Method for Managing Radioactive Waste

Abstract

To provide a radioactive waste management method which allows for reducing the wet storage period than before.SOLUTION: Radioactive waste is moved from wet storage to dry intermediate storage once the decay heat of the radioactive waste falls below an intermediate storage container design value.SELECTED DRAWING: Figure 1

Description

[Technical field]

[0001] The present invention relates to a method for managing radioactive waste. [Background technology]

[0002] As is well known, dry interim storage is expected to be a method of storing radioactive waste such as spent nuclear fuel. This dry interim storage has a lower heat removal capacity than a spent fuel pit (SFP: Spent Fuel Pit), so the decay heat of radioactive waste must be sufficiently reduced before it can be transferred to dry interim storage. Currently, a uniform period is set for the burnup limit value for fuel assembly removal, and only spent nuclear fuel that exceeds that period is stored in dry interim storage.

[0003] As an example, for spent nuclear fuel with a burnup limit of 48 GWd / t, a period of 15 years or the like has been set. At present, for spent nuclear fuel with a burnup limit of 55 GWd / t, the design of an intermediate storage container and the application for approval have not yet been made. Details of such dry intermediate storage of radioactive waste are described in, for example, the following non-patent document 1. [Prior art documents] [Non-patent literature]

[0004] [Non-Patent Document 1] Tomofumi Yamamoto, Hiroki Tamaki, Hideo Hode, Toshihiro Matsuoka, Yoshiharu Kamiwaki, and Kiminobu Hojo, "Development of Spent Fuel Transport and Storage Cask," Mitsubishi Heavy Industries Technical Review Vol. 43 No. 4: 2006 Summary of the Invention [Problem to be solved by the invention]

[0005] Incidentally, the thermal design of the dry storage container is carried out using a conservatively evaluated value of the decay heat of the spent nuclear fuel after the cooling period. Therefore, if the decay heat used in the design of the dry storage container is lower than that of each assembly, the spent nuclear fuel can physically be stored in the interim storage container. As a result, it becomes possible to end the wet storage of the spent nuclear fuel earlier.

[0006] For example, if all the fuel assemblies can be quickly transferred to dry storage at a nuclear power plant where decommissioning measures have been decided, the decommissioning measures for the building including the spent fuel pool can be started earlier, resulting in a large cost compression effect. Also, if the decay heat can be accurately evaluated to be below a certain value, denser storage becomes possible. Furthermore, when it becomes difficult to obtain helium gas used for heat removal in dry storage, it can be confirmed that there is no problem even if a gas with poor heat transfer is enclosed.

[0007] However, at present, there is no accurate evaluation method for the decay heat of radioactive waste that is relatively easy to introduce and has guaranteed reliability. Therefore, the decay heat of the spent nuclear fuel has to be conservatively evaluated, resulting in the problem of prolonging the wet storage at nuclear power plants.

[0008] The present invention has been made in view of the above circumstances, and an object thereof is to provide a method for managing radioactive waste that can shorten the wet storage period compared with the conventional method.

Means for Solving the Problems

[0009] To achieve the above object, in the present invention, as a first solution means for the method of managing radioactive waste, when the decay heat of the radioactive waste is lower than the design value of the interim storage container, the radioactive waste is transferred from wet storage to dry interim storage.

[0010] In the present invention, as a second solution means related to the method for managing radioactive waste, in the above first solution means, the decay heat is evaluated based on the radioactivity of a plurality of nuclides contained in the radioactive waste, and such a means is adopted.

[0011] In the present invention, as a third solution means related to the method for managing radioactive waste, in the above second solution means, the radioactive waste is spent nuclear fuel stored in a spent fuel pool, and a radiation detector is immersed in the shielding water of the spent fuel pool to detect the radiation emitted from the spent nuclear fuel, and the radioactivity is evaluated based on the radiation, and such a means is adopted.

[0012] In the present invention, as a fourth solution means related to the method for managing radioactive waste, in the above second solution means, the radioactive waste is spent nuclear fuel stored in a spent fuel pool, and the radiation emitted from the spent nuclear fuel is detected through a collimator embedded in the wall of the spent fuel pool, and the radioactivity is evaluated based on the radiation, and such a means is adopted.

[0013] In the present invention, as a fifth solution means related to the method for managing radioactive waste, in any one of the above second to fourth solution means, the plurality of nuclides include 90 Sr and 90 Y is at least included, and the bremsstrahlung emitted from the radioactive waste together with γ-rays is detected by a radiation detector, so as to evaluate the radioactivity of 90 Sr and the radioactivity of 90 Y, and such a means is adopted.

