JP2003098288A - Method of evaluating neutron emission rate from spent fuel, and program therefor - Google Patents

Abstract

PROBLEM TO BE SOLVED: To evaluate a neutron emission rate of Cm244 from a spent fuel. SOLUTION: A generation amount of a neutron emission nuclide Cm244 provided by combustion calculation based on respective data when loaded in a core of a nuclear reactor to be burnt by neutron irradiation is conformed substantially with an actual generation amount by regulating a neutron capturing cross-sectional area(s) of Am243 or Pu242, or Am243 and Pu242 using combustion calculation combined with a neutron spectrum calculation code 2 and a nuclear data library 3, based on a specification 1 of a fuel assembly as an evaluation object. A substantial consistency with an actually generated neutron emission rate (measured value) E is allowed thereby without being affected substantially by a neutron multiplying characteristic, and calculation precision 8 of the neutron emission rate based on Cm244 is enhanced thereby to conduct the neutron emission rate evaluation 9 of Cm244.

Description

Detailed Description of the Invention

[0001]

TECHNICAL FIELD The present invention relates to a spent fuel obtained by combustion calculation when a uranium fuel or a fuel assembly containing a mixed fuel of uranium and plutonium is loaded into a reactor core and burned by neutron irradiation. TECHNICAL FIELD The present invention relates to a method for evaluating the neutron emission rate from a satellite and a program for causing a computer to perform such evaluation.

[0002]

2. Description of the Related Art A fuel assembly containing a uranium (U) fuel or a mixed fuel (U / Pu fuel) of uranium (U) and plutonium (Pu) is loaded into a reactor core and burned by neutron irradiation. , Cm242, Cm244, etc. are generated, and the spent fuel taken out of the reactor mainly emits neutrons based on these.

Since Cm242 has a relatively short half-life (163 days), neutrons based on Cm242 can be almost ignored by cooling it for about 1.5 to 3 years, but Cm244 has a long half-life (18 years). It does not decay so much, so the neutron emission rate from spent fuel cooled for more than 2-3 years is almost Cm24.
You will be governed by 4.

By the way, spent fuel is cooled in a nuclear power plant for about 2 to 5 years, and then reprocessed or stored intermediately. In any case, spent fuel must be stored in a transportation container (cask) for transportation. At that time, in designing the transportation container, it is often important to confirm the neutron shielding problem, that is, it is important to confirm that the transportation container can sufficiently shield the neutrons generated from the spent fuel. Must be evaluated correctly. However, correct evaluation is not easy.

Since around 1975, the present inventors have been the first in the world to actually measure the neutron emission rate from a spent fuel and carry out research for comparison with theoretical calculation. Part of the research is Jou
rnalof Nuclear Science and Technology, vol.18, p.24
9 (1981) “Measurements and Analysis of Neutron Emi
ssion Rate for Irradiated BWR Fuel ”etc. At that time, the measured neutron emission rate from Cm244 was 30-40% different from the theoretical calculation for uranium fuel.

[0006] Since then, the cross-sections and decay schemes of many nuclides used for calculation have been actively improved worldwide, but the improvements have been comprehensively implemented, and particularly U235, U238 and Pu23.
Due to the focus on 9, Pu240, Pu241 and fission product (FP) nuclides, no significant improvement was observed in the neutron generation rate evaluation based on Cm244, and the calculated and measured values are stable at present. It's not easy to keep it within%.

Recent research “JAERI-Tech 2000-71“ Technical development on burnup credit of spent fuel of light water reactor ”(2
(October 000) ”, the amount of Cm244 produced is underestimated by 20-40% in the composition evaluation of the nuclide irradiated in the light water reactor by the combustion calculation combining the latest calculation code and nuclear data.

In addition, S. Catalau and A. Benslimimane “Q
ualification of JEF-2.2 CaptureCross Sections for
Heavy Nuclides and Fission Products in Thermal and
Epi-thermal Spectra ”Global95 pp.1537-1544 (199
In 5), it is pointed out that there is a considerable error in the capture cross section of actinide nuclides in the latest nuclear data.

