S. Breidokaitė, G. Stankūnas, and A. Tidikas

Laboratory of Nuclear Installation Safety, Lithuanian Energy Institute, Breslaujos 3, 44403 Kaunas, Lithuania

Received 14 October 2019; revised 25 October 2019; accepted 31 October 2019

Nuclear safety assessment in nuclear fusion devices relies on the Monte Carlo method based neutron transport calculations. This paper presents information about the calculation results of the activities and dose rates caused by neuron irradiation for the structural materials of the high flux test module sample holder of IFMIF-DONES. The neutron induced activities and dose rates at shutdown were calculated by means of the FISPACT-2010 code with data from the EAF-2010 nuclear data library. Neutron fluxes and spectra were obtained with MCNP neutron transport calculations. The activities and dose rates were calculated at the end of irradiation of the assumed device operation scenario for cooling times of 0 s – 1000 year. In addition, radionuclides with contribution of at least 0.5% to the total value of activation characteristics at the previously mentioned cooling times were identified. After the operation, the most active radionuclide is 55Fe, with an activity share ranging from 30% (M200) to 63% (M8), and at the end of the prediction it accounts for 86% of the total activity. The highest dose rates at the end of irradiation are attributed to 56Mn radionuclide. 54Mn and 60Co are the most dominant radionuclides during intermediate and long cool-down periods.

Keywords: IFMIF-DONES, fusion, neutron irradiation, MCNP, FISPACT

1. Introduction

Nuclear energy is a relatively clean energy source in terms of carbon footprint in comparison to widely used fossil fuels. There are two main nuclear processes that can result in net energy gain: fusion and fission. Although nuclear fission based technologies currently play an important role in the  energy market, they have issues with radiation protection, nuclear waste storage and limited fuel resources. Nuclear fusion offers the alternative that mitigates the  formerly mentioned issues to more manageable levels [1]. On the other hand, nuclear fusion technologies face their own challenges, for example: fusion devices are very complicated and achieving sustainable fusion on earth is not an easy task  [2]. In order to ensure safety of fusion devices, most times Monte Carlo method based neutron transport calculations are performed. The Monte Carlo method is based on simulation of individual particle histories in complex environment. Particle histories follow the established laws of physics; however, the outcome is decided by random events that are described by probability distributions [3].

There are many possible fusion reactions that result in energy release. At the moment, the most promising and feasible for the controlled thermonuclear process are reactions between hydrogen isotopes (deuterium and tritium) and helium [4]. Most of the energy released during D-T fusion reactions is carried by neutrons. Self-sustaining nuclear fusion can only be achieved if certain particle densities, temperatures and confinement times are present. The cross-section of the D-T reaction is the highest at the selected plasma temperature (around 100 million degrees) [5].

Based on the  preliminary engineering project of IFMIF (International Fusion Materials Irradiation Facility), the  European Union has proposed a  plan for the  development of the  DONES (DEMO-Oriented Neutron Source) nuclear fusion facility [6]. IFMIF-DONES neutrons will be generated during the Li (d, xn) reaction by liquid lithium target bombardment with deuterium ions (up to 40  MeV). The  device will be capable to produce energetic neutrons and will employ acceleration current (125 mA) in order to procure high neutron flux  [>1018 m−2 s−1] in order to achieve exposure conditions relevant to nuclear fusion for material testing and qualification. Neutron exposure in material samples will result in structural damage and activation. Activation will not be limited to samples as a large part of the facility will be exposed to neutron irradiation as well. The highest neutron intensity will be at the Test Cell (TC) (Fig. 1) [7] and its High Flux Test Module (HFTM) [8] where material samples are held (for more details see Figs. 2, 3) [9]. Estimation of activation in TC and HFTM is critical for the operation and maintenance of the device as it could lead to radiological hazards for staff and sensitive electronics.

2. Computational methods

The MCNP (Monte Carlo n-particle) particle transfer code was developed by the Los Alamos National Laboratory in the  United States. It was programmatically written using the FORTRAN and C programming languages.

