JPH01140092A - Reactor core of light water reactor - Google Patents

Abstract

PURPOSE:To obtain a reactor core in which a nuclear reactor operating characteristic supervising reactor monitor is properly arranged by separating the difference of height of reactor instrumentation and an inclusion area measuring the output in the time of high output operation such as a local output range monitor up to constant value or more while providing at least a low area (inclusion area) of fissionable nuclide concentration. CONSTITUTION:A fuel assembly is vertically arranged and the height H of its effective part (including an inclusion area) is 3.6-3.7m. In the case where the inclusion area of only a stage is arranged, the H/3-H/6 from fuel effective part upper end (TAF) is proper and this position makes the largest effectiveness from the point of the improvement of a reactor stop margin. Channels A-D of a local output range monitor 4 (LPRM) are arranged at regular intervals (not much exceeding H/4) and the increase of instrumentation errors can be prevented because the channels are separated from the inclusion area up to constant distance or more. Further, a neutron source range monitor (SRM) used in the time of nuclear reactor starting is arranged in the inclusion area and high sensitivity measurement in which rise is utilized can be performed by the inclusion area of thermal neutron flux.

Description

[Detailed Description of the Invention] [Objective of the Invention] (Industrial Field of Application) The present invention relates to the core of a light water reactor, particularly a light water reactor with a long operating cycle and a high shutdown margin, which is interposed in the core in a direction perpendicular to its axis. The present invention relates to a light water reactor core having a region formed therein and a neutron monitor appropriately placed with respect to such core.

(Prior Art) In the core of a boiling water nuclear reactor, cells each consisting of one cross-shaped control rod and four fuel assemblies surrounding it are normally arranged in a regular manner. That is, the fuel assemblies and control rods are arranged so that their axes are perpendicular and parallel to each other, and the cooling water, which functions as a moderator, flows from the bottom of the core to the top. There is.

No air bubbles are generated from the effective lower end of the core to the vicinity of the lower end of the heat generating section, but large air bubbles are generated from the center to the upper end of the core, and these generated air bubbles flow upwards of the core. When the volume ratio occupied by bubbles, that is, the void ratio increases, the neutron moderation characteristics decrease, resulting in a decrease in thermal neutron flux and a decrease in output. In order to avoid this, measures have been taken such as increasing the fission nuclide concentration, that is, the enrichment of the fuel, in areas with a high void ratio, or adding burnable poison to suppress the output in areas with a high void ratio.

Therefore, in a boiling water reactor, combustion in the upper part of the core tends to be delayed, which causes U -23!4 degrees to be relatively higher than in other parts, and voids can generate fissile nuclides such as Pu-239. Therefore, the margin for reactor shutdown tends to be tight in the upper part of the core. In addition, efforts are being made to lengthen the operating cycle and improve fuel burn-up, with the main purpose of improving economic efficiency. In this case as well, the enrichment of the fuel will inevitably increase, making the margin for reactor shutdown even tighter.

Next, typical examples of conventional fuel assemblies used in boiling water nuclear reactors and fuel assemblies expected to be used in the near future will be explained with reference to the drawings.

FIGS. 10(A) and 10(B) are a perspective view of a conventional fuel assembly and a schematic vertical sectional view of a fuel rod constituting the fuel assembly, respectively.

In FIG. 10(A), the fuel assembly includes a water rod (not shown) and a fuel rod 2 connected to an upper tie plate 4. Subeza 5. It is fixed by a lower tie plate 6 and surrounded by a channel box 1 on the outside. As shown in the same figure (B), the fuel rod 2 has fuel pellets 8 in the cladding tube 7.
A spring 9 is provided in the upper gas plenum, an upper end plug 10 is provided at the upper end, and a lower end plug 11 is provided at the lower end.

FIG. 11 is a cross-sectional view of the conventional fuel assembly shown in FIG. 10. There are 62 fuel rods 2 in the channel box 1.
and two water rods 3 are arranged to form a fuel assembly. The water rods 3 suppress the shortage of water, which is a moderator, inside the reactor assembly, but since these water rods 3 are uniform in the axial direction, there is a problem that there is an excess of water below the core and a lack of water above the core. There is a point.

The fuel assembly shown in FIG. 12 was developed to improve the characteristics of the fuel assembly, and one thick water rod 12 is placed inside the assembly to introduce non-boiling water. There is. However, even in this example, there is a problem in that there is an excess of water below the core and a shortage of water above the core.

