JP3041101B2 - Measurement method of effective multiplication factor of spent fuel assembly loading system - Google Patents

Description

DETAILED DESCRIPTION OF THE INVENTION

[0001]

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a method for measuring the effective multiplication factor of a spent fuel assembly loading system.

[0002]

2. Description of the Related Art Generally, spent fuel released from a nuclear reactor is sent to a reprocessing facility after a predetermined cooling period to be reprocessed, or after being stored for a long period of time, is subjected to any treatment. However, the global trend is that long-term storage is becoming increasingly important as more spent fuel is released from 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.

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 the spent fuel assemblies at a higher density. It becomes possible.

Therefore, the present inventors first disclosed in Japanese Patent Laid-Open Publication No. 1-1
No. 69398, “Spontaneous neutron multiplication method” for monitoring subcriticality without using an external (artificial) neutron source
Suggested.

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

[0008]

The use of the above two methods proposed by the present inventors makes it possible to evaluate the subcriticality of a system. A method of disposing a neutron detector in the axial direction of the fuel assembly has not yet been clarified, and this is a problem.

The present invention has been made in view of the above points, and is intended for spent fuel. By appropriately disposing a neutron detector with respect to the axial distribution of the composition, quality of the spent fuel can be improved. It is an object of the present invention to provide an effective multiplication factor measuring method capable of non-destructively determining the effective multiplication factor by two different methods and improving the reliability of the measurement result.

[0010]

According to the present invention, there is provided a method for measuring an effective multiplication factor of a loading system of a spent fuel assembly in which the spent fuel assemblies are loaded one by one. A neutron detector is arranged at the axial center of the system where the change in the relative distribution of neutron fluxes in the axial direction with the change in magnification is small, and the ratio of the neutron counting rate between the reference loading step and the step is determined. A spontaneous neutron multiplication method for obtaining a substantially average effective multiplication factor of the system, a neutron counting rate of a neutron detector arranged in the central portion in the axial direction, and an axial relative distribution of neutron flux with a change in the effective multiplication factor A subcritical axial spontaneous neutron flux distribution type that evaluates the effective multiplication factor of the system using a calibration curve prepared in advance from the ratio of the neutron detector and the neutron detector arranged at the axial end where the shape change occurs Law and
Characterized by the combination of

[0011]

In the method for arranging a neutron detector for subcriticality evaluation of a spent fuel storage system according to the present invention, the neutron detector comprises:
The neutrons that are disposed near the axial center of the spent fuel assembly and / or near the upper end or the lower end near the axial direction, and are accumulated in the spent fuel assembly by these neutron detectors. Spontaneous neutrons based on emitted nuclides are measured.

By the way, the positions near the upper end and the lower end in the axial direction of the spent fuel assembly are positions where neutron leakage to the upper end surface or the lower end surface is remarkable, while the position near the axial center of the spent fuel assembly is located. Is a position where the multiplication characteristic is approximately the average value of the spent fuel assembly. Therefore, the value of the relative standard deviation using the measured value of the neutron flux from these positions has the largest change due to the subcriticality, and the subcriticality evaluation can be performed with high accuracy.

[0013]

DESCRIPTION OF THE PREFERRED EMBODIMENTS The present invention will be described below with reference to a boiling water reactor (BWR).
The case where the spent fuel assembly (SFA) is stored in a storage system will be described as an example.

The present invention relates to a method for measuring the effective multiplication factor of a loading system of a spent fuel assembly in which the spent fuel assemblies are loaded one by one, comprising a spontaneous neutron multiplication method and a subcritical axial spontaneous neutron flux distribution. This is a method that is combined with the shape method.

In order to accurately execute the above two methods, correction calculation by a computer is necessary. Since the method is already disclosed in the above-mentioned application, it will be described below in order to facilitate the explanation. Only an approximate explanation will be given.

