JP2002022871A - Core for boiling water reactor and core monitor - Google Patents

Abstract

PROBLEM TO BE SOLVED: To increase tolerance of transient and stability by optimizing the axial output distribution and the cycle boiling length. SOLUTION: When the relative outputs of all fuel assemblies are arranged in descending order under rated output operating state, axial output distribution of fuel assemblies within a range of one third of all fuel rods is set not lower than 1.05 within a range of 5/24-9/24 node or the sub-cool boiling length is set not to exceed 2/24 node. A part of fuel pellets being filled in fuel rods within a range of 1/24-10/24 node is replaced by MOX fuel pellets containing Gd burning up to a middle of core operation cycle.

Description

DETAILED DESCRIPTION OF THE INVENTION

[0001]

The present invention relates to a reactor core for a boiling water nuclear power plant and a monitoring system for the reactor core.

[0002]

2. Description of the Related Art A boiling water nuclear power plant (hereinafter, referred to as a nuclear power plant)
The outline of “BWR” will be described with reference to FIGS. As shown in FIG. 14, a reactor pressure vessel 2 is provided in a reactor containment vessel 1, and a reactor core 3 and, for example, a coolant 4 of water are accommodated in the reactor pressure vessel 2.

[0003] As shown in FIG.
The fuel assembly includes a plurality of fuel assemblies 6, 6 arranged inside the fuel assembly 6, and control rods 7 inserted between the fuel assemblies 6, 6, and the like. A lower portion of the fuel assembly 6 is supported by a fuel support fitting 9 provided on a core support plate 8 at a lower portion of the shroud 5, and an upper portion of the fuel assembly 6 is formed by an upper grid plate 10 at an upper portion of the shroud 5.
Supported by

The coolant 4 flows in from the lower tie plate of the fuel assembly 6, passes through the fuel rod bundle and flows out of the upper tie plate, but is forcibly circulated by the recirculation system 11 shown in FIG. By receiving heat generated by the fission of uranium 235 (hereinafter referred to as “U235”), saturated water and saturated steam are mixed and move to the upper part of the core.

Gas and liquid are separated by a steam-water separator (not shown), the gas is dried by a steam dryer, and the dried steam is sent to a turbine 13 via a main steam pipe 12 connected to the reactor pressure vessel 2. Then, the turbine 13 is driven. The drive of the turbine 13 rotates the generator 14 to generate power.

The steam that has worked in the turbine 13 is introduced into a condenser 15 to be condensed, and the condensate is purified by a condensate filtration device and a condensate desalination device.
After flowing into 17 and being heated, the water is supplied again from the water supply regulating valve 20 into the reactor pressure vessel 2 through the water supply pipe 19 by the water supply pump 18.

[0007] In FIG. 14, a bypass pipe 21 is connected between the main steam pipe 12 and the condenser 15, and a bypass valve 22 is attached to the bypass pipe 21. A relief safety valve 23 is attached to the main steam pipe 12 in the containment vessel 1, and the relief safety valve 23 is connected to a vent pipe 24, and a tip end of the vent pipe 24 is immersed in pool water 25 a in a suppression pool 25. ing.

A jet pump 26 is provided in the reactor pressure vessel 2.
Is installed, and a reactor coolant recirculation loop 27 is connected between the jet pump 26 and the reactor pressure vessel 2, and a reactor coolant recirculation pump 28 is connected to the reactor coolant recirculation loop 27. Installed. The main steam pipe 12 is provided with a main steam isolation inner valve 29, a main steam isolation outer valve 30, a main steam stop valve 31, and a turbine steam control valve 32.

[0009] More specifically, the core 3 of the boiling water reactor is configured as shown in FIG. That is, a plurality of fuel assemblies 6 are inserted into the core 3, and each of the fuel assemblies 6 is covered by the channel box. In addition, in the core 3, in order to detect a neutron flux,
A plurality of local output area monitors (hereinafter, referred to as “LPRM”) 33 are arranged.

Further, the upper part of each channel box is
As shown in FIG. 15 and described above, the core 3 is supported by the upper lattice plate 10, the core support plate 8 and the upper lattice plate 10 are held by the cylindrical shroud 5, and the entire fuel assembly 6 is wrapped to constitute the core 3.
The fuel support 9 supports the lower end of the fuel assembly 6 in each of the upper four corner holes, and the control rod moves up and down in the center cross-shaped hole.

The coolant 4 flows into the channel box from below through the orifice of the fuel support 9 and the lower tie plate, is heated by the fuel assembly 6, generates steam (void) by boiling, and It becomes a liquid two-phase flow.
The fuel assembly effective length of a currently operating commercial BWR is about 3.7 m.

FIG. 17 is a typical operating characteristic diagram showing the relationship between the reactor power and the core flow rate of the BWR having the above configuration. Normal operation is performed on the rated power curve, the design flow control curve, the stability limit curve, the minimum pump speed curve, the cavitation limit curve, the maximum pump speed curve, the area surrounded by them and on the natural circulation curve. .

In the example of FIG. 17, the rated output is a core flow rate of 85%.
It has been achieved over ~ 105%. The operation of the reactor is usually near the 85% flow rate at the beginning of the cycle (point A1),
In order to utilize the reactivity gain due to the increase in the coolant flow rate, the flow rate increases toward the higher flow rate as the cycle burn-up advances (point A2), and approaches 105% flow rate at the end of the cycle (point A3).

The core power is normally designed so that the core axial power distribution has a lower strain in the early stage of the cycle, and the void fraction at the core outlet is increased. As a result, uranium 23
8 (hereinafter referred to as “U238”) promotes the production of fissile material plutonium 239 (hereinafter referred to as “Pu239”).

[0015] From the middle to the end of the cycle, the operation is such that Pu239 is burned in addition to U235, so that the power distribution in the axial direction of the core tends to be upwardly distorted. In a core with improved economy, optimization is performed by combining the above-mentioned core flow rate adjustment and a nuclear design that enables axial power distribution adjustment.

Incidentally, the boiling state in the core 3 is more specifically as follows. That is, the two-phase flow is classified into subcooled boiling in a thermally non-equilibrium state and saturated boiling in a equilibrium state.

Known Document 1, "THE THERMAL-HYDRAULICS
OF A BOILING WATER REACTOR-Second Edition
-By RTLahey, Jr. & FJMoody According to Fig.5-16 of "ANS (American Atomic Energy Society)", voids in subcooled boiling state are generated locally along the heat transfer surface,
The enthalpy of the liquid phase at the center of the stream has not reached the saturation point.

On the other hand, in the void in the saturated boiling state, the liquid phase reaches the saturated enthalpy h _f , and the difference from the enthalpy h _g of the gas phase (void) is constant with the latent heat of vaporization h _fg . is there.

For the heat flux from the gas phase on the heat transfer surface to the central liquid phase in the subcooled boiling state, two representative correlations have been proposed, which are used in thermal hydraulic analysis of a nuclear reactor. .
One is the Profile-Fit Model by Zuber-Staub et al., And the other is the Mechanisms Model by Lahey et al.

