JP5567904B2 - Method for measuring subcritical multiplication factor of irradiated fuel assembly, measuring apparatus, program for measurement, and method for verifying prediction accuracy of nuclide composition of irradiated fuel assembly - Google Patents

Description

  The present invention relates to a method for measuring a subcritical multiplication factor of an irradiated fuel assembly, a measuring apparatus and a program for measurement, and a method for verifying the prediction accuracy of the nuclide composition of the irradiated fuel assembly. More specifically, the present invention relates to a method for measuring a subcritical multiplication factor of an irradiated fuel assembly, a measuring apparatus, and a measurement method for measuring a subcritical multiplication factor by inserting a neutron source into the irradiated fuel assembly submerged in a storage pool or the like. And a method for confirming the prediction accuracy of the nuclide composition of the irradiated fuel assembly using the measurement method.

  In order to introduce burn-up credits for nuclear fuel transport, it is necessary to confirm by measurement that the predicted value of the nuclide composition in the assembly that determines the reactivity of the irradiated fuel assembly is correct.

  There is a subcritical multiplication factor ksub as a value determined by the nuclide composition of the irradiated fuel assembly. Accordingly, the subcritical multiplication factor ksub of the irradiated fuel assembly is measured, and the combustion value is determined depending on whether or not this measured value matches the value calculated based on the nuclide composition predicted from the management data of the irradiated fuel assembly. The accuracy of prediction of nuclide composition required for the introduction of credits can be confirmed.

  There is a neutron source multiplication method as a method considered to be applicable to the measurement of the subcritical multiplication factor ksub of the irradiated fuel assembly. In the neutron source multiplication method, it is necessary to obtain the neutron generation rate S of the entire fuel assembly. Therefore, based on the measured value of the neutron flux φ at one or more measurable points inside or outside the nuclear reactor or nuclear fuel assembly, It is considered to estimate the neutron generation rate S.

  An apparatus for measuring the subcriticality of an irradiated fuel assembly is, for example, Patent Document 1.

JP-A-6-59084

  However, the distribution of the neutron flux φ cannot be measured in detail in the nuclear reactor or nuclear fuel assembly. In the conventional method, it is assumed that the distribution of the neutron flux φ inside and outside the fuel assembly to be measured hardly changes, and the neutron generation rate S is evaluated by measuring the neutron flux φ at one or several points where measurement is possible. The distribution of φ varies strongly depending on the nuclide composition of the fuel assembly. For this reason, in order to accurately determine the distribution of the neutron flux φ, detailed calculation is performed based on information on subcriticality (combustion history for each fuel assembly evaluated by the reactor management system, predicted nuclide composition, etc.). Without this information, the neutron generation rate S cannot be obtained even if the neutron flux φ is actually measured at a measurable location, and the subcritical multiplication factor ksub cannot be obtained.

  The present invention uses the information on the subcriticality of the distribution of the neutron flux φ inside the irradiated fuel assembly to be measured (combustion history for each fuel assembly evaluated by the reactor management system, predicted nuclide composition, etc.) The subcritical multiplication factor measurement method, measuring apparatus and program for the irradiated fuel assembly, and the prediction accuracy of the nuclide composition of the irradiated fuel assembly that can determine the subcritical multiplication factor ksub from the measured data without calculation The purpose is to provide a verification method.

As a result of intensive studies on the nuclear characteristics of irradiated fuel assemblies, the present inventors have found that when a ²⁵² Cf neutron source is inserted into a fuel assembly stored in water, neutron leakage occurs regardless of the burnup of the fuel assembly. It was found that the rate L is constant. And by further research based on this knowledge, by applying this knowledge to the conventional "neutron source multiplication method", it was found that the neutron generation rate S of the irradiated fuel assembly can be calculated with high accuracy , The present invention has been completed.

That is, the method for measuring the subcritical multiplication factor of the irradiated fuel assembly according to claim 1 is a fuel assembly having the same composition and shape at the time of the new fuel as the irradiated fuel assembly to be measured and having a known nuclide composition. The neutron leakage rate L of the assembly is obtained in advance, and the external neutron absorption rate A is measured using a ²⁵² Cf neutron source with a neutron emission intensity S0 for the irradiated fuel assembly that is submerged, and the neutron Based on the leakage rate L and the neutron absorption rate A, the neutron generation rate S of the irradiated fuel assembly to be measured is calculated, and based on the neutron emission intensity S0 and the neutron generation rate S, The critical multiplication factor ksub is calculated.

Further, in the method for measuring the subcritical multiplication factor of the irradiated fuel assembly according to claim 2, when the neutron emission intensity S0 is unknown, a ²⁵² Cf neutron source having an unknown neutron emission intensity S0 is used to measure the irradiated fuel to be measured. A fuel assembly having the same composition and shape at the time of the new fuel and the new fuel, and submerging a fuel assembly having a known nuclide composition to measure the external neutron absorption rate Ax, and a fuel having the above nuclide composition The neutron leakage rate Lx and the subcritical multiplication factor ksubx for the aggregate are calculated, the neutron generation rate Sx is calculated based on the neutron absorption rate Ax and the neutron leakage rate Lx, and the neutron generation rate Sx and the subcritical multiplication factor ksubx The neutron emission intensity S0 is calculated based on the above.

The apparatus for measuring a subcritical multiplication factor of an irradiated fuel assembly according to claim 3 is a fuel assembly having the same composition and shape at the time of the new fuel as the irradiated fuel assembly to be measured that has been obtained in advance. External storage of neutron absorption using a ²⁵² Cf neutron source having a neutron emission intensity S0 with respect to a storage means for storing a neutron leakage rate L of a fuel assembly having a known composition and an irradiated fuel assembly to be measured that is submerged. An absorptivity measuring means for measuring the rate A, an input means for reading the neutron absorptance A and the neutron emission intensity S0, and neutron generation of the irradiated fuel assembly to be measured based on the neutron leakage rate L and the neutron absorption rate A And a calculating means for calculating the subcritical multiplication factor ksub of the irradiated fuel assembly to be measured based on the neutron emission intensity S0 and the neutron generation rate S.

The apparatus for measuring a subcritical multiplication factor of an irradiated fuel assembly according to claim 4 is a fuel assembly having the same composition and shape at the time of the new fuel as the irradiated fuel assembly being submerged, and the nuclide composition. A second absorptance measuring means for measuring an external neutron absorption rate Ax using a ²⁵² Cf neutron source whose neutron emission intensity S0 is unknown for a fuel assembly of which is known, and a second input for reading the neutron absorption rate Ax Neutron leakage rate Lx and subcritical multiplication factor ksubx for the fuel assembly having the above nuclide composition, and neutron generation rate Sx based on neutron absorption rate Ax and neutron leakage rate Lx, And a second calculating means for calculating the neutron emission intensity S0 based on the neutron generation rate Sx and the subcritical multiplication factor ksubx.

According to a fifth aspect of the present invention, there is provided a program for measuring a subcritical multiplication factor of an irradiated fuel assembly, at least a fuel having the same composition and shape at the time of new fuel as the irradiated fuel assembly to be measured that is stored in advance in the storage means. Means for reading the neutron leakage rate L of a fuel assembly whose nuclide composition is known, and the externally measured fuel submerged irradiated fuel assembly using a ²⁵² Cf neutron source of neutron emission intensity S0 Means for reading the neutron absorption rate A and the neutron emission intensity S0, and calculating the neutron generation rate S of the irradiated fuel assembly to be measured based on the neutron leakage rate L and the neutron absorption rate A, and the neutron emission intensity S0 And a neutron generation rate S for causing the computer to function as means for calculating the subcritical multiplication factor ksub of the irradiated fuel assembly to be measured.

