JP5787548B2 - Inventory calculation method and inventory calculation device - Google Patents

Description

  The present invention relates to an inventory calculation method and an inventory calculation device for calculating a radioactivity amount (inventory) of radioactive waste from a spent fuel assembly.

  Of the spent fuel assemblies generated from nuclear power plants, radioactive waste other than fuel is more radioactive than "high-level radioactive waste", which is composed of fission products and high uranium elements in the fuel. Low level, classified as “low level radioactive waste”. Among the low-level radioactive waste, “TRRU waste” containing TRansUranium elements, such as fuel rod parts, can be disposed in a relatively deep underground of 50 to 100 m, or “300 m” A disposal method called “geological disposal” that is disposed deep underground is considered.

  In order to design these TRU wastes on the environment and rational disposal design, it is necessary to accurately evaluate the amount of radioactivity (inventory) contained in the waste and determine the radiation source (source term). In determining these source terms, it is necessary to evaluate the production amount of important nuclides. Typical important nuclides include C-14, Cl-36, and Co-60. Although the evaluation of the source term is important, it varies depending on the operation history and the design of the fuel, which varies from power plant to power plant. Therefore, a lot of information is required to evaluate the amount of important nuclides produced from spent fuel.

Japan Federation of Electric Power Companies / Nuclear Fuel Cycle Development Organization, “TRU Waste Disposal Technology Review-Second TRU Waste Disposal Research and Development Summary”, 2005 A. G., Croff, “ORIGEN2: A Versatile Computer Code for Calculating the Nuclide Compositions and Characteristics of Nuclear Materials”, Nuclear Technology, 62, p.335-351, 1983 Y. Ando et al., “Development and Verification of Monte Carlo Burn up Calculation System”, ICNC2003, JAERI-Conf 2003-019 (Part II), p.494, 2003 Nuclear Safety Research Association Practical Text Series No. 3 “Light Water Reactor Fuel Behavior”, Nuclear Safety Research Association, July 1998

  Evaluation of the radiation source (source term) requires identification of trace elements in the base material that may vary from production lot such as cladding tube, fuel initial enrichment, burnup, operating history, and fuel structure information It is. Furthermore, activation calculations require advanced knowledge and techniques based on reactor physics. Therefore, it is difficult for engineers in other fields who are not familiar with these to evaluate source terms.

  On the other hand, in the disposal site design, it is necessary to determine the source term based on the amount of spent fuel generated under various scenarios, and to quickly evaluate the surrounding effects. For this reason, a simple inventory evaluation method is effective in practice.

  Therefore, an object of the present invention is to provide a simple inventory calculation method.

In order to achieve the above-described object, the present invention provides an inventory calculation method for calculating a total nuclide inventory of spent fuel assemblies taken out after loading a fuel assembly in which a plurality of cladding tubes containing fuel are bundled into a core. The initial enrichment and removal of the spent fuel assembly in a nuclide inventory function expressing the total inventory of nuclides per fuel assembly by a linear expression of any one of the initial enrichment and removal burnup of the fuel assembly A spent fuel assembly nuclide total inventory calculating step for substituting any of the burnups to calculate a total nuclide inventory of the spent fuel assembly, and the spent fuel assembly nuclide total inventory calculating step, Any one of the initial enrichment and the extracted burnup of the fuel assembly, the power density, the burnup, the fuel type, and the amount of the initial preceding nuclide in the component are input. A fuel effective portion nuclide inventory calculation step of calculating the nuclide inventory of the fuel effective portion is an axial region in which the fuel for each burn degree is housed, the output density of the fuel assembly, and burn or RaTeru morphism days Irradiance days calculating step, a fuel effective portion neutron flux calculating step for calculating the neutron flux of the fuel effective portion from the initial enrichment, power density and burnup of the fuel assembly, and the fuel effective portion neutron flux calculating A fuel effective external neutron that calculates a neutron flux outside the fuel effective portion that is an axially outer region of the fuel effective portion of the fuel assembly from the neutron flux calculated in the process and the axial neutron flux distribution of the fuel assembly The neutron flux calculated in the bundle calculation step, the fuel effective external neutron flux calculation step, the irradiation days calculated in the irradiation days calculation step, and the amount of initial leading nuclides in the components outside the fuel effective portion are input. A fuel effective external nuclide inventory calculation step for calculating a nuclide inventory outside the fuel effective portion for each burnup, a nuclide inventory of the fuel effective portion and a fuel effective external nuclide inventory calculated in the fuel effective portion nuclide inventory calculation step A nuclide inventory calculation step of calculating a total nuclide inventory of the fuel assembly by summing up the nuclide inventory outside the effective fuel portion calculated in the calculation step; a nuclide inventory of the fuel assembly calculated in the nuclide total inventory calculation step; And an inventory function deriving step of deriving the inventory function based on a correlation between the nuclide total inventory and either the initial enrichment or the extracted burnup of the fuel assembly.