According to these models, the heat flux q ″ _li from the bubble generating portion to the central subcool portion is different from the heat flux q ″ _w at the wall surface in addition to the central subcool portion enthalpy h _l and the mixed enthalpy

(Equation 1) , Quality X, etc. And q ''
_li is h _l or

(Equation 1) Phenomena that are greatly affected by

[0021] In contrast, in the nucleate boiling region are mainly used in BWR in saturated boiling, vapor bubbles are separated from the heat transfer surface, and the liquid phase enthalpy is kept at a constant value h _f Therefore, the heat flux q ″ _w from the wall surface becomes stable.

The subcool boiling start height is also summarized in TABLE 5-1 of the above-mentioned known document 1,
Force Balance Model by evy, E by Saha-Zuber
The mpirical model and the like are used in thermal hydraulic analysis of nuclear reactors.

The fuel assemblies in the boiling water reactor described above have been used since commercial power generation in Japan.
Row 7 column type, 8 row 8 column type, improved 8 row 8 column type, high burnup 8
A row 8 column type and a high burnup 9 row 9 column type have been adopted.

[0024] These improvements increase the fissile material capacity per fuel assembly, and realize high burnup and long-term operation cycle by optimizing the enrichment distribution in the assembly and the optimal arrangement of burnable poisons. Have been. Due to this and the effective use of the core flow rate adjustment width, the economic efficiency of the core is improved.

FIGS. 18 (a) to 18 (e) show a high burnup 8 rows and 8 columns type (hereinafter referred to as a high burnup 8 × 8 fuel assembly) among the fuel assemblies. 19 (a) to (e) are 9 rows and 9 rows
The basic structure of a row type (hereinafter referred to as a 9 × 9 fuel assembly) is shown.

FIG. 18A shows a high burnup 8 × 8 fuel assembly.
FIG. 18 (b) is an elevational view showing a part of FIG.
FIG. 18A is a plan view showing a conventional upper tie plate 36 supporting the upper end of the fuel rod 35 in FIG. 18A, FIG. 18C is a plan view showing the same round spacer 37, and FIG. FIG. 18 (e) is a transverse cross-sectional view of the same large diameter water rod 38 as shown in FIG.
FIG. 39 is a plan view showing a network section 40 into which 39 fuel rod end plugs are inserted.

FIG. 19 (a) is an elevational view showing a 9 × 9 fuel assembly 41 in a longitudinal section, and FIG. 19 (b) is an improved upper tie plate supporting the fuel rod 35 in FIG. 19 (a). 19 (c) is a plan view showing the same round spacer 43, and FIG. 19 (d) is a cross-sectional view showing the same two cross sections of the large diameter water rod 44. FIG. 19 (e) is a plan view showing the network section 46 of the improved lower tie plate 45 in the same manner. In FIGS. 19A and 19D, reference numeral 47 denotes a partial length fuel rod.

High burnup 8 × 8 fuel assembly shown in FIG.
In the 9 × 9 fuel assembly 41 evolved from 34 and shown in FIG.
Deterioration of transient characteristics and stability due to an increase in the absolute value of the void reactivity coefficient due to an increase in burnup and a long operation cycle is prevented by the use of two large diameter water rods 44 and the like.

The deterioration of stability due to thermo-hydraulic factors due to an increase in two-phase pressure loss due to an increase in the fuel assembly lattice is caused by a reduction in the spacer pressure loss coefficient, the eight partial length fuel rods 47 and the high pressure loss improved type. This is prevented by the adoption of the lower tie plate 45.

FIGS. 20 and 21 show examples of nuclear design when the 9 × 9 fuel assembly 41 is loaded. These are the same as those described in the publicly known document 2 “TLR-057“ Revision 2 for 9 × 9 fuel of boiling water nuclear power plant ”published by Toshiba Corporation”. Table 1 shows a legend for each fuel rod of FIGS.

In the case shown in FIG. 21, seven types of fuel rods are used. Fuel rod type 5 is a short fuel rod. The fuel rod types 6 and 7 are fuel rods containing gadolinia as a burnable poison, and the enrichment distribution and the arrangement in the fuel assembly of these gadolinia-containing fuel rods are the same as those of fuel rod types 1 to 4 in FIGS. Custody.

[0032]

[Table 1]

In the high burnup 8.times.8 fuel assembly 34, about one third of the number of fuel rods
A plan to load X fuel pellets into commercial reactors in Japan is in progress. MOX is an abbreviation for Mixed Oxide, which means a mixture of plutonium oxide and uranium oxide.

Here, the transient characteristics include the time change of process parameters such as output and pressure when an abnormal transient change occurs during operation such as shutting down a generator load or closing a main steam isolation valve in a plant. In addition, it means the thermal integrity of the fuel assembly. It is desirable that the fuel assembly can perform good heat removal during the transient, and it is engineeringly reasonable to design so that 0.1% or more of all the fuel rods do not become fiber boiling during the transient. .

A parameter relating to the heat removal performance of the fuel is a minimum critical power ratio (Minimum Critical Power Rati).
o, hereinafter referred to as MCPR).
M during operation so that PR does not fall below Safety Limit MCPR (hereinafter referred to as SMCPR).
CPR has restrictions.

Operation Limit MCPR
PR; hereinafter referred to as OLMCPR) analyzes various transient changes expected to occur during the life of the plant,
The change in MCPR (hereinafter, referred to as ΔMCPR) is obtained, and can be calculated by adding the largest change (hereinafter, referred to as ΔMCPRmax) to the SLMCPR.

OLMCPR = ΔMCPRmax + SLMCPR (1) ΔMCPRmax tends to increase in a plant in the early stage of BWR construction (hereinafter referred to as a conventional scrum plant) having a relatively low scram speed.

In particular, at the end of the cycle, the negative void reactivity coefficient increases and the scram characteristic deteriorates, so that the transient characteristic tends to deteriorate.
It is designed not to fall below CPR.

Using a three-dimensional nuclear thermal hydraulic analysis code,
For the 1.1 million kWe-class BWR5 conventional scrum plant, generator load shedding at rated output and bypass valve non-operation were performed at the beginning of the cycle (point A1 in FIG. 17), in the middle (point A2 in FIG. 17), and at the end (point A3 in FIG. 17). The examples analyzed in ()) are shown in FIGS.
It is shown in (c).

Here, when the load on the generator 14 shown in FIG. 14 is cut off, the turbine steam control valve 32 is rapidly closed by the operation of the power load unbalance relay, and the bypass valve 22 is usually rapidly opened. Is done. In this analysis, it is assumed that the bypass valve is not operated on the maintenance side.