In recent years, commercialization of mixed oxide fuel (MOX) in which plutonium is added to uranium is in progress, and the neutron emission rate of the spent MOX fuel is 5 to 5 times that of uranium fuel.
It is known to be 10 times higher, and ensuring the accuracy of neutron emission rate evaluation is becoming more important.

[0010]

The present invention has been made based on the above-mentioned technical background and impairs the results of comprehensive improvements in the cross-sectional areas and decay schemes of many nuclides used in calculations. Incorporate a method for evaluating the neutron emission rate from spent fuel that can stably and easily bring the evaluation value obtained by the calculation of the neutron generation rate based on Cm244 close to the measured value, and the evaluation method , Providing a neutron emission rate evaluation program which is stored and programmed to facilitate the evaluation work.

[0011]

The invention according to claim 1 is
Evaluate the amount of neutron-emitting nuclide Cm244 obtained by combustion calculation when the fuel assembly to be evaluated containing uranium fuel or mixed fuel of uranium and plutonium is loaded into the reactor core and burned by neutron irradiation In the neutron emission rate evaluation method from the spent fuel, by adjusting the neutron capture cross section of Am243 or Pu242, or Am243 and Pu242 at the stage of combustion calculation combining the neutron spectrum calculation code and the nuclear data library,
It is characterized in that the amount of Cm244 actually generated is made to substantially match.

According to the present invention, the calculated value of the neutron emission rate of Cm244 is calculated by using the spent fuel, and the neutrons of Am243, Pu242 or Am243 and Pu242 are adjusted so that the calculated value agrees with the measured value. Adjust to change the capture cross section. By this adjustment, the corrected trapping cross section is obtained, and the corrected trapping cross section is used to hardly affect the neutron multiplication characteristics obtained in the fuel combustion calculation, and the spent fuel other than the measured spent fuel can be obtained. It is possible to obtain the neutron emission rate of Cm244 accumulated in.

That is, the evaluation value by the calculation of the neutron emission rate based on Cm244 can be obtained without impairing the results of the improvement of the cross-sections of many nuclides used in the calculation and the decay scheme, which have been carried out globally. The measured value can be stably and easily approached.

Further, Am243 and Pu242 have almost no influence on the reactivity and neutron spectrum calculation in the nuclear calculation of the nuclear reactor, so there is no fear of impairing the results of the comprehensive improvements, and the neutrons based on Cm244 The calculation accuracy of the occurrence rate can be improved.

The invention according to claim 2 is obtained by a combustion calculation when a fuel assembly to be evaluated containing uranium fuel or a mixed fuel of uranium and plutonium is loaded into a core of a nuclear reactor and burned by neutron irradiation. Neutron emitting nuclide Cm
In the neutron emission rate evaluation method from spent fuel for evaluating the production amount of 244, Am24 is used at the stage of combustion calculation combining the neutron spectrum calculation code and the nuclear data library.
3, or Pu242, or Am243 and Pu242, the neutron absorption cross-section is made constant, and the neutron capture cross-section is adjusted to substantially match the amount of Cm244 actually generated.

The neutron absorption cross section is mainly represented by the sum of the fission cross section and the neutron capture cross section. The neutron multiplication factor, which is a problem in the core design, is closely related to the neutron absorption cross section. On the other hand, the neutron emission rate is more closely related to the neutron capture cross section than the fission cross section. Therefore, a method of changing the neutron emission rate without affecting the neutron absorption cross section is the invention according to claim 2.

According to the present invention, the neutron emitting nuclide Cm24
By adjusting the neutron capture cross section by keeping the neutron absorption cross section of Am243, Pu242 or Am and Pu242 constant, there is almost no effect on the neutron multiplication characteristics obtained by calculation. The generated amount can be made to substantially match. Further, since the neutron absorption cross section is retained, the influence on the neutron spectrum calculation code can be reduced.