Further, the present invention provides an inventory calculation apparatus for calculating a total nuclide inventory of a spent fuel assembly taken out after loading a fuel assembly in which a plurality of cladding tubes containing fuel are bundled into a core. A nuclide inventory calculation function storage unit storing a nuclide inventory function expressing a total nuclide inventory per one fuel assembly by a linear expression of any one of initial enrichment and removal burnup; and an initial of the spent fuel assembly An input unit for inputting either the enrichment level or the extracted combustion level, and a data input unit for inputting the initial enrichment level, power density, number of irradiation days, and initial leading nuclide amount outside the fuel effective part of the spent fuel assembly Combustion using as input the initial enrichment and removal burnup of the fuel assembly, power density, burnup, fuel type, and amount of initial leading nuclides in the component A fuel effective portion inventory calculation section that the fuel in each calculating the nuclide inventory of the fuel effective portion is an axial region which is accommodated, the irradiation days calculation for calculating an irradiation days from the output density, and burnup of the fuel assembly The neutron flux of the fuel effective portion from the initial enrichment, power density, and burnup of the fuel assembly, and the fuel assembly from the calculated neutron flux and the axial neutron flux distribution of the fuel assembly A neutron flux calculator that calculates the neutron flux outside the fuel effective part, which is the axially outer region of the fuel effective part of the body, the neutron flux calculated by the neutron flux calculator, and the irradiation days calculator and the irradiation of days, the fuel effective portion outer inventory calculation unit for calculating the nuclide inventory of fuel outside the effective portion of each burn the initial leading nuclide content in the components of the fuel outside the effective portion as an input, the The fuel assembly for calculating the total nuclide inventory of the fuel assembly by summing the fuel effective part nuclide inventory calculated by the fuel effective part inventory calculation part and the fuel effective part nuclide inventory calculated by the fuel effective part inventory calculation part And a body inventory calculation unit.

  According to the present invention, a simple inventory calculation method can be provided.

It is a flowchart of one embodiment of an inventory calculation method according to the present invention. It is a block diagram of an inventory calculation apparatus using an embodiment of an inventory calculation method according to the present invention. It is a flowchart of the derivation method of the nuclide inventory calculation function in one embodiment of the inventory calculation method according to the present invention. It is a graph which shows the example of the initial enrichment-nuclide total inventory function in one embodiment of the inventory calculation method according to the present invention. It is a graph which shows the example of the extraction burnup-nuclide total inventory function in one embodiment of the inventory calculation method according to the present invention.

  An embodiment of an inventory calculation method according to the present invention will be described with reference to the drawings. This embodiment is merely an example, and the present invention is not limited to this.

  FIG. 2 is a block diagram of an inventory calculation apparatus using an embodiment of the inventory calculation method according to the present invention.

  The inventory calculation apparatus 10 includes an input unit 11, a nuclide inventory calculation function storage unit 12, a nuclide inventory calculation unit 14, and a nuclide inventory integration unit 16. The inventory calculation device 10 also includes a nuclide inventory calculation function creation unit 20. The nuclide inventory calculation function creation unit 20 includes a data input unit 22, a neutron flux calculation unit 24, an irradiation days calculation unit 26, a fuel effective unit inventory calculation unit 28, a fuel effective outside inventory calculation unit 30, and a fuel assembly. A body inventory calculation unit 32 and a nuclide inventory calculation function deriving unit 34 are provided.

  The inventory calculation device 10 is constructed on, for example, one computer. Only the nuclide inventory calculation function creation unit 20 may be provided on an external computer.

  FIG. 1 is a flowchart of the inventory calculation method of the present embodiment.

  This inventory calculation method is a method for calculating the total nuclide inventory of spent fuel assemblies in order to determine the source terms of the landfill site. The fuel assembly is a bundle of a plurality of cladding tubes containing fuel. The spent fuel assembly is a fuel assembly that is taken out after being loaded into the reactor core.