The fuel assembly shown in FIG. 13 is also an improvement of the fuel assembly shown in FIG. The horizontal power distribution is flattened by creating a non-boiling cooling water region in the cross-shaped gap 14, but this type of fuel assembly also suffers from excess water below the core and insufficient water above it. There is a problem.

The fuel assembly shown in FIG. 14 was developed as an improved version of the fuel assembly shown in FIG. This fuel 11 assembly is composed of nine subassemblies 15, and each subassembly 15 is composed of nine fuel rods 2. A somewhat wide waterway 16 is provided between the subassemblies 15. In the case of this fuel assembly as well, the problem of excess and shortage of water in the upper and lower parts of the core has not been resolved.

(Problems to be Solved by the Invention) As mentioned above, bubbles do not occur at the lowest end of the fuel assembly, which is the heat generating part of a boiling water reactor (BWR), but bubbles occur everywhere else. Moreover, the generated bubbles flow upward (downstream) into the reactor core. Therefore, BWR
The bubble υ [void ratio] becomes higher above the core.

As a result, the rate of nuclear fission decreases because the moderation characteristics of neutrons decrease. That is, combustion proceeds below the core and is delayed above the core. Therefore, it has been proposed to increase the concentration of fission nuclides above the reactor core in order to suppress the decrease in power above the reactor core.

However, increasing the void ratio and fission nuclide concentration above the core will reduce the degree of subcriticality at the top of the core when the reactor is shut down.

On the other hand, in order to lengthen the operating cycle and improve economic efficiency, it is necessary to further increase the enrichment of the fuel, but this will make the subcriticality in the upper part of the core even shallower, and eventually There may be cases where the reactor cannot be shut down. In other words, this point has become a bottleneck, and conventional nuclear reactor cores have had the problem of not being able to extend the operating cycle.

The present invention was made to solve the above problems, and
Its purpose is to enable reactor shutdown even when the fuel enrichment level is high, and to deal with local distortions in the axial neutron flux distribution caused by the configuration introduced at that time. The object of the present invention is to provide a light water reactor core in which internal monitors are appropriately arranged.

[Structure of the Invention] (Means for Solving the Problems) In order to achieve the above object, the present invention provides a structure in which the axes of a fuel assembly consisting of a large number of fuel rods are arranged regularly so that they are perpendicular and parallel to each other. In the core of a light water reactor, a reactor core is arranged at At a distance above a certain distance, the proportion of the volume of bubbles in the space through which the coolant passes is relatively high, and in the area where the local subcriticality of the reactor becomes shallow when the reactor is shut down, At least one intervening region is provided that substantially reduces the fissile nuclide concentration over a certain range of height and width in the horizontal direction of the reactor core, and local distortion of the axial neutron flux distribution is generated by introducing the intervening region. In order to cope with this problem, the present invention is characterized in that the difference in height between the in-core instrumentation, such as a local power range monitor, that measures the output during high-power operation and the intervening area is set at a certain distance or more.

(Function) As described above, according to the core configuration of the present invention, the neutron interaction (coupling effect) in the upper and lower fuel regions sandwiching the region with low concentration of fissile nuclides (intervening region) is reduced, resulting in shutdown. The subcriticality of the reactor inside can be made larger (deeper), and unnecessary excessive reactivity during reactor operation can be suppressed.
At the end of the cycle, excess reactivity disappears and the coupling effect improves, making it possible to extend the operating cycle, thereby maintaining the red integrity of the fuel, and reducing the localization of the axial neutron flux distribution caused by the introduction of an intervening region. In order to deal with such distortions, the local power range monitor is placed at least a certain distance away from the intervening area, which prevents an increase in instrumentation errors, and the neutron source range monitor used during reactor startup is placed in the intervening area. This makes it possible to perform high-sensitivity measurements using the bulge caused by the intervening region of thermal neutron flux.

Next, the effect of the intervening region introduced in the present invention will be explained.

The effective multiplication factor of the core is k. Let it be ff. (Here, for simplicity, we use k in Table 6 and omit "eff.") Then, the modified first group model (which is simple and reliable as a description of k for light water reactors) can be expressed as follows.

k= k■ 1+M2B2 Here, k■...Infinite multiplication factor M2...Neutron transfer area B2...Buckling (cm-2 unit) k■ is usually calculated at each point ( It is treated as the ratio of the neutron emission rate and neutron absorption rate due to nuclear fission at a given volume point).

