JP3115092B2 - Effective gain measurement method and neutron detector placement method - Google Patents

Description

DETAILED DESCRIPTION OF THE INVENTION

[0001]

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to an irradiated fuel for securing criticality safety in a process of storing spent fuel used in a nuclear reactor, that is, irradiated fuel, in a fuel storage rack, a fuel transport storage container, or the like. The present invention relates to a method of measuring an effective multiplication factor of a loaded subcritical system and a method of arranging a neutron detector in the method of measuring an effective multiplication factor.

[0002]

2. Description of the Related Art In general, spent fuel discharged from a nuclear reactor is sent to a reprocessing facility after a predetermined cooling period and is reprocessed, or stored for a long period of time. However, as a global trend, the importance of long-term storage is gradually increasing because more spent fuel is released from nuclear reactors than reprocessing capacity.

[0003] In storage, it is economically advantageous to store at high density, but it is an absolute condition that there is no possibility of reaching criticality.

However, conventionally, no practical method for monitoring the subcriticality of the storage system has been developed, so that it is necessary to ensure that the criticality is not reached, so that excessive safety must be ensured.

[0005] Here, if the subcriticality of the system can be easily evaluated at an intermediate stage of the storage operation, it is not necessary to secure an excessive margin, so that it is possible to store spent fuel at a higher density. Become.

Therefore, the present inventors first disclosed in Japanese Patent Laid-Open No.
In Japanese Patent Publication No. 98, a "spontaneous neutron multiplication method" for monitoring subcriticality without using an external (artificial) neutron source was proposed.

The present inventors have also found that the neutron flux distribution form in the case of a spontaneous neutron source in a subcritical system is as follows:
It was confirmed through experiments and analysis that it had a unique relationship with the subcriticality, and in Japanese Unexamined Patent Publication No. 1-250898, a "subcritical axial spontaneous neutron flux distribution method" was proposed.

[0008]

[Problems to be Solved by the Invention] For example, JP-A-1-169398
The effective multiplication factor obtained by the eigenvalue calculation mode of the neutron transport / diffusion calculation in the spontaneous neutron inverse multiplication method disclosed in Japanese Unexamined Patent Publication is corrected based on the actually measured neutron flux count rate or the measured neutron flux count rate. Is not directly related to the calculated neutron flux obtained by the fixed source calculation mode of neutron transport / diffusion calculation using the calculated group constants, and it is unclear what the effective gain monitored by the conventional effective gain measurement method is. Met.

SUMMARY OF THE INVENTION The present invention has been made to solve the above problems, and an object of the present invention is to confirm the criticality safety of an irradiation fuel loading system based on the neutron multiplication characteristics of irradiation fuel, In the process of storing neutrons in a rack or container, the spontaneous neutron inverse multiplication method is directly related to the local neutron flux in the system where the neutron detector measures the effective multiplication factor to be monitored, and the measurement evaluation An object of the present invention is to provide an effective gain measuring method for monitoring the safest effective gain to be performed and a method of arranging a neutron detector in the effective gain measuring method.

[0010]

According to a first aspect of the present invention, a neutron flux count rate is measured at a predetermined position inside or outside a subcritical neutron multiplication system constructed by sequentially loading a large number of irradiation fuels. On the other hand, the group constant and neutron generation rate distributed in the axial direction are given for each irradiation fuel, and the neutron transport / diffusion calculation is performed by giving the shape and group constant of the irradiation fuel loading system, and the position corresponding to the measurement position is calculated. The neutron flux is calculated, and the irradiation fuel loading system is used as necessary so that the ratio between the calculated neutron flux and the measured neutron flux count rate is substantially constant at each loading step of the irradiation fuel. Corrected group constant, the corrected group constant that the ratio between the neutron flux obtained by the calculation and the neutron flux count rate obtained by the measurement becomes constant, and the neutron generation rate and the shape and size and the group constant of the irradiated fuel loading system Giving a neutral Performs transport and diffusion calculation, the effective multiplication factor measurement method of irradiation fuel loading subcritical system to determine the effective multiplication factor, calculations with zero fission cross-sectional area of the fuel region in the neutron transport and diffusion calculation to determine the neutron flux calculation value And the effective multiplication factor is measured from the two calculated neutron fluxes at the neutron flux count rate measurement position.