For neutron spectrum calculations the neutron transport equation must be solved:

1 v ϕ t +Ωϕ+ t ϕ( r,E )= 0 d E Ω s ( r, z ) f( r, E E, Ω Ω )ϕ( r, E , Ω ,t )d Ω + x( E ) 4π 0 d E Ω v f ( r, E ) ϕ( r, E , Ω ,t )d Ω +S( r,E,Ω,t ).( 1 )
Fig. 1. Model of the  Test Cell including the  High Flux Test Module [7].
Fig. 2. IFMIF-DONES High Flux Test Module. From left to right: a downstream view, an upstream view, a downstream view but cut in the middle of the container with the global coordinate system of IFMIF-DONES [8].
Fig. 3. MCNP model of the IFMIF-DONES HFTM sample holder.

The MCNP uses three processes to solve neutron transport equations:

(1) using the probability distribution determines the source parameters;

(2) neutron location, energy and direction tracking;

(3) data recording and analysis of results [10].

The FISPACT-2010 is describing the amounts of various nuclides in materials after given irradiation steps. It can be achieved by solving the  Bateman differential equations

d N i ( t ) dt = N i ( λ i +σϕ )+ N j ( λ ij + σ ij ϕ ) + S i ,( 2 )
S i = k N k σ k f ϕ Y ik , ( 3 )

where Ni is the amount of nuclide i at time t, λi is the decay constant of nuclide i, λij is the decay constant of nuclide j producing i, σi is the total cross section for reaction i, σij is the reaction cross-section for reaction j producing on i, σkf is the fission cross-section for reactions on actinide k, ϕ is the neutron flux, Si is the source of nuclide i from fission, and Yik is the yield of nuclide i from the fission of nuclide k. The  set of the  previously mentioned differential equations is solved using the Sidell method. Due to the dependence on the duration of irradiation and cooling sequences only nuclides with sufficiently long half-lives are calculated by this method, and nuclides with shorter half-lives are considered to be in equilibrium [11].

3. Calculation results

Neutron spectra (Fig. 4) were determined by using the IFMIF-DONES neutron source and geometry model. Neutron fluxes and energy distribution were obtained for the structure of the HFTM sample holder. All neutron transport calculations were done by using MCNP5 with the  JEFF-3.1.2  [12] nuclear library.

Activation analysis and radionuclide identification were performed with the  FISPACT-2010 code with the EAF-2010 nuclear data library [13].

JEFF-3.1.2 is the  general purpose fusion and fission nuclear data library, while EAF-2010 is the nuclear data library optimized for fusion applications. A concise structure of EAF-2010 enables a shorter computation duration.

For activation calculations the IFMIF-DONES irradiation scenario was employed that assumes facility operation for 10 years. Each year consists of 345 days of continuous irradiation and 20 days of downtime representing the maintenance work. Specific activities and dose rates were obtained and dominant radionuclides were identified for HFTM. Only radionuclides contributing more than 0.5% to the total value after any of the considered time interval are presented.

Activities and dose rates were calculated for the HFTM sample holder structure, which consists of 14 materials used for structural integrity and thermal insulation: SS316L (N), EUROFER 97 and several compounds such as M40 (EUROFER 97 | AISI 321 | MgO | NiCr 80/20 | INNOBRAZE ML 442), M200 (Silica Aerogel, Si (OCH3)4) and M8 (Fe, Cr, W, Mn, and some other elements). 11 of the investigated materials were iron based (see Table 1).

The specific activity dependence on time is shown in Fig. 5. At the end of irradiation, 55Fe is the most active radionuclide and represents 30% (M200) to 63% (M8) of the total activity. In addition to this, the contribution of 54Mn and 56Mn is 7% (M200) to 13% (M8) of total activities. Although the  half-life of the  55Fe iron isotope is relatively long (half-life is 2.734 yr), it remains the most relevant isotope for almost all prediction time period. A more detailed analysis of the obtained results shows that 55Fe is produced from the 54Fe neutron capture reaction with gamma radiation as a by-product. The neutron capture reaction is responsible for 56Mn production from 55Mn.