[0014] In the present invention, as a sixth solution means related to the method for managing radioactive waste, in any one of the above second to fifth solution means, the plurality of nuclides are 134 Cs, 137 Cs, 137m Ba, 154 Eu, 90 Sr, 90 Y, 144 Ce, 144 Pr, 106 Ru and106 Adopt the means of being Rh.

Advantages of the Invention

[0015] According to the present invention, it is possible to provide a method for managing radioactive waste capable of shortening the wet storage period compared with the prior art.

Brief Description of the Drawings

[0016]

Figure 1

Figure 2

Figure 3

Figure 4

Figure 5

Figure 6

Modes for Carrying Out the Invention

[0017] Hereinafter, an embodiment of the present invention will be described with reference to the drawings. The method for managing radioactive waste according to this embodiment targets spent nuclear fuel X well-known as radioactive waste. The spent nuclear fuel X is initially wet-stored (step S1) as shown in FIG. 1 after being used for power generation at a nuclear power plant or the like. That is, this spent nuclear fuel X is stored in a spent fuel pool A provided in a nuclear power plant or the like and immersed in shielding water B as shown in FIGS. 2(a) and (b).

[0018] Subsequently, in this management method, the decay heat of the spent nuclear fuel X stored in a wet state as described above is evaluated (step S2). This decay heat is evaluated by a method specific to the present embodiment described later, and is evaluated based on the radioactivity of a plurality of main nuclides contained in the spent nuclear fuel X.

[0019] The spent nuclear fuel X is an assembly of fuel rods and is irradiated nuclear fuel that emits γ-rays (gamma rays) to the surroundings along with β decay. Further, this spent nuclear fuel X contains high-speed electrons generated by β decay, and when the high-speed electrons are decelerated by nuclides, bremsstrahlung (a type of electromagnetic wave) is emitted to the surroundings. That is, the spent nuclear fuel X emits γ-rays and bremsstrahlung as radiation to the surroundings along with β decay.

[0020] Although details will be described later, the method for evaluating decay heat in the present embodiment detects γ-rays and bremsstrahlung of the spent nuclear fuel X, evaluates the radiation of a plurality of main nuclides contained in the spent nuclear fuel X based on the radiation, and non-destructively evaluates the decay heat of the spent nuclear fuel X based on the radiation of the plurality of nuclides.

[0021] The method for managing radioactive waste according to the present embodiment determines the timing of transition between wet storage and dry interim storage of the spent nuclear fuel X more appropriately than in the past by adopting the decay heat of the spent nuclear fuel X obtained by such a method for evaluating decay heat.

[0022] Here, FIGS. 2(a) and (b) are schematic diagrams showing a method for detecting radiation in the method for evaluating decay heat. As shown in FIGS. 2(a) and (b), the spent nuclear fuel X is stored in a spent fuel pool A as an assembly of fuel rods. In the method for detecting radiation in the present embodiment, radiation (γ-rays and bremsstrahlung) emitted from the spent nuclear fuel X immersed in the shielding water B of the spent fuel pool A is detected by a radiation detector C.

[0023] In this method for detecting radiation, with respect to the spent nuclear fuel X immersed in the shielding water B as described above, for example, as shown in Fig. 2(a), a radiation detector C is arranged near the spent nuclear fuel X in a state where it is housed in a watertight container D. Then, the radiation detector C detects the radiation (γ-rays and bremsstrahlung) emitted from the spent nuclear fuel X. That is, in the radiation detection method of Fig. 2(a), the radiation detector C is immersed in the shielding water B while being protected from the shielding water B (cooling water) in the spent fuel pool A by the watertight container D.

[0024] The watertight container D is composed of a material that hermetically houses the radiation detector C with respect to the shielding water B and allows radiation (γ-rays and bremsstrahlung) to pass through without attenuation. In such a radiation detection method, the radiation detector C detects the radiation (γ-rays and bremsstrahlung) emitted from the spent nuclear fuel X through the shielding water B and the watertight container D.

[0025] On the other hand, in the radiation detection method shown in Fig. 2(b), instead of being immersed in the shielding water B as described above, the radiation detector C detects the radiation (γ-rays and bremsstrahlung) through a collimator E embedded in the wall of the spent fuel pool A. The collimator E is embedded so as to penetrate the wall of the spent fuel pool A as shown in the figure, and the radiation (γ-rays and bremsstrahlung) emitted from the spent nuclear fuel X is made to enter the sensitive surface of the radiation detector C.