Although the neutron flux rises sharply due to the pressure increase due to the interruption of the steam flow, all the control rods 7 are suddenly inserted into the reactor (scram) by the quick closing signal of the turbine steam control valve 32. As a result, the power supply of the recirculation pump 28 is cut off (RPT) in order to alleviate the decrease in the void ratio in the core due to the pressure increase, so that the core flow rate decreases. The pressure increase in the reactor pressure vessel 2 is suppressed by opening the relief safety valve 23.

At this time, ΔMCPR is 0.04,
0.14, 0.27, and when these are added to SLMCPR = 1.07, OLMCPR becomes 1.1 at the beginning, middle, and end of the cycle.
1, 1.21, and 1.34. Thus, if the operating MCPR is maintained above 1.34 throughout the cycle, no more than 0.1% of the fuel rods will experience boiling fiber during all transient events.

The axial power distribution at the end of the cycle in this evaluation example is such that the core average is a peaking of 1.28 (peak node 14/24) (the conventional core indicated by a broken line in FIG. 5), and the subcooled boiling length is The core average is 9.5 / 24 nodes.

Next, the stability means that when the operating point is in a low flow rate / high output state at the time of starting or stopping the plant, or when a transient change such as a single recirculation pump trip occurs in the plant, the operation is stopped. It means the neutron flux vibration damping characteristics when the point shifts to low flow / high output.

The core is desirably stable over the entire operation range, and is confirmed by analyzing that the reduction ratio, which is a parameter for determining stability, is less than 1.0. The reduction ratio is indicated by the ratio of adjacent amplitudes in the response of the system when a step-like input disturbance is applied.

The operating region having a small margin with respect to the reduction ratio of 1.0 is selected rods (Selected Rods Insertion;
SRI) and stability limitation curve (straight line BD in FIG. 17).

The types of stability include channel stability focusing on the thermo-hydraulic stability of the highest power channel, core stability that is a neutron flux vibration in which the entire core phase is aligned,
There is a region stability that is a neutron flux oscillation having a symmetry axis in the core circumferential direction and 180 degrees out of phase. The sensitivity of each stability to the axial power distribution is in a severe direction as the core stability is generally flat, and the channel stability and the regional stability are in a severe direction as the distribution has a lower peak.

In the core stability, the difference between the other stability and the sensitivity to the axial power distribution is that the effect of the nuclear feedback is large in the core stability, which is obtained when the power peak is high at a high void fraction. Because it appears as a great influence.

It is considered that in the radial and circumferential directions of the core, the greater the distortion of the power distribution, the more unstable the power distribution becomes. As parameters representing the degree of distortion, the R0 value for the core stability and the R0 value for the region stability are used. The effectiveness of the R1 value has been clarified by the results of the private use of the actual stability of the machine and the sensitivity analysis using the analysis code.

The R0 value and the R1 value are formulated as follows.

(Equation 2)

Here, n is the number of fuel assemblies, and i is the fuel assembly number from 1 to n. Further, Ψ _i represents the radial direction power distribution peaking, and ψ _i represents the circumferential primary mode power distribution peaking.

Normally, the R0 value and the R1 value are the total fuel assembly number n.
= N. For example, when all the fuel assemblies are arranged in descending order of the relative output, by calculating the assemblies n = M in the range of 1/3, the power distribution is calculated. It can be used as an index with increased sensitivity.

Using the core three-dimensional nuclear thermal hydraulic analysis code,
When the recirculation pump speed was reduced from the rated output state, the neutron flux response of the average of all cores when a 10 ° reactivity was injected in a stepwise manner into the entire core was determined at the beginning of the cycle (point B in FIG. 17) and during the middle ( FIGS. 23 (a) to 23 (c) show examples of analysis at the point C) in FIG. 17 and at the end stage (point D in FIG. 17). The reduction ratio at this time is 0.1 at the beginning of the cycle shown in FIG. 23 (a), at the middle of the cycle shown in FIG. 23 (b), and at the end of the cycle shown in FIG. 23 (c).
38, 0.61, and 0.78.

In the axial power distribution at the end of the cycle at the lowest pump maximum power point in this evaluation example, the peaking is 1.14 (peak node 9/24) on the core average (dashed line in FIG. 7), and the subcooled boiling length is The core average is 9.5 / 24 nodes, and the subcool boiling height is within 2/24 on the core average.

As an example of a core three-dimensional nuclear thermal-hydraulic analysis code, see “TRACG TRANSIENT ANALYSIS
CODE -THREE-DIMENSIONAL KINETICS MODEL IMP
LEMENTATION AND APPLICABILITY FOR SPACE‐DEPEN
DENT ANALYSIS ”Nuclear Technology vol.105 pp1
62-183 Feb. 1994 ”.

As described above, in the core employing the 9 × 9 fuel assembly 41, it has been confirmed that the transient and stability deterioration is prevented by various design improvements. However, the MCPR and the core stability at the end of the cycle are confirmed. Sexual margins tend to decrease. This includes, besides the void coefficient, Pu239 generated at the core exit with a high void fraction at the end of the cycle.
This is related to the fact that the axial power distribution is slightly upwardly distorted due to the combustion of.

That is, if the axial power distribution becomes upwardly distorted, the scram characteristic deteriorates in terms of transient characteristics, and at the same time, when disturbance occurs, the subcooled boiling region in which the influence propagates to the downstream side becomes longer. There is a tendency. Also, in terms of core stability, the axial power distribution tends to be flat in the low flow rate / high power operation state, and in particular, the core stability reduction ratio tends to increase.

Conversely, when the axial power distribution in the core axial direction is set to the lower strain at the end of the cycle, the scram characteristic is improved and the core stability tends to be improved, but the economical efficiency of the core is sacrificed. Therefore, there is a technical problem that an excessive lower strain distribution is a factor that deteriorates the region stability and the channel stability.

Various proposals have been made to limit the power distribution in the axial direction of the reactor core from the viewpoint of transient and stability (for example, Japanese Patent Application Laid-Open No. 58-208692, "Power Control of Reactor"). Method and Apparatus ", JP-A-5-2569
No. 78, “Reactor protection device”).

However, they relate only to the integral of the channel, or to the axial power distribution of the average of the entire core, and greatly affect the transient and stability of the entire core. When the relative powers of the fuel assemblies are arranged in descending order, the characteristic of the core dynamic characteristics that the fuel assemblies are in a range of up to 1/3 of the total number of fuel assemblies has been overlooked.

FIG. 25 shows core stability analysis results at the minimum pump speed maximum power point of a typical BWR plant for various axial power distributions. For the evaluation, a frequency domain stability analysis code including the core dynamics equation is used. Figure 24 shows the applied axial power distribution.
Flat distribution (maximum value: 1.15; peak node 5)
/ 24), lower strain distribution (maximum value: 1.50; peak node 9 /
24), central distortion distribution (maximum value: 1.40; peak node 12 /
24).

The core stability reduction ratios were 0.72 and 0.4, respectively.
6, 0.34, and the flattened distribution has the largest width reduction ratio as described above. With these as basic cases, when the relative outputs of all the fuel assemblies are arranged in descending order, the axial output distribution of the fuel assemblies in the range of up to 1/3 of the total number of fuel assemblies is changed to the lower strain distribution or the central strain. The result of evaluating the case where the distribution is set as a distribution and the remaining 2/3 is made a flat distribution is plotted in FIG.