The invention according to claim 3 is obtained by a combustion calculation when a fuel assembly to be evaluated containing uranium fuel or a mixed fuel of uranium and plutonium is loaded into a core of a nuclear reactor and burned by neutron irradiation. Neutron emitting nuclide Cm
In the neutron emission rate evaluation method from spent fuel for evaluating the production amount of 244, Am24 is used at the stage of combustion calculation combining the neutron spectrum calculation code and the nuclear data library.
3 or Pu242, or by adjusting the neutron capture cross section of Am and Pu242 as the burnup-dependent amount, the production amount of Cm244 that is actually produced is substantially matched.

The neutron capture cross section is generally affected by burnup and changes with burnup. When the influence is small, the neutron capture cross section during combustion can be made almost constant in the neutron emission rate evaluation. According to this invention,
It is possible to deal with the case where the influence is not small.

The invention according to claim 4 is the first step of evaluating the neutron spectrum in the fuel to be evaluated by the neutron spectrum calculation code based on the combustion calculation data of the fuel assembly specifications to be evaluated, and the first step of the first step The second step of creating a 1-group reaction cross-section library of neutron cross-sections of actinide nuclides by combining the neutron spectrum calculation code with the nuclear data library, and nuclide generation from the 1-group reaction cross-section library created in this 2nd step Comparing the 3rd step of creating the annihilation calculation code and the calculated neutron emission rate (C) based on the nuclide generation and annihilation calculation code created in this 3rd step and the measured value (E) And calculation accuracy (C / E)
And a fourth step of evaluating the neutron emission rate of Cm244.

According to the present invention, in the first step, the neutron spectrum of the fuel to be evaluated is evaluated using the fuel assembly specifications and the irradiation history data, and the one-group reaction cross section used in the one-point combustion calculation code is evaluated. can do.

In the second step, the grouping reaction cross section of the neutron cross section of the actinide nuclide can be obtained by combining the neutron spectrum calculation code and the multigroup neutron reaction cross section.

In the third step, the nuclide generation amount can be obtained by using the nuclide generation / annihilation calculation code which takes into account the burnup dependency after the second step. In the fourth step, the neutron emission rate of Cm244 can be evaluated by calculating the calculation accuracy by comparing the calculated value of the neutron emission rate with the measured value of the neutron emission rate.

According to a fifth aspect of the present invention, the calculation accuracy evaluation value in the fourth step is input to a correction data table which is set in advance so that the measured values of the amount of Cm244 produced and the neutron source intensity match the calculated values. It is characterized by comprising a fifth step of

According to the present invention, by inputting the calculation accuracy in the fourth step, it is possible to set the measured amount of Cm244 and the measured value of the neutron source intensity so as to match the calculated value.

According to a sixth aspect of the present invention, a corrected one-group reaction cross section library prepared by revising the reaction cross section of Am243 or Pu242 or Am243 and Pu242 by adding a correction factor to the correction data table Is fed back to the nuclide generation / annihilation calculation code created in the third step.

According to the present invention, Am243, Pu242 or Am24 revised by adding a correction factor to the correction data table.
By inputting the neutron capture cross-sections of 3 and Pu242 in the third step, the neutron emission rate of Cm244 can be repeatedly obtained with almost no effect on the neutron multiplication characteristics obtained by the fuel combustion calculation.

The invention according to claim 7 provides a computer,
The first step of evaluating the neutron spectrum of the fuel to be evaluated by the neutron spectrum calculation code based on the combustion calculation data of the spent fuel created based on the specifications of the fuel assembly to be evaluated, and the neutron spectrum of this first step A second step of combining the calculation code with the nuclear data library to create a grouping reaction cross section of the neutron reaction cross section of the actinide nuclide, and a code for generating and eliminating nuclides from the grouping reaction cross section created in this second step The calculation value of the neutron emission rate is calculated based on the 3rd step of creating the neutron emission rate based on the nuclide generation / annihilation calculation code created in this 3rd step, and the calculation accuracy is calculated by comparing with the neutron emission rate of the measured value. Fourth evaluation of neutron emission rate
It is characterized in that steps and are executed.

