vicinity of the cross-sectional portion taken along the line CC in FIG. The upstream side of the coolant in the low-concentration region is an upstream main heat generating portion, and the downstream side is a downstream main heat generating portion, and the fuel is regularly arranged as shown in FIG.

  The low-concentration fuel region is different from the conventional example in that it is not a blanket. As shown in FIG. 18 (c), the absorbent segments using the enriched boron in which the concentration of boron 10 (B10) is increased as a neutron absorber are dispersed. Has been placed. The Pu concentration in the low concentration fuel region is about 1/2 to 1/4 of the concentration in the upstream and downstream main heat generating portions (usually the enrichment of Pu is 10 to 20%).

  On the downstream side of the fuel assembly, the AA arrow cross section in FIG. 17 (b) is a downstream blanket portion, and two rows of fuel rods are arranged along the inner surface of the channel box as shown in FIG. 18 (a). In addition, there is a space (moderator / coolant space, ie, water gap) in which about one row of fuel is removed, and a container 11 having a rectangular cross section is disposed inside the fuel rod without the fuel rod. .

  The space occupied by the vessel 11 is a low-scattering space for neutrons provided to suppress neutron scattering and to leak neutrons to the outside (downstream side) of the core as necessary. The purpose of the container 11 is to greatly suppress neutron scattering, but it is not necessary to force the inside to be evacuated. If deformation occurs during the operation of the reactor, the reactivity of the core will be affected, so that the structure is reinforced with the internal reinforcement 12.

  The thickness of the water gap is adjusted within the range of 0.5 cm to 3 cm, corresponding to the required amount of void coefficient adjustment, to the extent that neutrons leak downstream through the low scattering space. In FIGS. 18A, 18B and 18C, a large number of square members attached to the inner surface of the channel box 2 are water draining materials.

  The fuel assembly is configured by arranging a large number of fuel rods in a square channel box 2 in a hexagonal lattice shape, and water having the functions of a coolant and a neutron moderator fills the gap between the fuel rods and the fuel. It flows from the bottom to the top of the assembly.

[Second Embodiment of Reactor Core]
Next, a second embodiment of the reactor core according to the present invention will be described with reference to FIG. FIG. 20 corresponds to FIG. 17 (a). In FIG. 17 (a), a total of 44 fuel assemblies as shown in FIG. However, in the present embodiment, the fuel assembly shown in FIG. 17 (b) is basically arranged so that the core is cut into a cross shape, and the shortage is shown in FIG. 17 (b). It is the structure which put together 36 each and arranges a total of 36 bodies.

[Third embodiment of reactor core]
Next, a third embodiment of a reactor core according to the present invention will be described with reference to FIG. FIG. 21 corresponds to FIG. 17A, and such an arrangement is suitable for improving the void coefficient on a relatively large scale, and has a configuration in which the first embodiment is further emphasized. That is, the fuel assemblies shown in FIG. 17 (b) are paired by two, the core is cut into a cross shape, and six are disposed in the insufficient portion, for a total of 68.

  If it is necessary to further improve the void coefficient, for example, a fuel assembly as shown in FIG. 17B may be used for all fuel assemblies except for the periphery of the core. This fuel assembly can also be arranged on the outermost periphery. Further, when it is necessary to further improve the void coefficient, it is possible to use a single fuel assembly provided with a plurality of low concentration regions in the axial direction.

As mentioned above, although the embodiment of the core of the nuclear reactor according to the present invention has been described over many examples, there are many other examples. The void coefficient has been described assuming a boiling water reactor, but even in the case of a pressurized water reactor, if the cooling water boils, the same situation as the above-described void coefficient problem is assumed. As the strong neutron absorbing material has been described by taking only the _{B 4 C, EuB 6, Eu} 2 O 3 -Gd 2 O 3, HfH 2 , etc. can be used.

  Although the absorbent material segment is assumed to be disposed over the entire low concentration region, it may be shorter or longer. When the low-concentration region is relatively long, it is conceivable that the position of the absorbent material segment is preferably shifted to one side of the low-concentration region, as in the case where it is adjacent to the blanket. A low scattering segment, such as an internal gas plenum that excludes the fuel material, can also be placed around the absorbent segment.