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) (1) The value of k has been considered to be constant regardless of the location of the system in the conventional general common sense of reactor physics. For example, BW
It is easily understood by those involved in reactor physics that, when one RSFA is placed in water, there is little probability that neutron progeny emitted near the upper end will be found near the lower end.
Therefore, the upper and lower ends are not neutronically coupled together, and defining the whole with one k value means
It can be said that although it can be done mathematically, it is not physically significant. As the system approaches criticality, it tends to become united. Therefore, it can be said that Equation 1 locally defines k.

In the present invention, the "spontaneous neutron multiplication method" uses the above formula 1 as a basic principle. On the other hand, as the criticality is approached, there is a tendency that they are integrally coupled, and this is achieved at the criticality. The “subcritical axial spontaneous neutron flux distribution method” utilizes the phenomenon in which the subcritical axial neutron flux distribution pattern approaches the neutron flux distribution pattern (in this case, the meaning of the axial relative distribution pattern).

FIG. 1 shows the principle of the subcritical axial spontaneous neutron flux distribution method (hereinafter abbreviated as distribution method). The fuel assembly (FA) is assumed to have a uniform composition in the axial direction here for simplicity. Accordingly, the infinite multiplication factor (k _∞) is uniform in the axial direction. The spontaneous neutron source intensity in the FA is also uniform in the axial direction.

The FA is about 3.6 m, which is the same as the actual FA, and is divided into 24 equal parts in length, and is named nodes 1, 2,..., 24 from the lower end toward the upper end. One node is about 1
5cm.

If the system is critical, FIG. 1 (A),
(B) and (C) show that the distribution type is COS
When the subcriticality is deep, the COS distribution shape becomes a shape crushed except for both ends. As the system approaches criticality (subcriticality becomes shallower), the neutron flux distribution changes as shown by solid lines b, c, and d in FIGS. 1A, 1B, and 1C. As the degree becomes shallower, it approaches the COS distribution shape, and the flat portion at the center as shown in FIGS. 1A and 1B disappears.

[0022] As a method for monitoring the subcriticality monitors the distribution shape, as shown in FIG. 1 (A), placing the neutron detector to the position of the code _{d 1, d 3, d 5} . Then, when the ratio of the neutron count rates at the positions of the codes d ₁ and d ₃ (or the codes d ₅ and d ₃ ) is obtained, a correlation with the subcriticality is obtained. Further, even when a large number of neutron detectors are arranged and the relative standard deviation between the respective counting rates obtained by each detector is obtained, the correlation with the subcriticality can be obtained.

FIG. 2 schematically shows various characteristics of the SFA of the actual BWR. It should be noted that, in recent years, BWR fuels are often designed with a change in the axial direction under various ideas, but here, the axial direction is uniform in order to facilitate the description. Designed fuel is SFA
The example shown in FIG. However, even in the case of a design having a distribution in the axial direction, there is usually no significant change from the example of FIG.

FIG. 2A 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. 2B shows the distribution of the coolant void ratio. Voids begin to be generated around the node 3 and often become 40 to 50% near the center of the height and 70 to 75% near the upper end. Above, the rate of Pu generation is high due to the high void fraction, and the combustion of ²³⁵ U and ²³⁹ Pu tends to be delayed.

FIG. 2 (C) shows the axial relative distribution of fission nuclide (Fissil) concentration. At the upper and lower ends, as a result of delayed combustion, the residual fissail concentration is high,
The concentration is low in the high burn-up portion near the nodes 4 and 5. In the upper half except for the upper end, although the burnup is relatively high, the void fraction is high and the Pu generation rate is high, so that the residual fissail concentration is increasing further upward.

FIG. 2D shows the relationship between k _∞ and k _eff, and basically draws a curve similar to the curve of FIG. 2C. k _∞ is defined in the design calculation as the ratio between the local neutron generation rate with fission and the absorption rate. In FIG. 2 (D), the at broken line k _∞, the solid line is effective neutron multiplication factor k defined locally (k = k
_eff ).