[0026] That is, in the radiation detection method of Fig. 2(b), since the radiation detector C is not immersed in the shielding water B, a watertight container D is not required. That is, the radiation detector C in this radiation detection method detects the radiation (γ-rays and bremsstrahlung) emitted from the spent nuclear fuel X through the shielding water B and the collimator E.

[0027] Subsequently, in the radioactive waste management method according to this embodiment, the decay heat evaluated in step S2 is compared with the intermediate storage container design value (step S3). That is, in this step S3, the decay heat of the spent nuclear fuel X is compared with the intermediate storage container design value, which is the evaluation threshold for changing the management method of the spent nuclear fuel X. The intermediate storage container design value is the value of the decay heat used in the design of an intermediate storage container (a container called a "cask") used when dry-intermediate storing the spent nuclear fuel X.

[0028] When the decay heat evaluated in step S2 is smaller than the intermediate storage container design value, the intermediate storage container can stably intermediate store the spent nuclear fuel X as intended in the design. On the other hand, when the decay heat evaluated in step S2 is equal to or greater than the intermediate storage container design value, the intermediate storage container cannot intermediate store the spent nuclear fuel X in a stable state.

[0029] In the radioactive waste management method according to this embodiment, when the determination in step S3 is "Yes", that is, when the decay heat evaluated in step S2 is equal to or greater than the intermediate storage container design value, the wet storage of the spent nuclear fuel X in step S1 is continued (step S4). On the other hand, when the determination in step S3 is "No", that is, when the decay heat evaluated in step S2 is smaller than the intermediate storage container design value, the management method of the spent nuclear fuel X is shifted from wet storage in step S1 to dry-intermediate storage using the intermediate storage container (step S5).

[0030] That is, in the radioactive waste management method according to this embodiment, when the decay heat of the spent nuclear fuel X (radioactive waste) falls below the intermediate storage container design value, the spent nuclear fuel X (radioactive waste) is shifted from wet storage to dry-intermediate storage.

[0031] Next, with reference to FIGS. 3 to 6, the method for evaluating the decay heat in this embodiment will be described. Note that FIG. 3 is a flowchart showing a method for evaluating radioactivity that is initially implemented in the method for evaluating the decay heat of the spent nuclear fuel X. FIG. 5 is a flowchart showing the procedure for evaluating the decay heat based on such a method for evaluating radioactivity.

[0032] First, referring to FIGS. 3 and 4, the method for evaluating radioactivity in this embodiment will be described. In this method for evaluating radioactivity, by using a measuring device including the above-described radiation detector C, the spectral distribution (count rate distribution) of γ-rays in all energy ranges of γ-rays and bremsstrahlung emitted from nuclides contained in the spent nuclear fuel X is measured (step Sa1). Note that since the above measuring device is well-known in the technical field of radiation, a detailed description thereof will be omitted.

[0033] This spectral distribution shows the count rate of γ-rays in the energy range of 0 to 4 MeV, as shown in FIG. 4(a). From this measured spectral distribution, 106 γ-rays emitted from Rh and 144 Pr are counted (step Sa2). For example, 106 the energies of γ-rays emitted from Rh (energy imparted to the detector) are 2.366 MeV and 2.406 MeV, 144 and the energy of γ-rays emitted from Pr is typically 2.186 MeV.

[0034] Note that 106 Rh and 144 Pr also emit γ-rays having energies (energies imparted to the detector) different from the above-described 2.366 MeV, 2.406 MeV, and 2.186 MeV. Considering this, in step Sa2, γ-rays having the above different energies may be counted.

[0035] In this method for evaluating radioactivity, subsequently, the response function and detection efficiency in the measurement of β-rays (beta rays) and γ-rays are calculated (step Sa3). In this step Sa3, first, in the spent nuclear fuel X 90 Y, 106 Rh, and 144Evaluate the spatial distribution and energy spectrum related to the generation of beta rays and gamma rays emitted from Pr. For the above energy spectrum, data from the well-known "ICRU-56" (Explanation of External Beta Ray Dosimetry for Radiation Protection) and data from the equally well-known "Table of Isotopes" can be used.

[0036] Also, for the above spatial distribution, the following evaluation methods can be adopted. That is, assume that the space is uniform in the spent nuclear fuel X, and also assume that the spatial distribution in the spent nuclear fuel X can be represented by an analytical function (for example, a distribution proportional to a cosine function in the vertical direction, etc.). Furthermore, assume that the measured values of the burnup distribution and gamma ray dose distribution are similar to the spatial distribution of radioactivity.