The program for measuring the subcritical multiplication factor of the irradiated fuel assembly according to claim 6 is a fuel having the same composition and shape at the time of the new fuel as that of the irradiated fuel assembly to be measured that is stored in advance in the storage means. Means for reading the neutron leakage rate L of a fuel assembly whose nuclide composition is known, and the externally measured fuel submerged irradiated fuel assembly using a ²⁵² Cf neutron source of neutron emission intensity S0 Means for reading the neutron absorption rate A and the neutron emission intensity S0, and calculating the neutron generation rate S of the irradiated fuel assembly to be measured based on the neutron leakage rate L and the neutron absorption rate A, and the neutron emission intensity S0 Means for calculating the subcritical multiplication factor ksub of the irradiated fuel assembly to be measured on the basis of the neutron generation rate S, the irradiated fuel assembly to be measured submerged, and the composition and shape at the time of the new fuel Means for reading the neutron absorptance Ax measured using a ²⁵² Cf neutron source with unknown neutron emission intensity S0 for fuel assemblies having the same fuel assembly and known nuclide composition, a fuel assembly having the above nuclide composition The neutron leakage rate Lx and the subcritical multiplication factor ksubx for the body are calculated, and the neutron generation rate Sx is calculated based on the neutron absorption rate Ax and the neutron leakage rate Lx. This is to make the computer function as a means for calculating the neutron emission intensity S0 based on the above.

  Furthermore, the method for confirming the prediction accuracy of the nuclide composition of the irradiated fuel assembly according to claim 7 uses the subcritical multiplication factor measuring method for the irradiated fuel assembly according to claim 1 or 2, and The critical multiplication factor ksub is measured, and the subcritical multiplication factor ksub of the irradiated fuel assembly is calculated based on the nuclide composition predicted for management of the irradiated fuel assembly, and the calculated value and the subcritical multiplication factor ksub are calculated. The comparison with the measured values confirms that the prediction of the nuclide composition of the irradiated fuel assembly has sufficient accuracy in terms of critical safety.

  Here, the neutron emission intensity S0 means the number of neutrons generated from the neutron source due to radioactive decay per unit time. The neutron generation rate S is the number of neutrons generated due to radioactive decay of the neutron source inserted inside the fuel assembly per unit time, and the unit inside the fuel assembly due to the fission chain reaction caused by the neutron. The total number of neutrons generated per hour. Furthermore, the neutron leakage rate L refers to the ratio of the number of neutrons absorbed outside the assembly to the number of neutrons generated inside the fuel assembly.

The method for measuring the subcritical multiplication factor of the irradiated fuel assembly according to claim 1, the apparatus for measuring the subcritical multiplication factor of the irradiated fuel assembly according to claim 3, and the subcritical multiplication factor measurement of the irradiated fuel assembly according to claim 5. According to the program, when the neutron emission intensity S0 of the ²⁵² Cf neutron source to be inserted into the irradiation fuel assembly to be measured is known, the distribution of the neutron flux φ inside the irradiation fuel assembly to be measured is expressed as its subcriticality. The subcritical multiplication factor ksub of the irradiated fuel assembly can be obtained with high accuracy without calculation using information related to the above (combustion history for each fuel assembly evaluated by the reactor management system, predicted nuclide composition, etc.) It becomes possible.

A method for measuring a subcritical multiplication factor of an irradiated fuel assembly according to claim 2, a device for measuring a subcritical multiplication factor of an irradiated fuel assembly according to claim 4, and a subcritical multiplication factor of the irradiated fuel assembly according to claim 6. According to the measurement program, the neutron emission intensity S0 can be obtained even when the neutron emission intensity S0 of the ²⁵² Cf neutron source inserted into the irradiated fuel assembly to be measured is unknown. The subcritical multiplication factor ksub can be obtained with high accuracy using S0.

  Further, according to the method for confirming the prediction accuracy of the nuclide composition of the irradiated fuel assembly according to claim 7, the subcritical multiplication factor ksub and the nuclide composition correspond to each other, so that the administrative nuclide composition matches the actual one. It can be confirmed whether or not. Therefore, it is possible to confirm that the prediction accuracy of the nuclide composition of the irradiated fuel assembly has sufficient accuracy for criticality safety management, and it becomes possible to introduce burnup credits when transporting nuclear fuel.

It is a block diagram which shows 1st Embodiment of the subcritical multiplication factor measuring apparatus of this invention. It is a flowchart which shows 1st Embodiment of the subcritical multiplication factor measuring method of this invention. It is a block diagram which shows 2nd Embodiment of the subcritical multiplication factor measuring apparatus of this invention. It is a flowchart which shows 2nd Embodiment of the subcritical multiplication factor measuring method of this invention. It is a flowchart which shows the subroutine of step S40 of FIG. It is a conceptual diagram which shows an absorptivity measuring means. It is a figure which shows the relationship between the burnup of a fuel assembly, and the neutron leakage rate L (The leakage rate of the neutron from a fuel assembly (The leakage neutron from a calculation system is included in the external absorption neutron of an assembly.)). It is a figure for demonstrating the fuel rod which analyzed the fuel arrangement | sequence and nuclide composition of the PWR17 * 17 type | mold fuel assembly. It is an effective multiplication factor calculation system diagram of the state where the fuel assembly was stored alone in light water. It is a systematic diagram of the eigenvalue calculation paying attention to each aggregate node, and is a diagram regarding the calculation of the neutron generation rate / absorption ratio ratio in the aggregate part + periphery water region. It is a figure which shows the diffusion coefficient of each node of an irradiation fuel assembly. It is a figure which shows macroscopic absorption cross section Σa and fission neutron production cross section νΣf of each node of the irradiated fuel assembly. It is a conceptual diagram of the H (n, (gamma)) characteristic gamma ray measuring method for quantifying the neutron absorption rate in the outer peripheral part of an aggregate. It is a figure which shows the hydrogen neutron absorption rate distribution at the time of putting the point neutron source ²⁵² Cf in light water, and the reaction rate distribution of a minute ⁶ Li (n, t) or ¹⁰ B (n, α) detector.

  Hereinafter, the configuration of the present invention will be described in detail based on the form shown in the drawings. The present invention measures the subcritical multiplication factor ksub of an irradiated fuel assembly using a neutron source multiplication method characterized by measuring the number of neutrons leaking from the fuel assembly, and is submerged in irradiation. A neutron source is inserted into the fuel assembly and the subcritical multiplication factor ksub is measured.

  FIG. 1 shows a first embodiment of a subcritical multiplication factor measuring apparatus for irradiated fuel assemblies according to the present invention. In the present embodiment, a neutron source having a known neutron emission intensity S0 is used as a neutron source used for external measurement of the neutron absorption rate A and calculation of the neutron leakage rate L. An irradiation fuel assembly subcritical multiplication factor measurement device (hereinafter simply referred to as a subcritical multiplication factor measurement device) is a fuel having the same composition and shape at the time of new fuel as the previously determined irradiation fuel assembly 1 to be measured. A storage means 3 for storing a neutron leakage rate L of a fuel assembly 2 having a known nuclide composition, and a neutron source 4 having a neutron emission intensity S0 for an irradiated fuel assembly 1 to be measured that is submerged. Based on the absorptivity measuring means 5 for measuring the external neutron absorptance A using, the input means 6 for reading the neutron absorptivity A and the neutron emission intensity S0, the neutron leakage rate L and the neutron absorptivity A Calculation means for calculating the neutron generation rate S of the irradiated fuel assembly 1 to be measured and calculating the subcritical multiplication factor ksub of the irradiated fuel assembly 1 to be measured based on the neutron emission intensity S0 and the neutron generation rate S With 7 That. This subcritical multiplication factor measuring apparatus is also realized by executing a subcritical multiplication factor measuring program 8 of the irradiated fuel assembly 1 of the present invention on a computer. In the present embodiment, a case where a subcritical multiplication factor measurement program (hereinafter simply referred to as a measurement program) 8 of the irradiated fuel assembly 1 is executed on a computer will be described as an example. Hereinafter, the “irradiated fuel assembly 1 to be measured” is simply referred to as the measurement target fuel assembly 1, and “a fuel assembly having the same composition and shape at the time of the new fuel as the measurement target irradiated fuel assembly 1”. The fuel assembly 2 having a known nuclide composition is simply referred to as a known composition fuel assembly 2.