[0011] The first invention in The first few steps of the loading, measured by moving the measuring device for measuring the neutron flux count rate in the axial direction inside or outside of a predetermined position of the multiplication scheme At each measurement position, in addition to the usual neutron flux calculation values in the neutron transport / diffusion calculation in which the group constant has been corrected by this neutron flux count rate measurement value, both the neutron flux calculation values with the fission cross section of the fuel region set to zero From the axial direction-dependent effective multiplication factor, and the inclination when the axial direction-dependent reverse multiplication with respect to the value of the normal neutron flux calculation value in the first step as a function of the axial direction-dependent effective multiplication factor is maximum.
In the subsequent loading step, the neutron detector position is fixed at that height, and the reverse multiplication is measured and evaluated to determine the effective multiplication factor.

The second invention measures a neutron flux count rate at a predetermined position inside or at an outer peripheral portion of a subcritical neutron multiplication system constructed by sequentially loading a large number of irradiation fuels. The neutron transport / diffusion calculation is performed by giving the group constant and neutron generation rate distributed in the axial direction to the shape of the irradiated fuel, and the neutron flux at the position corresponding to the measurement position. Correct the group constant of the irradiation fuel loading system as necessary so that the ratio of the neutron flux obtained in the above and the neutron flux counting rate obtained by the measurement becomes almost constant at each loading step of the irradiation fuel. The corrected neutron flux obtained by the above calculation and the neutron flux count rate obtained by the measurement, the corrected group constant at which the ratio is constant, and the neutron generation rate and the shape and size and the group constant of the irradiated fuel loading system are given, and neutron transport is performed.・ Diffusion calculation Carried out, measured neutron flux count rate (below,
Measurements in the arrangement method of neutron detector for measuring the effective multiplication factor from the inverse multiplication of neutrons of flux), moves the neutron detector across the axial direction in the uppermost first few steps of loading, to measure the neutron flux, first Determine the axial position where the slope of the inverse multiplication with respect to the measured value at the step is the largest, and in the subsequent loading steps, fix the position of the neutron detector at that height and measure the inverse multiplication to determine the irradiated fuel. It is characterized by measuring the effective multiplication factor of the loaded subcritical system.

[0013]

The first aspect of the present invention evaluates the subcriticality of an irradiated fuel-loaded subcritical system. That is, fixed source mode calculation (neutron flux mode) of neutron transport / diffusion calculation performed by giving the neutron generation rate and the geometry and group constant of the irradiated fuel-loaded subcritical system
Calculate the neutron flux count rate at the measurement position of the neutron detector using the calculated values of the two neutron fluxes. Measure and evaluate the effective multiplication factor corresponding to the place where it is located.

A second invention is a method for arranging a neutron detector in an effective multiplication factor measuring method for evaluating the subcriticality of an irradiated fuel-loaded subcritical system, wherein the neutron detector comprises:
Between The first few steps of loading the spent fuel is moved in the axial direction, to measure the neutron flux to obtain the highest axial position opposite multiplication of inclination with respect to the measured value when the first step, In the following steps, the neutron detector is fixedly arranged at that height. The neutron detector measures a neutron flux due to spontaneous neutrons based on neutron emitting nuclides accumulated in spent fuel.

[0015]

DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS An embodiment of a method for measuring the effective multiplication factor of a subcritical system loaded with irradiated fuel according to the first invention will be described.

[0016] Fuel irradiated in a nuclear reactor (typically a fuel assembly, and hence hereinafter referred to as a fuel assembly) has a changed fuel composition due to nuclear fission. The change is already known by the combustion management method. However, it is not always clear whether the reality is accurate or not, and it depends on the accuracy of the combustion management method.

The spent fuel assembly irradiated in the nuclear reactor, that is, the irradiated fuel assembly (hereinafter referred to as the irradiated fuel assembly) has a changed fuel composition as described above, and the change depends on the axial direction of the irradiated fuel assembly. Has also changed. That is, the concentration of fissile nuclides changes in the axial direction.

Further, Cm-244, C
Nuclear nuclei of transuranium elements such as m-242, Pu-238 and Pu-240 are generated, and these emit neutrons.
These neutron emitting nuclides have substantially nothing to do with the neutron multiplication properties, and may be considered to constitute a neutron source. The concentration of neutron emitting nuclides varies depending on the axial direction of the irradiated fuel.

