Corresponding author: Artem V. Sobolev (SobolevArtem82@gmail.com)

Academic editor: Yury Kazansky

The paper discusses the stages of calculating the radiation safety of spent nuclear fuel (

All the problems presented in the paper are currently being solved by means of rather complex and voluminous calculations that take a long time. In order to be able to conduct a preliminary assessment of the radiation situation around the transport casks, the authors propose to create a methodology that will determine the type of interrelations between the maximum effective dose and input parameters, such as fuel burnup, decay, fuel composition, protection material in the

Today, ensuring nuclear and radiation safety during the

For a long time in the field of designing

Analyzing nuclear and radiation safety of

To develop this methodology, the following steps are required:

to estimate, based on existing data or calculations, the sources of gamma rays and neutrons that are formed during the burning of the most common types of fuel in Russian power reactors;

to analyze the dynamics of changes in the intensity of the sources and their spectral distribution depending on the burnup depth and cooling time;

to evaluate the contribution of FA materials to the source formation.

Let us now consider the problem of the formation of a source of neutrons and gamma rays. According to existing studies, neutron sources for uranium and MOX fuel irradiated in a VVER-1000 reactor with a three-year cooling time are formed due to the set of isotopes presented in Tab.

Isotopes determining a neutron source during three-year cooling (

Pu-238 | Pu-240 | Pu-242 | Cm-242 | Cm-244 | Cm-246 | Cf-252 |
---|---|---|---|---|---|---|

~0.1% | ~0.2% | ~0.1% | ~1% | ~96% | ~1% | ~1% |

A gamma source during ^{85}Kr, ^{90}Sr, ^{90}Y, ^{106}Rh, ^{125}Sb, ^{134}Cs, ^{137}Cs, ^{137m}Ba, ^{144}Ce, ^{144}Pr, ^{147}Pm, ^{154}Eu, ^{155}Eu, which are fission products (^{60}Co, which is contained in the stainless steel of the FA bottom fittings. The difficulty in estimating the dose rate from ^{60}Co is associated with the fact that there is no explicit normalization of the amount of ^{59}Co in the stainless steel composition of which the FA bottom fittings are made. According to GOST 5632-72 and GOST 5632-2014 (GOST 5632-72, GOST 5632-2014), the cobalt content is not explicitly standardized for austenitic steel grade 12X18H10T, and the limiting value of the mass fraction of 0.5% is given for nickel and iron-nickel alloys to which this steel does not belong. Therefore, there is a need to obtain information on the amount of ^{59}Co in steel grades used for manufacturing FA structural materials.

Despite the existence of a list of isotopes of gamma and neutron sources for a specific cooling time and burnup depth, there is no generalized information about the change in the group of isotopes and their contribution to the formation of gamma sources when these parameters change. From here follows the first important problem, i.e., the determination of the sources of neutron and gamma radiation for fuel assemblies with the values of fuel burnup, cooling time and compositions used at NPPs.

When calculating fuel burnup, it is possible to use software systems that calculate it either by solving a system of point burnup equations or taking into account the fuel geometry in the material, where it is possible to divide the burnable material into separate components that burn up in a specific neutron flux. It can take a large amount of time to form a new detailed model for each case or to calculate it using a computer. Therefore, it becomes necessary to solve the second problem, i.e., assessing the effectiveness of replacing a detailed model with a simplified one using homogenization. The solution to this problem is important both for calculating radiation sources and, therefore, calculating dose rates on the transport cask surfaces.

It is also necessary to clearly define a set of materials that are most suitable for protection against both neutron and gamma radiation. The protection material and its geometry are selected individually for each type of gamma radiation source. If we consider a point source, the expression characterizing the equivalent dose rate at the point behind the shield will look as follows (

where _{δ} is the constant kerma, aGy·m^{2}/(s×Bq); _{d}^{p.s.} is the dose accumulation factor of a point isotropic source;

Block diagram of a transport cask for the transportation and storage of spent nuclear fuel (the author is JSC Design Bureau for Special Machine-Building) (

At present, the attenuation of gamma and neutron radiation is most often calculated using the Monte Carlo-based software (

Due to uneven axial fuel burnup and the presence of ^{60}Co in the FA bottom fittings, the overall picture of the effective dose rate is uncertain and the position of the maximum value level can change. However, when designing transport casks, it is necessary to have information in advance about the region with the highest radiation level, which will make it possible to orient the cask design to the greatest attenuation of radiation at this level. For example, in Fig. ^{60}Co.

An example of the uneven axial effective dose rate for transport casks.

To fully cover the considered problem and its causes, we shall draw up a so-called ‘root cause analysis diagram’ (Fig.

The results of the regression analysis will make it possible to use the obtained dependences for initial estimates of a specific actual load of transport casks. Since ready-made dependences will be given, the estimation procedure will become more convenient and simple. Initially, the dependences can be obtained on the basis of numerical calculations of models with subsequent verification based on performed measurements. Since the scale of SFA shipments will grow due to the construction of new units, the emergence of a convenient estimation technique will accelerate the process of calculating radiation safety of

Root cause analysis diagram for assessing the effective dose on the transport cask surface.

The authors prepared the design model – TUK-141O (Fig.

Vertical and horizontal sections of the TUK-141O design model.

Based on the above, conclusions can be drawn regarding the problem of the calculation analysis of radiation safety of

* Russian text published: Izvestiya vuzov. Yadernaya Energetika (ISSN 0204-3327), 2019, n. 4, pp. 3–48.