JPS58179391A - Fuel flux having enrichment devided axially - 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 the Invention This invention relates to nuclear reactors and, more particularly, to fuel bundle configurations for boiling water nuclear reactors.

BACKGROUND OF THE INVENTION In a nuclear reactor, fissile material (e.g. uranium J3
The atom j) absorbs neutrons into its nucleus and undergoes nuclear decay, resulting in the production of an average of two low atomic mass fission fragments with high kinetic energy and several neutrons, also of high energy.

In a typical boiling water reactor (BWR), the nuclear fuel is in the form of fuel rods, each consisting of a plurality of sintered pellets contained within an elongated cladding tube.
Seven groups of such fuel rods are supported between upper and lower tie plates to form an independently replaceable fuel assembly or bundle. By arranging a sufficient number of such fuel assemblies in a matrix so that the whole approximates a right circular cylinder, a reactor core capable of a sustained nuclear fission reaction is formed. The kinetic energy of the fission products is dissipated as heat within the fuel rods. Neutrons, gamma rays and other radiation resulting from nuclear fission processes also release energy into the fuel structure and moderator.
The core is submerged in a coolant (such as water), and the heat removed can be extracted and used to perform useful work. When the coolant is water, it also serves as a neutron moderator, slowing down neutrons and making them more susceptible to nuclear fission reactions.

The fuel commonly used in power reactors that use water as a coolant and moderator consists of uranium dioxide, which is fissile uranium of about 0.7 to about 0.23 degrees and is the parent material of the fuel. Contains a mixture of uranium-23 and During operation of a nuclear reactor, uranium-23e, a parent fuel substance, is ?
A portion of this is converted to fissile plutonium, 239 and plutonium-62. Uranium-623 is also fissile, but it can only be fissioned by high-energy neutrons. Note that the fissile materials produced (e.g. plutonium, 239 and plutonium 2)
/) and fissile materials that are destroyed (e.g. uranium-23
The ratio between plutonium-239 and plutonium-2g/) is defined as the "conversion ratio".

When operating a nuclear reactor at a steady power level, the number of neutrons that induce fission must remain constant.

That is, each nuclear fission reaction should generate a net of seven neutrons, and the neutrons cause the next nuclear fission reaction, thereby sustaining operation. The operating state of the nuclear reactor is indicated by the effective multiplication factor keff, and for steady operation the value must be /.

It should be noted that the effective multiplication factor kcff is the neutron regeneration rate when considering the nuclear reactor as a whole, and needs to be distinguished from the local or infinite multiplication factor kink. The infinite multiplication factor kinr defines the neutron regeneration rate of an infinitely large system that has the same composition and properties throughout the local region of the core in question.

It is also true that during operation, fissile fuel is depleted, and some fission products may themselves constitute neutron absorbers or "poisons." To counter this, it is customary to initially load a nuclear reactor with an excess amount of nuclear fuel, resulting in an initial surplus of reactivity. The existence of such initial excess reactivity allows the effective multiplication factor to be maintained at / during reactor operation, while at the same time preventing the effective multiplication factor from decreasing below 7 when the reactor is required to shut down. A control system is required. In such control systems, it is customary to use neutron absorbing materials that help control the number of neutrons by non-fissile absorption or capture of neutrons. Although fissile fuel itself is also a neutron absorbing material in a sense, the term "-neutron absorption" in this specification is used to exclude fissile absorption.

At least a portion of the neutron absorbing material is incorporated into multiple control rods that can be selectively operated. A rod is inserted from the bottom of the core along the axial direction.
Combustible absorbent material can also be introduced into some fuel rods to reduce the required mechanical controls. A flammable absorbing material is a neutron absorbing material that is converted into a substance with low neutron absorption capacity by neutron absorption. A known combustible absorbent is gadolinium, usually in the form of cadlinium. Its odd-nuclear isotopes (cadrinium/jj and gadolinium/j7) exhibit very large capture cross sections for thermal neutrons.

However, these combustible absorbers that can be used during design are undesirable in that neutron absorption reactivity remains even at the end of the refueling cycle due to residual neutron absorption by isotope absorbers with small neutron absorption cross sections. For example, when gadolinium is used in combination with combustible absorbers, isotopes with large cross sections (gadolinium/3O and gadolinium/J2) are rapidly depleted, whereas even-nuclear isotopes (gadolinium 15g, gadolinium 15o) are depleted rapidly. Residual neutron absorption persists because neutron capture by Gadolinium and Gadolinium Bl'' continues.