  Most fuel assemblies are removed from the core near the designed average take-off burn-up, except for the initial loading core at power plant startup and the fuel assemblies that had to be removed early in the refueling plan. There are many cases. Average take-off burn-up is often designed to be the same for the same fuel type. Here, the fuel type is a type of the basic structure of the fuel assembly. In the case of a fuel assembly loaded in a boiling water reactor (BWR), there are fuel types called STEP-I, STEP-II, and STEP-III (see, for example, Non-Patent Document 4). Alternatively, there are fuel types called 8 × 8 fuel, 9 × 9 fuel, and the like.

  When the fuel assembly is improved, that is, when a new fuel type fuel assembly is introduced, the removal burnup generally increases due to the high burnup. This increase in the degree of burnout correlates with the initial enrichment of uranium stored in the fuel assembly. It was found that the correlation between the extracted burn-up or initial enrichment and the nuclide inventory can be expressed by a simple formula. Therefore, in this embodiment, the nuclide inventory is calculated using this simple correlation equation.

  First, in the inventory calculation method of the present embodiment, a nuclide inventory calculation function expressing the nuclide total inventory per spent fuel assembly is created (S1). The nuclide inventory function is created by the nuclide inventory calculation function creation unit 20 and stored in the nuclide inventory calculation function storage unit 12.

  As the nuclide inventory calculation function, an initial enrichment-nuclide total inventory function and a removal burnup-nuclide total inventory function are created. The initial enrichment-nuclide total inventory function is a function for calculating the nuclide total inventory from the initial enrichment. The extracted burn-up-nuclide total inventory function is a function for calculating the nuclide total inventory from the extracted burn-up.

These nuclide inventory calculation functions perform a large number of activation calculations (see, for example, PTL 2) based on reasonable initial conditions, and organize the results. The nuclide inventory function represents the nuclide total inventory ALLNUC [Bq] by a linear expression of either the initial enrichment or the extracted burnup. In other words, the initial enrichment-nuclide total inventory function is
ALLNUC = A1 + B1 × BU (1)
It is expressed as Here, A1 and B1 are constants, and BU is the burnup [GWd / t]. Also, the extracted burnup-nuclide total inventory function is
ALLNUC = A2 + B2 × E (2)
It is expressed as Here, A2 and B2 are constants, and E is the initial uranium 235 enrichment [%].

  When Co-60 is selected as the nuclide, the nuclide inventory function is as follows.

ALLCo60 = a1 + b1 × BU (3)
ALLCo60 = a2 + b2 × E (4)
Here, ALLCo60 is the total inventory [Bq] of Co-60. Depending on the material used for the fuel assembly, a nuclide inventory function is created for nuclides other than Co-60.

  Next, either the initial enrichment degree or the extracted combustion degree of the fuel assembly to be calculated is input (S2). The initial enrichment degree or the extracted combustion degree is input from the input unit 11.

  Further, the nuclide total inventory per fuel assembly is calculated using the input initial enrichment or extracted burnup (S3). The nuclide inventory calculation unit 14 calculates the nuclide total inventory using the nuclide inventory function stored in the nuclide inventory calculation function storage unit 12.

  When the initial enrichment of the spent fuel assembly is known, the nuclide inventory calculation unit 14 reads the initial enrichment-nuclide total inventory function from the nuclide inventory calculation function storage unit 12. Then, using the read initial enrichment-nuclide total inventory function, the initial enrichment is input, and the nuclide total inventory per fuel assembly is calculated.

  When the extracted burn-up degree is known, the nuclide inventory calculation unit 14 reads the extracted burn-up-nuclide total inventory function from the nuclide inventory calculation function storage unit 12. Then, using the read out burn-up-nuclide total inventory function, the n-type total inventory per fuel assembly is calculated using the extracted burn-up as an input.

  Next, the nuclide inventory integrating unit 16 calculates the total nuclide inventory of the buried disposal site (S4). When the total nuclide inventory per fuel assembly can be regarded as constant, the nuclide inventory integrating unit 16 multiplies the total nuclide inventory per fuel assembly by the number of spent fuel assemblies to obtain the total nuclide total at the buried disposal site. Calculate inventory. When the total nuclide inventory per fuel assembly cannot be regarded as constant, the nuclide inventory integration unit 16 integrates the total nuclide inventory per fuel assembly calculated by the nuclide inventory calculation unit 14 for each spent fuel assembly. Therefore, calculate the total inventory of nuclides in the landfill site.

