The investigation of the neutronic responses of components in the K-DEMO

The neutronic response of consisting materials of the Korean fusion DEMOnstration reactor (K-DEMO) in its fusion neutron environment is a crucial consideration factor from the conceptual design stage of the K-DEMO. Especially in the design of in-vessel components (IVCs) of the K-DEMO that are placed in the most extreme neutron irradiation field, neutronic damage of constituent components is a major limiting factor that determines the lifetime of IVCs. From the analysis of the neutronic response of IVCs of the K-DEMO, the neutron wall loading (NWL) related to the tritium breeding ratio and nuclear heating of IVCs can be quantified to assess the self-sufficient supply of tritium, and thermal energy transferred from fusion neutrons, respectively. The calculated NWL shows that the harmonizing design of the cooling configuration of each blanket segment with the corresponding NWL is critical thermos-hydraulic design issue for the efficient utilization of thermal energy in the blanket. Another finding is that the double null magnetic field configuration and related blanket configuration with a water-cooled pebble bed of the K-DEMO make a self-sufficient tritium supply challenging. The implicated lifetime of the first plasma-facing tungsten wall of the K-DEMO is around 2 full power years (FPY) in the severely neutron-irradiated region. For the reduced activation ferritic/martensitic steel layers in the blanket, the lifetime of it is estimated around 4 FPY in the inboard region. Based on the response analysis results of this study, optimization of the design of the K-DEMO will continue iteratively in the future.


Introduction
The electrical power generation using neutron energy comes from a fusion of deuterium (D) and tritium (T) nuclides adopts the concept of transferring neutron energy from the D-T fusion process to the primary cooling system and transferring it to the Original content from this work may be used under the terms of the Creative Commons Attribution 4.0 licence. Any further distribution of this work must maintain attribution to the author(s) and the title of the work, journal citation and DOI. next staged cooling system to produce electricity by driving a thermo-hydraulic turbine.
In the conventional fission nuclear power plant (NPP), most of the neutron energy (mean energy: ∼2 MeV) generated by nuclear fission is directly transmitted to the primary coolant and there is little concern about the structural damage of consisting components of the NPP by neutron bombardments. Contrarily, neutron energy (14.1 MeV) comes from nuclear fusion transferred to the structure named blankets that functions as the fuel breeding (by 6 Li (n, α) T reaction) and energy accumulation medium first.
The energy transferred to blanket structures is transferred to the next cooling medium (water, He, or LiPb, etc) by conduction, convection and radiation. The energy stored in these second cooling mediums can be reintroduced into the heat exchange system, such as the mimicked NPP's secondary cooling loop to drive a thermo-hydraulic turbine to produce electricity.
During this staged energy transfer, the neutrons generated by the D-T fusion impinge on the components that comprise the blanket structure and transmit its energy to increase the blanket's temperature in the first stage. The energy transferred by fusion neutrons can also destroy the mechanical integrity of each component of the blanket due to its higher energy compared to the energy of a nuclear fission neutron. In addition, they can produce gases such as helium and hydrogen by (n, α) and (n, p) reactions. These products can affect the reweldability and structural soundness of constituting material of blankets. The deterioration of the composition material of fusion facilities like the Korean fusion DEMOnstration rector (K-DEMO) [1] by fast fusion neutrons makes functional and technical problems related to the economics of the nuclear fusion facility.
The more serious concern related to the willing acceptance of nuclear fusion energy by the public is that dense highenergy fusion neutron flux in the nuclear fusion facility makes its components highly radioactive. Based on the level of radioactivity of the material, a certain fraction of radioactive material should be classified as radioactive waste (RW) as enforced by the law in most countries.
The roles in the generation of energy and fuel are a positive aspect of fusion neutrons, but the nuclear damage and the gas generation that destroy the mechanical integrity of constituent components of the nuclear fusion facility itself are one of the other pessimistic roles of fusion neutrons.
The more pessimistic role is making a huge amount of intermediate-level RW that strongly depends on the accumulated neutron irradiation time and fluence. The detailed conservative classification of RW in the K-DEMO with 20 full power years (FPY) of continuous irradiation duration was evaluated in [2].
A meticulous assessment of the pessimistic impact of fusion neutrons on mechanical integrity needs to be done to anticipate the operational lifetime of the nuclear fusion facility. The positive impact of fusion neutrons also needs to be evaluated to assess the facility's availability and costeffectiveness.
As a result, the evaluation of neutronic damages that affect the structural integrity of the blanket, vacuum vessel (VV), toroidal field coil (TFC), and others are considered essential prerequisites for the design of the K-DEMO.
The most important evaluation values for neutronic damages include displacement per atom (DPA), helium production rate, etc. Table 1 lists the suggested limits for these parameters and others related to the positive impact of fusion neutrons.
After the preliminary evaluation of DPA/FPY in the divertor of the K-DEMO [3], this work reports the results of a further detailed evaluation of the nuclear damages in the K-DEMO.
The Monte Carlo N-Particle radiation-transport (MCNP) code [4] was used as the basic code for neutron transport calculation and damage assessment for the present studies. Another widely used inventory calculation code, FISPACT-II [5], which provides the DPA/FPY and various gas production rates, was also used to compare the results. The discussion on the differences in results between the codes is described later.
Two codes have some distinctions in calculation procedures. The MCNP performs neutron transport and damage simulation within the code. So, it does not account for the change of material composition by the neutron reaction in the simulation.
However, FISPACT-II is an independent inventory simulation code requiring the neutron spectrum obtained by another independent simulation code, such as MCNP as a mandatory input.
The change in the material composition as a function of the specified irradiation time step can be estimated from the calculation results of FISPACT-II.