  FIG. 1 shows the overall configuration of the subcritical multiplication factor measuring apparatus according to this embodiment for executing the measurement program 8. This subcritical multiplication factor measuring device includes a control unit 9, a storage unit (storage unit) 3, an input unit (input unit) 6, and a display unit 10, which are connected to each other by a signal line 11 such as a bus. .

  The control unit 9 performs operations related to the control of the entire subcritical multiplication measuring device and the calculation of the subcritical multiplication factor ksub by the measurement program 8 stored in the storage unit 3. Device). The storage unit 3 is a device capable of storing at least data and the program 8, and is, for example, a hard disk.

  The input unit 6 is an interface for giving at least an operator command or the like to the control unit 9 or the like, and is, for example, a keyboard. The display unit 10 draws and displays characters, graphics, and the like under the control of the control unit 9, and is a display, for example.

  The control unit 9 of the subcritical multiplication factor measuring device executes a measurement program 8 to thereby read a leakage rate as a means for reading the neutron leakage rate L of the known composition fuel assembly 2 stored in the storage means 3 in advance. Reading unit 12 reads input data as means for reading external neutron absorption rate A and neutron emission intensity S0 measured by using neutron source 4 of neutron emission intensity S0 for submerged measurement target fuel assembly 1 The neutron generation rate S of the measurement target fuel assembly 1 is calculated based on the neutron leakage rate L and the neutron absorption rate A, and the measurement target fuel assembly is calculated based on the neutron emission intensity S0 and the neutron generation rate S. A calculation unit (calculation means) 7 is configured as means for calculating one subcritical multiplication factor ksub.

  A receiver 14 is connected to the subcritical multiplication factor measuring device by a signal line 11 such as a bus. For example, the neutron absorption rate A outside the irradiated fuel assembly 1 measured by the absorption rate measuring means 5 is The data is stored in the storage unit 3 via the receiving device 14. Here, for example, the absorptivity measuring means 5 installed in a pool in which the irradiated fuel assembly 1 is submerged and the receiving device 14 are connected by a communication line 15, and the measured value of the absorptance measuring means 5 (the outside of the fuel assembly) The neutron absorption rate A) may be automatically supplied to the receiving device 14, or the neutron absorption rate A may be supplied to the receiving device 14 by an operator's transmission operation. Alternatively, the operator may directly input the neutron absorption rate A by operating the input unit 6 and store it in the storage unit 3.

  The neutron emission intensity S0 of the neutron source 4 is directly input by, for example, the operator operating the input unit 6 and stored in the storage unit 3.

  As the absorptivity measuring means 5, a measuring device that implements a known neutron absorptivity measuring method is used. As a known neutron absorption rate measurement method, for example, there is a neutron emission rate measurement method of a neutron emitter disclosed in Japanese Patent Application Laid-Open No. 2007-315909, and the “neutron emission rate” of the method is “neutron absorption rate” in the present invention. It is.

  In the present embodiment, the absorptivity measuring means 5 is provided using a pool for storing irradiated fuel assemblies. The absorptivity measuring means 5 is shown in FIG. The measurement target fuel assembly 1 is submerged in the pool, and the neutron source 4 is inserted into, for example, a control rod guide tube of the measurement target fuel assembly 1 using an interpolating apparatus. For example, since 24 control rod guide tubes are provided in a 17 × 17 type PWR fuel assembly, the neutron source 4 is inserted into all 24 tubes. Here, the neutron source 4 may be inserted into all of the 24 control rod guide tubes at the same time, and measurement may be performed. However, the neutron source 4 is only applied to some (one or more) control rod guide tubes. It is also possible to perform measurement for all of the 24 control rod guide tubes by repeatedly performing this measurement while replacing the inserted control rod guide tube. When the measurement is divided into multiple times, the measured values are totaled.

  For example, a neutron source 4 having a sufficient length is used and inserted over the entire length of the fuel assembly 1. Thus, by using the neutron source 4 having a length that can be inserted over the entire length of the fuel assembly 1, the neutron absorption rate A (neutron emission rate) can be measured in a short time. However, it is not always necessary to use a neutron source 4 having such a sufficient length, and it is possible to use a neutron source shorter than a point-like one. When a shorter neutron source 4 is used, measurement is performed while moving the neutron source 4 at a constant speed throughout the entire length of the fuel assembly 1 in the tube, and the above-described sufficient length is obtained by summing the measured values. Measurement can be performed under substantially the same conditions as when the neutron source 4 is used. In addition, even when the measurement is performed by stopping the neutron source at a plurality of points at regular intervals and the measurement values are added, if the number of measurement points is increased, the conditions are substantially the same as when the neutron source 4 having a sufficient length is used. Can be measured.

  The absorption rate measuring means 5 supplies the calculated neutron absorption rate A to the receiving device 14. Alternatively, the operator inputs the neutron absorption rate A calculated by the absorption rate measuring means 5 by operating the input unit 6. The neutron absorption rate A is temporarily stored in the storage unit 3.

  The measurement target fuel assembly 1 is, for example, a PWR irradiated fuel assembly 1. The irradiating fuel assembly 1 of the PWR can be inserted with the neutron source 4 from above, and is stored while standing in the storage pool, so that the absorption rate measuring means 5 can measure the neutron absorption rate A. Easy. However, if it is possible to insert the neutron source 4 and measure the neutron absorption rate A, a fuel assembly other than the PWR, for example, a fuel assembly such as a BWR may be used as a measurement target.

The neutron source 4 has a rod shape that can be inserted into a control rod guide tube, for example. In the present invention, ²⁵² Cf which is a spontaneous fission neutron source is used as the neutron source 4 .

  The neutron leakage rate L is obtained by numerical analysis and stored in the storage unit 3 in advance. As a result of intensive studies on the nuclear characteristics of the irradiated fuel assembly 1, the present inventors have found that the neutron leakage rate L of the fuel assembly is constant regardless of the burnup of the nuclear fuel (at least in the calculation of the subcritical multiplication factor ksub). Is constant to such an extent that it can be regarded as constant). The neutron leakage rate L of the fuel assembly can be calculated by numerical analysis if the nuclide composition of the nuclear fuel is known. Therefore, in the present invention, the neutron leakage rate L is calculated for a fuel assembly 2 that has the same composition and shape at the time of the new fuel as the measurement target fuel assembly 1 and has a known nuclide composition. The rate L is used for the fuel assembly 1 to be measured. The calculation of the neutron leakage rate L is performed using a calculation code. As the calculation code, any code that can calculate the three-dimensional neutron flux distribution can be used. For example, CITATION code as a multigroup diffusion method, DANTSYS as a multigroup deterministic transport method, etc. It is possible to use KENO etc. as the Monte Carlo method code and MCNP-5, MVP-II, etc. as the continuous energy Monte Carlo code. The neutron leakage rate L is calculated for the same system as the measurement system of the absorptivity measuring means 5.

  As the fuel assembly 2 with a known composition, there is an unused new fuel assembly. Furthermore, there is a fuel assembly that can analyze the current nuclide composition by referring to management data or the like even though there is an irradiation history. If the neutron leakage rate L is obtained using the composition known fuel assembly 2, it is not necessary to obtain the neutron leakage rate L every time measurement is performed as long as the initial nuclide composition is the same as the measurement target.

  The calculated neutron leakage rate L is directly input by, for example, an operator operating the input unit 6 and stored in the storage unit 3.