[0037] Furthermore, as another method, under such assumptions, actually calculate the spatial distribution related to the generation of the above beta rays and gamma rays. For this calculation, the collective burnup calculation function of the SCALE code system developed by the Oak Ridge National Laboratory in the United States and calculation codes such as SWAT and MVP-BURN developed by the Japan Atomic Energy Agency (JAEA) can be used. These calculations can be made possible by inputting the time change of the thermal output of the spent nuclear fuel X, which is an assembly of fuel rods, the type of the assembly, and the density of the cooling water.

[0038] Response function R for electrons e (r, E e → h) and the response function R for gamma rays γ (r, E γ → h) represent the probability that an electron beam or gamma ray with energy E generated at position r is detected as energy h by the radiation detector C. Also, the detection efficiency ε(h) is expressed by the following equations (1) and (2) when the absolute value of the gamma ray generation spectrum is S γ (r, E γ ), and the absolute value of the electron beam generation spectrum is S e (r, E e ).

[0039]

Equation

[0040] Before measurement, the absolute value S γ (r, E γ ) and S e (r, E e ) are unknown. However, if the relative value can be evaluated or assumed, the detection efficiency ε(h) can be calculated. For the calculation of this detection efficiency ε(h), codes such as the MCNP code developed by the Los Alamos National Laboratory in the United States, the EGS code developed by the Japan Atomic Energy Agency and the High Energy Accelerator Research Organization, and the PHITS code by JAEA incorporating the functions of the EGS code can be used. Any of these codes inputs the nuclear fuel to be measured, the position, composition, and density of the HP-Ge detector and the atmosphere, and the absolute value S γ (r, E γ ) and S e (r, E e ) or the absolute value S γ (r, E γ ) and S e (r, E e ). By inputting the relative value of S

[0041] In this method for evaluating radioactivity, subsequently 106 the number of generated γ-rays related to Rh and 144 Pr is calculated (step Sa4). In this step Sa4, first, the peak count rate C(h) of Rh and 106 Rh and 144 Pr obtained in step Sa2 is divided by the detection efficiency ε(h) obtained in step Sa3 to 106 obtain the number of generated γ-rays at the energy specific to Rh and 144 Pr. Subsequently, the absolute value S 106 of the γ-ray generation spectrum related to Rh and 144 Pr γ (r, E γ ) is calculated.

[0042] Here, the above energy h indicates the peak of the γ-ray energy. For example 106For the γ-rays emitted from Rh, h = 2.3660 MeV, 2.406 MeV, 144 For the γ-rays emitted from Pr, h = 2.186 MeV. Also, since the counting rate C(h) is given by the following equation (3), 106 Rh and 144 The number of γ-rays generated for Pr is given by the following equation (4), which is obtained by dividing the above counting rate C(h) by the detection efficiency.

[0043]

Number

[0044] In this method for evaluating radioactivity, subsequently 106 Rh and 144 The absolute value S e (r, E e ) of the electron beam generation spectrum for Pr is calculated (step Sa5). In this step Sa5, based on the relationship between the absolute value S γ (r, E γ ) of the γ-ray generation spectrum obtained in step Sa4 and the absolute value S e (r, E e ) of the electron beam generation spectrum, which can be understood from the relevant data such as those in the above-mentioned "Table of Isotopes" and "ICRU-56", the absolute value S e (r, E e ) of the electron beam generation spectrum is calculated.

[0045] In this method for evaluating radioactivity, subsequently, the counting rates of Rh and Pr in the energy range of h = 1.6 to 1.7 MeV are calculated (step Sa6). In this calculation of the counting rate, the above-mentioned MCNP code, EGS code, and PHITS code are used, and the system information, the absolute value S 106 Rh and 144 (r, E γ ) of the γ-ray generation spectrum obtained in step Sa4, and the absolute value S γ ) of the electron beam generation spectrum obtained in step Sa5 are input to perform the calculation. e (r, E e )

[0046] More specifically, the absolute value S of the gamma-ray generation spectrum γ (r, E γ ) and the absolute value S of the electron-ray generation spectrum obtained in step Sa5 e (r, E e ) are used as inputs, and the detection efficiency ε of gamma rays γ (h) and the detection efficiency ε of electron rays e (h) are calculated based on the following formulas (5) and (6), and the integrated value of the absolute value S of the gamma-ray generation spectrum γ (r, E γ ) and the product of the absolute value S of the electron-ray generation spectrum e (r, E e ) are added together. In formulas (5) and (6), the energy range of h is 1.6 to 1.7 MeV.

