JPS608474B2 - Method for measuring burnup of fuel assembly - Google Patents

Description

DETAILED DESCRIPTION OF THE INVENTION The present invention relates to a method for measuring burnup of a fuel assembly, and more particularly to a method for measuring burnup using neutrons emitted from irradiated fuel.

The burnup of a fuel assembly is a measure of how much thermal energy is extracted from the fuel when the fuel assembly is loaded into a nuclear reactor and burned. The fuel assembly is loaded into a nuclear reactor and irradiated with neutrons to become irradiated fuel, and transuranic elements that emit neutrons are generated in the irradiated fuel. The burnup measurement methods using neutrons emitted from irradiated fuel are listed below.

(11 A method that uses the relationship between burnup and the number of neutrons generated. '21 A method that uses an external neutron source of known intensity to determine the number of neutrons generated from the relationship between burnup and the number of neutrons generated.

(3' A method of determining neutrons emitted from transuranium elements Cm-242 and Cm-244 from the relationship between the ratio of the two and the burnup by distinguishing them from neutrons emitted from other nuclides.

These three conventional measurement methods can be called indirect methods because they indirectly determine the burnup by measuring the amount of transuranium elements produced by absorption of neutrons generated by nuclear fission.

The object of the present invention is different from the above-mentioned indirect method, and is a direct method, in which the combustion of a fuel assembly is determined by measuring the neutron spectrum formed in or around the fuel assembly in the form of neutron ratio. To provide a method to measure the degree of

The principle of the present invention will be explained in detail below using mathematical formulas.

First, for a system in which a neutron moderator such as light water and irradiated fuel exist, the following equation can be obtained by expressing the behavior of neutrons using the two-group diffusion theory used in the theory of thermal neutron reactors. That is, (2 Cape 10 ZsdF + DFBF2)○F; (2f)
F◇F+(shi2f)TOT+S[11(2aT+DTB
T2 )○T=2sdF◇F [21 here S=
Neutron source strength F = Index for fast neutron group T = Index for thermal neutron group Z is ZaT; Absorption cross section of F, T group 23 dF = Deceleration cross section of F group ○ F, DT = Diffusion of F, T group With coefficient BF2, B = F, during buckling of T group F, 07 = F, neutron east of T group (small F is called fast neutron east, ?T is called thermal neutron east) (Li2f) F? F, (2f) TOT = neutron generation rate by neutrons of F, T group (this neutron generation rate is the neutron generation rate when there is a neutron source S in the system and induced fission occurs based on the neutron east formed Both ?F and ?T are composed of a neutron source and induced nuclear fission.

) The following '3' formula can be obtained from the '2-equation ◇F/OT: (2aT+DTBT2 )/2SdF=(
ZaT/2sdF) (10L2&2) ■Kokobi = DT/2aT = Thermal neutron diffusion area L Since the value of T2 is generally considerably or very small compared to 1? P/
The influence of L Swamp T2 on GuT is 4.

Therefore, the value of F/JT obtained separately by theoretical calculation etc. will not be a cause of error. The principle of the present invention is [3]
It is based on the formula.

That is, as nuclear fission progresses (combustion progresses), the amount of fissile nuclides such as U-235 and Pu-239, which have particularly large thermal neutral absorption cross sections, is depleted, so the value of 2aT gradually decreases. This phenomenon will be explained in more detail below. Considering the case where uranium is used as fuel, as combustion progresses, U-235 is depleted and Pu-239, P
Fissile nuclides and fission fragments such as u-240 and Pu-241 are generated. Depletion of U-235 reduces 2aT, but production of Pu-240, Pu-241, fission fragments, etc. conversely increases 28 guns. The production of Pu-240 further reduces the value of 2sdF, albeit slightly. Taking these things into consideration, the value of JF/OT does not matter whether it increases or decreases as combustion progresses, but it does not matter whether the value of JF/OT increases or decreases as combustion progresses. If it progresses more quickly than the production of -239 etc., the reduction effect of 2aT is superior, and? F/? The value of T decreases. The absorption cross section of fission fragments is small except for Xe-135, and Xe-135 has a short half-life, so it disappears 10 days after the reactor is shut down. For this reason, 10 days after the nuclear reactor is shut down, the effect of the production of fission fragments that contributes to an increase in 2aT is small. Pu-24
When 0 is generated, the value of 2sdp decreases, albeit slightly, so whether the value of F/◇T decreases or increases is ultimately determined by the difference in the speed of decrease of SaT and 2sdF. As the enrichment of U-235 increases, U-235 increases as it burns.
Because the depletion of -235 proceeds much faster than the production of Pu-239 etc.? F/? The value of T decreases. In the fuel enriched with plutonium in uranium, the Pu-239 that was enriched from the beginning was generated from the U-2 powder.
Since the depletion of -239 progresses faster with combustion, the value of JF/◇T often decreases, but the value of Pu-24
If the proportion of 0 in plutonium is particularly large, the value of F/JT does not necessarily decrease. In the case of fast breeder reactors whose conversion ratio is said to be greater than 1, highly enriched uranium or highly enriched plutonium is used for the reactor core, and depleted uranium is used for the plunger part. The fact that the conversion ratio is greater than 1, that is, that fissile nuclides are multiplied during combustion, is when considering the reactor core and the plunket as a whole. fissile nuclides are depleted, and Pu-239 increases in the plunket part.Therefore, if OF/OT is measured focusing on the fuel in the reactor core, this value will decrease; On the contrary, it increases.Next is the boiling water reactor (hereinafter abbreviated as BWR).