According to the present invention, the fuel assembly specification, the first
Neutron spectrum calculation code of step, nuclear data library, 1st group reaction cross section library of 2nd step,
By storing and programming the nuclide generation / annihilation calculation code of the third step and the calculation accuracy of the fourth step,
The work of evaluating the amount of Cm244 produced becomes easy.

Further, since a correct correction database for nuclear data can be constructed in a wide range by accumulating, storing and programming a large number of measured values, the neutron emission rate can be obtained with high accuracy in a region where there are no measured values. The method of predicting can be stored and programmed.

The invention according to claim 8 provides a computer,
The calculation accuracy evaluation value in the fourth step is previously set to Cm.
It is characterized in that the fifth step of inputting into the correction data table which is set so that the measured value and the calculated value of the generation amount of 244 and the neutron source intensity match.

According to the present invention, by inputting the calculation accuracy in the fourth step, it is possible to store and program the method for setting the measured amount of Cm244 and the measured value of the neutron source intensity so as to match the calculated value. .

The invention according to claim 9 provides a computer,
Am243 by adding a correction factor to the correction data table,
Alternatively, the corrected one-group reaction cross-section library prepared by revising the cross-sections of Pu242 or Am243 and Pu242 is fed back to the nuclide production / annihilation calculation code created in the third step.

According to the present invention, a correction factor is added to the correction data table to add Am243, Pu242, or Am24.
Corrected 1 created by revising the reaction cross section of 3 and Pu242
The grouping reaction cross section library is fed back and input to the nuclide generation / annihilation calculation code in the third step. This makes it possible to store and program the method for repeatedly determining the neutron emission rate of Cm244 with almost no effect on the neutron multiplication characteristics obtained by the fuel combustion calculation.

[0035]

DESCRIPTION OF THE PREFERRED EMBODIMENTS FIG. 1 shows a combustion chain of an actinide nuclide in a uranium fuel or a mixed fuel of uranium and plutonium. The process indicated by the bold line in FIG. 1 is a process that is particularly important for the production of the neutron emitting nuclide Cm244. When assessing the neutron source strength from spent fuel within a few decades after removal monitored at transport and reprocessing facilities, Cm
It can be seen from these figures that it is important to accurately evaluate the production amount of 244.

Among the processes shown by the bold line in FIG. 1, the neutron absorption cross section of the nuclide having a short half-life has little influence on the amount of Cm244 produced even if adjusted. On the other hand, the neutron emission rate is closely related to the neutron capture cross section of the neutron absorption cross section.

Therefore, in the embodiments of the invention according to claims 1 to 3, in evaluating the production amount of Cm244 without affecting the production amount of the nuclides less than Pu241, the neutron shown in FIG. Am243 or Pu242, or Am243 and Pu2 at the stage of combustion calculation combining the spectrum calculation code 2 and the nuclear data library 3.
Adjust the neutron capture cross section at 42. As a result, the evaluation generation amount of Cm244 obtained by the calculation for this adjustment is
The amount of Cm244 actually generated can be made to substantially match.

Since the composition data from U235 to Pu241 is important data that has a great influence on the conventional nuclear design and criticality safety evaluation, adjustment of the neutron capture cross section that affects these nuclide composition evaluations. Does not.

The neutron source intensity of spent fuel in a boiling water reactor (BWR) is shown in FIGS. 2 and 3. FIG. 2 is a comparative diagram of neutron source components in the spent UO ₂ fuel, and FIG. 3 is a comparative diagram of neutron source components in the spent MOX fuel.

The composition in UO ₂ fuel and MOX fuel is assumed in a typical case, U 235 enrichment is about 3 wt% for UO ₂ fuel and Pu enrichment is about 4 for MOX fuel.
It was set to wt% (Puf ratio of about 68%).

As is clear from FIGS. 2 and 3, in order to improve the evaluation accuracy of the neutron source intensity from the spent fuel,
It can be seen that it is effective to improve the evaluation accuracy of the production amount of Cm244.