[0047]

Equation

[0048] That is, by calculating the following formula (7) based on the above formulas (5) and (6), the counting rates of Rh and 106 Pr in the energy range of h from 1.6 to 1.7 MeV can be obtained. Here, the integrated value of the absolute value S of the above gamma-ray generation spectrum 144 (r, E γ (r, E γ ) and the integrated value of the absolute value S of the electron-ray generation spectrum e (r, E e ) include all the gamma rays and bremsstrahlung radiated from Rh and 106 Pr. 144 Pr.

[0049]

Equation

[0050] In this method for evaluating radioactivity, subsequently 90Calculate the counting rate of Y (step Sa7). In this step Sa7, within the range where the energy h obtained in step Sa2 is 1.6 - 1.7 MeV 106 Rh and 144 subtract the counting rate of Rh and 106 Pr obtained in step Sa6 from the counting rate of Rh and 144 Pr to 90 calculate the counting rate of Y.

[0051] Here, the calculation process of the counting rate of Y in step Sa7 90 includes correction processing to further improve the evaluation accuracy (calculation accuracy) of the radioactivity of Y in the next step Sa8. Therefore, 90 when it is acceptable to sacrifice the evaluation accuracy of the radioactivity of Y to a certain extent, without subtracting the counting rate of Rh and 90 Pr obtained in step Sa6 from the counting rate of Rh and 106 Pr, that is, the counting rate of Rh and 144 Pr measured by the above-mentioned measuring device, the counting rate of Rh and 106 Pr measured by the measuring device can be directly used as 144 the counting rate of Y without subtracting the counting rate of Rh and 106 Pr obtained in step Sa6. 144 Pr measured by the measuring device can be directly used as 106 the counting rate of Y. 144 Pr measured by the measuring device can be directly used as 90 the counting rate of Y.

[0052] In this method for evaluating radioactivity, subsequently 90 calculate the Y radioactivity (step Sa8). In this step Sa8, divide the counting rate of Y obtained in step Sa7 by the detection efficiency ε(h) evaluated in step Sa3 to 90 calculate the β-ray generation rate related to Y, and then divide the β-ray generation rate related to Y by the (electron-ray generation rate per decay) described in the above-mentioned "Table of Isotopes" etc. to 90 calculate the Y radioactivity. 90 divide the β-ray generation rate related to Y by the (electron-ray generation rate per decay) described in the above-mentioned "Table of Isotopes" etc. to 90 calculate the Y radioactivity.

[0053] Furthermore, in this method for evaluating radioactivity, 90Calculate the Sr radioactivity (step Sa9). In this step Sa9, by using the condition of secular equilibrium described above, 90 the Sr radioactivity 90 is determined to be the same amount as the Y radioactivity.

[0054] The method for evaluating radioactivity in this embodiment is realized by a series of processes consisting of the above-described steps Sa1 to Sa9. According to such a method for evaluating radioactivity, for the spent nuclear fuel X (radioactive waste), at least 90 the Sr radioactivity and 90 the Y radioactivity can be non-destructively evaluated.

[0055] Further, Fig. 4(a) shows the calculation results of the wave height spectrum of the braking line based on γ-rays and β decay for the spent nuclear fuel X with a burnup of 52 GWd / t and a cooling period of 13 years. From this calculation result, it can be seen that in the region where the energy imparted to the detector (radiation detector C) is 1.6 to 1.7 MeV, most of the counting rate 90 is occupied by the radioactivity due to Y.

[0056] Further, Fig. 4(b) shows the ratio of the 90 Y braking line in the counting rate in the region where the energy imparted to the detector (radiation detector C) is 1.6 to 1.7 MeV. This Fig. 4(b) shows that the ratio of the 90 Y braking line in the region where the energy imparted is 1.6 to 1.7 MeV becomes about 94% regardless of the burnup when the cooling exceeds 13 years. Therefore, for the spent nuclear fuel X with such a cooling period, simply obtaining the count where the energy imparted is 1.6 to 1.7 MeV and dividing by the detection efficiency, the 90 Y radioactivity and 90 the Sr radioactivity can be obtained within an error of 6%.

[0057] Therefore, taking the spent nuclear fuel X with a cooling period of 13 years or more as the evaluation target, and also 90 the Sr radioactivity and 90When allowing a 6% measurement error regarding γ radioactivity, the correction process in step Sa7 may be omitted. That is, without performing the correction process on the counting rates of 106 Rh and 144 Pr, the γ radioactivity may be calculated based only on the counting rates of 106 Rh and 144 Pr measured by the measuring device in step Sa8. 90 Y radioactivity may be calculated.