Set-up for neutronic response simulation with MCNP
The input model for MCNP simulation is developed and upgraded in series as described in detail [2] for the neutronic damages and gas production rate calculation in the K-DEMO. With this model, other neutronic parameters like tritium breeding ratio (TBR), and nuclear heating on the TFC are also evaluated.
The overall structural configuration of the K-DENO is presented in figure 1. The toroidal and poloidal symmetry of the K-DEMO structural configuration makes this simple 11.25-degree half model sufficient for the analysis of neutronic responses. The reflecting boundary conditions are used on both toroidal and upper horizontal surfaces.
The detailed description of the model is as follows. The main component of the water-cooled pebble bed tritiumbreeding blanket (TBB) is the 90% 6 Li enriched tritium breeding layer (TBL) within the supporting reduced activation ferritic/martensitic (RAFM) steel structure, as shown in the There are ten RAFM steel layers and nine TBLs with the mixed beryllium oxide ceramic bed between the RAFM steel layers in the TBB of the IB region. As the first direct plasmafacing wall, tungsten (W) with vanadium (V) filler material is installed in the front of the RAFM steel layer. There are three poloidal segments (PSs) of TBB structures along the poloidal direction in the IB region.
The first segmented TBB is a poloidal segment number (PSN) 1 that is placed just under the equatorial line of the IB region. The next PSN increases clockwise in figure 1. Four segmented TBB structures along the poloidal direction are placed in the outboard (OB) region, as shown in figure 1.
The PSN 16 is located just under the equatorial line of the OB region. Behind the TBB, the layered 2% borated stainless steel (B-SS) within the structural container made by the RAFM steel including the water constitutes the shielding blanket (SB). The VV region is enclosing the SB. The 16 TFC structures are outside of VV.

Set-up for the calculation of the neutronic damages and gas production rate with FISPACT-II
For the inventory calculation including the DPA/FPY and gas production rate using FISPACT-II with the TENDL2017 nuclear data library, pre-required neutron flux spectra were simulated by using MCNP with FENDLE 3.0 nuclear library.
The representative spectra in the concerned region are presented in figure 2. Due to the limited space in the IB region, SB's dimension in this region is smaller than that in the OB region. From this, it follows that the neutron flux of the IB VV and TFC is more intense than those of the OB VV and TFC. The neutron irradiation time is the input parameter of FISPACT-II with the neutron flux spectrum and total neutron intensity.
The accumulative irradiation times are specified from 1 s to 20 FPYs. For the specific mass (1 kg) of the concerned material in the interested irradiation region, the time evolution of neutron damage, the gas production rate, and material composition are estimated.