As is known, combustible absorbent materials such as gadolinium act as self-shielding agents when incorporated into fuel rods in sufficient concentrations.
That is, when exposed to a neutron flux, neutron absorption occurs essentially at the outer surface of the absorber-containing twisted rod. As a result, the volume of the absorbent material contracts along the radial direction;
Its speed depends on the concentration of absorbent. In such cases, by choosing the number of absorbent-containing twist rods and the absorbent concentration therein appropriately, it is possible to vary the absorption value as desired over seven or more reactor operating cycles. .

During operation, since the proportion of steam voids increases towards the top of the reactor, the degree of deceleration in the upper region decreases and the power distribution is therefore skewed towards the lower region of the core. To compensate for this, methods are known in which the combustible absorbent material is distributed unevenly along the axial direction. That is, combustible absorbent material is introduced into some of the fuel rods with a distribution that increases toward the axial region where it exhibits maximum reactivity during hot operation. Typical configurations of the diet are described in US Pat. Nos. 3,799 and 39.

Still no U.S. patent! Another method is described in the Gugu 2 Zagu specification. In this case, the axial power peaking is compensated to a certain extent by increasing the uranium-23.3 enrichment or plutonium enrichment in the upper region of the core.

By the way, the situation is completely different during cold shutdown.

Specifically, in the cold state, the upper part of the irradiated BWR core exhibits a higher reactivity than the lower part.

This is because during operation, more plutonium is produced in the upper part and less uranium-23j is destroyed.
(That is, the upper part of the core has a high conversion ratio and low burnup). In fact, during cold shutdown, the steam voids in the upper part of the core are eliminated, so the upper part of the core exhibits a higher degree of reactivity than the lower part. According to typical approval standards, even if any seven control rods are pulled out of the reactor core, the shutdown margin for 03 and the staff (099 Otsu 2)
kcff) is required to exist. Note that in order to take prediction uncertainty into account, it is customary to use /% (keff less than 099) as the design standard for the predicted shutdown margin to be provided by control rods and combustible absorbers.

Adjustment of the axial power distribution can be achieved by increasing the amount of combustible absorber or reducing its enrichment in the lower part of the core, but the optimal absorber and fuel distribution for the axial power distribution at maximum power is It does not help maintain downtime margin. It is usually necessary to design with an excess of residual neutron absorber to meet cold shutdown constraints, but this has disadvantages for initial enrichment and uranium ore requirements. At the same time, it increases the fuel cycle cost of the nuclear reactor.

Another problem is that gadolinia reduces the thermal conductivity of the fuel rods and increases fission gas release. As a result, gadolinia-containing fuel rods are often the most restrictive fuel rods in a fuel assembly. Therefore,
It is necessary to set the rated power of the gadolinia-containing fuel rod low, and the local power distribution will be adversely affected accordingly. Although the extent to which the rated output should be reduced depends on the gadolinia concentration, the design of long-term combustion fuel bundles and (yet) high-energy cycle designs where high gadolinia concentrations are required to provide sufficient cold shutdown margin. This becomes a serious problem.

Thus, the conditions required during wet operation and cold shutdown impose competing constraints on the design of a nuclear reactor core, which tend to prevent the achievement of an optimal core configuration.

SUMMARY OF THE DISCOVERY The present invention is directed to the construction of fuel assemblies for nuclear reactors that enable the achievement of cold shutdown margins with minimal reduction in operating efficiency. In implementing the present invention,
Little or no special fuel pellets or the like are required.

Generally speaking, a fuel assembly according to the present invention includes a fissile material component distributed over a substantial axial extent thereof, such fissile material component exhibiting maximum reactivity during cold shutdown. an upper axial zone (i.e. "low enrichment zone") which corresponds to at least a portion of the axial zone;
has a relatively reduced concentration.

Such fuel assemblies typically consist of a first group of fuel rods having a reduced enrichment in a low enrichment zone and a Ωth group of fuel rods having a uniform enrichment along the axial dimension. In order to minimize the number of fuel rods that should have non-uniform enrichment along the axial direction, the first group of fuel rods preferably includes the highest enrichment fuel rods used in the fuel assembly. Further, it is preferable that the enrichment in the low enrichment zone corresponds to the enrichment exhibited by any of the uniform fuel rods. The fuel rods in the first group can then be manufactured from fuel pellets with standard enrichment.

The presence of such low enrichment zones causes the axial power peak to be slightly higher toward the bottom of the fuel assembly, but this can be compensated for by increasing the amount of combustible absorbent material in the corresponding lower zone.