[0058] Furthermore, as a result of investigating γ-rays that cause interference in the imparted energy region (pulse height region) of 1.6 to 1.7 MeV, for spent nuclear fuel X of 2 cycles = 26 GWd / t or more, it is found that it is 144 Ce → 144 Pr and 106 Ru → 106 Rh, and when adding up the components of the bremsstrahlung lines of this two-component and 90 Sr → 90 Y, it is found that it accounts for 98% of the counts of 1.6 to 1.7 MeV.

[0059] 144 Ce → 144 Pr and 106 Ru → 106 Rh emit γ-rays with a peak structure, so it is easy to measure and quantify this radioactivity. From the counts of 1.6 to 1.7 MeV, based on the measurement, 144 Ce → 144 Pr and 106 Ru → 106 Rh components are subtracted, and further divided by the detection efficiency, 90 Sr → 90 Y radioactivity can be measured and evaluated from a stage earlier than when the cooling period is 13 years. For example, in Fig. 4(b), it shows that the radioactivity of 90 Sr → 90 Y can be evaluated from the first year of the cooling period.

[0060] Next, with reference to FIGS. 5 and 6, an evaluation method for decay heat in this embodiment will be described. This evaluation method for decay heat is based on the above-described evaluation method for radioactivity, and uses the radioactivity of 90 Sr and 90 Y, etc., to evaluate the decay heat of spent nuclear fuel X (radioactive waste).

[0061] In this evaluation method for decay heat, first 106 the radioactivity of Rh and 144 the radioactivity of Pr are calculated (step Sa10). That is, in this step Sa10, the 90 radioactivity of Sr and 90 the radioactivity of Y obtained by the evaluation method of 106 Rh and 144 Pr are used. The number of γ-rays generated per decay of Rh and Pr is divided by the number of γ-rays generated per decay described in the "Table of Isotopes" to obtain 106 the radioactivity of Rh and 144 the radioactivity of Pr.

[0062] In this evaluation method for decay heat, subsequently 106 the radioactivity of Ru and 144 the radioactivity of Ce are calculated (step Sa11). That is, in this step Sa11, 106 Rh and 144 Pr are each in 106 Ru and 144 a condition of secular equilibrium with Ce. Therefore, the radioactivity of 106 Rh and 144 Pr obtained in step Sa10 is used as the same amount to obtain 106 the radioactivity of Ru and 144 the radioactivity of Ce.

[0063] In this evaluation method for decay heat, by using the measuring device in step Sa1 described above, a spectrum including γ-rays emitted from 134 Cs, 137m Ba, and 154 Eu of spent nuclear fuel X is obtained, and from this, 134 Cs, 137m Ba, and 154Count the γ-rays emitted from Eu (step Sa12).

[0064] The peak energies of the γ-rays are, for example 134 0.605 MeV and 0.796 MeV for Cs, 137m 0.662 MeV for Ba, and 154 1.274 MeV for Eu, but multiple γ-rays are generated per decay 134 Cs and 154 Eu are not limited to this. Due to such circumstances, in step Sa12, 134 Cs, 137m Ba and 154 Count the peak spectra of the γ-rays related to Eu.

[0065] In this method for evaluating the decay heat, subsequently calculate the response function and the detection efficiency (step Sa13). That is, in this step Sa13, by using the same method as step Sa3 described above, obtain the response function and the detection efficiency in the measurement of γ-rays.

[0066] In this method for evaluating the decay heat, subsequently 134 Cs, 137m Ba and 154 Calculate the number of γ-rays emitted from Eu (step Sa14). That is, in this step Sa14, by dividing the peak count obtained in step Sa12 by the detection efficiency obtained in step Sa13, 134 Cs, 137m Ba and 154 Obtain the number of γ-rays emitted from Eu.

[0067] In this method for evaluating the decay heat, subsequently 134 Cs, 137m Ba and 154 Calculate the radioactivity of Eu (step Sa15). That is, in this step Sa15, by dividing the number of γ-rays generated obtained in step Sa14 by the number of γ-rays generated per decay described in the "Table of Isotopes", 134 Cs, 137m Ba and154 Obtain the radioactivity of Eu.

[0068] In this decay heat evaluation method, subsequently 137 Calculate the radioactivity of Cs (step Sa16). 137m For Ba and 137 Cs, the condition of secular equilibrium holds, so 137 The radioactivity of Cs is the same as that of 137m Ba obtained in step Sa15.