It is also possible to place a moderator segment around the absorbent segment, for example downstream of the coolant flow. Water, ZrH ₂ , ZrD ₂ , HfH ₂ , C, Be, BeO, Al ₂ O ₃ , CeO ₂ or the like can be used for the moderator of the moderator segment.

  When introducing a moderator segment containing hydrogen or deuterium into a part of the fuel rod, it must be separated from the fuel region and airtight with respect to the fuel region so that those gases do not enter the fuel part. However, the other moderators can be separated from the fuel material by interposing a wool metal material.

(A) is a schematic diagram schematically showing the axial configuration of the fuel assembly in the first embodiment of the fuel assembly according to the present invention, and (b) is a cross section in the direction of arrow BB in (a). The figure, (c) is CC arrow direction cross-sectional view of (a), (d) is the DD arrow direction cross-sectional view of (a). The graph figure which plots and shows the calculated value of the infinite multiplication factor for demonstrating the effect | action of this invention. The graph figure which plots and shows the void reactivity effect calculation value for demonstrating the effect | action of this invention. (A)-(f) is a figure for demonstrating the axial characteristic of the fuel assembly of the 1st Embodiment of this invention typically, (a) is a fission nuclide density | concentration distribution map, (b). (C) is a void ratio distribution map, (d) is an infinite multiplication factor distribution map, (e) is an output distribution map, and (f) is an neutron importance distribution map. The longitudinal cross-sectional view which expands and shows FIG.1 (d). The cross-sectional view which shows the principal part of 2nd Embodiment of the fuel assembly which concerns on this invention. The transverse cross section which shows the principal part of 3rd Embodiment of the fuel assembly which concerns on this invention. The cross-sectional view showing a fourth embodiment of the fuel assembly according to the present invention. The longitudinal cross-sectional view which expands and shows FIG.1 (b). The cross-sectional view showing a fifth embodiment of the fuel assembly according to the present invention. The cross-sectional view showing the sixth embodiment of the fuel assembly according to the present invention. The schematic diagram which shows roughly 7th Embodiment of the fuel assembly which concerns on this invention. The schematic diagram which shows roughly 8th Embodiment of the fuel assembly which concerns on this invention. The schematic diagram which shows roughly 9th Embodiment of the fuel assembly which concerns on this invention. (A)-(d) is a figure for demonstrating embodiment of the various fuel rods integrated in the 1st fuel assembly of this invention shown in FIG. 1, (a) is a fuel rod arrange | positioned in the channel box vicinity. (B) is a longitudinal sectional view schematically showing a shortened fuel rod used for forming a water gap on the downstream side, and (c) is a diagram showing a B ₄ C pellet stack. The longitudinal cross-sectional view which shows schematically the fuel rod accommodated in the part, (d) is the longitudinal cross-sectional view which shows schematically the fuel rod which has the gas plenum in the MOX low concentration area | region arrange | positioned inside a fuel assembly and the downstream of a coolant Figure. (A)-(e) is a figure for demonstrating embodiment of the standard fuel rod used for the fuel assembly which concerns on this invention, (a) is a longitudinal cross-section which shows a 1st fuel rod roughly (B) is a longitudinal sectional view schematically showing a second fuel rod, (c) is a longitudinal sectional view schematically showing a third fuel rod, and (d) is a schematic showing a fourth fuel rod. (E) is a longitudinal sectional view schematically showing a fifth fuel rod. (A) is a core layout diagram for explaining a first embodiment of a nuclear reactor core according to the present invention; (b) is a longitudinal direction of a fuel assembly according to the present invention indicated by “1” in (a). The schematic diagram which shows a cross section, (c) is a schematic diagram which shows the longitudinal direction cross section of the conventional fuel assembly. (A) is a cross-sectional view in the direction of arrows AA in FIG. 17 (b), (b) is a cross-sectional view in the direction of arrows BB in FIG. 17 (b), and (c) is in FIG. 17 (b). CC arrow direction cross-sectional view. (A)-(c) is a figure for demonstrating embodiment of the various fuel rods integrated in the fuel assembly which concerns on this invention in FIG.17 (b), (a) is a fuel rod arrange | positioned in the channel box vicinity. FIG. 4B is a longitudinal sectional view schematically showing a fuel rod having a MOX low concentration region disposed inside the shortened fuel assembly, and FIG. 4C is a B ₄ C pellet. The longitudinal cross-sectional view which shows schematically the fuel rod which accommodated the stack in part. The core layout which shows 2nd Embodiment of the core of the reactor which concerns on this invention. The core layout which shows 3rd Embodiment of the core of the reactor which concerns on this invention.