Simulation results of neutron wall loading (NWL) and TBR
3.1.1. NWL. Figure 3 shows the NWL according to the PSN of the TBB that is simulated by the model of figure 1 at the nuclear fusion power of 2.2 GW.
From these relative differences in the NWL along the poloidal direction, the cooling channel design of each PS of the TBB needs to be harmonized to align the outlet temperature of the coolant. The detailed specifications of cooling channel size, coolant velocity, and pressure are under design evolution.
The value of * F2 tally of MCNP in the W wall surface of each PS is used to infer these NWL variances along the poloidal direction. The peak NWL in the PSN 16 is 2.48 MW m −2 and the average NWL is 1.73 MW m −2 . The similarity of the NWL profiles can be found in the previous result that was estimated by in-house code [11]. .5-degree model with the pure material composition [12]. Even though the model was optimized as described in [12] to increase the TBR, the TBR in the old model is under the suggested value of 1.05. In the new model, there was some modification in the equatorial port area in the OB region to enhance the neutron shielding. In addition, the exclusion of the tritium-breeding region considering the engineering challenges for the installation of the complicated tritium-breeding region just around the port is incorporated into the new model. As shown in the TBR of PSN 16 in the new model with light water, it reduces the TBR in this region up to −0.0421. The TBR in the divertor region was also decreased due to the re-positioning of the tritium breeding zone from the old position shown in [3] to the region far from the plasma. As result, the total TBR in the new model is less than that of the old model.
The change of cooling water from light water to heavy water to increase the TBR is a typical measure. The impact of this change is also listed in table 2. The TBR 1.061 is larger than the recommended one. But the heavy water coolant makes another concern on the generation of tritium in the coolant by D(n, γ)T reaction [13]. The safety management related to the water detritiation system of coolant will be an extra burden to be resolved.
The gap between the neighbouring toroidal modules of the TBB of the K-DEMO is optimized to 1 cm for enhancing the TBR and neutron shielding.
But this gap still needs further optimization to facilitate access for the remote handling (RH) equipment and respond to the thermal expansion of the blanket module. The double null magnetic configuration of the K-DEMO is the main reason for this low value of TBR as compared to the optimistically recommended value of 1.05 with light water coolant. The design optimization of the K-DEMO to increase the TBR will be a major effort in the ongoing design evolution if not the tritium-breeding concept of the K-DEMO adopts another one as suggested in [14].
The variation of the TBR in each TBL along the radial direction is also investigated to see the dependence of the TBR on the radial neutron spectra changes. The inventory calculation to investigate the detailed variation of the atomic composition of constituent materials including the production of tritium in the TBL along the radial direction in PSN 16 is performed. The used flux spectra of the neutron in the interested layers for FISPACT-II inventory simulation are presented in figure 4.
The calculations of the DPA and the helium production in the RFFM steel layers are also done using the spectrum in figure 4. The increasing moderation effect according to the increase of radial distance from the plasma is clearly shown in the cumulative fraction distribution curve for each component in PSN 16. The reduction in the low energy fraction of less than 0.01 MeV neutron energy in the TBL is distinguishable because of the tritium breeding reaction. The evolution of the atomic concentrations of the representative isotopes as a function of irradiation time is shown in figure 5. Before the irradiation, the concentration of 9 Be is 8.12 × 10 5 appm and that of 6 Li is 4.35 × 10 4 appm. After 20 FPYs of irradiation, the concentrations of 6 Li in 1st TBL, 6th TBL, and 10th TBL are 6.9784 × 10 3 appm, 2.38 × 10 4 appm, and 4.19 × 10 4 appm, respectively. The change of the 6 Li fraction during operation may alter the TBR according to the operation time because the relative fraction of 6 Li and 9 Be has an impact on the TBR. After 20 FPYs of operation, the 6 Li atomic fraction in the 1st TBL is 16% of the initial one. The reduced fraction of 6 Li atoms transmuted to tritium ( 3 H).
Even after 5 FPYs of operation, the 6 Li atomic fraction in the 1st TBL is 61% of the initial one. But the 6 Li atomic fraction in the 10th TBL is 96% of the initial one even after 20 FPYs of operation.
Based on the supporting assessment of Li depletion, the replacement of the Li ceramic pebble bed in a high irradiated layer (1st TBL) with that in a less irradiated layer (10th TBL) to maintain the overall TBR in table 2 may be required for The helium and tritium are produced in the tritium breeding reaction with depletion 6 Li. But the number of helium is 2 or 3 times larger than that of tritium because there are various pathways to produce helium except for the tritium breeding reaction as shown in figure 5.