  As described above, in this embodiment, once the nuclide inventory calculation function is derived, it is a primary expression of the initial enrichment of the spent fuel assembly or the extracted burnup degree. Can be calculated. At this time, activation calculations are not required, so advanced knowledge based on reactor physics, identification of trace elements in the base material that may vary from production lot such as cladding, initial fuel enrichment, Information on burnup, operating history and fuel structure is not required. As a result, even engineers in other fields who are not familiar with advanced knowledge and technology based on reactor physics can easily evaluate the source terms, and the spent fuel assemblies generated under various scenarios The source term is determined based on the amount generated, and the surrounding influence can be quickly evaluated.

  Next, a method for deriving a nuclide inventory calculation function will be described.

  FIG. 3 is a flowchart of a method for deriving a nuclide inventory calculation function in the present embodiment.

  The nuclide inventory calculation function inputs the initial enrichment of the spent fuel assembly, power density, burnup, the amount of initial leading nuclides in the fuel assembly components such as the cladding, and the axial neutron flux distribution of the fuel assembly. Is derived by the nuclide inventory calculation function creation unit 20. Here, for example, when the target nuclide to be calculated is Co-60, Co-59 is the preceding nuclide. These pieces of information to be input are given for a plurality of fuel types.

  First, these pieces of information are input via the data input unit 22 (S11). Next, the fuel effective part inventory calculation part 28 calculates the fuel effective part nuclide inventory for each burnup by activation calculation (S12). Here, the fuel effective portion is a region in which fuel (nuclear fuel material) in the axial direction of the fuel assembly is stored.

  Moreover, the irradiation days calculation part 26 calculates the number of irradiation days (S13). Since the burnup [GWd / t] is obtained by multiplying the power density [GW / t] by the number of irradiation days [day], the number of days of irradiation can be easily calculated from the burnup and the power density.

  Next, the neutron flux calculator 24 calculates the neutron flux of the fuel effective part (S14). The initial uranium amount and fission cross section are uniquely determined by the initial enrichment. The power density can be easily calculated from the neutron flux [n / sec / cm 2], the fission cross section [/ cm], the energy generated per fission [J], and the initial uranium amount [g / cc]. In a light water reactor, the energy [J] generated per fission is almost constant. Therefore, the neutron flux can be calculated from the initial enrichment and the power density.

  In this way, the neutron flux and irradiation days of the cladding tube portion (fuel effective portion) can be calculated from the initial enrichment, power density, and burnup.

  Further, the neutron flux calculation unit 24 receives the axial neutron flux distribution and the neutron flux as input, and calculates the neutron flux outside the fuel effective part (S15). Next, the effective fuel outside inventory calculation unit 30 receives the effective fuel outside neutron flux, the number of irradiation days, and the amount of initial preceding nuclides outside the effective fuel portion as inputs, and by activation calculation, the effective fuel outside region for each burnup is calculated. Calculate the nuclide inventory.

  The fuel assembly inventory calculation unit 32 calculates the total nuclide inventory of the fuel assembly (S16). Specifically, the fuel effective part nuclide inventory calculated by the fuel effective part inventory calculation part 28 and the fuel effective part nuclide inventory calculated by the fuel effective part inventory calculation part 30 are summed up, and the fuel for each burnup is calculated. The total nuclide inventory is calculated.

  Thereafter, the nuclide inventory calculation function deriving unit 34 derives a nuclide inventory calculation function (S17). That is, the initial enrichment-nuclide total inventory function is created from the correlation between the nuclide total inventory and the initial enrichment. Further, the nuclide inventory calculation function deriving unit 34 creates an extracted burnup-nuclide total inventory function from the correlation between the nuclide total inventory and the extracted burnup.

  FIG. 4 is a graph showing an example of the initial enrichment-nuclide total inventory function in the present embodiment. FIG. 5 is a graph showing an example of the extracted burnup-nuclide total inventory function in the present embodiment. 4 and 5 show the total inventory of one nuclide.

  4 and 5, it can be seen that the correlation between the nuclide total inventory and the extracted burnup or initial enrichment can be well expressed by a linear expression of the extracted burnup or initial enrichment. The same applies to other nuclides.