We will quantitatively explain the case where a standard fuel assembly (uranium fuel with an average concentration of 2.5%) is burned. 1st
The figure is a diagram of a calibration curve showing the relationship between the burnup of the BWR fuel assembly and the neutron east ratio.

To create this calibration curve, it is convenient to calculate the relative change and measure the absolute value using a fuel assembly with known burnup, such as new fuel. In the figure, the value of (ZaT/23dF) (shown as a relative value) decreases almost linearly, but when 20
The decrease in curve 1 becomes gradual at burn-up of ~3 Wd/t or higher. The main causes are Pu-240 and Pu-2
This is because the production of 41 suppresses the decrease in 2aT. The burnup actually achieved is at most about 3 Wd/t for the fuel assembly used in this calculation example, so the curve 1
Correction of the curve 2 can be performed by theoretical calculation. In a fuel assembly designed to obtain a higher burnup than this example, the enrichment of U-235 or the like is set to be greater than 2.5%. Since the curve 2 of the curve shifts to the high burnup side as the enrichment level increases, the above-mentioned curve correction can be performed by theoretical calculation for the same aircraft as in the above example. When the enrichment is about 2 to 2.5%, bend 2 shifts to the low burnup side, but in such a combustion case, the maximum burnup that can be achieved also decreases, so the correction for the bend is still the same. can be done. When attempting to obtain a relatively high burnup by loading fuel with an enrichment of about 1.5% or less into a reactor together with fuel with a relatively high enrichment, the OF/
Since the change in pressure becomes complex and small, the principle of the '3' formula cannot be applied as is.

In such cases, it is best to focus only on neutrons with energy below Hoko V and use the relationship between the neutron spectrum and burnup to determine the burnup from the measured value of the neutron spectrum.As a neutron detector, for example, When a foil of disprosium (Dy) and Pu general 9 is inserted inside a fuel assembly and activated by thermal neutrons, Pu-239 etc. are generated in the fuel in the advanced fuel. , absorbs many neutrons near 0.$V, and the neutron east near 0.$V decreases, but neutrons below ~0.1eV are not absorbed much, so the Pu-23$ tube of the neutron detector and DX The ratio of reaction rate to foil decreases as combustion progresses.
Pu-239 foil has relatively high energy thermal neutrons (
0.1~0. &V), and Dy tends to be sensitive to relatively low-energy thermal neutrons (approximately 0.1 eV or less). Therefore, the ratio of reaction rates between the Pu-23 group and the DX foil can be regarded as the ratio between thermal neutrons with relatively high energy and those with relatively low energy. Since it is difficult to express such a ratio analytically in the range of thermal neutrons like a glue formula, numerical calculation codes (e.g. THERMOS code) based on the already highly developed thermal neutron multigroup transport calculation theory are used.
You can use In the above explanation, P is used as a neutron detector.
Although U-239 and Dy foils were used, P generated in the fuel
Foils such as ln and Rh, which are sensitive to the decrease in neutron energy due to u-240, Am-241, etc., may be used in place of the Pu-23 foil, or may be used in combination. Moreover, it does not have to be limited to the state of foil. Next, consider the case where the BWR fuel assembly is installed in a water pool.