M2 is expressed as M2 = τ+L2. τ is the Fermi age, which is called the moving area of fast neutrons in core design.

Note that τ=τF+τE (or τ1+τ2), and τ
In this way, even faster neutrons and epithermal (ep)
(thermal) neutrons, but in the present invention there is almost no need to explain them separately.

L2 is the thermal neutron transfer area (given by the ratio of thermal neutron diffusion coefficient to absorption cross section), M = bar is the neutron transfer distance, L
= I'U is the thermal neutron movement distance, and B2 is the buckling (in cm-2 units), 32.3 2 +822
It can be expressed as 3.2 is a radial buckling, and B22 is an axial buckling.

By the way, in terms of neutron reactor physics, the value of k is defined for the entire system in the vicinity of criticality, but in this invention, we are from the standpoint of core design that incorporates the spatial dependence of K, so the value of k is also defined using the above formula. , we will treat it as incorporating spatial dependence.

In addition, in power reactors, M2 B2 is 0.03 to 0.0
5, B2 is 0.0001 to 0.0002 (cm-
2) The axial direction is usually flat, and the height of the core is Z
Axial reflector saving (sometimes referred to as axial outer end distance)
If we let δ=δ or +δ− (meaning upper side and lower side), then it is given by.

(The sign was changed in the definition of reactivity to handle subcritical systems) ? The value of is defined for the entire system in the critical vicinity in neutron reactor physics, but in this invention, we take the position of core design that incorporates spatial dependence into ■, so (k■→→?)? The values are treated as incorporating spatial dependence.

Therefore, the degree of subcriticality in the present invention refers to a spatially dependent degree of subcriticality for the reasons described above. If there is a place in the core system where the degree of subcriticality is low (close to criticality), it becomes an indicator that that place is likely to become critical.

Next, the operation of the present invention will be explained with reference to FIG. As shown in the same figure (^), it is assumed that there are two fuel regions I and II having a rectangular cross section, and a water gap of width W exists between them. The relationship between the water gap W and the neutron multiplication factor at this time is as shown in the same figure (B). The solid line is when cold (
The dotted line shows the dependence of the keff water gap on the width of the water gap when the reactor is generating almost no heat, such as when the reactor is shut down, and the broken line shows the dependence of the keff on the water gap width when the void ratio is high at high temperatures. It shows. For easy comparison w=
The lines are standardized so that they match at Q. Fuel assemblies are often designed in a state where the deceleration is slightly insufficient near the optimum deceleration, that is, the minimum critical mass, so when W increases slightly from O, the keff value remains almost constant or slightly decreases (
+), but if W is further increased, kef
The f value becomes smaller and smaller. In other words, the fuel region (I>
The neutron binding effect of (II) becomes noticeably smaller.

One guideline is the thermal neutron diffusion distance (~2.5 cm in water at 20°C).

Next, if the system is high temperature and has a high void fraction, for example, BW
The upper part of the R core is approximately 286℃ (the density of water is approximately 0.74 >
In some cases, the void ratio exceeds 60%.

The effective density of water in this case is ~ compared to the case of 20℃
0.7/IX(1-0,E)) = approximately 0.3.

As can be easily understood from the definition equation, the diffusion distance of thermal neutrons is approximately proportional to the reciprocal of the change in density (1→0.3).

Therefore, in this case, the thermal neutron diffusion distance is about 8 cm.

When the water gap is this wide, k even at high temperatures and high void fractions. The ff value becomes slightly smaller than the maximum value. The keH value reaches its maximum when it is about half of that, that is, about 4 to 5 cm. The reason why the maximum value appears on the broken line is that there is a shortage of moderator in the high temperature, high void state. In other words, the moderator is in a state of insufficient moderation (under-moderate state), and by separately introducing water, kef
f can be increased. This state shows that the two fuel regions (1) and (2) are in the optimum state.

The present invention is a clever application of the characteristics described above. That is, the area where almost no fuel exists is approximately the width W of the water gap.
Corresponding to this, when the reactor is not producing output, the reactor's k due to the gap. This is a state in which the keff value clearly decreases and the bonding region is weakened), and in the case of high temperature and high voids, the keff value either increases or hardly decreases due to the introduction of the GiV-tube. Since water is introduced into areas that lack water during high-temperature, high-void conditions, neutron moderation characteristics are improved, thereby improving output.