  Next, the subcritical multiplication factor measuring method of the irradiated fuel assembly 1 of the present invention will be described. This subcritical multiplication factor measurement method (hereinafter simply referred to as subcritical multiplication factor measurement method) of the irradiated fuel assembly 1 is obtained by previously obtaining a neutron leakage rate L of the known composition fuel assembly 2 as shown in FIG. 2, for example. (Step S31), an external neutron absorption rate A is measured for the submerged fuel assembly 1 using the neutron source 4 having a neutron emission intensity S0 (Step S32). The neutron generation rate S of the measurement target fuel assembly 1 is calculated based on the absorption rate A (step S33), and the subcritical multiplication factor of the measurement target fuel assembly 1 is calculated based on the neutron emission intensity S0 and the neutron generation rate S. ksub is calculated (step S34).

  First, numerical analysis is performed on the fuel composition 2 with a known composition to obtain the neutron leakage rate L, which is stored in advance in the storage unit 3 (step S31). The leakage rate reading unit 12 of the control unit 9 reads this neutron leakage rate L from the storage unit 3 and supplies it to the calculation unit 7.

  Next, the absorption rate measuring means 5 measures the neutron absorption rate A outside the measurement target fuel assembly 1, and the measured neutron absorption rate A is stored in the storage unit 3. The neutron emission intensity S0 of the neutron source 4 used for the measurement is also stored in the storage unit 3. The input data reading unit 13 reads the neutron absorption rate A and the neutron emission intensity S0 from the storage unit 3 and supplies them to the calculation unit 7.

  The calculation unit 7 calculates the neutron generation rate S of the fuel assembly 1 to be measured based on the neutron leakage rate L and the neutron absorption rate A (step S33), and measures based on the neutron emission intensity S0 and the neutron generation rate S. The subcritical multiplication factor ksub of the target fuel assembly 1 is calculated (step S34). Specifically, the neutron generation rate S is calculated based on Formula 1, and the subcritical multiplication factor ksub is calculated based on Formula 2.

<Equation 1>
Neutron generation rate S = neutron absorption rate A / neutron leakage rate L outside the assembly
<Equation 2>
Subcritical multiplication factor ksub = (neutron generation rate S−neutron emission intensity S0) / neutron generation rate S

  The calculated subcritical multiplication factor ksub is stored in the storage unit 3 and displayed on the display unit 10 (step S35). Thereafter, the measurement program 8 ends.

  Note that, in the second and subsequent measurements, the neutron leakage rate L is already stored in the storage unit 3, so that step S31 is skipped and the process is executed from step S32.

  In the present invention, when the neutron emission intensity S0 of the neutron source 4 is known, the measurement of the fuel assembly 1 to be measured is not performed without referring to the combustion history of each assembly evaluated by the reactor management system, the predicted nuclide composition, and the like. The critical multiplication factor ksub can be obtained directly.

  Next, a second embodiment of the subcritical multiplication factor measuring apparatus of the present invention will be described. The same members and steps as those in the above-described embodiment are denoted by the same reference numerals, and description thereof will be omitted.

  FIG. 3 shows a subcritical multiplication factor measuring device. This subcritical multiplication factor measuring device enables the neutron source 4 whose neutron emission intensity S0 is unknown to be used for measuring the neutron absorption rate A, and in addition to the above components, the submerged composition is known. The second absorptivity measuring means 16 for measuring the external neutron absorption rate Ax using the neutron source 4 whose neutron emission intensity S0 is unknown for the fuel assembly 2 and the second input means 17 for reading the neutron absorption rate Ax. And neutron generation based on the neutron absorption rate Ax and the neutron leakage rate Lx, as well as calculating the neutron leakage rate Lx and the subcritical multiplication factor ksubx for the fuel assembly having the known nuclide composition (assumed fuel assembly) And a second calculating means 18 for calculating the rate Sx and calculating the neutron emission intensity S0 based on the neutron generation rate Sx and the subcritical multiplication factor ksubx.

  In the present embodiment, the second absorptivity measuring means 16 is also used by the absorptivity measuring means 5 described above. That is, the neutron absorption rate Ax is measured using the above-described absorption rate measuring means 5. The measured neutron absorption rate Ax is stored in the storage unit 3 in the same manner as the neutron absorption rate A.

  In the present embodiment, the second input means 17 is also used by the input means (input unit) 6 described above. That is, the operator may input the neutron absorption rate Ax by operating the input unit 6 described above.

  Further, by executing the measurement program 8, in addition to the above means 12, 13, and 7, the control unit 9 is provided with a neutron source 4 whose neutron emission intensity S 0 is unknown for the submerged composition known fuel assembly 2. The second input data reading unit 19 as means for reading the measured neutron absorption rate Ax, the neutron leakage rate Lx and the subcritical multiplication factor for the fuel assembly (assumed fuel assembly) having the known nuclide composition As a means for calculating ksubx, calculating neutron generation rate Sx based on neutron absorption rate Ax and neutron leakage rate Lx, and calculating neutron emission intensity S0 based on neutron generation rate Sx and subcritical multiplication factor ksubx The second calculation unit (second calculation means) 18 is configured.

  Next, the subcritical multiplication factor measuring method will be described. For example, as shown in FIG. 4, the subcritical multiplication factor measuring method includes a step S40 for obtaining the neutron emission intensity S0 of the neutron source 4. In this embodiment, after obtaining the neutron leakage rate L of the fuel assembly 2 with a known composition (step S31), step S40 is executed.

  In the routine of step S40, for example, as shown in FIG. 5, the neutron source 4 whose neutron emission intensity S0 is unknown is used to submerge the composition known fuel assembly 2 and measure the neutron absorption rate Ax (step S41). The neutron leakage rate Lx and the subcritical multiplication factor ksubx for the fuel assembly having the known nuclide composition are calculated (step S42), and the neutron generation rate Sx is calculated based on the neutron absorption rate Ax and the neutron leakage rate Lx. (Step S43) The neutron emission intensity S0 is calculated based on the neutron generation rate Sx and the subcritical multiplication factor ksubx (Step S44).

  In step S41, the external neutron absorption rate Ax is measured by the absorption rate measuring means 5 for the fuel assembly 2 with a known composition using the neutron source 4 whose neutron emission intensity S0 is unknown. The measured neutron absorption rate Ax is stored in the storage unit 3 via the receiving device 14 or the second input unit 17 (input unit 6). The second input data reading unit 19 of the control unit 9 reads the neutron absorption rate Ax from the storage unit 3 and supplies it to the second calculation unit 18.

  The second calculating unit 18 calculates a neutron leakage rate Lx and a subcritical multiplication factor ksubx assuming a fuel assembly (assumed fuel assembly) having the same nuclide composition as the known composition fuel assembly 2 measured in step S41. (Step S42). The calculation of the neutron leakage rate Lx and the subcritical multiplication factor ksubx is performed using a calculation code. As the calculation code, any code that can calculate the three-dimensional neutron flux distribution can be used. For example, CITATION code as a multigroup diffusion method, DANTSYS as a multigroup deterministic transport method, etc. It is possible to use KENO etc. as the Monte Carlo method code and MCNP-5, MVP-II, etc. as the continuous energy Monte Carlo code. The neutron leakage rate Lx and the subcritical multiplication factor ksubx are calculated for the same measurement system as that of the absorptance measuring means 5.

  The second calculating unit 18 calculates the neutron generation rate Sx of the assumed fuel assembly based on the neutron absorption rate Ax and the neutron leakage rate Lx (step S43), and the neutron generation rate Sx and the subcritical multiplication factor ksubx. Based on the above, the neutron emission intensity S0 of the neutron source 4 used in step S41 is calculated (step S44). Specifically, the neutron generation rate Sx is calculated based on Formula 3, and the neutron emission intensity S0 is calculated based on Formula 4. Formula 4 is a modification of Formula 2.

<Equation 3>
Neutron generation rate Sx = neutron absorption rate Ax outside assembly / neutron leakage rate Lx
<Equation 4>
Neutron emission intensity S0 = neutron generation rate S (1-subcritical multiplication factor ksubx)

  Then, after the calculated neutron emission intensity S0 is stored in the storage unit 3, the process returns to step S32 in FIG. 4 and proceeds to step S32 → step S33 → step S34 → step S35. The subcritical multiplication factor ksub is obtained and displayed on the display unit 10. Thereafter, the measurement program 8 ends.