Fig. 4. Neutron spectrum.
Table 1. Chemical composition of the investigated materials.
M1 % M2 % M3 % M5 % M8 % M30 % M31 % M32 % M33 % M40 % M200 %
Fe 62.974 Fe 65.4 Fe 47.442 Fe 61.645 Fe 87.974 Fe 84.777 Fe 84.857 Fe 84.94 Fe 85.02 Fe 74.9 Fe 26.6663
Cr 18 Cr 16.998 O 33.035 Cr 18.5 Cr 9.5 Cr 9.155 Cr 9.163 Cr 9.172 Cr 9.181 Cr 10.65 O 24.1705
Ni 12.5 Ni 11.998 Ca 7.103 O 13 W 1.199 Na 3.634 Na 3.544 Na 3.449 Na 3.358 Ni 7.723 C 17.2439
Mo 2.699 Mo 2.499 Ti 5.434 Mn 2.499 Mn 0.6 W 1.155 W 1.156 W 1.158 W 1.159 Mg 2.298 Si 10.2879
Mn 2 Mn 2 Si 2.575 Mn 2 V 0.25 Mn 0.578 Mn 0.579 Mn 0.579 Mn 0.58 O 1.52 Cr 7.4234
Cu 1 Si 0.75 Al 2.348 Cu 1 Ta 0.15 V 0.241 V 0.241 V 0.241 V 0.242 W 0.916 Ni 5.9955
Si 0.5 Nb 0.1 Mg 0.934 Si 1 C 0.12 Ta 0.144 Ta 0.145 Ta 0.145 Ta 0.145 Mn 0.686 H 4.3041
Other 0.322 Other 0.255 Other 1.129 Other 0.357 Other 0.19 Other 0.295 Other 0.295 Other 0.296 Other 0.296 Other 1.283 Other 3.8842
Fig. 5. Specific activity at HFTM for M40.

The main radionuclides that are gamma radiation sources are presented in Fig. 6. As we can see, the largest contribution to the equivalent dose rate is provided by 56Mn, but due to its short half-life (2.64 h), 58Co (half-life is 70.86 d.) and 54Mn (half-life is 312.2 d.) are relevant for longer times. From 10 to 100 years due to the long half-life (5.273 m), 60Co emits almost 100% of the total dose.

Fig. 6. Dose rate at HFTM for M40.
Fig. 7. Specific dose rate at HFTM for EUROFER 97.

In Fig. 7, we can see that the main dominant radionuclides are the  same as in the  material in Fig. 6; however, in Fig. 7 there are some differences in radionuclides responsible for the total dose rate. The  main contributors are 56Mn and 54Mn, while 60Co is insignificant and amounts to less than 0.5% of the total dose rate.

As seen in Fig. 8, in the  first second after the  end of irradiation, the  increase in the  iron content increases the activity of the 55Fe isotope fraction in the  material. The  highest activity (272.8 TBq/kg) is reached in the AISI SS316L material with 65% iron content. 51Cr has the greatest influence on SS316L. For the activity of M40 (concentration of Fe in material is ~75%), in addition to 51Cr, 56Mn and 54Mn are also significant sources.

Figure 9 depicts the  iron percentage and related specific activity values present for 181 days of the observation time. During this time period, the influence of iron activation products (mainly 55Fe) increases with regard to the total specific activities. It should also be noted that the  activity of iron isotopes (55Fe and 53Fe) ranges from 65% (M200) to 86% (M8). The most active substance is M8, containing 87.97% of iron, and has a total activity of 159.1 TBq/kg after 181 d.

Fig. 8. Specific activity dependence on the iron content in the material (1 second of cooling).
Fig. 9. Specific activity dependence on the iron content in the material (181 days of cooling).