Displacement of atoms.
The average displacement energy threshold (E d ) to produce an atomic defect pair in the bulk of the material is required to calculate the final DPA [15]. A detailed assessment of the DPA/FPY of the W and the RAFM steel which are the major material consisting of the TBB is performed.
MCNP and FISPACT-II provide the simulation result on DPA depending on the flux spectrum, neutron source intensity, and material information. FISPACT-II uses E d = 55 eV for the W reflecting the latest research result [5]. This value differs from the historically accepted value of 90 eV which is recommended in table II of [15] and used for MCNP simulation in this study.
The in-depth study of this issue is beyond the scope of the current study, but further research in this field with the additional enhanced experimental evidence is strongly required to make a conclusion on the E d of the W. Depending on the conclusion of this issue, the expected lifetime of the W as the first plasma-facing material of nuclear fusion power plants may shorten by nearly half.
The calculation results of the DPA/FPY with two codes are presented in table 4 for the W and the 1st RAFM steel just behind the W in the PS of the TBB. The underestimation of the DPA/FPY of the W by MCNP is tabulated compared to the values that are calculated by FISPACT-II. The harshest neutron irradiation environment of these layers along the poloidal direction will reveal the worst values of the DPA and helium production that have adverse effects to shorten the lifetime of the TBB.
Another concern impacting the DPA/FPY calculation results is the composition of the material used in the simulation. The DPA/FPY values using the material composition including the realistic industrial-level impurities in this study are larger than those of the previous results [3] with the pure W. The expected lifetimes are also presented in table 4. The lifetime is inversely proportional to the NWL shown in table 2. Based on these estimations of the lifetimes, the maintenance and replacement plan of each PS of the TBB can be established.
Replacement by one or two years of the W wall can be planned for the PS which is heavily irradiated by neutrons if the recommended limit on the DPA is maintained.
In the case of the RAFM steel structures of the TBB in the IB and OB region, the replacement period varies from 2 years to 4 years.
The variation of the DPA/FPY along the radial direction is also estimated to see the impact of deep moderation of the neutron spectrum along this direction as shown in figure 6. There are dramatic changes in the DPA/FPY in the RAFM steel layers in PSN 16. As far as from the plasma, there is a fast decrease of DPA/FPY according to the distance from the plasma. So, the replacement period also varies depending on the installed position of the corresponding RAFM steel layers. So, manufacturing strategy of RAFM steel structures or coolant tubing structures must take into account these dramatic variations in the lifetime of RAFM steel layers.
For the SS316LN of the VV in the IB region, the DPA and helium concentration after 20 full power years is estimated at 4.03 × 10 −3 DPA, and 8.758 × 10 −2 appm, respectively. Two values are below the recommended limits in table 1. The reduction ratio of these values compared to those of the RAFM steel layers in the TBB is proportional to the reduced ratio of neutron fluxes between them.