In a BWR SFA having a remarkably deep subcriticality, since the effective neutron travel distance (including the descendant neutrons) is short, the effect of neutron leakage from the end face near the upper and lower ends is limited to the vicinity of the end face. _Therefore , only the vicinity of the end face causes a decrease in k _eff . Near the end face, the neutron leakage occurs in the radial direction and the end face, so the probability of leakage is high, and if it is 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 only leakage in the radial direction is effective. Although not shown in FIG. 2, when the subcriticality becomes shallow, the influence of neutron leakage from the end face extends to a relatively long distance.

FIG. 2E shows primary neutrons, that is, SFs.
Neutron emission rate from transuranium nuclides accumulated in A (S)
Are shown in comparison with the relative distribution of burnup (BU). S is similar to the BU distribution type, but is lower than the BU distribution type below the SFA in the axial direction, and is upper above the SFA. This is mainly because the higher the void ratio, the higher the value of the S distribution type in the case of the same burnup.

FIG. 2 (F) shows a spontaneous neutron source (primary neutron source) inherent in the SFA governed by the above equation 1 mainly by the k _eff curve of FIG. 2 (D) and the S curve of FIG. 2 (E). 3) shows the axial distribution of the spontaneous neutron flux based on ()). This distribution shape has a shape such that the k _eff curve in FIG. 2D is emphasized by the S curve in FIG. 2E.

The method of quantifying the primary neutron emission rate S and the multiplication factor k _eff by placing one SFA in water has already been established by the present inventors and overseas researchers, and has entered the practical stage. ing. Then, quantification of (A) to (F) except for (B) of FIG. 2 is possible. However, a method for obtaining these in a system including a plurality of SFAs has not been developed.

FIG. 3 schematically illustrates the axial neutron flux distribution based on spontaneous neutrons in the BWR SFA storage system.

FIG. 3A is a view showing FIG. 2F, that is, BW
The axial distribution pattern of one RSFA is shown. For comparison,
This curve A is also shown by a broken line in FIGS. 3B and 3C. When the fuel (SFA) loading on the system increases, the neutron multiplication factor increases, so that an increase in the neutron flux based on Equation 1 and a change in the distribution shape based on FIG. 1 appear. The progress is shown by curves A, B and C in FIGS. 3 (A), 3 (B) and 3 (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.

FIG. 4 shows the neutron flux or the ratio of the neutron flux between the measurement points to the neutron effective multiplication factor k depending on the location of the measurement position.
_The change due to _eff is shown as normalized by the value at k _eff = 0.7. The measurement points d ₁ ~ shown below
d ₅ corresponds to the measurement point d ₁ to d ₅ shown in FIG.

A curve A shown in FIG. 4A indicates a change in the curve shape in FIG. 1 and a ratio of the neutron flux at the measurement point d ₁ or d _{5 to} the neutron flux at the measurement point d ₃ as a relative change. . In fuel storage systems, k _eff values are typically between 0.8 and
Since it is necessary to be able to monitor between 0.9 and the curve has changed sufficiently between them, it can be seen that the basic condition of the monitor is possessed.

The curve (b) shown in FIG.
This characteristic can be used when there is a difference in the value of k _eff between the measurement points d ₂ and d ₄ as shown in A. If both measurement points d ₂ and d ₄
If there is no difference between the curves, the curve b becomes a flat straight line. In the SFA of a pressurized water reactor (PWR), it is expected that the case will be almost flat and the case will change as shown by the curve B in FIG. 4B due to the influence of the insertion control rod.

A curve C shown in FIG. 4B shows a neutron flux change characteristic that changes as the k _eff value changes in the equation (1). The curve c of this shape is obtained by measuring points d ₂ , d ₃ ,
d ₄ appear either. Although similar curves appear at the measurement points d ₁ and d ₅ , the effects of the shape change shown in FIG. 1 tend to appear relatively remarkably, so that the measurement points d ₁ and d ₅ are not very convenient for applying the formula 1. .