4. Summary

The objective of this study was to calculate the activities and dose rates in the HFTM sample holder structure at the IFMIF-DONES induced by neutron irradiation. Subsequent activities and dose rates at shutdown were calculated by means of the  FISPACT-2010 code using the  irradiation scenario specified for the  IFMIF-DONES. After the end of irradiation, the activities and dose rates were calculated at the cooling time of 0 and 1 s, 5 and 30 min, 1, 3, 5 and 10 h, 1 and 3 days, 2, 4 and 8 weeks, 181 days, 1, 10, 100, 300 and 1000 years.

The investigation of the  iron based materials shows that the total dose rate values range from 2.5 ∙ 104 to 4 ∙ 104 Sv/h within the first second after the end of irradiation and drops to 1.05 · 10–3– 5.92 · 10–2 Sv/h, respectively, after 1000 years of cooling.

The specific activity of 55Fe in the metal samples ranges from 4.92 · 1013 to 1.57 · 1014 Bq/kg at the first second of the cooling time and from 1.54 · 108 to 4.95 · 108 Bq/kg after 50 years of the cooling time. The highest activity value belongs to M8. After 50 years, no activity of 55Fe was observed.

The total specific activity in the metal samples ranges from 1.63 · 1014 to 2.74 · 1014 Bq/kg at the first second of the cooling time and from 1.72 · 108 to 5.21 · 1010 Bq/kg after 50 years of the cooling time. The highest activity values belong to M8.

At 0 second of the cooling time, the most active radionuclide is 55Fe: its activity ranges from 30% (M200) and 63% (M8) of the total activity up to 86% at the end of the observation. 56Mn is the radionuclide emitting the highest doses, but it decays fast and due to the longer half-life the dominant nuclides at the end of the cooling time are 54Mn and 60Co.


This work has been carried out within the frame work of the EUROfusion Consortium and has received funding from the  Euratom Research and Training Programme 2014–2018 and 2019–2020 under Grant Agreement No. 633053. The  views and opinions expressed herein do not necessarily reflect those of the European Commission.


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S. Breidokaitė, G. Stankūnas, A. Tidikas

Lietuvos energetikos institutas, Kaunas, Lietuva


Vertinant branduolių sintezės prietaisų saugumą yra atliekami neutronų pernašos lygties skaičiavimai. Dalis jų pagrįsti Monte Karlo metodu. Straipsnyje pateikiami aktyvumų ir dozės galių skaičiavimų rezultatai, kuriuos lemia neutronų apšvita IFMIF-DONES didelio neutronų srauto testavimo modulio bandinių laikiklio struktūrinėse medžiagose. Bandinių laikiklį sudaro 14 medžiagų (11-oje iš jų pagrindinis elementas yra geležis): SS316L (N), EUROFER 97, M40 (EUROFER 97 | AISI 321 | MgO | NiCr 80/20 | INNOBRAZE ML 442), M200 (Silica Aerogel, Si (OCH3)4), M8 (Fe, Cr, W, Mn ir kiti elementai). Neutronų sukelti aktyvumai ir dozės galios įrenginio eksploatavimo metu buvo apskaičiuoti pritaikius FISPACT-2010 kodą su EAF-2010 branduolinių duomenų biblioteka. Neutronų spektras buvo gautas naudojant MCNP kodą neutronų pernašos lygties skaičiavimams. Aktyvumai ir dozės galios buvo apskaičiuotos aušimo laikams nuo 0 s iki 1000 metų. Taip pat buvo nustatyti radionuklidai, turintys mažiausia 0,5 % bendro aktyvumo vertės anksčiau minėtais aušimo laikais. Stebėjimo pradžioje geležies tipo medžiagose aktyviausias radionuklidas buvo 55Fe, kurio aktyvumo dalis medžiagose svyruoja nuo 30 % (M200) iki 63 % (M8), o stebėjimo pabaigoje sudaro 86 % bendro aktyvumo. Didžiausia dozės galia pasižymi radionuklidas 56Mn, o ilgesniais aušimo laiko periodais – 54Mn ir 60Co.