Next, a case where the core is divided into three in the axial direction and two intervening layers are present as shown in FIG. 3(A) will be described. In the figure, the subcriticality of each core piece is P+, 5'2,
P3 is calculated as follows.

'? 1 = 3 - - 1 to 1 B12-82 + 872 B22 = 8 2 + Bz□2 B32 = 8 2 + 8232 Here, δi is the reflector saving that takes into account the effects of the upper and lower adjacent core pieces, and B and 2 are the reflector savings for each core piece. Common. When the core temperature rises, the temperature of the moderator increases and voids occur. Then, ■i changes slightly and decreases on average in light water reactors), Mi2 increases, and B21 increases due to the increase in δi.
2 decreases.

If the thicknesses dl and d2 of the intervening layers are appropriately selected, δi increases significantly. Then, each core piece is integrally connected, and the axial buckling is integrally formed. On the other hand, when the core temperature decreases, a feature appears in which each core piece appears to be divided by an intervening layer. This is because as the core temperature falls, the density of water increases, and the water in the intervening layer acts to divide and shield the core into upper and lower regions. In this way, the intervening layer increases the ability to separate the core pieces into upper and lower parts during cold temperatures (when the reactor is shut down).
At high temperatures, the separation function is weakened (as the density of water decreases, there is an effect that dl and d2 become substantially smaller. This can be said to be a bonding effect). dl,
If the value of d2 is appropriately selected, the effect of compensating for the lack of water (moderator) at high temperatures will appear, and there will be no intervening layer (
dl and dz parts also have fuel) k than in the case. It becomes possible to increase the ff value to some extent.

The thickness of the intervening layer is optimally greater than the distance traveled by thermal neutrons at cold temperatures (when the reactor is shut down), and approximately the same or slightly smaller than that at high temperatures (furthermore, when voids occur in BWR). A preferable range as a specific value is about 3 to 8 CIIl. If it is less than 2 cm, the separation function will not occur when it is cold, and if it is 10 cm or more, the separation function will be weakened at high temperature.
This is disadvantageous because the effective multiplication factor of the core is reduced compared to when no intervening layer is present, leading to a reduction in the operating cycle. BWR
For PWR, the thickness of the intervening layer is preferably 3 to 3 cm, and for PWR, the thickness of the intervening layer is preferably 3 to 5 cm. That's BW
In R, the density of water is 1/3 even when it goes from cold to high temperature, so the neutron travel distance is tripled, but in PWR, even when it goes from cold to high temperature, the density of water is about 0.65. Therefore, the distance traveled by neutrons only increases by a factor of two or more. One of the reasons for the difference in the thickness of the intervening layer is, as mentioned above, the moderator density near the intervening layer,
This is because the separation and combination effects are affected by the moderator to fuel volume ratio, the concentration of fissile nuclides in the fuel, etc.

However, in the cold state when the intervening layer exerts a separation effect, B
Due to the rapid increase in Zi2, 1 decreases in each core piece, and P
, (subcriticality) increases. During high-temperature operation when the intervening layer exhibits a binding effect, the value of BZi2 rapidly decreases, and in a suitable state, it becomes almost as if there is no intervening layer, and k rapidly increases (in this case, the (can be approximately equal or slightly larger).

Next, a specific calculation example will be shown in which the intervening layer exhibits a separation effect and a bonding effect.

In BWR, initial average enrichment is 3.7%, burnup is 28
It is assumed that the control rod is not partially inserted in the GWd/l system.

In such a WR, the degree of subcriticality is the lowest during reactor shutdown near the 174-length length from the top of the core, so an intervening layer is placed horizontally across the entire core at that point, and the thickness of the intervening layer is changed to calculate S1. did. There are two calculation systems: cold (20°C) and high temperature operation (286°C, with void distribution), and for each system, the effective increase in the core when the thickness of the intervening layer is set to zero. The magnification keff was used as a reference. The graph in FIG. 3(b) shows an example of this calculation. The following can be seen from this graph.

(2) In the cold state, it suddenly decreases to ~5 cm (this is the range in which the separation effect increases significantly), and then approaches a saturated state. This asymptotic value is the core intercept (in this example, 3 below the core
74 part) k. ff value.