As the Pu and Am isotopes having a long half-life in the nuclides after Pu241 shown in FIG. 1, Pu242 (half-life:
3.733 × 10 ⁵ years), Am241 (half-life: 432.2 years), Am24
2m (half-life: 141 years), Am243 (half-life: 7370) 4
A change in the amount of Cm244 produced when the capture cross section of the nuclide was increased by 10% was evaluated by combustion calculation.

FIG. 4 is a relative change diagram of the amount of Cm244 produced when the actinide capture cross section is increased by 10% with UO ₂ fuel, and FIG. 5 is a diagram for the same MOX fuel.

FIGS. 4 and 5 show that the process of the bold line shown in FIG. 1 is particularly important for the production of Cm244. In the embodiment according to the present invention, the effect of changing the trapping cross-sectional areas of Pu242 and Am243 by a certain amount during combustion will be described.

FIGS. 4 and 5 show that when the capture cross section of Pu242 or Am243 is increased by 10%, the Cm244 production amount is 8-10% and 7-10% respectively in the case of UO ₂ fuel, and in the case of MOX fuel. It shows that the amount of Cm244 produced can be increased by 8 to 10% and 6 to 10%, respectively, and that the amount of Cm244 produced in the latest research is 20 to 40.
% Of events are Pu242 or Am243, or Pu24
It is shown that this can be solved by increasing the corrected cross-sectional area of 2 and Am243 by about 20 to 50%. In order to improve the accuracy when the evaluation error of the Cm244 production amount has burnup dependency,
This can be solved by incorporating the burnup dependency in the cross-sectional area adjustment rate.

Next, a concrete embodiment of the method for evaluating the neutron emission rate from the spent fuel according to the present invention and the program for storing the method in the program will be explained with reference to FIG.
That is, as shown in FIG. 6, in the present embodiment, as a countermeasure when there is no measured value, the fuel core design code is set in the first step based on the combustion calculation data of the fuel assembly specification 1 to be evaluated. The multi-group neutron spectrum is evaluated by the lattice calculation code of the neutron spectrum calculation code 2 used.

That is, neutron spectrum calculation code 2
Is a group used to evaluate the neutron spectrum of the fuel to be evaluated by using the geometrical shape data of the combustion calculation data, the composition and temperature of the fuel and moderator, and the irradiation history data in fuel assembly specification 1 and used in the one-point combustion calculation code Evaluate the chemical reaction cross section.

Here, the neutron spectrum calculation code is a neutron flux of the calculation system under given calculation conditions (geometrical shape, composition and temperature of fuel and moderator, irradiation history) using the nuclear data library described later. It is a calculation code that can evaluate energy and spatial distribution.

The lattice calculation code is a neutron spectrum calculation code applicable to a fuel assembly system arranged in a lattice, and is a calculation code used for nuclear design of a fuel assembly.
Included as part of the former.

Next, in the second step, a group of neutron cross sections of actinide nuclides, which is important for evaluating the neutron emission rate by combining the multi-group neutron reaction cross sections registered in the nuclear data library 3, Chemical reaction cross section
It is calculated by equation (1).

Here, the nuclear data library is processed by the energy group structure set in the neutron spectrum calculation code so that various neutron reaction cross sections for nuclides handled in combustion calculation can be used in neutron spectrum calculation, This is a library of the information necessary for evaluating the nuclear characteristics in the combustion calculation.

The multi-group neutron reaction cross section is the upper limit of the neutron energy (eg, 20
MeV) is a neutron reaction cross section obtained by dividing into several tens to several hundreds.

The 1-group reaction cross-section library is a combination of the neutron flux and atomic number density obtained by the combustion calculation using the neutron spectrum calculation code, and the multi-group reaction cross-section and the 1-group reaction cross-section are used. This is a library of cross-sectional areas grouped so that the reaction rate is preserved in.
Note that nuclides and reactions that are of nuclear importance in the 1-group reaction cross-section are created as a function of burnup.

[0054]

[Equation 1]

Next, in the third step, the one-group reaction cross section obtained by the lattice calculation code of the neutron spectrum calculation code 2 is converted into a library 4 and is described later by the nuclide generation / annihilation calculation code 5. Perform calculation evaluation.