Assessment of RW during the replacement of com-
ponents of the TBB. For the optimistically assumed 2 year replacement period, the radioactivity, decay heat, and contact dose rate of each PS in the IB and OB regions are estimated by FISPACT-II.
The history of activity growth-up according to the operation time of the K-DEMO before the replacement of the W wall of PSN 3 is shown in figure 7.
The noticeable concern is the generation of intermediatelevel waste (ILW) by Korean regulation on RW classification and self-disposal standards.
Among PSs in the IB and OB region, the W and 1st RAFM steel layers of PSN 3 are also classified as ILW after 4 months of operation. The growth-up histories of 94 Nb activity in other PSs are also presented in figure 8.
Even in the lowest irradiated PSN 6 in the divertor, the 94 Nb specific activities in the W and 1st RAFM steel layers exceed the low-level activity limit of 94 Nb (111 kBq kg −1 ) after 1 year of operation of the K-DEMO at the nuclear fusion power of 2.2 GW.
The crucial cause of the generation of ILW is the Nb impurity (0.001 wt% [2,16]) added in the W and the RAFM steel for the conservative assessment of RW related to impurities in this study.
The detailed overall compositions of the W and the RAFM steel can be found in [2] and its references. Compared to the previous study with pure W [11], the generation of ILW is a public safety-related concern regarding the replacement of the components of the TBB.
Its frequent replacement of the W and the RAFM steel dramatically increases ILW weight and volume from the early stage of the K-DEMO operation. In addition, the half-life of 94 Nb (2.03 × 10 4 years) makes it troublesome for long-term storage in the temporal hot cell at the nuclear fusion facility site.
The ultimate solution to this concern is to use Nb-free raw material from the initial mill manufacturing stage of the W and the RAFM steel.
Another impurity worth mentioning is hydrogen (0.0005 wt%), which can exist in the W as a constituent of H 2 O or in different forms.
From the hydrogen, tritium is produced by 1 H(n, γ) 2 H(n, γ) 3 H reaction, and also through various nuclear reactions in different pathways from the W. If tritium is not properly extracted and recovered from the W, this layer may also act as an atmospheric radioactive source that has critical concerns on public health because tritium being included in the W dust generated during replacement of the TBB or decommissioning of the K-DEMO. As a result, the treatment of the W dust has become a major issue from the nuclear fusion facility design to final decommissioning.
Another concern is the contact dose rate of the W and the RAFM steel for RH including the decay heat generated from it. The contact dose rate of the representative W wall of PSN 16 is presented in figure 9. The cumulative dose limit for the RH equipment to ensure its performance in a radioactive environment is set between 10 to 100 MGy [17].
Using the contact dose rate for the corresponding component, the time that the RH equipment can operate in the β, γ radiation field can be estimated from the contact dose rate calculation result.
In addition, the dose rate limit for access according to the class of RH equipment is set as follows: 10 −6 Sv h −1 Figure 9. The time evolution of the contact dose rate with major contributing isotopes in the W wall of PSN 16. for hand-on RH equipment, 10 −2 Sv h −1 for conservative RH equipment, and 10 4 Sv h −1 for advanced RH equipment [18].
Based on these kinds of dose rate analyses [19] with the corresponding dose limit, one can estimate when maintenance equipment can access a certain component for fragmentation, complete dismantling, and other inspections. For example, it can be expected that conservative RH equipment will be allowed to access the W layer of PSN 16 4 d after the shutdown after 2 years of operation for maintenance or replacement.
If the possible access time to all components can be checked in advance, it can be used to plan the spatial layout of hot cells, and the order of work, and thus an efficient replacement management plan can be established.
The decay heat generated from the irradiated component after irradiation stop is the physical quantity that serves as an important basis for determining the need and duration for active cooling. Active cooling is an important measure to prevent temperatures from rising above the designed limit by self-heating, thereby preventing component damage for re-use and reducing adverse effects on the surrounding environment.
The specific decay heat from the W wall of PSN 16 is presented in figure 10. Depending on emitting radiation by the nuclide (ex: 185 W), the contribution of a certain radionuclide to the decay heat may be large while the contribution to the activation and dose is small. The specific heat of the W in PSN 16 is 112 W kg −1 at the start of cooling and downs to 4.43 W kg −1 after 6 months. The total mass of the W in the PSN 16 of the K-DEMOs is about 9 tonnes. And the total mass of the W is about 120 tonnes for all PSs. So, the cooling facility corresponding to this total mass needs to be equipped.
These kinds of assessments for the components in each PS will be required for the establishment of a replacement plan for the corresponding components.
Information on the amount of heat generated per unit mass of the component may be used to calculate the temperature change of the component using a thermal flow analysis code together with information such as the total weight of the concerned component and its specific heat [11]. This calculation result can be used to determine the amount and the duration of active cooling needed for limiting the temperature increase of the component.
When a certain component requires active cooling during maintenance or replacement, it implies an increase in waste disposal costs, a delay in time, and the possibility of accidents due to cooling failure, all of which have a profound impact on safety management.