The concentration of fission nuclide (Fissail) is directly related to the effective neutron multiplication factor, and the neutron emitting nuclide is not related to the neutron multiplication factor. If these are shown in a subcritical system to which a simple one-point reactor approximation can be applied, the neutron flux (φ) is calculated as neutron generation rate (neutron emission intensity) (S)
And the effective multiplication factor (k _eff ) is given by φ = αS / (1−k _eff ) (1) via a proportional coefficient (α).

The value to be determined is k _eff , but the measurable value is the neutron count rate proportional to φ. If the unit of the proportional coefficient is appropriately selected, φ can be regarded as the neutron counting rate. α is determined by the position and sensitivity of the neutron detector.

Since the value of φ depends on k _eff and S, if the evaluation of S is performed correctly or the effect of S can be substantially eliminated, k _eff can be evaluated from the measurement of φ. However, correctly evaluating the absolute value of S and the absolute value of α is generally very troublesome and not practical.

The main principle of the spontaneous neutron inverse multiplication method employed in this embodiment is described in detail in Japanese Patent Application Laid-Open No. 1-169398, and is outlined below. That is, JP-A-1-169398
According to the publication, the above-mentioned one-point reactor approximation formula is applied to both the measurement and the calculation of the fixed source mode of the neutron transport / diffusion calculation, and the measured neutron flux φ _j ^M in the loading j-th step,
The nuclear constant is corrected so that (φ _j ^C / φ _j ^M ) becomes constant at the calculated value φ _j ^C.

By using the corrected nuclear constants and calculating the eigenvalue mode of the neutron transport / diffusion calculation, the effective multiplication factor (R) at the actual j step is obtained, the subcriticality is evaluated, and the neutron flux φ _{1 in} the first step is obtained. the ratio of the relative first j steps neutron flux phi _j, shown as a function of the effective multiplication factor R, it is expected to secure and "critical are expected" loading step of subcriticality Figures of so-called reverse multiplication.

[0024] However, more and more demanded neutron flux were to measure to anchorage mode calculation in the description are by neutrons generated from nuclei of transuranic elements Cm-244 or the like generated by the combustion of the fuel, neutron detector The value differs depending on the position of.

On the other hand, the generated neutrons are not used in the eigenvalue mode calculation, and the neutron flux distribution obtained at the same time as the effective multiplication factor is completely different from the neutron flux distribution obtained by the measurement or the fixed source mode calculation. There is no direct relationship between the neutron flux being evaluated and the effective multiplication factor, and it was unclear what the effective multiplication factor was.

Therefore, in the present embodiment, when the fission cross-section of the fuel region is set to zero, k _eff = 0 and the obtained neutron fluxes φ ₀ , φ ₀ = αS (2) From the equation (1), k _eff = 1− (φ ₀ / φ) (3), and the eigenvalue mode of the neutron transport / diffusion calculation in which the nuclear constant is modified so that (φ _j ^C / φ _j ^M ) is constant. The calculation is added to the case where the fission cross section is set to zero, and the calculation is performed at the detector measurement position x. k _{eff j} (x) = 1 − [(φ ₀ (x) / φ (x)) _j ] (4)

FIG. 1 is a flow chart showing the flow of one embodiment of the first invention. In FIG. 1, the portion indicated by the broken line is k
FIG. 9 is a supplementary block diagram for improving measurement evaluation accuracy of _eff .

First, the combustion characteristics of the irradiated fuel body are usually
Average axial burnup of irradiated fuel assemblies (BU), the initial enrichment epsilon, given the cooling time t _c is.

In the present embodiment, the effective multiplication factor k _eff is determined for each loading step while a three-dimensional analysis of the irradiated fuel loaded subcritical system is advanced. Therefore, an axial distribution is required for the irradiated fuel body. Therefore, when the axial distribution is given, it may be used, but in FIG. 1, it is devised so that it can be implemented even when it is not given.

Although the relative distribution in the axial direction depends somewhat on the reactor type of the reactor, there is an ideal distribution, and the reactor is designed and operated with the aim of realizing it. Therefore, there is usually no problem even if a typical distribution form is used as the axial distribution form (relative value).

Therefore, in this embodiment, as shown in FIG. 1, the average burn-up (BU) of the irradiated fuel body for each irradiated fuel body
_j is determined, and the axial burnup absolute value distribution (BU) _ij is determined from (BU) _j and the relative distribution form. Here, i is an axial position index, and j is a loading step index.