DESCRIPTION OF SYMBOLS 10 ... Inventory calculation apparatus, 11 ... Input part, 12 ... Nuclide inventory calculation function storage part, 14 ... Nuclide inventory calculation part, 16 ... Nuclide inventory integration part, 20 ... Nuclide inventory calculation function creation part, 22 ... Data input part, 24 Neutron flux calculator, 26 irradiation day calculator, 28 fuel effective part inventory calculator, 30 fuel external inventory calculator, 32 fuel assembly inventory calculator, 34 nuclide inventory calculation function derivation unit

Claims (3)

In an inventory calculation method for calculating a total nuclide inventory of spent fuel assemblies taken out after loading a fuel assembly in which a plurality of cladding tubes containing fuel are bundled into a core,
The initial enrichment and removal combustion of the spent fuel assembly are expressed in a nuclide inventory function that represents the total nuclide inventory per fuel assembly by a linear expression of either the initial enrichment or the removal burnup of the fuel assembly. A spent fuel assembly nuclide total inventory calculating step for calculating a total nuclide inventory of the spent fuel assembly by substituting any of the degrees,
Have
The spent fuel assembly nuclide total inventory calculation step includes:
One of the initial enrichment level and the extracted burnup level of the fuel assembly, the power density, the burnup level, the fuel type, and the amount of initial leading nuclides in the constituent members as inputs. A fuel effective part nuclide inventory calculating step for calculating a nuclide inventory of a fuel effective part;
And irradiating days calculation step of calculating the output density, and burn or RaTeru morphism days of the fuel assembly,
A fuel effective part neutron flux calculating step of calculating the neutron flux of the fuel effective part from the initial enrichment, power density, and burnup of the fuel assembly;
From the neutron flux calculated in the fuel effective part neutron flux calculation step and the axial neutron flux distribution of the fuel assembly, the neutron flux outside the fuel effective part, which is the axially outer region of the fuel effective part of the fuel assembly, is obtained. A fuel effective external neutron flux calculating step to calculate,
Fuel effective part for each burnup using as input the neutron flux calculated in the fuel effective part neutron flux calculating step, the irradiation days calculated in the irradiation day calculating part, and the amount of initial leading nuclides in the components outside the fuel effective part A fuel effective external nuclide inventory calculation step for calculating an external nuclide inventory;
The nuclide of the fuel assembly is obtained by summing the nuclide inventory of the fuel effective part calculated in the fuel effective part nuclide inventory calculating step and the nuclide inventory outside the fuel effective part calculated in the fuel effective part nuclide inventory calculating step. Nuclide total inventory calculation process for calculating total inventory;
An inventory function deriving step for deriving the inventory function based on the correlation between the nuclide total inventory of the fuel assembly calculated in the nuclide total inventory calculation step and either the initial enrichment or the removal burnup of the fuel assembly; ,
Features and to Louis inventory calculation method to have a. Multiplying the total nuclide inventory by the number of the spent fuel assemblies to be disposed at the landfill site, and calculating the total nuclide inventory of the landfill site,
The inventory calculation method according to claim 1, further comprising: In an inventory calculation device for calculating a total nuclide inventory of spent fuel assemblies taken out after loading a fuel assembly in which a plurality of cladding tubes containing fuel are bundled into a core,
A nuclide inventory calculation function storage unit storing a nuclide inventory function expressing a total nuclide inventory per one fuel assembly by a linear expression of any one of the initial enrichment and the extracted burnup of the fuel assembly;
An input unit to which one of the initial enrichment and the extracted burnup of the spent fuel assembly is input;
A data input unit for inputting the initial enrichment of the spent fuel assembly, the power density, the irradiation days, and the amount of initial leading nuclides outside the fuel effective part;
One of the initial enrichment level and the extracted burnup level of the fuel assembly, the power density, the burnup level, the fuel type, and the amount of initial leading nuclides in the constituent members as inputs. a fuel effective portion inventory calculation unit for calculating the nuclide inventory of certain fuel effective portion,
An irradiation day calculation unit for calculating the number of irradiation days from the power density of the fuel assembly and the burnup;
The neutron flux of the fuel effective part is calculated from the initial enrichment, power density, and burnup of the fuel assembly, and the fuel of the fuel assembly is calculated from the calculated neutron flux and the axial neutron flux distribution of the fuel assembly. A neutron flux calculator that calculates the neutron flux outside the fuel effective part, which is the axially outer region of the effective part,
Fuel effective part for each burnup using as input the neutron flux calculated by the neutron flux calculating part, the irradiation days calculated by the irradiation day calculating part, and the amount of initial leading nuclides in the components outside the fuel effective part A fuel effective external inventory calculation unit for calculating an external nuclide inventory,
A fuel that calculates the total nuclide inventory of the fuel assembly by summing the effective fuel nuclide inventory calculated by the effective fuel inventory calculator and the effective fuel nuclide inventory calculated by the non-fuel effective inventory calculator An aggregate inventory calculation unit;
An inventory calculation apparatus comprising:

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