As a result, the intermediate zone with high enrichment and low gadolinium content constitutes the part of the reactor core with the greatest reactivity.

It is recognized that the low enrichment zone as described above has the effect of reducing the reactivity during cold shutdown without having a very slight effect on the excess reactivity in the wet state. As a result, more cycle energy can be obtained with the same amount of uranium, 233, in the fuel assembly because the fuel is used more efficiently. The large control width made possible by the present invention provides margin for significant unplanned fluctuations in cycle energy, so flexibility can be obtained even for reactor operations that deviate from the originally planned cycle energy. Become.

In order that the nature and advantages of the invention may be more clearly understood, a more detailed description will be given below with reference to the accompanying drawings.

DESCRIPTION OF THE PREFERRED EMBODIMENT FIG. 1 shows a boiling water reactor 1 using water as a coolant and moderator.
0 is a partially cutaway side view schematically illustrating FIG.

Such a nuclear reactor includes a pressure vessel 11 within which a reactor core 12 is submerged in a moderating coolant, such as light water. The core 12 consists of a plurality of fuel cells 13 surrounded by an annular shroud 14. Each fuel cell includes a fuel assembly (or fuel tube) 15 and a control rod 16. The fuel cells are held apart from each other by an upper core grid 18 and a lower core plate 19, each supported at the bottom by a suitable support 20. Control rods can be selectively inserted between fuel assemblies to control core reactivity. Each control. Figure 2 shows the core 1
FIG. 2 is a road view showing the arrangement of fuel cells 13 inside the vehicle. It should be understood that a typical reactor core includes fuel assemblies of the order of 30[theta] to 900.

A portion of the pressure vessel 11 below the core 12 forms a coolant supply chamber 22 , while a portion below the core 12 includes a steam separation and drying device 25 .

In operation, the coolant recirculation pump 27 pressurizes the coolant in the coolant supply chamber 22, thereby forcing the coolant upwardly through the reactor core 12. Such coolant absorbs the heat generated by the nuclear fission reactions occurring within the reactor core, with the result that some of the coolant is converted to steam. This vapor is separated and dried. The exhaust steam is condensed by a condenser 32 in series with the turbine, and the condensate thus obtained is returned to the input side of the coolant recirculation pump 270 as feed water by a condensate return pump 35.

Turning now to FIG. 3, the structure of seven fuel cells 13 is shown. The control rod 16 has a cruciform cross-section, and each of the control vanes 40 defined thereby is disposed between λ adjacent fuel assemblies. Each fuel assembly 15 consists of a plurality of elongated fuel rods 42 held between an upper tie plate 45 and a lower tie plate 46 and surrounded by a tubular channel box 48 of rectangular cross section. The lower part of the fuel assembly is equipped with a nose 50 having an opening 52 through which coolant flows upwardly along the fuel rods 42 within the channel box 48. In addition, nose 5
0 is adapted to fit into a corresponding socket (not shown) in the fuel assembly support 20.

As is known, each fuel rod may be comprised of a cylindrical cladding tube and a plurality of enriched uranium oxide and/or plutonium oxide sintered pellets contained therein.
The enrichment varies from fuel rod to fuel rod within the seven fuel assemblies, and typically ranges from about 02 to θ (weight)% with an average value of about /! ~3. ! (by weight)% (for natural uranium, it is θ2 (weight)%). The diameter of the fuel rod is approximately A inch and the length is approximately /, 2~/! It can be feet.

Figures gA and 98 show the lateral (or horizontal) orientation of fuel rods in two types of replacement fuel assemblies according to the present invention.
Figure 2 is a schematic diagram showing the distribution. As explained below, these types of fuel assemblies differ primarily in the number of gadolinia-containing fuel rods, and a typical core will include both. The fuel rods 42 are arranged in a matrix of holes, with the two central fuel rod positions occupied by cooling water channels 55. Note that each fuel rod is shown as a circle with a code number written inside.

FIG. 96 shows the longitudinal composition of such a fuel rod. Each fuel rod has a blanket of natural uranium in the top six inch area 60, but the warping blanket will be ignored for the remainder of the discussion. The fuel rods with the second ~ (aside from the blanket)
The first fuel rod has a segmented enrichment along the axial direction, while the first fuel rod has a uniform enrichment along the axial direction over the entire length of the first fuel rod. More specifically, the first fuel rod has a lower enrichment in the upper 3 foot zone 62 (the "low enrichment zone") compared to the lower 9 foot zone 65. The second and second fuel rods contain gadolinia as a combustible absorbent material. The concentration is divided along the axis, with an upper 7-foot area 68
The concentration is higher in zone 67, j feet below (the "high gadolinia zone"). As a result, the core can be considered to consist of three axial zones: a low enrichment zone 62, an intermediate zone 70 (which is the overlap of zones 65 and 68), and a high gadolinia zone 67.