The next axial neutron generation rate distribution (for each irradiated fuel body) S _ij is obtained by using the correlation obtained by calculation or the like from the initial enrichment εj and the cooling time t _{cj in} addition to (BU) _ij .
In order to improve the accuracy of S _ij , the neutron flux φ _ij is measured as shown by the broken line to measure S _ij (because it may be partial), and S _ij obtained by using a correlation such as a calculation is calculated. It can be modified.

In performing the three-dimensional neutron flux mode calculation, the geometry and group constant of the system are required in addition to the axial neutron generation rate distribution S _ij for each irradiated fuel body. The group constant needs an axial distribution for each irradiated fuel body, and gives the fuel specifications in advance and calculates the grid burning of the irradiated fuel body using a standard nuclear design method to calculate the burnup and fissile nuclide concentration. Ask for it. In order to distribute these group constants to three-dimensional positions in the loading system, the grid combustion calculation was performed using the burnup (BU) _ij and εj obtained as the axial distribution for each irradiated fuel body ( BU) and a group constant corresponding to ε may be obtained.

Therefore, in this embodiment, as shown in the flow chart of FIG. 1, the above-mentioned grid combustion calculation is performed using the fissile nuclide concentration (Fis) _ij obtained based on (BU) _ij , ε j, and t _cj. A method using a group constant corresponding to the obtained fissile nuclide concentration is shown. That is, the neutron generation rate distribution S
Given _ij , the group constant Σ _ij, and the system geometry, three-dimensional neutron flux mode calculations can be performed.

In this calculation, the neutron source intensity distribution (S _ij ) is given instead of _obtaining k _eff, and the absolute value of the neutron flux in the subcritical neutron source multiplication system corresponding to the value is obtained. As the value of k _eff approaches criticality, a small k
_{Even if the eff} changes, the neutron flux (measurable) changes greatly, so that accurate k _eff can be evaluated from the neutron flux measurement.

[0036] In the three-dimensional neutron flux mode calculation, since the neutron flux of the desired position is obtained, the same by measuring the neutron count rate in the corresponding position, by comparing the calculated value phi _j ^c a measure phi _j ^M Can be.

It should be noted that in an elongated system, good results can be obtained even in a two-dimensional calculation in a direction perpendicular to the axis.

[0038] In this embodiment, as both a proportional relationship, that is, φ _j ^c / φ _j ^M ratio to modify the group constant sigma _ij so as not to be changed by j. Using the corrected group constants, the three-dimensional neutron flux mode calculation is performed again, and the calculation with the fission cross-section of the group constants added to zero is added.The neutron detector measures according to equation (4). The effective multiplication factor Rj corresponding to the neutron flux is calculated for each loading step of the irradiated fuel body.

FIG. 2 shows a measurement example and a calculation example of the relationship between φ _i / φ _j and step j. That is, φ _i /
From the extrapolation of φ _j, the “step at which criticality is expected” is extrapolated. Since Σ _ij is corrected so that the calculated value is normalized to the measured value, the “step width at which criticality is expected” becomes narrow. Therefore, if it is found that the agreement with the measured value is good in the middle, it is possible to accurately predict which step criticality is likely to be reached only by calculation without measurement.

Next, the position of the neutron detector used in the present embodiment will be described by taking as an example the case where the irradiated fuel body (SFA) of a boiling water reactor (BWR) is stored in a storage system.

FIG. 3 schematically shows various characteristics of an actual BWR SFA. It should be noted that, in recent years, BWR fuels are relatively often designed to have a change in the axial direction under various ideas, but here, in order to facilitate the description, the axial direction is uniform. An example in which the designed fuel is SFA is shown. However, even if the design has a distribution in the axial direction, if it is an SFA,
There is usually no significant change from the above example.

FIG. 3A shows the relative distribution of the axial burnup (BU). The upper and lower ends are lowered due to axial leakage of neutrons from the core, but the other parts are generally flat. This is because the operation is performed so as to be flat or the fuel is designed.

FIG. 3B shows the distribution of the coolant void ratio. Voids start to be generated from around node 3, and often become 40 to 50% near the center of the height and 70 to 75% near the upper end. Since the void ratio is high in the upper part, the generation rate of Pu is high, and the combustion of ²³⁵ U and ²³⁹ Pu tends to be delayed.

FIG. 3C shows a fission nuclide (Fissail:
(Fis) shows the axial relative distribution of concentration. As a result, combustion is delayed at the upper and lower ends, resulting in a high concentration of residual fissail,
The concentration is low in the high burn-up portion near the nodes 4 and 5. Although the burnup is relatively high in the upper half excluding the upper end, the residual fissail concentration is increasing more and more upward due to the high void ratio and the high Pu generation ratio.