Here, the nuclide generation / annihilation calculation code is 1
Using the grouping reaction cross section library 4, the combustion equation of nuclides is solved considering all the generation and annihilation processes that occur during combustion, and the characteristics such as atomic number density, activation amount, calorific value, neutron source intensity of the nuclide are analyzed. A calculation code to be evaluated, which is ORIGE
The N code is a typical example.

In the third step, the neutron source intensity 12 can be obtained. The neutron reaction cross section generally changes with combustion, but the ones that have to consider the burnup dependency of the reaction cross section in the nuclear property evaluation are limited, so the 1 group reaction cross section in the second step Library 4
Is a typical burnup, and the nuclide reaction cross section for which burnup dependence should be considered is evaluated as a function of burnup.
It is used in the nuclide generation / annihilation calculation code 5 in the step.

The equation solved by the nuclide generation / annihilation calculation code 5 is expressed by the equation (2). dN _i (t) / dt = Σλ _{m → i} N _m (t) + ΣY _i ^j N _j (t) σ _f ^j φ <t> + N _k (t) σ _{k → i} (t) φ (t) + N _i (T) {Σλ _{m → i} + (σ _a ^j + σ _{n, 2n} ^j ) φ (t)} (2) where N _i (t): atomic number density λ _{m → i} of nuclide i at time t. : Decay constant Y _i ^j of nuclide m into nuclide i: Fission yield of nuclide i by nuclide j σ _f ^j : Fission cross section σ _a ^{j of} nuclide j σ _{n, 2n} ^j : Nuclide (n, 2n) cross section φ <t> of j: neutron flux at time t

As the nuclide generation / annihilation calculation code 5, for example, ORIGEN2 (AGCroff, "ORIGEN2-A Versatil
e Computer Code for Calculating the Nuclide Compos
itions and Characteristics of Nuclear Materials,
"Nucl. Technol., Vol. 62 pp.335-352, September 19
83.), the neutron emission rate (C) 6 from the fuel can be obtained. By comparing the neutron emission rate (C) obtained by this calculation with the neutron emission rate measurement value (E) 7 in the fourth step, the calculation accuracy (C /
E) 8 is obtained.

In the fourth step, Am243 or Pu242 or Am243 is used in the combustion calculation stage in which the neutron spectrum calculation code 2 and the nuclear data library 3 are combined as described above as the embodiment of the invention according to claims 1 to 3.
And the change in the neutron emission rate when the reaction cross section is corrected by adjusting the neutron capture cross section of Pu242 in advance in the correction data table 10. Then, the calculation accuracy (C
By correcting such that / E) 8 becomes 1, the neutron emission rate evaluation 9 of Cm244 can be performed with high accuracy.

The neutron emission rate (measurement value E) 7 is measured by measuring the neutrons generated from the fuel rod with a BF ₃ detector (converting the (n, α) reaction occurring in the detector into an electric signal) The absolute intensity is determined by comparison with the radiation source. Also, by irradiating the gold foil with neutrons and measuring the activation rate,
The source intensity can be determined.

The values calculated in the fourth step are input to the correction data table 10 in the fifth step. The correction data table 10 indicates whether the cross-sectional area of which nuclide should be changed in the preliminary examination to change the neutron emission rate. Information is registered.
That is, the correction data table 10 is preset so that the measured values and calculated values of the amount of Cm244 produced and the neutron source intensity match.

Therefore, C / E: correction of calculation accuracy,
In order to be able to carry out in combination with the embodiment of the invention according to claims 1 to 3, it is necessary to send C / E: calculation accuracy to the correction data table 10.

Then, the correction factors of the correction data table 10 are added to revise the cross-sectional areas of Am243, Pu242 or Am243 and Pu242, and the corrected one-group reaction cross-sectional area library 11 is Am243, Pu242, or Am243 and Pu242. The corrected capture cross section is fed back to the nuclide generation / annihilation calculation code 5 and input.

Here, the corrected 1-group reaction cross-section library is a predetermined amount of a part of the neutron cross-section stored in the 1-group reaction cross-section library used in the nuclide production / annihilation calculation code. It is a library.