  In the second and subsequent measurements, since the neutron emission intensity S0 of the neutron source 4 is already stored in the storage unit 3, the subcritical multiplication factor ksub is obtained by the measurement method of FIG.

  In this embodiment, the neutron emission intensity S0 can be obtained even if the neutron emission intensity S0 of the neutron source 4 is unknown, and the subcritical multiplication factor ksub can be obtained using the obtained neutron emission intensity S0. That is, even if the neutron emission intensity S0 of the neutron source 4 is unknown, if an assembly with a known composition such as a new fuel assembly is used, the combustion evaluated by the reactor management system for the fuel assembly to be measured The subcritical multiplication factor ksub of the irradiated fuel assembly 1 can be directly obtained without referring to the history, the predicted nuclide composition, and the like.

  Next, a method for confirming the prediction accuracy of the nuclide composition of the irradiated fuel assembly of the present invention will be described.

  This verification method for the prediction accuracy of the nuclide composition of the irradiated fuel assembly (hereinafter simply referred to as the verification method) measures the subcritical multiplication factor ksub of the irradiated fuel assembly using the above-mentioned subcritical multiplication factor measurement method. At the same time, the subcritical multiplication factor ksub of the irradiated fuel assembly is calculated based on the nuclide composition predicted for the irradiated fuel assembly, and the irradiated fuel assembly is compared with the measured value of the subcritical multiplication factor ksub. This confirms the prediction accuracy of the nuclide composition of the aggregate.

  The nuclide composition of the fuel assembly is managed by, for example, a reactor core management system. The subcritical multiplication factor ksub of the fuel assembly is calculated on the basis of the administrative nuclide composition, and this calculated value is compared with the measured value of the actual subcritical multiplication factor ksub. It can be confirmed that the prediction has sufficient accuracy in terms of criticality safety.

  Calculation of the subcritical multiplication factor ksub of the fuel assembly based on the administrative nuclide composition is performed using a calculation code. As the calculation code, any code that can calculate the three-dimensional neutron flux distribution can be used. For example, CITATION code as a multigroup diffusion method, DANTSYS as a multigroup deterministic transport method, etc. It is possible to use KENO etc. as the Monte Carlo method code and MCNP-5, MVP-II, etc. as the continuous energy Monte Carlo code.

  The actual measurement of the subcritical multiplication factor ksub for the fuel assembly is performed by the subcritical multiplication factor measuring method of the present invention. By comparing this measured value with the subcritical multiplication factor ksub (calculated value) based on the nuclide composition managed by the core management system, the accuracy of the composition prediction by the core management system can be verified. That is, when the measured value and the calculated value based on the core management system are the same or different but within the allowable range, the nuclide composition by the core management system and the actual nuclide composition of the irradiated fuel assembly 1 are (The evaluation that the prediction of the nuclide composition of the irradiated fuel assembly has sufficient accuracy for critical safety management). On the other hand, when the measured value and the calculated value based on the core management system are different and the difference exceeds the allowable range, the nuclide composition by the core management system and the nuclide composition of the actual irradiated fuel assembly 1 are different. If there is a flaw between the two, it can be evaluated (evaluation that the prediction of the nuclide composition of the irradiated fuel assembly does not have sufficient accuracy for critical safety management). Thus, it is effective as means for confirming the presence or absence of wrinkles.

  By verifying the subcritical multiplication factor ksub of the fuel assembly 1 in this way, it is possible to introduce burnup credits when transporting nuclear fuel.

  The above-described embodiment is an example of a preferred embodiment of the present invention, but is not limited thereto, and various modifications can be made without departing from the scope of the present invention.

  For example, in the method shown in FIG. 2, step S32 is executed after step S31. However, step S31 may be executed after step S32.

  In the above description, the neutron source 4 is inserted into the measurement target fuel assembly 1 at the time of measurement with respect to all the control rod guide tubes. It is not necessary, and it may be performed only for some control rod guide tubes. For example, when there are 24 control rod guide tubes, the neutron source 4 may be inserted into a plurality of 23 or less or one control rod guide tube.

  A numerical experiment was conducted on the relationship between the neutron leakage rate L of the fuel assembly and the burnup. The experiment was conducted on a 17 × 17 type fuel assembly (FIG. 6). The assembly has 24 control rod guide tubes that are easily accessible from the top in the vertical direction. Further, about 2 to 4 fuel assemblies loaded in the nuclear reactor are used by inserting a rod group of a neutron source into the control rod guide tube.

Assuming the case where a cylindrical ²⁵² Cf and Sb-Be neutron source with a fuel rod effective length of 366 cm is inserted into this control rod guide tube, the multiplication of neutrons and the neutron absorption rate A outside the assembly are Numerical experiments were performed with the continuous energy Monte Carlo code MCNP-5.

  Here, the fuel assembly was intended for new fuel and fuel used in the reactor for 1-4 cycle cycles. In this numerical experiment, the vertical nuclide composition distribution for each fuel rod of the actual irradiated fuel assembly is evaluated in detail using a commercial reactor design code, and the evaluated nuclide composition is used. The result is shown in FIG.

^{In the 252} Cf neutron source, it was confirmed that the neutron leakage rate L is almost independent of the burnup and is constant. On the other hand, in the case where the Sb-Be neutron source was inserted, the neutron leakage rate L tended to decrease slightly as the burnup increased compared to the case of the ²⁵² Cf neutron source, but it was found that it did not change significantly.

As described above, the neutron leakage rate L is calculated for the fuel assembly 2 having the same composition and shape at the time of the new fuel as the measurement target fuel assembly 1 and having a known nuclide composition. It was found that the neutron generation rate S of the measurement target fuel assembly 1 can be calculated using the measurement target fuel assembly 1. In particular, when ²⁵² Cf was used as the neutron source, it was also found that the neutron generation rate S can be calculated with high accuracy since the neutron leakage rate L is almost independent of the burnup and is constant.

(1) Using the commercial PWR replacement core design code, we will analyze the detailed nuclide composition distribution of typical irradiated fuel and perform numerical simulations using this to determine the neutron and γ-ray fields around the PWR irradiated fuel. Was revealed. Based on this, a method for measuring the subcritical multiplication factor ksub was examined and proposed.

(2. Analysis of basic characteristics of irradiated fuel)
(2.1 Detailed calculation of fuel assembly composition)
In this embodiment, the fuel loading pattern of an equilibrium core satisfying the safety requirements of a nuclear reactor is intended for a three-loop PWR core loaded with 157 17 × 17 type fuel assemblies with a maximum burnup of 55 MWD / kg HM. Analyzed. In this case, the fuel was used under the condition that a uranium standard fuel assembly and a fuel assembly (Gd-containing fuel) loaded with 16 10 wt% gadolinia-added uranium fuel rods were used. The core thermal power was 2652 MW, the core cycle burnup was 14.8 MWd / kgHM, and the cycle combustion days were about 410 days.

  In order to obtain typical nuclide composition distribution in this fuel, the aggregate calculation code AEGIS developed by Nuclear Engineering Co., Ltd., and the three-dimensional detailed mesh core calculation code SCOPE2 developed by Nuclear Fuel Industry Co., Ltd. (hereinafter NFI) The calculation using was performed.

  AEGIS calculates the effective cross-sectional area of each fuel rod by ultra-detail group resonance calculation, performs neutron transport calculation of heterogeneous assembly system by Characteristics method (MOC), and based on the obtained microscopic reaction rate, Krylov part It is a code that performs combustion calculation by the spatial method.