FIG. 3D shows the relationship between k∞ and k _eff, and basically a curve similar to the curve of FIG. 3C is drawn. k∞ is defined in the design calculation as the ratio between the local neutron generation rate with fission and the absorption rate. In FIG. 3D, the broken line is k 破 線, and the solid line is the locally defined neutron effective multiplication factor k (k =
k _eff ).

[0046] In SFA of subcriticality significantly deeper BWR, since neutrons effective travel distance (including up to offspring neutrons) is short, the effect of neutron leakage from the end face in the vicinity of upper and lower ends is limited to the vicinity of the end face As a result, only the vicinity of the end face causes a decrease in k _eff .

In the vicinity of the end face, neutron leakage occurs in the radial direction and the end face, so that there is a high probability of leakage. If the neutron leaks slightly from the end face, for example, 20 to 30 cm or more, the influence of the neutron leak from the end face almost disappears. However, only leakage in the radial direction is effective. Although not shown, as the subcriticality becomes shallower, the influence of neutron leakage from the end face becomes relatively far away.

FIG. 3E shows primary neutrons, ie, SFA
The relative distribution form of the neutron emission rate (S) from the transuranium nuclides accumulated in Fig. 3 is shown in comparison with the relative distribution form of the burnup (BU). The S distribution is similar to the BU distribution, except that SF
A is below the BU distribution shape below A in the axial direction, and is above above the BU distribution type. This is mainly because the value of S in the case of the same burnup increases as the void ratio increases.

FIG. 3F mainly shows the spontaneous neutron source (primary neutron) inherent in the SFA controlled by the above equation (1) by the k _eff curve of FIG. _3D and the S curve of FIG. _3E . 3 shows the axial distribution of the spontaneous neutron flux based on the source. This distribution shape has a shape such that the k _eff curve in FIG. _3D is emphasized by the S curve in FIG. 3E.

A method for quantifying the primary neutron emission rate S and the effective multiplication factor k _eff by placing one SFA in water has already been established and is in the practical stage, except for (B) in FIG. A) to (F) can be quantified. However,
There is no disclosure of a method for obtaining these in a system composed of a plurality of SFAs.

[0051] FIG. 4 is an schematically shows an axial neutron flux distribution based on the spontaneous neutron SFA storage system of BWR.

FIG. 4A shows FIG. 3F, that is, BWR
The axial neutron flux distribution pattern for one SFA is shown. For comparison, this curve A is also shown by a broken line in FIGS. 4B and 4C. As the fuel (SFA) loading on the system increases, the neutron multiplication factor increases.
The neutron flux increase based on (1) appears. The progress is shown by curves A, B and C in FIGS. 4 (A), (B) and (C). In order to facilitate the comparison of the curves, the graph is shown except for the proportional increase in S accompanying the increase in the number of SFA bodies, that is, the case of the same neutron source intensity.

FIGS. 3 and 4 show a fuel length of about 3.6 m, which is divided into 24 equal parts in total length, and named as nodes 1, 2,..., 24 from the lower end toward the upper end. In this case, one node corresponds to about 15 cm.

The method of arranging the neutron detector in this embodiment can employ the method disclosed in JP-A-3-249397 as it is, or the safest effective gain to be measured and evaluated can be obtained. .

As can be seen from FIG. 4, the neutron flux measurement values at each axial height in the first step of loading increase differently as the loading progresses in the second and third steps. It depends on the axial dependence of the local effective multiplication factor shown in (D). Therefore, the inverse multiplication φ _i (Z) / of the neutron flux φ (Z) at the axial height position Z of the neutron detector at the j-th loading step with respect to the first step is
The axial dependence of φ _j (Z) corresponds to the height-dependent effective multiplication factor k _eff (Z) of the neutron detector shown in equation (4).

Therefore, the inverse multiplication φ _i (Z) / φ
When plotting _j (Z) with the effective multiplication factor k _{eff j} (Z) as the horizontal axis, the first few steps move the neutron detector in the axial direction and identify the axial position where the tilt is maximum, in subsequent loading step of measuring without movement of the neutron detector. FIG. 5 shows this state.

The effective multiplication factor k _eff (Z) thus determined for each loading step is the safest value of the effective multiplication factors to be measured and evaluated. Step is also required with high accuracy, and the reliability of subcriticality evaluation and subcriticality securing is greatly improved.