■At high temperatures, k up to ≦10Cm. The ff value is hardly reduced by the intervening layer. This is due to the PA coupling effect, and especially in the vicinity of 3 to 5 cm, keff increases on the contrary. This is because the intervening layer effectively replenishes the lack of thermal neutron flux due to lack of moderator during high-temperature voids.

Next, the relationship between the in-core instrumentation and the intervening area will be explained.

For example, as already explained, the BWR core is shown in Figure 10 (B
) The fuel rods shown in ) are regularly arranged to form a fuel assembly as shown in the same figure (＾). As shown in FIGS. 11 to 14, various arrangements of fuel rods are used. Usually, as shown in Figure 7, the core 2
0 indicates four fuel assemblies 22 surrounding a cross-shaped control rod 21
is configured as one cell. The cell corner portion includes a neutron source 23 (indicated by ■ in the figure), a neutron source range monitor 24 (hereinafter referred to as SRM, indicated by M in the figure), 2 a local output range monitor 25 (hereinafter referred to as PRM,
(indicated by ■ in the figure) etc. are arranged. Note that 22a is a combustion rod and 22b is a channel box. S
FIG. 8 shows the relationship between a fuel assembly centered on RM or LPRM and a cross-shaped control rod. As shown in FIG. 8, the instrumentation pipe 26 is arranged in a water gap section 28 at the center of the four fuel assemblies 22.

FIG. 9 is a detailed view of the instrumentation tube arrangement portion of FIG. 8. As shown in the figure, the LPRM25 has A, B, and C.

There are four channels D, which are arranged in the order of A to D from the bottom at approximately equal intervals in the core axis direction. Roughly speaking, this figure shows the detector (TTP) main body 27a and the C detector 25c of the LPRM 25 being at the same height, with the C detector 25C cut in the horizontal direction (perpendicular to the core axis). be. Therefore, A, B, and D have cable parts that are thinner than the main body part cut off. Further, the detector 27 is configured such that a detector main body 27a can freely travel in the axial direction within a detector guide tube 27b having a watertight structure. In addition, the instrumentation pipe 26
In other parts of the interior, water flows axially.

FIG. 5 shows an example of the core axial thermal neutron flux distribution obtained when the detector is moved in the axial direction (during high power operation). That is, the distribution diagram in FIG. 5(B) corresponds to the axial direction of the fuel assembly in FIG. 5(A). This figure shows that the introduction of the intervening region locally generates a sharp thermal neutron flux peak. - In general, thermal neutron flux peaks generate local power peaks and adversely affect the health of the fuel, so measures to suppress power peaks are implemented to prevent this. An example of such a countermeasure is the local placement of burnable poisons. This significantly suppresses the power peak and thermal neutron flux peak. However, at the end of the cycle, when there is no problem even if the output peak occurs, it is a good idea to intentionally generate the peak. In this case, when looking at the thermal neutron flux peak throughout the cycle, the relative peak magnitude changes significantly. Therefore, if the LPRM is placed near the location where the thermal neutron flux peak occurs, it will be directly affected by changes in the thermal neutron flux peak, which will be a major hindrance in accurately determining the output near the LPRM.

However, since the range in which the thermal neutron flux peak spreads is generally narrow, if the distance from the intervening region is, for example, 10 cm or more, the rise in thermal neutron flux becomes minute, and J
: There is no possibility of interference with output monitoring.

In the present invention, the distance between the end of the effective part of the LPRM and the end of the interposed region is set at least 10 cm to prevent a decrease in the measurement accuracy of the LPPM.

Further, FIG. 6 shows an example of the axial thermal neutron flux distribution when the nuclear reactor is started from a stopped state.

In other words, the distribution diagram in Figure 6 (B) corresponds to the axial direction of the fuel assembly in Figure 6 (A), and although the axial thermal neutron flux distribution during reactor shutdown is small, the It can be seen that a large thermal neutron flux peak exists even during shutdown. Neutron flux monitoring at startup is performed by SRM (neutron source range monitor). Since the neutron flux level at startup is low, the SR
It is a good idea to place M at a highly sensitive position, so that the reactor can be started while monitoring changes in the neutron flux level with sufficient reliability. As is clear from FIG. 6, if the SRM is placed in the intervening region, the counting rate of the SRM increases. As the counting rate increases, statistical accuracy improves, allowing highly reliable G) activation.