The width reduction ratio is 0.50 for the former and 0.45 for the latter. From these results, it can be seen that an aggregate having a low relative output also affects stability, but its contribution is small and an aggregate having a high relative output controls most of the stability.

[0064]

The fuel assemblies for boiling water reactors proposed so far have not only improved the economical efficiency of the reactor core, but also achieved a reduction in the void reactivity coefficient and pressure loss. Various improvements have been made to prevent the deterioration of transient and stability. However, higher burnup,
In order to improve the economic efficiency of the core by using a long operation cycle, it is indispensable to manufacture a core that comprehensively considers the transient and stability evaluation results throughout the core operation cycle.

An object of the present invention is to take into consideration the intra-cycle variation of the axial power distribution in a core with improved economy, and particularly to a fuel having a high relative power which is important for transient characteristics and stability under rated power operation. Transient characteristics and stability are improved by optimizing the axial power distribution, subcooled boiling length, and subcooled boiling start height in the aggregate in terms of transient and stability, and the margin is effectively increased. An object of the present invention is to provide such a boiling water reactor core.

[0066]

The invention according to claim 1 is
In a boiling water reactor core in which a fuel assembly in which fuel rods filled with low-enriched uranium oxide are arranged in a square lattice shape is loaded in the core, 1 / The average uranium enrichment in the cross section of the fuel assembly in the lower region of 24 to 10/24 is higher than the average uranium enrichment in the cross section of the fuel assembly in the upper region of 11/24 to 24/24 in the axial direction, and Loading the core with a burnable poison having an average cross-sectional burnable poison amount of the fuel assembly in the region from 10/24 to 10/24 greater than the average burnable poison amount in the cross-sectional area of the fuel assembly in the upper 11/24 to 24/24 region In the rated output operation state, when the relative outputs of all the fuel assemblies are arranged in descending order of the relative output, 1
The peaking of the axial power distribution of the fuel assembly in the range of up to / 3 is not less than 1.05 in the range of 5/24 to 9/24, or the subcooled boiling length generated in the lower part of the core is in the axial direction. It is characterized in that the area does not exceed a length of 8/24.

According to the first aspect of the present invention, the scram characteristic at the time of occurrence of an abnormal transient change during operation is improved,
It is possible to improve the neutron flux in-phase attenuation characteristics when the operating point shifts to a low flow rate or high output state.

According to a second aspect of the present invention, there is provided a boiling water reactor core comprising a fuel assembly in which fuel rods filled with a mixed oxide of uranium and plutonium are arranged in a square lattice, the core being loaded in the core. The average uranium enrichment cross section of the fuel assembly in the lower region of 1/24 to 10/24 below the axial region filled with the oxide is the cross section of the fuel assembly in the region of 11/24 to 24/24 in the upper axial direction. Amount of burnable poison that is higher than the average uranium enrichment and average cross section of the fuel assembly in the lower 1/24 to 10/24 area is higher than average fuel section in the upper 11/24 to 24/24 area. By loading a larger fuel assembly into the core, the axial peaking of the fuel assemblies in the range from the highest relative output of all the fuel assemblies in the core to one third is reduced to the axial region 5 /. Not less than 1.05 in the range 24 to 9/24, or Wherein the Bukuru boiling length does not exceed the length of 8/24 in axial region. According to the second aspect of the present invention, it is possible to satisfy the restriction on the core axial power distribution or the restriction on the subcooled boiling length.

A third aspect of the present invention is to provide a partial length fuel rod filled with a mixed oxide of uranium and plutonium containing no burnable poison from 1/24 to 10/24 below the axial region,
A fuel rod with a partial-length fuel rod filled with low-enriched uranium oxide containing no burnable poison in the upper axial section 11/24 to 24/24 and a full-length fuel rod filled with low-enriched uranium oxide are squared. In a boiling water reactor core in which fuel assemblies arranged in a lattice are loaded in a core, burnable poison is applied to 1/24 to 10/24 of at least four fuel rods around the fuel rods. By disposing the full-length fuel rods filled with the low-enriched uranium oxide containing, the axial peaking of the fuel assemblies in the range from the highest relative output of all the fuel assemblies in the core to one third is axially reduced. 1. In the range 5/24 to 9/24.
It is characterized in that it does not fall below 05 or the subcooled boiling length does not exceed a length of 8/24 in the axial region.

According to the third aspect of the present invention, it is possible to prevent the core axial power distribution from becoming extremely lower strain distribution from the early stage to the middle period of the cycle, and to reduce the core axial power distribution from flat to lower peak from the middle to the end of the cycle. Can be maintained. As a result, it is possible to prevent deterioration of channel and region stability from the early stage to the middle stage of the cycle, and to improve scram characteristics and core stability from the middle stage to the end stage of the cycle.

According to a fourth aspect of the present invention, when the core inlet flow rate is automatically or manually reduced from the rated output operation state to 50% or less of the rated flow rate, or in the rated output operation state, the power of the recirculation pump motor is lost and the core inlet flow rate is reduced. When the flow rate falls below the 50% rated flow rate, or when the relative outputs of all the fuel assemblies are arranged in descending order, the axial power distribution of the fuel assemblies within 1/3 of the total number of fuel assemblies It is characterized in that the peaking does not fall below 1.15 in the range of 4/24 nodes to 8/24 nodes and does not exceed 1.75 in the range of 3/24 nodes to 5/24 nodes. According to the invention of claim 4, the attenuation characteristics of the fundamental mode neutron flux and the higher mode neutron flux at the time of low flow rate or high output operation can be improved.

According to a fifth aspect of the present invention, when the core inlet flow rate is automatically or manually reduced from the rated output operation state to 50% or less of the rated flow rate, or in the rated output operation state, the power of the recirculating pump motor is lost and the core inlet flow rate is reduced. When the flow rate falls below the 50% rated flow rate, the relative output becomes one-third of all fuel assemblies.
The subcooled boiling length of the fuel assembly in the range up to 10 /
It does not exceed 24 nodes and the subcool boiling point is 2 /
The feature is to not go below the lower end of 48 nodes.

According to the fifth aspect of the present invention, attention is paid to the fact that the subcool guide generated at the lower part of the core is deeply involved in the stability of the core, and the relative output is within a range of up to 1/3 of the total number of fuel assemblies. Subcooled boiling length of fuel assembly at 10/24
Subcooled boiling length starting point is 2 /
You can impose restrictions that do not go below the bottom of the 48 nodes.

According to a sixth aspect of the present invention, the core operation cycle burn-up period which imposes a limit on the axial power distribution of the fuel rods in the fuel assembly, or a limit on the subcooling boiling length or the saturation boiling start point, The period is from the time when all the control rods are pulled out to the end of the cycle.