When this embodiment adopts the neutron detector arrangement method disclosed in Japanese Patent Laid-Open Publication No. Hei 3-249397, this method incorporates the features of the BWR, and the arrangement position of the neutron detector is optimized. There is no significant difference from the position, and there is no practical problem.

Although this method can be applied not only to the fuel assembly but also to the fuel rods, at present, the fuel for power reactors mainly deals with the unit of the assembly, so that the method is practically used. Is effective when applied to the loading of the irradiated fuel body into a rack or a transport / storage container (cask).

Next, an embodiment of the second invention will be described by taking as an example a case where an irradiated fuel body (SFA) of a boiling water reactor (BWR) is stored in a storage system.

The second invention relates to a spontaneous neutron measurement method using neutrons spontaneously released from SFA,
This is a method of specifying the SFA axial arrangement of a neutron detector for performing the above-mentioned spontaneous neutron inverse multiplication method.

In order to accurately execute the above method, correction calculation by a computer is necessary. Since the method is already disclosed in the above-mentioned application, for the sake of simplicity, the following approximation will be used. Only a brief explanation.

When a neutron source (intensity S) is given in an infinite homogeneous system or a one-point system, the subcritical neutron flux (φ) is an effective neutron multiplication factor (k _eff : hereinafter abbreviated as k).
It is well known that there is a relationship between φ = αS / (1-k) (5)

[0064] The value of k, when the conventional general knowledge of the furnace physics have been considered constant regardless of the location of the system, for example, placed one body the SFA of the BWR in water, near the top end It is easily understood by those involved in reactor physics that there is almost no probability that the neutron progeny released at the time will be found near the lower end. Therefore, the upper end and the lower end are not neutronically coupled together, and it can be said that it is mathematically possible but physically meaningless to define the whole with one k value. . As the system approaches criticality, it tends to become united. Therefore, the above equation
It can be said that (1) locally defines k.

In the present embodiment, the "spontaneous neutron inverse multiplication method" utilizes the above equation (5) as a basic principle.

FIG. 3 schematically shows various characteristics of the SFA of the actual BWR. It should be noted that, in recent years, BWR fuels are relatively often designed to have a change in the axial direction under various ideas, but here, in order to facilitate the description, the axial direction is uniform. An example in which the designed fuel is SFA is shown. However, even if the design has a distribution in the axial direction, if it is an SFA,
There is usually no significant change from the above example.

FIG. 3A shows the relative distribution of the axial burnup (BU). The upper and lower ends are lowered due to axial leakage of neutrons from the core, but the other parts are generally flat. This is because the operation is performed so as to be flat or the fuel is designed.

FIG. 3B shows the distribution of the coolant void ratio. Voids start to be generated from around node 3, and often become 40 to 50% near the center of the height and 70 to 75% near the upper end. Since the void ratio is high in the upper part, the generation rate of Pu is high, and the combustion of ²³⁵ U and ²³⁹ Pu tends to be delayed.

FIG. 3C shows a fission nuclide (Fissail:
(Fis) shows the axial relative distribution of concentration. As a result, combustion is delayed at the upper and lower ends, resulting in a high concentration of residual fissail,
The concentration is low in the high burn-up portion near the nodes 4 and 5. Although the burnup is relatively high in the upper half excluding the upper end, the residual fissail concentration is increasing more and more upward due to the high void ratio and the high Pu generation ratio.

FIG. 3D shows the relationship between k∞ and k _eff . Basically, a curve similar to the curve in FIG. 3C is drawn. k∞ is defined in the design calculation as the ratio between the local neutron generation rate with fission and the absorption rate. In FIG. 3D, the broken line is k∞, and the solid line is the neutron effective multiplication factor k locally defined.

In an SFA of a BWR whose subcriticality is extremely deep, the effective neutron movement distance (including the descendant neutrons) is short, so that the effect of neutron leakage from the end face near the upper and lower ends is limited to the vicinity of the end face. As a result, only the vicinity of the end face causes a decrease in k _eff . In the vicinity of the end face, neutron leakage occurs in the radial direction and the end face, so there is a high probability of leaking.Slightly away from the end face, for example, 20 to 30 cm or more, the effect of neutron leakage from the end face almost disappears, and the radial direction Only leaks will work. Although not shown in FIG. 3, when the subcriticality becomes shallow, the effect of neutron leakage from the end face extends to a relatively long distance.