  SCOPE2 is a code for calculating the three-dimensional neutron flux distribution and power distribution of a nuclear reactor using the SP3 transport node method using a calculation mesh that is divided in detail up to one unit cell including fuel rods. The fuel rod vertical direction is divided into 24 equidistant calculation mesh = nodes. The local change of the neutron flux in the node due to the grid etc. is explicitly considered by the analytical polynomial expansion node method. In the equilibrium core analysis in this example, the number of neutron energy groups was nine, and the assembly nuclear constant created by the AEGIS code was used.

  By calculating the SCOPE2 code, the output history and water density history of the fuel taken out of the equilibrium core were calculated. As a result, the fuel with the highest removal burnup was the uranium standard fuel, and the extraction was at the end of the 4th cycle. As for the loading position of the fuel in the furnace, 1, 2 cycles are the outermost periphery, and 3, 4 cycles are inside.

This assembly burnup at the end of each cycle is shown in Table 1 (the cycle burnup of the fuel assembly with the highest burnup at the time of removal in the PWR 3-loop equilibrium core).

  In this embodiment, the fuel assembly is used as a typical example of PWR irradiated fuel and used in the subsequent analysis.

  With respect to the fuel, the average value (aggregate node average value) of the burn-up fuel burnup, the output history, and the water density history was obtained for each node having the same height in the vertical direction. Using these histories, aggregate combustion calculation was performed using AEGIS code for 264 fuels per aggregate node. In this combustion calculation, the number of energy groups was 172, and the actinide 28 nuclides and the fission product nuclides 193 nuclides were considered as the combustion chain. The number of combustion steps per cycle was 10. Thereby, the nuclide composition distribution in the fuel rod at the end of each cycle was obtained. The nuclide composition was evaluated for every 24 nodes in the vertical direction for 38 lines in FIG. 8 assuming 1/8 symmetry in the fuel assembly.

The nuclide composition evaluated here is at the time of each cycle stop, but the nuclide number density changes due to radioactive decay during the cooling period thereafter. This was calculated with the MVP-II code, and the detailed nuclide composition distribution in the aggregate after 10 years from the irradiation was evaluated. The target nuclides are shown in Table 2 (nuclides used in nuclear calculations), and actinide nuclides and fission products (FP nuclides) with a large neutron absorption cross section were selected. Prior studies have calculated the contribution of individual FP nuclides accumulated in PWR fuel assemblies with a burnup of 30 MWd / kg HM to the neutron absorption using the ORIGEN-2 code. According to this, the ratio of the neutron absorption by the FP nuclides selected in Table 2 to the absorption by all FP nuclides is 91%. From this ratio, the types of FP nuclides considered in this example are sufficient for the purpose of the intended measurement method development. In this example, the analysis using this composition was advanced.

(2.2 Calculation of eigenvalues of irradiated fuel assembly)
The calculation of nuclear properties of irradiated fuel was advanced using the continuous energy Monte Carlo code MCNP-5. In the analysis of the fuel assembly, a system composed of 264 cylinders composed of 264 fuel rods, a control rod guide tube, and an instrumentation guide tube was used. The fuel rod locations were given the composition calculated in the previous section every 24 nodes in the vertical direction. The inside of the control rod, the instrumentation guide tube, and the space between the fuel rods are assumed to be filled with light water. Only the effective length of the fuel rod was considered, and the grid, plenum, end plug, and upper and lower nozzles were ignored. The outer peripheral portion and the upper and lower portions of the aggregate were light water having a thickness of 50 cm (see FIG. 9). The temperature of light water and fuel was 20 ° C., and the density of light water was 1 gcm ⁻³ .

  In the Monte Carlo calculation, ENDF / B-VI. 8-base ACTI and ENDF66 were used, and SAB2002 data library was used for thermal neutron inelastic scattering cross section correction of hydrogen.

The effective multiplication factor keff of the place where one fuel assembly was stored in light water was calculated. Table 3 (effective multiplication factor keff of a system in which the fuel assembly is stored alone in light water) is summarized. The PWR new fuel assembly has high reactivity, and the keff exceeds 0.9 for one fuel assembly. However, as the combustion progresses, the reactivity of the assembly decreases to about 70%.

(3. Confirmation method of reactivity decrease by combustion)
Neutron source multiplication has been used for criticality control when handling fuel assemblies or for predicting critical states for reactor startup. The neutron source multiplication method utilizes the fact that the total neutron generation rate S obtained by multiplication with fuel is expressed by Equation 5 when a neutron source having a neutron emission intensity S0 is inserted into the fuel assembly.

<Equation 5>
S = neutron emission intensity S0 / (1-subcritical multiplication factor ksub)

When applying the neutron source multiplication method to the measurement of subcriticality when a single fuel is stored in light water, the ratio of the neutron count rate to the total neutron generation rate S, so that the detection efficiency does not depend on the burnup Development of a new neutron detection method and evaluation of the standard subcriticality or neutron emission intensity S0 are problems. A method to solve this will be examined.

(3.1 Neutron source multiplication method)
(3.1.1 Neutron leakage rate and total neutron generation rate)
The neutron multiplication and the neutron absorption rate outside the assembly were calculated with the MCNP-5 code for the system (Fig. 9) where one irradiated fuel was stored in light water. As neutron sources, ²⁵² Cf, Sb-Be and an actinide neutron source accumulated in irradiated fuel were considered.

²⁵² Cf is used as a neutron source for fuel-criticality control and reactor start-up of the initially loaded PWR core, and is called a primary neutron source PS. On the other hand, the Sb-Be neutron source is called a secondary neutron source SS, and is used when the cycle is started after the fuel is replaced. SS is a neutron source that generates ¹²⁴ Sb with a half-life of 60.4 days by neutron capture reaction of antimony ¹²³ Sb in the previous cycle operation, and emits neutrons from ⁹ Be by photonuclear reaction with the γ rays. ^{In the} actual machine, ²⁵² Cf and Sb-Be are enclosed in a cylindrical rod and inserted into one to four control rod guide tubes shown in FIG. 8, but in this analysis, the same number of 24 control rod guide tubes is used. The conditions were such that a strong neutron source was inserted, and the vertical direction of the assembly was uniform.

  For the actinide neutron source, the distribution of occurrence was taken into account, taking into account the distribution of fuel rods divided into 24 nodes in the vertical direction.

FIG. 7 shows the ratio of the neutron absorption rate to the outer circumference and the upper / lower water region in FIG. 9 and the leakage rate L with respect to the neutron generation rate. The leakage rate L when an actinide neutron source or ²⁵² Cf neutron source is inserted is constant at about 40% regardless of the burnup of the fuel assembly (the maximum / minimum ratio is 1.012 for the actinide neutron source and 1.007 for ²⁵² Cf) )Met. On the other hand, when Sb-Be was used, the neutron leakage rate L decreased in a system in which the burnup progressed and the subcritical multiplication factor decreased.

If the neutron absorption rate in the outer periphery and the upper / lower water region of FIG. 9 is measured, the primary + secondary neutron generation rate S caused by actinide neutrons, or the primary + secondary neutron generation rate S by insertion of ²⁵² Cf, It can be estimated by using a leak rate that is unrelated to the burnup of the fuel assembly. In addition, when Sb-Be is inserted, the error is larger than when ²⁵² Cf is inserted, but the leakage rate can be estimated.

(3.1.2 Analysis of neutron leakage rate)
(3.1.2.1 Factors with a constant leakage rate)
The fuel assembly has a sufficiently large horizontal buckling compared to the vertical buckling, and neutrons leak mainly in the horizontal direction. According to the group diffusion theory, the neutron leakage in the lateral direction of the bare rectangular core is proportional to the product of the horizontal geometric buckling B _h ² and the core diffusion coefficient D. On the other hand, the absorption rate at the core is proportional to the neutron absorption cross section Σa at the core average. Therefore, the constant neutron leakage rate in the lateral direction can be explained if Equation 6 does not depend on the burnup. Since the aggregate shape is constant, the geometric buckling B _h ² is substantially constant. Here, D and Σa were investigated.