According to the sixth aspect of the present invention, the cycle burn-up period for imposing a limit on the core axial power distribution or a limit on the saturation boiling start point is a period from the time when all the control rods are withdrawn to the time when the control rod is terminated. Can be limited to

According to a seventh aspect of the present invention, the limitation on the axial power distribution of the fuel rods in the fuel assembly or the limitation on the subcooled boiling length or the saturated boiling start point is as follows:
When the relative outputs of all the fuel assemblies are arranged in descending order, the average axial power distribution or the average subcooled boiling length of the fuel assemblies in the range of up to 1/3 of the total number of the fuel assemblies, or It is characterized by imposing on the average subcool boiling start height.

According to the seventh aspect of the present invention, a limit is imposed on the average axial power distribution, average subcooled boiling length or average subcooled boiling height for a fuel assembly having a relative output ranging from the highest to 1/3. Can be imposed.

The invention according to claim 8 is characterized in that when relative outputs of all fuel assemblies are arranged in descending order, one out of the total number of fuel assemblies
For fuel assemblies in the range up to / 3, the result (R0 value) obtained by dividing the sum of the squares of the peaking of the radial output distribution by the number of fuel assemblies is 1.75 or less, and When the absolute value of the product of the peaking of the radial power distribution and the peaking of the circumferential first-order mode power distribution is arranged in descending order, the fuel assemblies in the range of up to 1/3 of the total number of the fuel assemblies have the radial output. Result of adding the absolute value of the product of the peaking of the distribution and the peaking of the circumferential first-order mode output distribution and dividing by the number of fuel assemblies (R1 value)
Is set to 2.00 or less, and restrictions on the axial power distribution, the subcooled boiling length and the subcooled boiling start point described in claims 1 to 6 are imposed. According to the invention of claim 8, it is possible to impose restrictions on the axial power distribution, the subcool boiling length and the subcool boiling start point described in claims 1 to 6.

According to the ninth aspect of the present invention, a local power monitor signal disposed in a core, a reactor pressure, a core inlet flow rate,
A process computer for inputting a heat balance condition including a feedwater temperature and a feedwater flow rate, and a core monitoring unit for inputting a core power distribution, a subcool boiling start height and a saturation boiling start height signal from the process computer. ,
The core monitoring unit includes a determination unit, an information storage unit and a display unit connected to the determination unit, and a system that generates an alarm in a central control room when a value from the core monitoring unit exceeds a predetermined value. It is characterized by having.

According to the ninth aspect of the invention, it is possible to monitor the restrictions in the core axial direction, the radial direction, and the circumferential direction. That is, when the axial output distribution of the fuel assembly whose relative output is within a predetermined range of all the fuel assemblies falls below or exceeds a predetermined value within a predetermined range, the subcooling is performed. Central control when the boiling length exceeds a predetermined value, when the subcool boiling start height falls below a predetermined value, when the R0 value exceeds a predetermined value, or when the R1 value exceeds a predetermined value. An alarm can be generated in the room.

[0081]

DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS A first embodiment of a boiling water reactor core according to the present invention will be described with reference to FIGS.
In each of the drawings, the same portions as those described in FIGS. 14 to 21 are denoted by the same reference numerals, and description of overlapping portions will be omitted.

FIGS. 1 and 2 show the structure of the present embodiment.
× 9 fuel assembly axial power distribution and subcooled boiling length,
7 is an example of a fuel assembly nucleus design for realizing a restriction on a subcool boiling start height. FIG. 1 is a cross-sectional view of a fuel assembly, and FIG. 2 shows enrichment or gadolinia concentration of each of the fuel rod types 1 to 7 in FIG. Fuel rod type 6 node 1/24
~ 10/24 is filled with MOX fuel pellets (M) containing the highest gadolinia concentration a +.

[0083]

[Table 2]

Due to the gadolinia effect, the output generation in the lower region of the fuel rod type 6 having high reactivity value is suppressed in the first half of the cycle, and the output ratio to the entire bundle output in the same region is increased in the second half of the cycle when the gadolinia burns out.

This prevents the core axial power distribution from becoming extremely low strain distribution from the initial stage to the middle stage of the cycle, and maintains the core axial power distribution from flat to the lower peak from the middle stage to the end of the cycle. It is possible to do.

As a result, it is possible to prevent the channel / region stability from being deteriorated in the early to middle stages of the cycle, and to improve the scram characteristics and core stability in the middle to late stages of the cycle.

FIG. 3 shows the results of the area safety analysis in the core in FIG. 1 in the axial power distribution, and FIGS. 4 (a) and 4 (b) show examples of nuclear design with and without gadolinia. . FIG. 4A shows that, at the beginning of the cycle, the recirculation pump speed is lowered from the operating state of the rated output / 85% rated flow rate loaded with the 9 × 9 fuel assembly of the nuclear design example,
The operating point is defined as the intersection of the stability limit curve and the flow control curve (Fig. 17
The response (region stability) of the neutron flux at the point of symmetry of the reactor core with respect to the reactivity disturbance when moved to the point B) is evaluated by the core three-dimensional analysis code.

When gadolinia is contained only in a + in a mixed oxide fuel pellet of uranium and plutonium, when the relative output is arranged in descending order, the fuel within the range of up to 1/3 of the total number of fuel assemblies As shown in FIG. 3, the average axial power distribution in the aggregate is 1.60 (peak node 4/24) with peaking, but when no gadolinia is contained, peaking is 1.80 (peak node 3/24).
It has become.

At this time, the subcool boiling start heights are within the 2/48 node and the 1/48 node, respectively. When the average axial power distribution is 1.6 (peak node 4/24), the neutron flux is attenuated at a reduction ratio of 0.8. On the other hand, when the average axial power distribution is 1.8 (peak node 3/24), the neutron flux is oscillating (limit cycle state).

FIG. 6 shows a conventional scrum plant loaded with the 9 × 9 fuel assembly of the above nuclear design example, and the rated output / 105% rated flow rate when the control rods were fully pulled out at the end of the cycle (FIG. The analysis result of the core load three-dimensional code at 17 points A3) of the generator load shedding / bypass valve inactivation is shown by comparing the present invention with the conventional example.

The initial core axial power distribution at this time is shown in FIG. 5 in comparison with the conventional example. The average axial power in the fuel assembly whose relative power is in the range from the highest to one third is shown. The distribution exceeds 1.10 in the range of 5/24 nodes to 9/24 nodes (the maximum value is 1.15 for 16/24 nodes), and the average subcooled boiling length is 7.5 / 24 nodes.

The ΔMCPR during the transient change is 0.21, and the margin of ΔMCPR is increased by 0.07 as compared with the conventional example. This is considered to be due to a combination of improved scrum performance and a shorter subcooled boiling butterfly compared to the conventional example.

Next, FIG. 8 shows that at the end of the cycle, the operating point is moved to the minimum pump speed maximum output point (point D in FIG. 17) by lowering the recirculation pump speed from the operation state of the rated output / 105% rated flow rate. The response of the core average neutron flux to the reactivity disturbance at the time of the reaction is evaluated by the core three-dimensional analysis code.