Helium production.
The fusion neutrons, which have higher fluxes and higher energies compared to those of fission neutrons, bombard the material and are captured by the material. After then the emission of the α-particle ( 4 He 2+ ) is followed sometimes.
These cascade reactions, often written as (n, α), makes problems with the material integrity by swelling or embrittlement of it. The helium production rates of the W, 1st RAFM steel layers of each PS consisting of the TBB are enumerated in table 5.
The result obtained by FISPACT-II is the case for 1 sec neutron irradiation time.
FISPACT-II calculates the rate equation with the timevarying atomic composition according to the specified irradiation time step.
As such, the helium production rate in each time step is also varying because the atomic composition of the concerned material at the specified irradiation time step is varied by nuclear reaction as shown in figure 11. But the helium production rate obtained from MCNP is not varying because this value is calculated just once. The relative differences in the results between the codes for the He production rate in the W wall of PS are about 30 %.
This difference may come from the difference in the used nuclear cross-section library between the codes.
The detailed investigation of this issue is beyond this study, but the pessimistic results will be utilized in the conservative conceptual design of the K-DEMO if there is no concrete evidence of estimated values for optimistic ones.  The accumulated helium concentration in the concerned layer can be estimated by FISPACT-II as shown in figure 12.
The ten years of full power operation are assumed for estimation to see the long-term evolution of helium fraction in the concerned layer.
The helium production in the RAFM steel layer is one order of magnitude higher than that of the W layer. Most of the helium generated in the RAFM steel layers comes from 54 Fe(n, α) 53 Cr and 56 Fe(n, p) 56 Mn reactions. These reactions are energy threshold reactions occurring above 3.7 MeV and 2.9 MeV of neutron energy, respectively.  As presented in figure 4, the moderation of neutron energy in the 1st RAFM steel layer is not sufficient to avoid these energy threshold reactions.
Even after 2 full power-year operations, the helium concentration in the 1st RAFM steel layer is about 482.5 appm but 10.4 appm in the W layer located in the front of it. As the position of the concerned layers is far from the plasma, the energy spectrum becomes softer, consequently, the helium production reaction rate decreases as shown in figure 13.
After 2 full power years of operation, the helium concentration in the 2nd RAFM steel layer is 333.7 appm but He concentration in the 11th RAFM steel layer is 1.9 appm. This is the combined consequence of the reduction of flux and high energy fraction (>2.9 MeV) of neutrons that can be seen in figure 4.
If there is no mechanical partition between the layered RAFM steel layers, there is little possibility for the reuse of the even less irradiated layers because there is no possibility of re-welding. The main reason is that the helium concentration in all RAFM steel layers in PSN 16 is over the recommended limit of 1 appm for re-welding.

Nuclear heating
The map of deposit energy by neutron and photon heating on the K-DEMO at the nuclear fusion power of 2.2 GW is plotted in figure 14 using the FMESH tally of MCNP.
A detailed investigation of the nuclear heating on the TBB is required to assess energy utilization and the assessment of nuclear heating on the TFC is essential to guarantee the operation soundness of the TFC.

3.3.1.
Heating on the TBB. The distribution of the deposited nuclear heating according to the radial arrangement of components in the PSN 16 is depicted in figure 15. The total heating on all PSN 16 for the full torus is 366.8 MW.
As a suggestion to enhance the energy multiplication efficiency of the K-DEMO, the efficient utilization of the components in the rear parts of the TBB to increase the energy density with the addition of fissile material such as thorium can be considered. But, it makes another safety related issues requiring re-evaluation of neutronic responses.
The heating on each PS along the poloidal direction for 11-25 degree half model is tabulated in table 6. The sum of heating power over the full torus on the TBB in the IO and OB The energy multiplication factors M E is defined [8] as the ratio of the total nuclear heating power over the fusion neutron power which is 80% of 2.2 GW in the case of K-DEMO. The M E of the K-DEMO is 1.31.
These deposited heats for each component will be used as the initial input parameters for the overall thermo-hydraulic analysis which is similar to that was done for the old model [20].