The vertical direction of FIG. 10 was extended to 366 cm without changing the composition, and the fixed source calculation was performed at a position of 70 cm from the upper end of the ²⁵² Cf neutron source with the upper and lower surfaces as the vacuum boundary. In this system, the average neutron flux φCf (z) and neutron current JCf (z) of the aggregate cross section were obtained for each vertical position (z). Further, the diffusion coefficient D of the aggregate portion was evaluated according to Fick's law expressed by Equation 7, and is shown in FIG. Despite statistical variations, the change in diffusion coefficient due to combustion is small.

On the other hand, the average neutron flux and neutron absorption rate in the aggregate portion in the node were calculated by eigenvalue calculation using the aggregate node system of FIG. 10, and the neutron absorption cross section Σa was obtained and shown in FIG. The figure also shows νΣf as the fission neutron production cross section. It was found that the change in Σa was small, although νΣf decreased with the progress of burnup. This is thought to be due to the fact that the absorption cross-sectional area of ²³⁵ U and the like is reduced by combustion, but the increase in the absorption effect of the accumulated uranium element and FP nuclide is about the same.

  From the above, in the irradiated fuel, the change in the diffusion coefficient and the absorption cross-sectional area due to combustion is small, so the change in combustion of the neutron leakage rate L is also small.

Here, we review the selection of FP nuclides in Section 2.1. For FP nuclides, the error associated with considering only the representative nuclides particularly affects the neutron absorption rate of the 4-cycle irradiated fuel. Table 4 (ratio of the neutron absorption rate of each element in the assembly to the primary neutron generation rate when the neutron source ²⁵² Cf is inserted into the 4-cycle irradiated fuel) summarizes the neutron absorption destinations in the 4-cycle irradiated fuel. The ratio of FP nuclides to the total absorbed amount is less than 12%. As mentioned above, according to the previous study, the neutron absorption rate of the FP nuclides selected in Table 2 is 91%, and the neglected FP nuclide absorption effect is about 1% of the main system It is evaluated. This result does not affect the conditions for establishing a constant leak rate.

(3.1.2.2 Neutron leakage rate in boric acid water)
In the criticality safety evaluation of irradiated fuel, it is often assumed that the fuel is submerged in light water. However, loading of the transport cask at the PWR power plant is performed in high-concentration boric acid water. Therefore, the neutron leakage rate in boric acid water was calculated and evaluated when a ²⁵² Cf neutron source (24, fuel rod length) was inserted. Table 5 (corresponding to the total length of 24 control rod guide tubes for spent PWR fuel) The neutron source ²⁵² Cf was inserted and the boron concentration and the neutron leakage rate (the relationship between the total neutron generation rate and the neutron absorption rate to the outer and upper and lower water regions).

  The increase in boron concentration increases the neutron leakage rate. Some of the neutrons transmitted from the fuel assembly to the outside re-enter into the assembly while being decelerated due to multiple scattering, but when the boron concentration increases, the neutrons that have been decelerated in the peripheral water region are absorbed by boron. The It is considered that the amount of neutron absorption inside the aggregate decreases due to the decrease in the amount of neutron re-inflow into the aggregate, and the neutron leakage rate increases under high boron concentration. However, at the same boron concentration, the difference in leakage rate due to the number of irradiation cycles of the aggregate is as small as less than 1%. If the boron concentration is obtained by measurement, it is possible to estimate the total neutron generation rate by measuring the neutron absorption rate outside the fuel assembly and then considering the leakage rate according to the boron concentration.

(3.1.2 Examination of leakage neutron measurement technique for subcritical multiplication measurement)
(3.1.2.1 Neutron leakage rate and neutron source multiplication method)
According to the calculation in Section 3.1.1, the neutron absorption rate outside the fuel assembly in a state where ²⁵² Cf or Sb-Be is inserted into the assembly is obtained. The total neutron generation rate S is obtained by using the neutron leakage rate L for the new fuel in common with the irradiated fuel, ksub is estimated from the neutron source multiplication method (Equation 5), and ksub Table 6 (ksub error when subcritical multiplication factor ksub and common neutron leakage rate L are used when neutron source ²⁵² Cf and Sb-Be corresponding to the full length are inserted into 24 control rod guide tubes of irradiated PWR fuel) Compared. ^{When 252} Cf is used, this error is less than 0.5%, and ksub can be obtained with high accuracy using a common L regardless of the burnup. That is, if a ²⁵² Cf neutron source with a cylindrical and effective fuel length is inserted into 24 control rod guide tubes of the fuel assembly and the neutron absorption rate outside the assembly is measured, ksub can be obtained without information on the burnup, It can be used to confirm the decrease in reactivity due to combustion of the fuel assembly.

In addition, when Sb-Be is used, ksub estimation by the neutron source multiplication method is possible, although the ksub evaluation accuracy for the three- and four-cycle irradiated fuel is inferior compared with the case where ²⁵² Cf is used.

  In the following, a method for measuring neutron absorption outside the fuel assembly was examined.

(3.1.2.2 H (n, γ) ray measurement method)
In the calculation for the system in which ²⁵² Cf was inserted into the 4-cycle irradiated fuel, 99.5% of the neutrons absorbed in the peripheral water region were absorbed into hydrogen. In this reaction H + n → D + γ, 2.223 MeVγ rays are generated. Concept of a technique for evaluating the neutron absorption rate in the peripheral water region by measuring the 2.223 MeV characteristic γ-ray (referred to as H (n, γ) -ray measurement method in this embodiment; see Japanese Patent Laid-Open No. 2007-315909). Is shown in FIG.

A γ-ray spectrometer detector is placed at the position where the neutron leaked from the fuel assembly is expected. The detector is shielded from the aggregate so that the interior of the aggregate is not expected. The absorptance A in the neutron region of interest is expressed by Equation 8 when the neutron flux is φ, the absorption cross section is Σa, and water and hydrogen are H ₂ O and H, respectively.

  The expected value c of the characteristic γ-ray flux at the detector point for one characteristic γ-ray generated at the position r is given by Equation 9 using the distance d between r and the detector.

  Here, Ω is a solid angle of the detector viewed from the position r, and μ is an attenuation coefficient of 2.223 MeVγ ray in water. When 2.223 MeVγ rays are selected by spectrum measurement, it is only necessary to consider the zero-th collision event, so that the evaluation of c is easy. Since the characteristic γ-ray flux G to be measured and evaluated is given by Equation 10, the absolute value of φΣa, H can be obtained from G by obtaining the absorption distribution φΣa, H of neutrons into hydrogen.

As a method for evaluating the absorptance distribution φΣa, H, there is a method using a similar reaction as a method that does not require composition information of irradiated fuel. As shown in FIG. 14, the ⁶ Li (n, t) reaction and the ¹⁰ B (n, α) reaction show similar distributions as the H (n, γ) reaction, so a ¹⁰ B proportional counter or ⁶ Li fiber detector is used. It is conceivable to actually measure and evaluate the H (n, γ) reaction rate distribution.

(3.1.3.3 Subcriticality measurement method based on neutron source multiplication method for irradiated fuel and new fuel)
The H (n, γ) ray measurement method is characterized in that a thermal neutron detector is used after the absolute value is calibrated by a characteristic γ ray count rate with a γ ray spectrometer. On the other hand, if it is used only for the neutron source multiplication method, an absolute value is not necessary. Insert ²⁵² Cf for new fuel with reliable composition information. In this case, the neutron absorption rate in the peripheral water region per neutron from ²⁵² Cf can be accurately calculated. This system is measured with a ¹⁰ B proportional counter or a 6Li (n, t) fiber detector, and the relative detection efficiency of these detectors is obtained (calibrated) based on the ²⁵² Cf neutron emission rate. After that, if the same ²⁵² Cf is inserted into the target irradiated fuel and neutron detection is performed with a calibrated ¹⁰ B proportional counter or ⁶ Li (n, t) fiber detector, the external neutron absorption rate and The total neutron generation rate is obtained as a relative value of the ²⁵² Cf neutron emission rate. From these, ksub is obtained.