FIG. 3E shows primary neutrons, ie, SFA
The relative distribution form of the neutron emission rate (S) from the transuranium nuclides accumulated in Fig. 3 is shown in comparison with the relative distribution form of the burnup (BU). The S distribution is similar to the BU distribution, except that SF
A is below the BU distribution shape below A in the axial direction, and is above above the BU distribution type. This is mainly because the value of S in the case of the same burnup increases as the void ratio increases.

FIG. 3F mainly shows a spontaneous neutron source (primary neutron) inherent in the SFA controlled by the above equation (1) by the k _eff curve of FIG. _3D and the S curve of FIG. _3E . 3 shows the axial distribution of the spontaneous neutron flux based on the source. This distribution shape has a shape such that the k _eff curve in FIG. _3D is emphasized by the S curve in FIG. 3E.

A method for quantifying the primary neutron emission rate S and the multiplication factor k _eff by placing one SFA in water has already been established and is in the practical stage. Then, FIG.
(A) to (F) except for (1) can be determined. However, there is no disclosure of a method for obtaining these in a system composed of a plurality of SFAs.

FIG. 4 schematically shows the axial neutron flux distribution based on spontaneous neutrons in the SWR storage system of the BWR.

FIG. 4A shows FIG. 3F, that is, BWR
The axial neutron flux distribution pattern for one SFA is shown. For comparison, this curve A is also shown by a broken line in FIGS. 4B and 4C. As the fuel (SFA) loading on the system increases, the neutron multiplication factor increases.
The neutron flux increase based on (1) appears. The progress is shown by curves A, B and C in FIGS. 4 (A), (B) and (C). In order to facilitate the comparison of the curves, the graph is shown except for the proportional increase in S accompanying the increase in the number of SFA bodies, that is, the case of the same neutron source intensity.

FIGS. 3 and 4 show a fuel length of about 3.6 m, which is divided into 24 equal parts in length, and named as nodes 1, 2,..., 24 from the lower end toward the upper end. In this case, one node corresponds to about 15 cm.

As can be seen from FIG. 4, the neutron flux measurement value at each axial height in the first step of loading is different from the second and third steps in the way of increasing as the loading progresses. This depends on the axial dependence of the local effective multiplication factor shown in FIG. Therefore, the inverse multiplication φ1 (Z) / φn (Z) of the measured neutron flux φn (Z) at the height Z of the first step with respect to the measured neutron flux φ1 (Z) at the height Z of the n-th step depends on the axial direction. There is.

In this embodiment, in consideration of this point, φ1 (Z) / φ
In the first few steps, the neutron detector is moved in the axial direction to specify the axial position where the inclination of n (Z) is the maximum, and the detector is not moved in the subsequent loading step. Measure.

In this embodiment, during the first few steps, FIG.
As shown in (1), it is possible to specify the position of the detector that gives the reverse multiplication curve that is the safest side in the measurement evaluation of the approach to the criticality or the effective multiplication factor by the neutron flux reverse multiplication.

Since the various characteristics of the SWR of the BWR shown in FIG. 3 do not greatly depend on the axial design based on various ideas of the BWR fuel in recent years, if the SFA becomes a SFA, the characteristics greatly change from the side of FIG. Is usually not present.

Therefore, the axial position of the neutron detector can be determined by finding the position where the slope of the inverse multiplication curve obtained from the neutron flux measurement value due to the axial movement of the neutron detector during the first few steps becomes the maximum. It does not change to subsequent loading steps.

The present invention can be used not only for BWRs but also for PWRs and other nuclear reactors during loading of spent fuel storage.

[0084]

According to the first aspect of the present invention, in the process of proceeding from the first step to the n-th step where the subcriticality is sufficiently deep, the neutron detector is directly related to the neutron flux being measured and should be measured and evaluated. The safest effective multiplication factor can be obtained, and the fuel can be loaded without obtaining an unnecessary excess margin and safely and securely obtaining the effective multiplication value.

According to the second aspect of the invention, in the neutron flux reverse multiplication measuring method for measuring and evaluating the effective multiplication factor in the process of proceeding from the first step having sufficiently deep subcriticality to the n-th step, It is possible to specify the position of the neutron detector.

[Brief description of the drawings]

FIG. 1 is a flowchart showing a flow of an embodiment in the first invention.