Figures 3A and 5B show the overall axial distribution of enrichment and gadolinia content (averaged over all fuel rods) for the fuel assemblies of Figures &A and 98, respectively. Although it is recognized that in these examples the enrichment loss in zone 62 is equal to about 0.2 (by weight) H, the preferred range for such enrichment loss is about 0.7! ~θ
3 (weight). This range has an average enrichment of about 3
0% (by weight) and the enrichment of each fuel rod is generally within the range of 0% to 0% by weight. For the embodiment shown in Figures
Therefore, the concentration reduction rate corresponds to about j to 10. Similar reduction rates should be used in more enriched fuel assemblies, so that the (absolute amount of) enrichment reduction in zone 62 increases correspondingly with enrichment.

To achieve the desired degree of axial segmentation while minimizing the number of fuel rods that need to be segmented along the axial direction, the axial segmentation should be done for the fuel rods with the highest working enrichment. It is preferable to apply The low enrichment in zone 62 preferably corresponds to a uniform enrichment in the other fuel rods in the fuel assembly, so as to avoid requiring pellets of a more delicate or special enrichment. That is, for the particular embodiment illustrated, the 330 (by weight) enrichment in section 62 of the first fuel rod corresponds to the uniform enrichment of the second fuel rod.

Diagram B A-GC relates to a nuclear reactor that includes a core consisting of fuel assemblies such as those shown in FIGS. KA and 4B and is operated with all control rods withdrawn.
BOC), mid-cycle (MOC), and end-of-cycle (EOC) relative power as a function of axial position are shown in graphs 80a-80C, respectively. These graphs are normalized using average output as a unit. Also, for comparison, a corresponding graph 82 of the axial power distribution for a core with uniform enrichment.
A to 8-2C are indicated by dotted lines in Fig. gA-C. In BOC, as can be seen from the diagram, area 67
As a result of the increased amount of absorbent material (gadolinia) in zone 62 counteracting the effect of the enrichment loss in zone 62, the operating characteristics obtained with the present invention are approximately those of a corresponding core with uniform enrichment along the axial direction. It will be the same. A significant improvement in the MOC is that the axial power distribution is significantly flatter compared to a uniform enrichment core.
In an EOC, the axial power distribution obtained by the present invention is approximately the same as in a uniform enrichment core.

Figures 2A-2C are graphs plotting the relative power at BOC, MOC, and EOC as a function of axial position for the core at cold shutdown with all control rods inserted.
5a to 85C are shown, respectively. These graphs are also normalized using average output as a unit. For comparison, corresponding graphs 87a-87C of the power distribution for a core with uniform enrichment along the axial direction are shown in dotted lines in FIGS. 7A-2C. In addition, the second
It should be noted that the absolute neutron flux for diagrams A-7C is generally about 7 orders of magnitude lower than during wet operation in diagram A-gc (compared to 1013 neutrons/c4/sec in the wet state). In the cold state, it is 106 pieces/Ca/sec).

The effect of the axial division of enrichment is manifested as a dramatic reduction in the power output of zone 62 at all times during the cycle.

By the way, graphs 85a to 85C and 87a to 87C
Since is normalized, it is not possible to derive information about the absolute reduction in neutron flux due to axial division of enrichment. This effect can be understood by examining the diagram. Figures 9 and 9 show graphs 90a-90C plotting keff over the entire operating cycle, with corresponding graphs 92a-92c for the case of uniform enrichment along the axial direction also shown in dotted lines. Graphs 90a and 92a relate to wet operation (HARO) with all control rods pulled out, graphs 90C and 92C relate to cold shutdown (CARI) with all control rods inserted; 9
0b and 92b relate to the cold shutdown when seven control rods were pulled out. As can be easily seen from these graphs, the core of the present invention, which is segmented along the axial direction, achieves cold shutdown both when all control rods are inserted and when seven control rods are withdrawn. Increased margin (
keff) is obtained.

In summary, the present invention provides a configuration of a fuel assembly that minimizes the influence on excess reactivity in the wet state, reduces the output during cold shutdown, and thereby enables more efficient use of nuclear fuel. It can be seen that it is provided by Although the preferred embodiments of the present invention have been described in detail and fully above, various modifications, alternative constructions, and equivalent components may be used without departing from the spirit and scope of the invention. is possible. For example, although the described embodiment uses a one-inch natural uranium blanket, other dimensions may be used. Also, such blankets can be placed on the top, bottom, or both the top and bottom.