The correction factor is a correction factor used when a corrected one-group reaction cross-section library is created, so that a calculation error of the Cm244 production amount obtained by using the nuclide production / annihilation calculation code does not occur. Used to As the correction factor, measured values and calculated values of the Cm244 production amount are stored in a database for various irradiation histories (burnup rate / void ratio).

According to the present embodiment, a correct correction database for nuclear data can be constructed in a wide range by applying it to a large number of measured values, so that the neutron emission rate can be obtained with high accuracy in a region without measured values. It can be predicted and can be programmed and stored.

The processing of each step relating to the evaluation of the neutron emission rate in each of the embodiments described above is realized by a computer-executable program, and the program is stored in a computer-readable storage medium. It can also be realized.

[0069]

According to the present invention, the amount of Cm244 produced, which is obtained by the combustion calculation used to evaluate the neutron emission rate, is actually produced without substantially affecting the neutron multiplication characteristic obtained by the combustion calculation. It can be almost matched with the production amount. In addition, the evaluation values based on Cm244-based calculation of the neutron production rate can be stably and easily maintained without compromising the results of the comprehensive improvement of the cross-section areas and decay schemes of many nuclides used in the calculation. You can get closer to the measured value.

Further, Am243 and Pu242 have almost no influence on the reactivity and neutron spectrum calculation in the nuclear characteristic calculation of the nuclear reactor, so there is no fear of impairing the results of the comprehensive improvement, and based on Cm244. The calculation accuracy of the neutron production rate can be improved. Furthermore, by storing various data, codes and libraries in the embodiment of the present invention and programming them, it becomes easy to evaluate the neutron emission rate from the spent fuel.

[Brief description of drawings]

FIG. 1 is an explanatory diagram of a neutron capture chain generated in a uranium and plutonium fuel for explaining an embodiment of the present invention.

FIG. 2 is a characteristic diagram showing relative intensity of neutron emission rate from spent UO ₂ fuel.

FIG. 3 is also a characteristic diagram showing relative intensity of neutron emission rate from spent MOX fuel.

FIG. 4 is a characteristic diagram similarly showing a relative difference in the amount of Cm244 produced in UO ₂ fuel.

FIG. 5 is a characteristic diagram showing a relative difference in the amount of Cm244 produced in MOX fuel.

FIG. 6 is a flow chart for explaining an embodiment of a method for evaluating a neutron emission rate from a spent fuel and a program therefor according to the present invention.

[Explanation of symbols]

1 ... Fuel assembly specification, 2 ... Neutron spectrum calculation code, 3 ... Nuclear data library, 4 ... Grouping reaction cross-section library, 5 ... Nuclide generation / annihilation calculation code, 6 ... Neutron emission rate (calculated value C), 7 … Neutron emission rate (measured value E), 8
... C / E (calculation accuracy), 9 ... Neutron emission rate evaluation of Cm244, 10 ... Correction data table, 11 ... One grouped reaction cross section library after correction, 12 ... Neutron source intensity.

   ─────────────────────────────────────────────────── ─── Continued front page    (72) Inventor Kenichi Yoshioka             2-1, Ukishima-cho, Kawasaki-ku, Kawasaki-shi, Kanagawa             Ceremony Company Toshiba Hamakawasaki Factory (72) Inventor Tsuyoshi Kikuchi             2-1, Ukishima-cho, Kawasaki-ku, Kawasaki-shi, Kanagawa             Ceremony Company Toshiba Hamakawasaki Factory F term (reference) 2G075 CA38 DA07 FA20 GA21

Claims (9)