(Example) Embodiments of the present invention will be described with reference to the drawings.

FIG. 1 is a schematic diagram of an embodiment of the present invention, and this embodiment is applied to a boiling water reactor.

As shown in the figure, the fuel assembly is arranged vertically, and the height H of its effective part (including the intervening area) is 3.6 to 3.7 m.
That's about it. When arranging only one stage of intervening region in a boiling water reactor, from the upper end of active fuel part (TAF), 173H
~176H is suitable, and this position has the greatest effect in terms of improving the reactor shutdown margin. However, these values vary depending on core design conditions. In addition, in a pressurized water reactor, the intervening area is 17 mm above the active fuel part (TAF).
It is effective to provide it between 2H and 1/311.

In addition, the in-core instrumentation includes a neutron source range monitor (SRM).
), an intermediate range monitor (IRM, not shown), and a local output range monitor (LPRM>), but the intermediate range monitor is not related to the present invention and is therefore not shown.
For M, 4 channels A, B, C, and D are approximately equally spaced (1
/40) and are arranged as shown in the figure. In this figure, the distance h1 between the boundary between the D channel and the intervening region, and the distance 1ath2 between the boundary between the C channel and the intervening region are both 10CIn or more at least at the position of the intervening region or LPPM. One of the two is shifted as necessary so that the influence of the thermal neutron flux peak can be evaluated within a range that does not cause errors.

The SRM is placed in the intervening region to further increase the neutron count rate at low neutron flux levels. This is because the counting rate is improved if the SRM is placed at the thermal neutron flux peak position. It is obvious that the counting rate can be improved by placing the neutron source at the same height as the SRM, but
As the distance between the SRM and the neutron source decreases, the probability of counting directly transmitted neutrons that have not been affected by the multiplication effect of the fuel, that is, the background level, tends to become relatively worse, and the multiplication characteristics at reactor startup tend to deteriorate. In this example, SRM
An example is shown in which the distance between the two is increased somewhat by adding hs to the heights of the neutron source and the neutron source. Usually, such considerations are not necessary.

FIGS. 4(^), (C), and (D) respectively show different longitudinal cross-sectional views of the fuel assembly according to the present invention. In other words, the figure (A) is 174H to 176 from the top.
This is a case where one step of the intervening layer is provided at a position of about H, and the same figure (
B) is a diagram showing the void fraction distribution in the core axis direction and the reactor shutdown margin in the same figure (^). If there is no intervening layer, the void fraction distribution in the core axis direction is as shown by the solid line A in the figure, so the reactor shutdown margin becomes particularly severe near 1/4 tl to 1/6 H. However, by inserting the intervening layer, the harshness of that part is greatly reduced as shown in the figure, and the reactor shutdown margin is improved. Figure (C) is an example in which the intervening layer is provided in two stages.

The same figure ([)] is an example in which the intervening layer is provided in three stages.

In a BWR, the shutdown margin is generally severe above the center of the core, so this configuration (C and D in the figure) is a suitable example. However, if more intervening layers are provided in the upper half of the core, the shutdown margin tends to become severe in the lower half of the core, which is not preferable.

In PWR, the intervening layers are generally distributed in the axial direction.

[Effects of the Invention] As explained above, with the configuration of the present invention, it is possible to effectively increase the subcriticality of the nuclear reactor, and at the same time, the height of the intervening region and the height of the neutron detector for in-reactor instrumentation can be increased. By appropriately selecting the neutron source range, it is possible to prevent an increase in error in the local power range monitor during reactor power operation, and to improve the sensitivity of the neutron source range monitor during reactor startup. This has the effect of being able to monitor the furnace startup. Broadly speaking, the present invention has the following effects by introducing the intervening region. (1) When the reactor is shut down, there are no voids, the water temperature is low, and the water density is high, so the diffusion distance of thermal neutrons is small, but according to the core configuration of the present invention, the concentration of fissile material is low. region(
Neutron interaction (intervening region) in the upper and lower fuel regions (
As a result, the degree of subcriticality of the reactor during shutdown can be increased (deepened) by J.

(2) During high-temperature, high-void operation, the average density of water decreases significantly, and as a result, the thermal neutron diffusion distance increases significantly (2nd to 3rd place)
Extends to. As a result, the coupling effect across the intervening region is improved, and the effective multiplication factor can be increased, even if only slightly, despite the fact that there is a region where the concentration of fissile material is significantly reduced. There are no disadvantages due to the introduction of the intervening region.