It is quite difficult to express the curve A shown in FIG. 4 (A) by an analytical expression. The curve A is obtained by numerical calculation, and k _eff is evaluated from the measured value of the neutron flux ratio. The k _eff value obtained here is difficult to define as a local value in a system in which the k _eff value changes in the axial direction as shown in FIG. 2, and is considered to be an approximate SFA average value. On the other hand, the curves b and c shown in FIG.
Can be used to estimate the local k _eff value from the measured neutron flux.

When the relative standard deviation is obtained from the measured values of the neutron flux at a plurality of points separated by space, it is known that there is a correlation with the subcriticality. At this time, in order to increase the change of the value of the relative standard deviation due to the subcriticality,
It is necessary to select a point where the change in the distribution form due to the subcriticality is as large as possible as a measurement point. For this purpose, it is necessary to select at least one of the measurement points d ₁ and d ₅ near both ends and the intermediate measurement point d ₃ at a minimum.
If possible, it is desirable to include the measurement points d ₂ and d ₄ .

FIG. 5 shows an example of a cask fuel basket for accommodating a BWR SFA. The difference from the conventional basket is that the neutron detector insertion hole is provided with a half in which the SFA is not inserted inside the outer cylinder. This is a point provided in the notched lattice portion.

That is, as shown in FIG. 5, the cask fuel basket includes a cylindrical outer cylinder 10 in which a grid-like SFA partitioned by a lattice material 11 is provided. A storage space 12 is provided. In the outer cylinder 10, neutron detector insertion holes 13 are provided at, for example, four places in a semi-notched lattice portion that does not store the SFA.

Note that the detector insertion position need not be limited to the position shown in FIG. 5 as long as the method according to the present invention is carried out, but it is necessary to change the horizontal position of the detector during the SFA loading stage. Is advantageous in that there is no

As described above, at least two of the measurement points d _{1 to} d ₅ are selected as the axial neutron detector positions.

Now, when the neutron flux is measured at points other than the measurement points d ₁ and d ₅ near the upper and lower ends (spontaneous neutron multiplication method), Equation 1 is established as described above. Distinguishing the reference system as A and the system to evaluate the subcriticality as B,
The following equation is obtained.

[0045] _{φ A = α A S A /} (1-k A) ......... (2) φ B = α B S B / (1-k B) ......... (3) subcriticality, [rho _A = 1 / k _A -1, ρ _B = 1 / k _B -1
Thus, the ratio between Equation 2 and Equation 3 is

[0046]

(Equation 1) Can be represented by

Now, if A is a system of one SFA, ρ
How to measure the _A thus k _A is already developed, it can be measured. φ _B / φ _A is a measurable quantity, (α _B /
α _A ) is obtained by calculation for a given system. S
_A measurement method for the value of _A has already been developed. Target B
Sum of S of a plurality of SFAs included in the system, that is, S
_B is obtained as the sum of the measured values of each body. Therefore, the only unknown quantity in Equation 4 is k _B, and k _B can be determined from the measurement of φ _B / φ _A. One of these methods, two
By sequentially repeating the body, three bodies,..., The k _eff value at each loading stage can be evaluated.

The neutron flux ratio at the measurement points d ₄ and d ₂ is:
It is obtained as the ratio of Equation 4. It is not easy to analytically express the neutron flux ratio between the measurement points d ₁ and d ₅ and the measurement point d _3. Therefore, a calibration curve is prepared by calculation for a given system, and the neutron flux ratio is calculated. The multiplication factor is determined from the measured value.

Further, relative standard deviations are obtained from measured values of neutron fluxes at a plurality of points at positions where a change in the neutron flux distribution is characteristic, such as at measurement points d _{1 to} d ₅ , separated by space. The neutron multiplication factor of the system can be evaluated from the calibration curve obtained by calculation or the like.

In the above embodiment, mainly the BWR
Although the storage of SFA has been described, the present invention can be similarly applied to storage of other furnace-type SFA such as PWR.