[0069] In this decay heat evaluation method, finally calculate the decay heat of the spent nuclear fuel X (radioactive waste) (step Sa17). From the data described in "JENDL / DDF - 2015", the calculated value Q β of the total β - ray energy per β - decay heat and the calculated value Q γ of the total γ - ray energy are read out, and radioactivity × (Q β +Q γ ) is 134 summed for Cs, 137 Cs, 137m Ba, 154 Eu, 90 Sr, 90 Y, 144 Ce, 144 Pr, 106 Ru and 106 Rh. This sum represents the heat generation amount from nuclides that account for 75 - 90% or more of the decay heat of the spent nuclear fuel X.

[0070] According to such a decay heat evaluation method, 134 Cs, 137 Cs, 137m Ba, 154 Eu, 90 Sr, 90 Y, 144 Ce, 144 Pr, 106 Ru and 106 Rh, the decay heat regarding the spent nuclear fuel X is calculated based on the sum of the energies obtained from the respective radioactivities, so it can be obtained non - destructively.

[0071] Here, for typical light water reactor fuels, the radioactivity was calculated using the ORIGEN-ARP module of the SCALE-6.1 code developed by Oak Ridge National Laboratory in the United States when burned at burnup levels of 13 GWd / t, 26 GWd / t, 39 GWd / t, and 52 GWd / t for 406 days, 2×406 days, 3×406 days, and 4×406 days, respectively, and then cooled (removed from the core and simply stored).

[0072] Figure 6(a) shows the proportion of the decay heat of the whole obtained based on this calculation accounted for by 90 Sr → 90 Y. As shown in Figure 6(a), the proportion of the decay heat of these two nuclides is the largest during the 1 to 20 years after removal, indicating that the radioactivity evaluation of these two nuclides is essential for decay heat evaluation.

[0073] Figure 6(b) shows the proportion of the decay heat of measurable nuclides in the total decay heat. In this Figure 6(b), for the total value of the decay heat, the proportion of the decay heat that can be evaluated from the γ-ray counting rates of 134 Cs, 137 Cs, and 154 Eu measured by SKB in Sweden is shown by a line, and the decay heat that can be evaluated from the radioactivity of 244 Cm by neutron measurement and the radioactivity of 90 Sr → 90 Y, 106 Ru → 106 Rh, 144 Ce → 144 Pr evaluated by the radioactivity evaluation method in this embodiment is shown by dots.

[0074] According to this Figure 6(b), for the previous three nuclides, the proportion of the measurable decay heat was about 40% even after 5 years of cooling, but it becomes about 90% after 5 years of cooling. This indicates that the proportion of the decay heat that must be supplemented by numerical calculation is significantly reduced. Therefore, by evaluating the decay heat using the radioactivity of each nuclide obtained by the radioactivity evaluation method in this embodiment, the reliability of the decay heat evaluation by radiation measurement can be significantly improved.

[0075] As described above, in the radioactive waste management method according to the present embodiment, the decay heat of the spent nuclear fuel X obtained by the decay heat evaluation method peculiar to the present embodiment is compared with the design value of the interim storage container. When the decay heat of the spent nuclear fuel X falls below the design value of the interim storage container, the management method of the spent nuclear fuel X is shifted from wet storage to dry interim storage.

[0076] According to such a radioactive waste management method, since the reliability of the decay heat obtained by the decay heat evaluation method in the present embodiment is higher than that of the prior art, the timing of the shift from wet storage of the spent nuclear fuel X to dry interim storage can be appropriately evaluated. Therefore, it is possible to shorten the wet storage period as compared with the conservative evaluation in the prior art.

[0077] In particular, according to the present embodiment, in the spent nuclear fuel X, a plurality of main nuclides 134 Cs, 137 Cs, 137m Ba, 154 Eu, 90 Sr, 90 Y, 144 Ce, 144 Pr, 106 Ru and 106 Rh, the decay heat evaluated based on the radioactivity is used to evaluate the shift of the spent nuclear fuel X from wet storage to dry interim storage. Therefore, it is possible to accurately shorten the wet storage period.

[0078] Note that the present invention is not limited to the above embodiment, and for example, the following modification examples can be considered. (1) In the above embodiment, the spent nuclear fuel X (radioactive waste) stored in the spent fuel pool A is the management target, but the present invention is not limited thereto. That is, the present invention targets various radioactive wastes typified by the spent nuclear fuel X.

[0079] (2) In the above-described embodiment, as shown in FIGS. 2(a) and 2(b), the radiation detector C was immersed in the shielding water B of the spent fuel pool A, or the radiation (γ-rays and bremsstrahlung) emitted from the spent nuclear fuel X was detected through the collimator E embedded in the wall of the spent fuel pool A. However, the present invention is not limited to this.