(3) It is easy to add a burnable poison to a limited area in or adjacent to the intervening area and set it so that the poisonous effect disappears near the end of the operating cycle; in that case, Until near the end of the cycle, the burnable poison absorbs thermal neutrons, suppressing the binding effect even during high-temperature, high-void operation. In other words, unnecessary mistakes while driving! 1
The reactivity is suppressed and the excess reactivity disappears at the end of the cycle, and at the point when the cycle has no choice but to end, the binding effect improves and the excess reactivity is supplied. As a result, the operating cycle can be extended.

(4) In the present invention, since the burnable poison is effectively placed in a limited portion of the fuel in or axially adjacent to the intervening region, power peaks (power spikes) do not occur and therefore the fuel integrity Maintains sexuality.

Furthermore, since the vertical coupling is improved by increasing the void ratio, it has the property of alleviating an excessively negative void coefficient.

[Brief explanation of the drawing]

Figure 1 is a schematic configuration diagram of an embodiment of the present invention, Figure 2 (^)
and the same figure (B) is a figure for detailed explanation of the present invention,
3(A) and (B) are diagrams for explaining the characteristics of the reactor core according to the present invention, and FIG. 4(A). (C) and ([)) are respectively different vertical cross-sectional views of fuel assemblies according to the present invention, and (B) is a diagram showing the void fraction distribution and reactor shutdown margin in the core axis direction of (A). , 5th
Fig. 6 and Fig. 6 are distribution diagrams comparing the axial thermal neutron flux of the conventional fuel assembly and the fuel assembly of the present invention during high power operation and reactor shutdown, respectively, and Fig. 7 shows the neutron source and neutron source range. BWR with monitor and local output range monitor
A plan view of the reactor core, FIG. 8 is a partially enlarged view of FIG. 7, FIG. 9 is a detailed view of the instrumentation tube arrangement part of FIG. 8, and FIGS. A perspective view of a fuel assembly and a schematic longitudinal sectional view of fuel rods constituting the fuel assembly, No. 11
The figure is a cross-sectional view of the conventional fuel assembly shown in Figure 10.
2 to 14 are cross-sectional views of different conventional fuel assemblies. 1.22b...Channel box 2.22a...Fuel rod 3...Water rod 12...Thick water rod 13...Small desennel box 14.16.28...Gap 15... Subassembly 20...core 21...cruciform control rod 22...fuel assembly 23...neutron source 24...neutron source range monitor 25...local power range monitor 26...instrumentation Tube 27...Detector (8733) Agent Patent Attorney Yoshiaki Inomata (Idiot)
1 person) (A) Water N-77/Yu Q2 (w) (B) Figure 2 I P -/l (gb ``y knee c-'-&← CB) (A) <C)
(D) Con 4 @ Con 5 Z 72 ] S ; 39 Co 1 [@ Figure 12

Claims (3)

[Claims] (1) A fuel assembly consisting of a large number of fuel rods bundled together in an orderly manner
The in-core gauge is arranged so that its axes are perpendicular and parallel to each other, and is installed inside the fuel assembly or in the gap between adjacent fuel assemblies to measure the output during high-output operation, such as a local output monitor. A reactor core in which a neutron source area monitor for measuring neutron flux during reactor startup is arranged inside the fuel assemblies or in the gap between adjacent fuel assemblies, and the fuel is disposed between the fuel rods. In a light water reactor core configured such that light water, which is a coolant, flows from the bottom to the top of the rods, a fissile core of a predetermined width is divided such that a part or the entire core is vertically divided into at least two core segments. The intervening region where the nuclide concentration has been significantly reduced is placed across a certain height difference from the height where the in-core instrumentation that measures the output during high-power operation, such as the local power range monitor, is located. The width of the intervening region is equal to or within twice the thermal neutron diffusion distance during power operation of the light water reactor, and is equal to or larger than the thermal neutron diffusion distance at room temperature of the light water reactor. Reactor core. (2) The core of a light water reactor according to claim 1, wherein the height of the neutron source area monitor in the core is substantially the same as the height of the intervening layer. (3) The difference between the axial height of the arrangement of the intervening region and the height at which in-core instrumentation such as the local power monitor for measuring output during high-power operation is arranged is about 15 cm or more. A core of a light water reactor according to claim 1.