FIG. 7 shows the core axial power distribution at the lowest pump speed maximum power point at this time. When the relative power is arranged in descending order, it is within 1/3 of the total number of fuel assemblies. The average axial power distribution in the fuel assembly has a peaking of 1.25 (peak node 5/24), from 5/24 to 10/24.
In the range of 1.10, the average subcool boiling length is 7.
8/24 nodes, average subcool boiling start height within 2/24 nodes. The reduction ratio of the neutron flux response is 0.65, which is 0.13 more than the conventional example.

Next, a second embodiment of the boiling water reactor core according to the present invention will be described with reference to FIGS. 9 and 10. 9 and 10 show a second embodiment for realizing restrictions on the axial power distribution of the fuel rods, the saturation boiling start height, and the subcool boiling start height. Table 3 is a legend for FIGS. 9 and 10.

[0096]

[Table 3]

The nodes 1/24 to 10/24 of the fuel rod type 6 are filled with MOX fuel pellets having the maximum plutonium enrichment X including the highest gadolinia concentration a +, and
24 is filled with MOX fuel pellets of plutonium enrichment Y containing gadolinia concentration b.

Due to the gadolinia effect, the output generation in the lower region of the fuel rod type 6 having high reactivity value is suppressed in the first half of the cycle, and the output ratio to the total bundle output in the same region in the second half of the cycle when the gadolinia burns out is increased. ing.

Next, a third embodiment of the boiling water reactor core according to the present invention will be described with reference to FIGS. 11 and 12. Figure
FIGS. 11 and 12 show a third embodiment for realizing the restrictions on the axial power distribution, the saturated boiling start height, and the subcooled boiling start height. Table 4 shows FIG.
13 and the legend of FIG.

[0100]

[Table 4]

The nodes 1/24 to 10/24 of the fuel rod type 6 are filled with MOX fuel pellets (M) not containing gadolinia, and the enrichment without gadolinia is also contained on 11/24 to 23/24. A is filled with uranium fuel pellets.

Then, the fuel rod type 7 node 1/24 to
10/24 is filled with uranium fuel pellets of enrichment B containing the maximum gadolinia concentration a, and from 11/24 to 23/24 is filled with uranium fuel pellets of enrichment B also containing gadolinia concentration b. .

As shown in FIG. 11, at least four fuel rod types 7 are arranged around the fuel rod type 6. Fuel rod type 6 with high reactivity value in the lower area,
Also, by enclosing the fuel rod type 7 having a high gadolinia concentration in the lower region, the ratio of the lower power of the fuel rod type 6 to the entire bundle output in the latter half of the cycle is increased.

The present embodiment is different from the 9 × 9 fuel assembly having two large diameter water rods 44 described above in addition to the high burnup 8 × having one large diameter water rod.
8 fuel assemblies, 9 × 9 fuel assemblies or 10 × 10 fuel assemblies with one square water channel, 9 × 9 fuel assemblies or 10 × 10 fuel assemblies with one cruciform water channel Is also applicable.

Further, nuclear (power distribution) in the core axis direction,
Restrictions on thermo-hydraulic enemies (subcooled boiling length and subcooled boiling start height) are imposed on the fuel assemblies in the range of up to 1/3 of the total number of fuel assemblies when the relative power is arranged in descending order. Sufficient effectiveness can be expected, but by further adding nuclear restrictions in the core radial direction and circumferential direction, transient characteristics and stability can be more reliably improved.

This is because, when the relative outputs are arranged in descending order, the R0 value of the fuel assemblies in the range of up to 1/3 of the total number of fuel assemblies is 1.75 or less, and the radial output distribution of all the fuel assemblies is 1.75. When the absolute value of the product of peaking and the peaking of the circumferential first-order mode output distribution is arranged in descending order, the R1 value of the fuel assembly in the range of up to 1/3 of the total number of fuel assemblies is set to 2.
Achieved by setting it to 00 or less.

Next, a first embodiment of the core monitoring apparatus according to the present invention will be described with reference to FIG. FIG. 13 shows the axial power distribution and subcooled boiling start height, saturated boiling start height,
FIG. 2 shows a block diagram of a core monitoring device for monitoring limits on radial power distribution and circumferential power distribution.

The core monitoring apparatus according to the present embodiment is similar to that shown in FIG.
As shown in (1), a process computer 48 for inputting the LPRM signal and the heat balance condition, a core monitoring unit 49 for inputting the core power distribution, the subcool boiling start height and the saturated boiling start height output from the process computer 48, Consists of

The core monitoring unit 49 includes a determination unit 50 and the determination unit 50.
An information storage unit 51 and a display unit 52 are connected to each other. The core monitoring unit 49 is connected to a system (not shown) for generating an alarm in a central control room (not shown) when the value from the core monitoring unit 49 exceeds.

That is, the process computer 48 executes the LPRM
Signal, reactor pressure P, core inlet flow rate W, feedwater temperature T _FW ,
With the heat balance condition consisting of the feedwater flow rate F _FW as input, the core power distribution consisting of the radial output peaking Ψ (r) and the axial output peaking Φ (z), the reactor heat output Q,
The core entrance enthalpy H is calculated.

[0111] The core monitoring unit 49 includes: Ψ, Φ, Q, W, P, H
Is an input signal, and the determination unit 50, the information storage unit 51, and the display unit
Consists of 52. The information storage unit 51 stores a relative output range (for example, 1 /
3), the range of the axial power distribution (for example, 5/24 to 9/24 node, 3/24 to 5/24 node) that imposes the limit, the output distribution limit value (for example, 1.05 and 1.75), and the limit. Subcool boiling length to impose (eg, for 8/24 nodes), subcool boiling start height to impose limits (eg, 2/24 node bottom), and R0 value (eg, 1.75), R to impose limits
One value (for example, 2.00) is input in advance.

Next, an example of the operation when the core monitoring apparatus having the above configuration is operated at the end of the operation cycle will be described. The determination unit 50 determines that the axial output distribution of any one of the fuel assemblies in the range from the maximum relative output to 1/3 is from 5/24 to 9 /
If it becomes 1.04 in any of the 24 nodes, or 3 /
If it becomes 1.76 in any node from 24 to 5/24, an alarm is issued to the central control room.

When the subcooled boiling length of any one of the fuel assemblies in the range from the maximum relative output to 1/3 is 9/24 node, or the subcooled boiling start point is 1/24
An alarm is also issued to the central control room when there are 48 nodes.

Further, when the relative outputs are arranged in descending order, when the R0 value of the fuel assemblies within 1/3 of the total number of fuel assemblies becomes 1.76, When the absolute value of the product of the peaking of the power distribution and the peaking of the circumferential first-order mode power distribution is arranged in descending order, the R1 value of the fuel assembly within 1/3 of the total number of fuel assemblies becomes 2.01. In the event of a warning, a warning is issued to the central control room.