FIG. 2 is a characteristic diagram showing a relationship between a reverse multiplication factor φ _i / φ _j and a loading step in FIG. 1;

FIG. 3 is a BWR SFA in the first and second inventions.
3 is a graph showing distributions of various characteristics in the axial direction of (A), burnup, (B) void fraction, (C) fission nuclide, (D)
Indicates k _eff and k∞, (E) indicates the primary neutron, and (F) indicates the characteristic distribution of the neutron flux.

FIG. 4 shows a BWR SFA in the first and second inventions.
It is a graph showing the axial neutron flux distribution of the storage system,
(A) is an axial distribution form of one BWR SFA,
(B) shows the axial distribution when the number of SFAs is slightly increased, and (C) shows the axial distribution when the number of SFAs is further increased.

FIG. 5 is a characteristic diagram showing a relationship between φ _i (Z) / φ _j (Z) and an effective multiplication factor of a height Z in a detector axial direction at a loading j-step in the first invention.

FIG. 6 is a characteristic diagram showing a relationship between φ _i (Z) / φ _j (Z) and a loading step (effective multiplication factor) in the second invention.

[Explanation of symbols]

Curves a in FIGS. 4 (B) and 4 (C)... Comparison curves of the axial distribution type.

──────────────────────────────────────────────────続 き Continuation of the front page (56) References JP-A-1-250897 (JP, A) JP-A-1-199195 (JP, A) JP-A-57-201891 (JP, A) JP-A-57-201891 57294 (JP, A) JP-A-1-169398 (JP, A) JP-A-1-250898 (JP, A) (58) Fields investigated (Int. Cl. ⁷ , DB name) G21C 17/06 G21C 19 / 40

Claims (2)

(57) [Claims] 1. A neutron flux counting rate at a predetermined position inside or outside a subcritical neutron multiplication system constructed while sequentially loading a large number of irradiation fuels, while measuring an axial direction for each irradiation fuel. Gives the group constant and neutron generation rate distributed to
Further, a neutron transport / diffusion calculation is performed by giving a shape dimension and a group constant of the irradiation fuel loading system, a neutron flux at a position corresponding to the measurement position is obtained, and the neutron flux calculated by the calculation and the neutron flux count obtained by the measurement are obtained. Correct the group constant of the irradiation fuel loading system as necessary so that the ratio with the rate becomes substantially constant at each loading step of the irradiation fuel, the neutron flux obtained by the calculation and the neutron flux obtained by the measurement. Given a corrected group constant at which the ratio with the counting rate is constant, and the neutron generation rate and the shape and size and group constant of the irradiated fuel loading system, neutron transport / diffusion calculation is performed to determine the effective multiplication factor. in effective multiplication factor measurement method of the critical system, add the calculations zero fission cross-sectional area of the fuel region in the neutron transport and diffusion calculation for obtaining the neutron flux calculation value, the in neutron flux count rate measurement position Effective multiplication factor measurement method characterized by measuring the effective multiplication factor from One of the neutron flux calculation value. 2. A neutron flux count rate at a predetermined position inside or at a peripheral portion of a subcritical neutron multiplication system constructed by sequentially loading a large number of irradiation fuels, and, on the other hand, an axial direction of each irradiation fuel is measured. Gives the group constant and neutron generation rate distributed to
Further, a neutron transport / diffusion calculation is performed by giving a shape dimension and a group constant of the irradiation fuel loading system, a neutron flux at a position corresponding to the measurement position is obtained, and the neutron flux calculated by the calculation and the neutron flux count obtained by the measurement are obtained. Correct the group constant of the irradiation fuel loading system as necessary so that the ratio with the rate becomes substantially constant at each loading step of the irradiation fuel, the neutron flux obtained by the calculation and the neutron flux obtained by the measurement. By giving the corrected group constant that the ratio with the counting rate is constant, the neutron generation rate and the shape and size and group constant of the irradiated fuel loading system, neutron transport / diffusion calculation is performed, and the effective neutron flux from the inverse multiplication is measured. In the method of arranging a neutron detector for measuring the multiplication factor, the neutron detector is moved in the axial direction in the first few steps of loading, the neutron flux is measured, and the inclination of the inverse multiplication with respect to the measured value in the first step is measured. There obtains the axial positions of the largest,
In the subsequent loading step, the neutron detector is arranged at a fixed height and the reverse multiplication is measured, and the effective multiplication factor of the irradiated fuel-loaded subcritical system is measured. .