[Claims] 1. A neutron-emitting nuclide Cm244 obtained by combustion calculation when a fuel assembly to be evaluated containing uranium fuel or a mixed fuel of uranium and plutonium is loaded into a reactor core and burned by neutron irradiation. In the method of evaluating the neutron emission rate from spent fuel for evaluating the amount produced, adjust the neutron capture cross-section of Am243 or Pu242 or Am243 and Pu242 at the stage of combustion calculation combining the neutron spectrum calculation code and the nuclear data library. Therefore, the method for evaluating the neutron emission rate from spent fuel is characterized in that the production amount of Cm244 actually produced is substantially matched. 2. A neutron-emitting nuclide Cm244 obtained by a combustion calculation when a fuel assembly to be evaluated containing uranium fuel or a mixed fuel of uranium and plutonium is loaded into a reactor core and burned by neutron irradiation. In the neutron emission rate evaluation method from the spent fuel to evaluate the production amount, the neutron absorption cross section of Am243 or Pu242, or Am243 and Pu242 is set to be constant at the stage of combustion calculation combining the neutron spectrum calculation code and the nuclear data library. ,
A method for evaluating the neutron emission rate from a spent fuel, characterized in that the production amount of Cm244 actually produced is made substantially equal by adjusting the neutron capture cross section. 3. A neutron-emitting nuclide Cm244 obtained by a combustion calculation when a fuel assembly to be evaluated containing uranium fuel or a mixed fuel of uranium and plutonium is loaded into a reactor core and burned by neutron irradiation. In the method for evaluating the neutron emission rate from spent fuel for evaluating the amount produced, the neutron capture cross-section of Am243 or Pu242, or Am and Pu242 was burned up at the stage of combustion calculation combining the neutron spectrum calculation code and the nuclear data library. Cm244 actually generated by adjusting as a dependent amount
A method for evaluating the neutron emission rate from spent fuel, which is characterized in that the amount of neutrons produced is approximately the same. 4. A neutron spectrum of a fuel to be evaluated is evaluated by a neutron spectrum calculation code based on combustion calculation data of a fuel assembly specification to be evaluated.
Step, a second step of combining the neutron spectrum calculation code of the first step with a nuclear data library to create a one-group reaction cross-section library of neutron cross-sections of actinoid nuclides, and one group created in this second step 3rd step of creating nuclide generation / annihilation calculation code from the chemical reaction cross-section library, and calculated neutron emission rate based on the nuclide generation / annihilation calculation code created in this 3rd step And a fourth step for evaluating the neutron emission rate of Cm244 by comparing the calculated neutron emission rate with Cm244 to evaluate the neutron emission rate from the spent fuel. 5. The method further comprises a fifth step of inputting the calculation accuracy evaluation value in the fourth step into a correction data table which is set in advance so that the calculated values of the amount of Cm244 produced and the neutron source intensity match the calculated values. The method for evaluating the neutron emission rate from spent fuel according to claim 4, wherein 6. The corrected one-group reaction cross-section library prepared by revising the reaction cross-sections of Am243 or Pu242 or Am243 and Pu242 by adding a correction factor to the correction data table in the third step. The method for evaluating the neutron emission rate from spent fuel according to claim 5, wherein the method is fed back to the created nuclide generation / annihilation calculation code. 7. A first step in which a neutron spectrum of a fuel to be evaluated is evaluated by a neutron spectrum calculation code on a computer based on combustion calculation data of a spent fuel prepared based on specifications of a fuel assembly to be evaluated. , The second step of combining the neutron spectrum calculation code of the first step with a nuclear data library to create a one-group reaction cross section of the neutron reaction cross section of the actinide nuclide, and the one-group reaction created in the second step 3rd step of creating nuclide production / annihilation calculation code from the cross-sectional area, and calculated neutron emission rate based on the nuclide production / annihilation calculation code created in this 3rd step, and compare with the measured neutron emission rate And calculate the calculation accuracy, Cm24
A program for executing the fourth step of evaluating the neutron emission rate of 4 and. 8. The computer inputs the calculation accuracy evaluation value in the fourth step into a correction data table which is set in advance so that the calculated values of the amount of Cm244 produced and the neutron source intensity match the calculated values. The program according to claim 7, which is for executing a step. 9. A computer adds a correction factor to the correction data table and Am243 or Pu242, or A
Corrected 1 created by revising the cross-sectional area of m243 and Pu242
The program according to claim 7, wherein the grouping reaction cross-section library is fed back to the nuclide generation / annihilation calculation code created in the third step.