[0051]

As described above, according to the present invention, the neutron detector is arranged at the axial center of the system where the change in the axial relative distribution of the neutron flux is small with the change in the effective multiplication factor.
A spontaneous neutron multiplication method for obtaining a substantially average effective multiplication factor of the system from a ratio of a neutron counting rate between a reference loading step and the step, and a neutron counting method for a neutron detector arranged at the central portion in the axial direction. Using a calibration curve prepared in advance from the ratio of the neutron detector arranged at the axial end where the relative distribution of the neutron flux in the axial direction changes with the change in the effective multiplication factor. Since the effective multiplication factor is measured in combination with the subcritical axial spontaneous neutron flux distribution method for evaluating the effective multiplication factor of the system, the non-destructive calculation of the effective multiplication factor using two methods of different quality is required. It is possible to confirm the presence or absence of a systematic difference due to the difference in the measurement method, so that the reliability of the effective multiplication factor obtained in the measurement can be improved. Thus, according to the present invention, the subcriticality at each stage of storing the spent fuel assembly in the system can be evaluated, and the subcriticality of the system can be maintained at a certain value or more. Unnecessary excess margins can be eliminated, more spent fuel assemblies can be stored in confined spaces, and excessive neutrons causing increased cost of storage facilities Absorbents can be eliminated and economical storage is possible.

[Brief description of the drawings]

1A and 1B are graphs showing the principle of a subcritical axial spontaneous neutron flux distribution method, wherein FIG. 1A shows a distribution form when the subcriticality is deep, and FIG. 1B shows a distribution form when the subcriticality is slightly shallow. , (C) respectively show distribution shapes when the subcriticality is shallow.

FIG. 2 is a graph showing distributions of various characteristics in the axial direction of a BWR SFA, where (A) is a burnup, (B) is a void ratio,
(C) shows a fission nuclide, (D) shows k _eff and k _∞ , (E) shows primary neutrons, and (F) shows a neutron flux characteristic distribution.

FIG. 3 is a graph showing the axial neutron flux distribution of the BWR SFA storage system, where (A) is the axial distribution pattern of one BWR SFA, and (B) is the axis when the number of SFA is slightly increased. (B) shows the axial distribution when the number of SFAs is further increased.

FIG. 4 is a graph showing a change in a neutron flux and a neutron flux ratio depending on a neutron multiplication factor, wherein (A) shows a relative change in φ (d ₁ ) or a ratio of φ (d ₅ ) to φ (d ₃ ); (B) is φ
Relative change in the ratio between (d ₄ ) and φ (d ₂ ) and φ
The relative change of (d ₃ ) is shown.

FIG. 5 is a configuration diagram showing an example of a cask fuel basket for storing a BWR SFA.

[Explanation of symbols]

10 outer tube 11 grid member 12 SFA receiving space 13 neutron detector insertion hole _{d 1, d 2, d 3} , d 4, d 5 measurement points

────────────────────────────────────────────────── ─── Continuation of the front page (56) References JP-A 1-250898 (JP, A) JP-A 1-169398 (JP, A) Makoto UEDA et al, “Subcriticality Measurement Methodology Enrollment Neuroscience "J. of Nucl. Sci. and Tech. , 27 [12] p. 1102-1114 (1990) (58) Fields investigated (Int. Cl. ⁷ , DB name) G21C 17/06 G21C 19/40 JICST file (JOIS)

Claims (1)

(57) [Claims] 1. A method for measuring an effective multiplication factor of a loading system of a spent fuel assembly in which the spent fuel assemblies are loaded one by one, wherein a change in a relative distribution of neutron fluxes in the axial direction accompanying a change in the effective multiplication factor is determined. A neutron detector is arranged at the axial center of the small system, and a spontaneous neutron multiplication for obtaining a substantially average effective multiplication factor of the system from a ratio of a neutron counting rate between a reference loading step and the step. The neutron counting rate of the neutron detector arranged at the axial center and the neutron detector arranged at the axial end where the change in the axial relative distribution of the neutron flux with the change of the effective multiplication factor occurs. A method for measuring an effective multiplication factor comprising a combination of a subcritical axial spontaneous neutron flux distribution method for evaluating an effective multiplication factor of the system from a ratio with a neutron counting rate using a calibration curve prepared in advance.