Explanation of symbols

DESCRIPTION OF SYMBOLS 1 ... Fuel assembly, 2 ... Channel box, 3 ... Fuel rod, 4 ... Absorber segment, 5 ... Water gap, 6 ... Low scattering space, 7 ... Fuel reduction segment (V rod), 8 ... Moderator segment (W (Rod), 9 ... large diameter hollow tube, 10 ... small diameter tube, 11 ... container, 12 ... internal reinforcement, 13 ... downstream main heating region, 14 ... downstream blanket, 15 ... low concentration region, 16 ... upstream side main heating region, 17 ... upstream blanket, 18 ... main heating region, 19 ... cladding tube, 20 ... lower end plug, 21 ... lower DUO _2, 22 ... upstream MOX pellets, 23 ... MOX low density pellets, 24 ... Downstream MOX pellet, 25 ... Upper DUO ₂ , 26 ... Gas plenum, 27 ... Upper end plug, 28 ... Short fuel rod, 29 ... Burnishing part, 30 ... Metal tube, 31 ... Thermal insulation pellet, 32 ... Long gas plenum, 33 ... B ₄ C pellet, 34 ... metal wool, 35 ... full length main heating type MOX fuel, 36 ... fuel reduction Segment, 37 ... intermediate plenum, 38 ... moderator segment.

Claims (10)

In a fuel assembly in which the coolant is light water or heavy water, used in a high conversion or breeder reactor with a moderator to fuel volume ratio of 1.0 or less, and the nuclear fuel material is MOX fuel,
The fuel assembly includes an upstream blanket disposed at a lower end thereof, a downstream blanket disposed at an upper end thereof, and an upstream having a high Pu concentration disposed between the upstream blanket and the downstream blanket. A fuel assembly comprising: a side main heat generating portion and a downstream main heat generating portion; and a low concentration region having a low Pu concentration disposed between the upstream main heat generating portion and the downstream main heat generating portion. .   2. The fuel assembly according to claim 1, wherein the Pu concentration in the low concentration region is 1/2 to 1/4 of the Pu concentration in the upstream main heat generating portion and the downstream main heat generating portion.   The fuel assembly according to claim 1 or 2, wherein the absorber segments of the fuel assembly are arranged in a dispersed manner.   The fuel assembly according to any one of claims 1 to 3, wherein a moderator / coolant space is provided in a downstream side of the fuel assembly and in the downstream blanket.   The fuel assembly according to claim 4, wherein a cross-sectional width of the moderator / coolant space is 0.5 to 3 cm.   The fuel assembly according to any one of claims 1 to 5, wherein a gas plenum-like space is provided on the downstream side of the fuel assembly and on the upper portion of the fuel rods arranged on the downstream blanket.   6. A plurality of large-diameter hollow tubes and a plurality of small-diameter tubes are arranged in a space surrounded by a moderator / coolant space in a downstream blanket portion of the fuel assembly. The fuel assembly according to claim 1.   The fuel according to any one of claims 1 to 5, wherein a container having a substantially square cross section is arranged in a space surrounded by a moderator / coolant space in a downstream blanket portion of the fuel assembly. Aggregation. The neutron absorber of the neutron segment includes natural boron or concentrated boron enriched with B10, boron carbide (B ₄ C), europium hexaboride (EuB ₆ ), mixed oxide of Europia and gadolinia (Eu ₂ O ₃ − The fuel assembly according to any one of claims 1 to 8, wherein the fuel assembly is made of Gd ₂ O ₃ ), HfB ₂ or a mixture thereof, or a hydride of hafnium.   A nuclear reactor core, wherein the fuel assembly according to any one of claims 1 to 9 is loaded on at least a part of the core.

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