(Simple method of 3.1.2.4 H (n, γ) ray measurement)
From the method described in 3.1.2.1, a measurement method that omits the absorption rate distribution φΣa and H evaluation was studied. A ²⁵² Cf neutron source is arranged for the irradiated fuel assembly, or only the actinide neutron source stored in the irradiated fuel is given, and the vertical position at the detector position in FIG. 13 is the center height of each vertical node on the assembly side. The gamma ray flux at the position was calculated and added up. This is equivalent to continuously measuring the H (n, γ) characteristic γ rays by scanning the γ ray spectrometer in the vertical direction. The total value of the γ-ray flux was divided by the neutron generation rate in the aggregate and summarized in Table 7 (distance of 2.223 MeV γ-radiation at a distance relative to the amount of neutron generation in the aggregate).

In the case of actinide neutron source and ²⁵² Cf, the distribution of neutron absorption outside the fuel assembly is different due to the difference in the shape of the neutron source distribution, so there is a systematic difference in the γ-ray flux total value / neutron generation rate ratio. There are also changes depending on the burnup. However, in the neutron source multiplication method using ²⁵² Cf, the total neutron generation rate is calculated from the combined γ-ray flux value in each system by using the combined γ-ray flux value / neutron generation rate ratio for each new fuel. When ksub was estimated, the error was 2.5%. This can sufficiently separate the ksub difference of the 3 and 4 cycle irradiated fuels in Table 6. This simple method that does not measure the absorptance distributions φΣa and H is also promising for neutron generation rate measurement.

DESCRIPTION OF SYMBOLS 1 Fuel assembly to be measured 2 Fuel assembly having the same composition and shape at the time of new fuel as the fuel assembly to be measured and whose nuclide composition is known 3 Storage means 4 Neutron source 5 Absorbance measurement means 6 Input Means 7 Calculation means 16 Second absorption rate measurement means 17 Second input means 18 Second calculation means 19 Second input data reading unit

Claims (7)

The neutron leakage rate L of a fuel assembly having the same composition and shape at the time of the new fuel as the measurement target irradiated fuel assembly and a known nuclide composition is obtained in advance, and the measurement target that is submerged is submerged. An external neutron absorption rate A is measured for the irradiated fuel assembly using a ²⁵² Cf neutron source having a neutron emission intensity S0, and the irradiated fuel to be measured is measured based on the neutron leakage rate L and the neutron absorption rate A. A neutron generation rate S of the assembly is calculated, and a subcritical multiplication factor ksub of the irradiated fuel assembly to be measured is calculated based on the neutron emission intensity S0 and the neutron generation rate S. Method for measuring subcritical multiplication factor of aggregate. When the neutron emission intensity S0 is unknown, a ²⁵² Cf neutron source with an unknown neutron emission intensity S0 is used, and the irradiated fuel assembly to be measured and the fuel assembly having the same composition and shape at the time of new fuel are used. And submerging a fuel assembly having a known nuclide composition to measure an external neutron absorption rate Ax, and calculating a neutron leakage rate Lx and a subcritical multiplication factor ksubx for the fuel assembly having the nuclide composition, A neutron generation rate Sx is calculated based on the neutron absorption rate Ax and the neutron leakage rate Lx, and the neutron emission intensity S0 is calculated based on the neutron generation rate Sx and the subcritical multiplication factor ksubx. The method for measuring a subcritical multiplication factor of an irradiated fuel assembly according to claim 1. Storage means for storing the neutron leakage rate L of the fuel assembly having the same composition and shape at the time of the new fuel as the measurement target irradiated fuel assembly and the known nuclide composition An absorptivity measuring means for measuring an external neutron absorption rate A using a ²⁵² Cf neutron source having a neutron emission intensity S0 for the irradiated fuel assembly to be measured, submerged, and the neutron absorption rate A and the neutron emission Based on the input means for reading the intensity S0, the neutron generation rate S of the irradiated fuel assembly to be measured based on the neutron leakage rate L and the neutron absorption rate A, the neutron emission intensity S0 and the neutron An apparatus for measuring a subcritical multiplication factor of an irradiated fuel assembly, comprising: a calculating means for calculating a subcritical multiplication factor ksub of the irradiated fuel assembly to be measured based on the occurrence rate S. The ²⁵² Cf neutron source with unknown neutron emission intensity S0 for a fuel assembly having the same composition and shape at the time of new fuel as the measurement target irradiated fuel assembly and a known nuclide composition is submerged. A second absorptivity measuring means for measuring an external neutron absorptance Ax, a second input means for reading the neutron absorptance Ax, and a neutron leakage rate Lx for a fuel assembly having the nuclide composition A subcritical multiplication factor ksubx is calculated, a neutron generation rate Sx is calculated based on the neutron absorption rate Ax and the neutron leakage rate Lx, and the neutron generation rate Sx and the subcritical multiplication factor ksubx are calculated. The apparatus for measuring a subcritical multiplication factor of an irradiated fuel assembly according to claim 3, further comprising a second calculating means for calculating the neutron emission intensity S0. At least means for reading the neutron leakage rate L of the fuel assembly having the same composition and shape at the time of the new fuel as the measurement target irradiation fuel assembly stored in advance in the storage means and the nuclide composition being known Means for reading the external neutron absorption rate A measured with a ²⁵² Cf neutron source of neutron emission intensity S0 and the neutron emission intensity S0 for the irradiated fuel assembly to be measured submerged, the neutron leakage rate A neutron generation rate S of the irradiated fuel assembly to be measured is calculated based on L and the neutron absorption rate A, and the irradiated fuel to be measured is calculated based on the neutron emission intensity S0 and the neutron generation rate S. A program for measuring the subcritical multiplication factor of an irradiated fuel assembly for causing a computer to function as a means for calculating the subcritical multiplication factor ksub of the assembly. At least means for reading the neutron leakage rate L of the fuel assembly having the same composition and shape at the time of the new fuel as the measurement target irradiation fuel assembly stored in advance in the storage means and the nuclide composition being known Means for reading the external neutron absorption rate A measured with a ²⁵² Cf neutron source of neutron emission intensity S0 and the neutron emission intensity S0 for the irradiated fuel assembly to be measured submerged, the neutron leakage rate A neutron generation rate S of the irradiated fuel assembly to be measured is calculated based on L and the neutron absorption rate A, and the irradiated fuel to be measured is calculated based on the neutron emission intensity S0 and the neutron generation rate S. Means for calculating the subcritical multiplication factor ksub of the assembly, the submerged irradiated fuel assembly and the fuel assembly having the same composition and shape at the time of the new fuel, and the nuclide composition already existing Means for reading neutron absorption rate Ax measured using a ²⁵² Cf neutron source with unknown neutron emission intensity S0 for known fuel assemblies, neutron leakage rate Lx and subcritical multiplication factor for fuel assemblies having the nuclide composition ksubx is calculated, a neutron generation rate Sx is calculated based on the neutron absorption rate Ax and the neutron leakage rate Lx, and the neutron emission intensity S0 is calculated based on the neutron generation rate Sx and the subcritical multiplication factor ksubx. A program for measuring a subcritical multiplication factor of an irradiated fuel assembly for causing a computer to function as a means for calculating the value.   A subcritical multiplication factor ksub of the irradiated fuel assembly is measured using the method for measuring a subcritical multiplication factor of the irradiated fuel assembly according to claim 1 or 2, and a nuclide predicted for management of the irradiated fuel assembly. Based on the composition, the subcritical multiplication factor ksub of the irradiated fuel assembly is calculated, and the predicted value of the nuclide composition of the irradiated fuel assembly is critical by comparing this calculated value with the measured value of the subcritical multiplication factor ksub. A method for verifying the prediction accuracy of the nuclide composition of an irradiated fuel assembly, characterized by verifying that it has sufficient accuracy for safety management.

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