The axial power distribution, saturated boiling start height,
The subcool boiling start height, the R0 value, the R1 value, and the determination result are displayed on the display unit 52 in numerical value, color, color tone, color brightness,
The image is displayed on the screen by identifying the color intensity.

According to the present embodiment, the nuclear thermo-hydraulic restriction in the axial direction of the reactor core can be monitored.
It is possible to notify the operator in the central operation room, and it is possible to promote appropriate monitoring parameters by operating the control rod.

The core monitoring apparatus according to the present invention can be easily combined with another stability monitoring apparatus. For example, the core monitoring apparatus can be combined with a boiling water reactor monitoring control apparatus described in Japanese Patent Application Laid-Open No. 11-38175. Thereby, it is possible to more reliably prevent the deterioration of stability.

[0118]

According to the core of the present invention, nuclear limitations (limits on power distribution) in three dimensions (axial, radial, circumferential) and thermal hydraulic limitations (subcooled boiling length and By applying the subcooling boiling start height), it is possible to increase the margin of thermal integrity of the fuel when a transient change occurs in the rated output operation state, and at the same time, to reduce the flow rate / high output. It is possible to increase the margin for the occurrence of neutron vibration in the operating state.

The nuclear thermo-hydraulic restriction in the axial direction of the core is based on a fuel rod filled with a low-enriched uranium oxide or a fuel rod filled with a mixed oxide of uranium and plutonium arranged in a square lattice. In the assembly, this is achieved by optimizing the axial distribution of the average uranium enrichment and the amount of burnable poisons and the arrangement in the cross section of the fuel assembly.

According to the core monitoring apparatus of the present invention, it is possible to monitor the nuclear thermo-hydraulic restriction in the axial direction of the core, and to notify the operator in the central operating room if the restriction is deviated, It is possible to promote appropriate control of monitoring parameters by operating the control rod.

[Brief description of the drawings]

FIG. 1 is a cross-sectional view showing a fuel assembly according to a first embodiment of the present invention including a part of a control rod.

FIG. 2 is an arrangement diagram of different fuel rods in the fuel assembly in FIG. 1;

FIG. 3 is an axial power distribution diagram showing a result of an area safety analysis in the core shown in FIG. 1;

4A is a waveform curve diagram showing an example of a nuclear design with gadolinia in FIG. 1; FIG. 4B is a waveform curve diagram without gadolinia in FIG. 1;

FIG. 5 is a core axial direction power distribution curve diagram showing a transient change analysis result in a core according to the present invention and a conventional example.

FIG. 6 is a curve diagram showing a comparison between analysis results by a three-dimensional core code of the present invention and a conventional example.

FIG. 7 is a curve diagram showing a comparison of core stability analysis results in a core axial direction power distribution between a core according to the present invention and a core according to a conventional example.

FIG. 8 is a comparison curve diagram in which the response of the average neutron flux in the core according to the present invention and the conventional example is evaluated by the core three-dimensional analysis code.

FIG. 9 is a cross-sectional view showing a state including a part of a control rod of a fuel assembly according to a second embodiment of the present invention.

FIG. 10 is an array diagram of different fuel rods in the fuel assembly in FIG. 9;

FIG. 11 is a cross-sectional view showing a fuel assembly according to a third embodiment of the present invention, including a part of a control rod.

FIG. 12 is an arrangement diagram of different fuel rods in the fuel assembly in FIG. 11;

FIG. 13 is a block diagram showing a first embodiment of a reactor core monitoring apparatus according to the present invention.

FIG. 14 is a schematic configuration diagram for explaining a conventional boiling water nuclear power plant.

FIG. 15 is a longitudinal sectional view schematically showing the core in FIG.

FIG. 16 is a plan view of the core in FIG.

17 is an operation characteristic diagram showing a relationship between a reactor power and a core flow rate in FIG.

18 (a) is an elevation view showing a conventional 8 × 8 fuel assembly, FIG. 18 (b) is a plan view showing an upper tie plate part of FIG. 18 (a), and FIG. 18 (c) is a round cell type of FIG. Plan view showing a spacer,
(D) is a cross-sectional view showing the large diameter water rod part of (a), and (e) is a plan view showing the lower tie plate part of (a).

19A is an elevation view showing a conventional 9 × 9 fuel assembly, FIG. 19B is a plan view showing an upper tie plate part of FIG. 19A, and FIG. 19C is a round cell type of FIG. Plan view showing a spacer,
(D) is a cross-sectional view showing the large diameter water rod part of (a), and (e) is a plan view showing the lower tie plate part of (a).

FIG. 20 is a cross-sectional view of the fuel assembly in FIG. 19 including a part of control rods.

FIG. 21 is an arrangement diagram of different fuel rods in the fuel assembly in FIG. 20;

22A is a transient change analysis diagram at the beginning of a cycle in a conventional example, FIG. 22B is a transient change analysis diagram at the middle of the cycle, and FIG. 22C is a transient change analysis diagram at the end of the cycle.

23A is a stability analysis diagram at the beginning of a cycle in a conventional example, FIG. 23B is a stability analysis diagram at the middle stage of the cycle, and FIG. 23C is a stability analysis diagram at the end of the cycle.

FIG. 24 is an axial output distribution diagram showing a stability analysis result in a conventional example.

FIG. 25 is a characteristic diagram showing a width reduction ratio and an axial output distribution as in FIG. 24;

[Description of Signs] 1 ... Reactor containment vessel, 2 ... Reactor pressure vessel, 3 ... Core,
4 ... Coolant, 5 ... Shroud, 6 ... Fuel assembly, 7 ... Control rod, 8 ... Core support plate, 9 ... Fuel support fitting, 10 ... Upper grid plate, 11 ... Recirculation system, 12 ... Main steam piping, 13 ... turbine,
14 ... generator, 15 ... condenser, 16 ... condensate pump, 17 ... feed water heater, 18 ... feed water pump, 19 ... feed pipe, 20 ... feed water regulating valve,
21: bypass piping, 22: bypass valve, 23 ... relief safety valve, 24 ... vent pipe, 25 ... suppression pool, 25a ...
Pool water, 26… Jet pump, 27… Reactor coolant recirculation loop, 28… Reactor coolant recirculation pump, 29… Main steam isolation inner valve, 30… Main steam isolation outer valve, 31… Main steam stop valve , 32
... turbine steam regulator, 34 ... high burnup 8x8 fuel assembly, 35 ... fuel rod, 36 ... conventional upper tie plate, 37 ... round spacer, 38 ... large diameter water rod, 39 ... lower tie plate, 40… Network part, 41… 9 × 9 fuel assembly,
42… Improved upper tie plate, 43… Round spacer, 44…
Large diameter water rod, 45 ... Improved lower tie plate, 46
… Network part, 47… Part length fuel rod.