[0080] (3) In the above-described embodiment, 134 Cs, 137 Cs, 137m Ba, 154 Eu, 90 Sr, 90 Y, 144 Ce, 144 Pr, 106 Ru and 106 Rh were used to determine the transition from wet storage to dry interim storage of the spent nuclear fuel X based on the decay heat evaluated based on the radioactivity of Rh. However, the present invention is not limited to this. That is, the main nuclides governing the decay heat vary depending on the type of radioactive waste. Therefore, the nuclides for evaluating the decay heat are 134 Cs, 137 Cs, 137m Ba, 154 Eu, 90 Sr, 90 Y, 144 Ce, 144 Pr, 106 Ru and 106 Rh are not limited to.

[0081] (4) In the above-described embodiment, regarding the early transition technology from wet storage to dry interim storage of spent nuclear fuel based on FIG. 1, in "S2: Evaluation of the decay heat of spent nuclear fuel", a method using radiation measurement from the target spent fuel assembly was described with reference to FIGS. 3 and 5. However, the present invention is not limited to this.

[0082] For example, as a conventional method, there is a technique of using a calorimeter. The calorimeter places a container that completely covers the spent nuclear fuel in the spent fuel pit, passes water from the lower part to the upper part of the container, and waits until the water temperatures at the inlet and outlet of the container become stable. Then, the decay heat is evaluated by taking the product of the difference in water temperatures at the inlet and outlet, the amount of water flowing inside the container, and the specific heat of water. Although this technique has issues such as the use of a large-scale device, the calorimeter, and the need for time until the water temperature stabilizes, making it difficult to introduce easily, the calorimeter is an established technology.

[0083] Such a calorimeter is publicly known as described in the document "Measurements of decay heat in spent nuclear fuel at the Swedish interim storage facility, Clab", SKB Rapport R-05-62, 2006. This document can be viewed at the URL "https: / / www.skb.se / publikation / 1472024 / R-05-62.pdf".

[0084] Also, regarding the technology for the early transition from wet storage to dry interim storage of spent nuclear fuel based on Figure 1, in "S2: Evaluation of decay heat of spent nuclear fuel", there is another method that does not involve physical measurements on the spent fuel.

[0085] When using nuclear fuel in a nuclear reactor, the time change in the output of each fuel in the reactor is measured monthly by the core management system, and the density of the cooling water corresponding thereto is calculated. Given the type of the fuel and the time change in the output of each fuel as inputs, the radioactivity of each nuclide in the nuclear fuel can be calculated using the SCALE code system created at the Oak Ridge National Laboratory in the United States or similar codes.

[0086] The decay heat can also be obtained by multiplying the radioactivity of each nuclide calculated in this way by the heat generation amount per decay and summing up the products for all nuclides. Note that to use this evaluation method, it is necessary to ensure that the core management system is correct and that there is no misidentification of the spent nuclear fuel.

Description of Symbols

[0087] A Spent Fuel Pool B Shielding Water C Radiation Detector D Watertight Container E Collimator X Spent Nuclear Fuel (Radioactive Waste)

Claims

1. When the decay heat of the radioactive waste falls below the design value of the interim storage container, the radioactive waste is transferred from wet storage to dry interim storage, The decay heat is evaluated based on the radioactivity of a plurality of nuclides contained in the radioactive waste, The plurality of nuclides at least include 90Sr and 90Y, The radioactivity of 90Sr and the radioactivity of 90Y are evaluated by detecting the bremsstrahlung emitted together with γ-rays from the radioactive waste with a radiation detector, A method for managing radioactive waste, characterized in that.

2. The radioactive waste is spent nuclear fuel stored in a spent fuel pool, The method for managing radioactive waste according to claim 1, characterized in that radiation emitted from the spent nuclear fuel is detected by immersing a radiation detector in the shielding water of the spent fuel pool, and the radioactivity is evaluated based on the radiation.

3. The radioactive waste is spent nuclear fuel stored in a spent fuel pool, The method for managing radioactive waste according to claim 1, characterized in that radiation emitted from the spent nuclear fuel is detected through a collimator embedded in the wall of the spent fuel pool, and the radioactivity is evaluated based on the radiation.

4. The plurality of nuclides are 134 Cs, 137 Cs, 137 mBa, 154 Eu, 90 Sr, 90 Y, 144 Ce, 144 Pr, 106 Ru and 106 Rh, and the method for managing radioactive waste according to any one of claims 1 to 3.

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