In FIG. 2a, the fuel assembly 3 consists of a large number of nuclear fuel rods 4, and these fuel rods 4 are housed inside a channel box 5. Each fuel twill is occupied by water 6 from a water pool, which acts as a moderator. The outside of the channel box 5 is also filled with pool water 7, and this water acts as a reflective material. Figure 2b is the arrow A- in Figure 2a.
It is a figure which shows the neutron east distribution in the direction of A. In the figure, thermal neutrons East◇T are greatly raised inside the reflector 7, but this is mainly because the fast neutrons leaking from within the fuel assembly 3 are decelerated by the reflector 7 and converted into thermal neutrons. It is. Thermal neutrons generated inside the fuel assembly leak and strike the reflective material 7.
Since only a small amount of thermal neutrons come out in the reflector, the thermal neutrons in the reflector are almost proportional to F in the fast neutrons inside the fuel assembly.
Therefore, it can be seen that the ratio of thermal neutron east within the reflection village to thermal neutron east inside the fuel assembly is approximately proportional to the rigid JF/OT. Next, a method of measuring fast neutrons that are not decelerated in the reflector 7 will be explained.

Since the fast neutron flux in the reflector is proportional to that inside the fuel assembly, the ratio between the fast neutron flux in the reflector and the thermal neutron flux in the fuel assembly is similarly expressed in equation {3}? F/◇Proportional to T. The next method is to place materials that easily guide thermal neutrons (hereinafter referred to as thermal neutron derivatives) such as lead, carbon, heavy water, cavities, etc. in the fuel assembly, and aggregate thermal neutrons inside the fuel assembly. Thermal neutrons are measured outside the assembly by being guided outside, while fast neutrons are guided into the water region that makes up the reflective material and decelerated to become thermal neutrons, and the ratio of thermal neutrons in both regions is [ 3) ◇F/? The burnup is determined from the value that is proportional to the value of T.

Fast neutron east may be measured directly outside the assembly, or in this case fast neutron east may be measured within the thermal neutron derivative. FIG. 3 shows an embodiment in which a neutron detector is installed inside or around at least one of the fuel assemblies 3, and FIG. It is a diagram in which a rod-shaped neutron detection device 8 equipped with a neutron detector such as gold foil is attached. FIG.
are installed in the reflective material 7 around the fuel assembly 3.
FIG.
is installed inside the reflective material 7. Figures 3a, b and c are representative examples of the arrangement of neutron detection devices in the method described above. FIG. 4 is a diagram showing the rod-shaped neutron detection device 8 installed in the core of a BWR nuclear reactor that is stopped.

During reactor shutdown, the cross-shaped control rods 10 are fully inserted into the reactor core, and one control rod controls four fuel assemblies surrounded by broken lines 11. Inside the body, the detection device 8b is installed outside the fuel assembly.Since there is no neutron leakage from the reactor core, neutrons from 20 to 5 needles can be obtained compared to the position 8 in FIG. 3a. The method according to the invention can be applied to burnup measurements of fuel assemblies of nuclear reactors of the same type.

If the neutron emission rate from the irradiated fuel is low, a neutron source may be artificially attached from the outside. Furthermore, the method of the present invention can be applied not only to fuel assemblies but also to measuring the burn-up of each nuclear fuel ray that constitutes the assembly.

[Brief explanation of the drawing]

Figure 1 shows how the ratio of the thermal neutron absorption cross section to the fast neutron moderation cross section changes depending on the burnup of the fuel in the fuel assembly of a BWR reactor, and Figure 2 a shows the fuel assembly of a BWR reactor. Cross-sectional view of the body "Figure b is a diagram showing the neutron east distribution curves of fast neutrons and thermal neutrons on the straight line A-A.
Figures a to c are cross-sectional views of a neutron detection device installed inside or around a fuel assembly, and FIG. 4 is a view of the neutron detection device installed inside the reactor core while it is stopped. 1...Burnup curve, 2...Bend~ 3...Fuel assembly, Tortoise...Nuclear fuel rod, 5...0 channel box, 6...Moderator~? ...Reflective material, 8,8a...
Neutron detection device, 9... thermal neutron derivative, ]0...
...control. Gay figure 2 figure figure 3 figure figure Chimura

Claims (1)

[Claims] 1. In a system composed of a fuel assembly consisting of a large number of nuclear fuel rods and a medium, two types of neutron fluxes having different energetic properties are detected by a neutron detector inside or outside of the fuel assembly. measure,
A method of determining the neutron flux ratio from these neutron fluxes, and using the neutron flux ratio to measure the burnup of the fuel assembly from a calibration curve showing the relationship between the burnup and the neutron flux ratio.