Continued on the front page (51) Int.Cl. ⁷ Identification symbol FI Theme coat II (Reference) G21C 7/04 G21C 3/30 Y 17/00 17/00 SV (72) Inventor Shiho Miyamoto Isogo-ku, Yokohama-shi, Kanagawa 8 Shin-Sugitacho Inside Toshiba Yokohama Office (72) Inventor Tomomi Yamamoto 66-2 Horikawa-cho, Saiwai-ku, Kawasaki-shi, Kanagawa Prefecture Inside Toshiba Engineering Co., Ltd. 66-2 Horikawacho Toshiba Engineering Co., Ltd. F-term (reference) 2G075 CA08 CA38 DA20 EA01 FA06 FA12 FA19 FB07 FB08 FB09 FB16 FD07 GA15

Claims (9)

[Claims] 1. A boiling water reactor core, in which fuel assemblies in which fuel rods filled with low-enriched uranium oxide are arranged in a square lattice are loaded in a core, an axial direction in which the oxides are filled. The average uranium enrichment of the cross section of the fuel assembly in the lower region of 1/24 to 10/24 below the region is higher than the average uranium enrichment of the cross section of the fuel assembly in the upper region of 11/24 to 24/24 in the axial direction, And a fuel assembly in which the average burnable poison amount in the cross section of the fuel assembly in the lower 1/24 to 10/24 region is larger than the average burnable poison amount in the cross section of the fuel assembly in the upper 11/24 to 24/24 region. By loading the reactor core, peaking of the axial power distribution of the fuel assemblies in the range of up to 1/3 of the total number of fuel assemblies when arranged in the order of the relative power of all the fuel assemblies in the rated power operation state Is in the range of 5/24 to 9/24.
A boiling water reactor core, wherein the core length is not less than 05 or a subcooled boiling length generated in a lower portion of the core does not exceed a length of 8/24 in an axial region. 2. A boiling water reactor core, in which a fuel assembly in which fuel rods filled with a mixed oxide of uranium and plutonium are arranged in a square lattice is loaded in a core, the oxide is filled. 1/24 to 10 / below the axial area
24, the average uranium enrichment cross section of the fuel assembly in the lower region is higher than the average uranium enrichment cross section of the fuel assembly in the region of 11/24 to 24/24 in the upper part in the axial direction, and the lower uranium enrichment is 1/24 to 10/24.
The average burnable poison amount of the cross section of the fuel assembly in the region is 11/24
By loading a fuel assembly larger than the average burnable poison amount of the cross section of the fuel assembly in the region from 24 to 24 to the core, the relative output of all the fuel assemblies in the core is increased by 1 from the highest order.
The axial peaking of the fuel assembly in the range of up to 3/3 does not fall below 1.05 in the axial region 5/24 to 9/24 or the subcooled boiling length is less than 8/24 in the axial region. A boiling water reactor core characterized by not exceeding. 3. A part-length fuel rod filled with a mixed oxide of uranium and plutonium free of burnable poisons from 1/24 to 10/24 below the axial region, and from the upper axial portion 11/24.
A fuel assembly in which 24/24 fuel rods connected to low-enriched uranium oxide containing no burnable poisons and fuel rods connected to partial-length fuel rods and full-length fuel rods filled with low-enriched uranium oxide are arranged in a square lattice In a core for a boiling water reactor in which a body is loaded in a core, a low-enriched uranium oxide containing a burnable poison is added to 1/24 to 10/24 of at least four fuel rods around the fuel rod. By arranging the filled full-length fuel rods, the axial peaking of the fuel assemblies in the range from the highest relative output of all the fuel assemblies of the core to one third is increased from the axial region 5/24 to 9/9. A boiling water reactor core, characterized in that the boiling range is not less than 1.05 in the range of 24 or the subcooled boiling length does not exceed 8/24 in the axial region. 4. When the core inlet flow rate is automatically or manually reduced from the rated output operation state to 50% or less of the rated flow rate,
Or, when the power of the recirculation pump motor is lost in the rated output operation state and the core inlet flow rate becomes 50% or less of the rated flow rate, or when the relative outputs of all the fuel assemblies are arranged in descending order, 1 in the total number of fuel assemblies The peaking of the axial power distribution of the fuel assembly in the range of up to / 3 is from 4/24 node to 8 /
A boiling water reactor core characterized in that it does not fall below 1.15 in the range of 24 nodes and does not exceed 1.75 in the range of 3/24 nodes to 5/24 nodes. 5. When the core inlet flow rate is automatically or manually reduced from the rated output operation state to 50% or less of the rated flow rate,
Or, when the power of the recirculation pump motor is lost in the rated output operation state and the core inlet flow rate becomes 50% or less of the rated flow rate,
The subcooled boiling length of the fuel assembly whose relative output is within 1/3 of all fuel assemblies does not exceed 10/24 nodes,
A boiling water reactor core characterized in that the subcooling boiling start point does not fall below the lower end of the 2/48 node. 6. All the control rods are pulled out of the core operation cycle burnup period which imposes a restriction on the axial power distribution of the fuel rods in the fuel assembly or a restriction on the subcooled boiling length or the saturation boiling start point. 6. A period from the end of the cycle to the end of the cycle.
A boiling water nuclear reactor core as described. 7. The restriction on the axial power distribution of the fuel rods in the fuel assembly, or the restriction on the subcooled boiling length or the saturation boiling start point, is set when the relative outputs of all the fuel assemblies are arranged in descending order. , 1/3 of all fuel assemblies
The boiling water atom according to claim 1, wherein the average axial power distribution, the average subcool boiling length, or the average subcool boiling start height is set for the fuel assembly in the range of up to: Furnace core. 8. When the relative outputs of all the fuel assemblies are arranged in descending order, the fuel assemblies in the range of up to 1/3 of the total number of fuel assemblies are added for the square of the peaking of the radial power distribution. Is divided by the number of fuel assemblies (R0 value) is 1.75 or less, and the absolute value of the product of the peaking of the radial power distribution and the peaking of the circumferential primary mode power distribution of all the fuel assemblies is determined in descending order. When arranged, the fuel assemblies in the range up to 1/3 of the total number of fuel assemblies are:
The result (R1 value) obtained by dividing the sum of the absolute value of the product of the peaking of the radial power distribution and the peaking of the circumferential primary mode power distribution by the number of fuel assemblies (R1 value) is 2.00 or less. A core for a boiling water reactor, which imposes restrictions on the axial power distribution, the subcooled boiling length, and the subcooled boiling start point described in the items 1 to 6. 9. A process computer for inputting a local power monitor signal arranged in a reactor core and heat balance conditions including a reactor pressure, a reactor core inlet flow rate, a feedwater temperature and a feedwater flow rate, and a core power distribution from the process computer A core monitoring unit for inputting a subcool boiling start height and a saturation boiling start height signal, wherein the core monitoring unit includes a determination unit, an information storage unit and a display unit connected to the determination unit. And a system for generating an alarm in a central control room when a value from the core monitoring unit exceeds a predetermined value.