Further, the values of concentration and absorbent concentration are exemplary only. For other possible core configurations, please refer to Figures 9A-9L, which illustrate other possible designs that provide the improved control width of the present invention. In these examples as well, the upper limit of the enrichment is set at 39j (weight)%, but the present invention can also be implemented with respect to a core having a higher enrichment. Therefore, the scope of the invention, as defined by the appended claims, is not to be considered limited by the above description and illustrations.

[Brief explanation of the drawing]

Figure 1 is a schematic diagram of a nuclear reactor that uses water as a coolant and moderator.
The figure is a schematic plan view showing the overall configuration of the reactor core, Figure 3 is a partially cutaway isometric view showing seven fuel cells in the reactor core, and Figures A and 98 are inside the respective fuel assemblies. FIG. 98 is a schematic diagram showing the lateral distribution of fuel rods in the fuel assemblies of FIGS. FA and 4B; FIGS.
Figures 6A, gB and 30 are graphs showing the relative power during wet operation at the beginning, middle and end of the cycle; Figures 7A, 7B and Diagram 7C is the initial stage of the cycle.
．． Graphs showing the relative power during cold shutdown in the middle and end stages; Figures 1 and 2 are graphs showing keff over the entire operating cycle for both wet and cold; Figures 9A-9L are other available graphs. 1 is a schematic diagram showing fuel assembly composition. In the figure, 10 is a boiling water reactor, 11 is a pressure vessel, 12 is a reactor core, 13 is a fuel cell, 15 is a fuel assembly, 16 is a control rod, 40 is a control vane, 42 is a fuel rod, and 45 is an upper type rate, 46 is the lower tie plate, 48 is the channel box, 50 is the nose, 62 is the low enrichment zone, 67 is the high gadolinia zone, and 'rO is 3, F representing the middle zone.
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P/Hi, 9 and/&: 9L. Continued on page 10 Inventor Roger Zoll Williams 15260 Brison Court, Morgan Hill, California, United States 0 Inventor Kenneth Earl Gardner 5899 Trap Court, San Jose Towley, California, United States 0 Inventor Munir Hamel Published by California, USA San Jose Carmor Avenue Number 2 5652

Claims (1)

[Claims] / There is a substantial proportion of steam voids and an associated reduction in deceleration towards the top of the core during operation, and as a result of such reduction in deceleration leading to slow combustion in the upper core region, In a fuel assembly to be used in a core of a boiling water reactor in which a peak of reactivity occurs in the upper core region during a cold shutdown in which the deceleration in the upper core region increases relatively, the fuel assembly the body comprises a fissile material component distributed over a substantial axial extent thereof, said fissile material component in an axial region corresponding to at least a portion of the axial region in which a peak of reactivity exists during cold shutdown; A fuel assembly characterized in that it has a reduced enrichment. (S) Claim 1 additionally includes a neutron absorbing material component distributed such that it has a high concentration in an axial region corresponding to at least a part of the axial region where the reactivity during hot operation is maximum. fuel assembly. 3. wherein the fuel assembly includes a plurality of spatially spaced substantially parallel fuel rods, a first group of the fuel rods having an axial distribution characterized by a reduced fissile material enrichment in the axial section; 2. A fuel assembly according to claim 1, wherein said fuel rods of the Ωth group exhibit an axial distribution characterized by a uniform fissile material enrichment. 4. The fuel assembly of claim 3, wherein the fuel rods in the first group have a higher enrichment outside the axial region than the enrichment of the fuel rods in the second group. 2. The fuel of claim 1, wherein the fuel rods in the first group have an enrichment in the axial section that corresponds to the enrichment of at least seven fuel rods in the second group. Aggregation. (b) the fissile material component is present over an axial dimension of about 70 to 100 ft; and the axial area is about Ω to
2. A fuel assembly according to claim 1, having an axial dimension of . 2. The fuel assembly of claim 1, wherein the enrichment within said axial zone is 5 to 10 percentage points lower than the enrichment outside said axial zone. The enrichment within the axial zone is greater than the enrichment outside the axial zone by about θ/! The fuel assembly according to claim 1, which is lower by ~θ30 (weight). ? 2. A fuel assembly according to claim 1, wherein the enrichment within said axial section is uniform along the axial direction.