3.3.2.
Heating on the TFC. The deposited heat by the neutron and photon on the TFC are tabulated in table 7. The poloidal angle in table 7 is depicted in figure 14. To investigate the peak nuclear heating limit in the TFC winding pack in table 1, the heating on the more irradiated inner fraction of the TFC conductor and TFC SS316LN jacket is listed in table 7.
The estimated heating per volume in the high irradiated IB region is below the recommended limit of 5 × 10 −5 W cm −3 [7]. The estimated DPA/FPY of this region is 1.31 × 10 −5 DPA/FPY. For a 20 year fullpower operation, the total neutron fluence estimation is 1.5 × 10 18 cm −2 . This value exceeds the recommended limit in table 1.
Because of the enhancement of neutron shielding around the equatorial port and the change of plunging material by B-SS [21], the total nuclear heating on sixteen TFCs is reduced to 3.4 kW. If there are changes in the port plugging configurations for use as like the neutral beam injection port, these kinds of assessments will be necessary again to check the excessive heating on the TFC in detail. B.C. Kim Table 6.
The nuclear heating on the components of the TBB in the 11.25-degree half model of the K-DEMO with impurities in the material composition using light water coolant.
I. The nuclear heating on the radial layers of the TBB in the IB and OB regions. The radial layer number increases according to the increase of the plasma minor radius increase.
The nuclear heating on the radial layers (kW) for the 11.25-degree half model PSN 1_W 2_V 3 The total sum of heating on the components of TBB in the IB region The total sum of heating on the component in the divertor 6.713

429.7
The layers from 2 to 10 are the RAFM steel layers with 40 vol% water except for layers 8 of PSN 4, PSN 12, and layer 7 of PSN 8. These 3 layers are TBLs.

Conclusion
The K-DEMO will be operated in the extreme neutron environment never experienced in the NPP. The neutronic damages of the constituent materials irradiated by the fast and dense neutron field of the K-DEMO are critical considering factors in the early design stage of the K-DEMO. Through this study, the major nuclear damages of the candidate materials consisting of the K-DEMO are assessed to estimate the lifetime of the components. The defects represented by the DPA/FPY and the helium production rate are calculated to support the lifetime assessment.
The lifetimes of the severely irradiated components such as the W and 1st RAFM steel layers in the TBB vary from 2 years to 4 years. But the rear half part of the components of the TBB survives more than 10 years.
The careful control of impurities in the consisting materials of the K-DEMO is critical to help the willing acceptance of nuclear fusion energy by the public. Especially, niobium contents in the raw material from the early design of the K-DEMO and the manufacturing stage of materials for the K-DEMO is a critical impact on the reduction of ILW generation during the replacement of components or final decommissioning of the K-DEMO. Another finding is that the self-sufficient tritium supply is a challenging task in the K-DEMO magnetic configuration and tritium breeding concept. The design improvement to increase the TBR in the K-DEMO will main driving force of the optimization of the current K-DEMO design.
The nuclear heating on the TBB is assessed at about 2282 MW. The nuclear heating on another IVC, SB, is 15.5 MW. The deposited energy for each PS is varying from 10.4 MW to 391.0 MW so the design of the harmonized cooling layout to maintain the aligned outlet temperature has a critical impact on the effective utilization of nuclear heating energy on the TBB for the generation of electrical energy.
The nuclear heating on the TFC is also assessed to guarantee the sound performance of the TFC. The anticipated nuclear heating on the TFC is under the recommended limit. But this estimation is under the assumption that there is no opening in the port to install the heating and diagnostic devices that are not yet fully designed, so further evaluation of the nuclear heating on the TFC still needs.
According to the design evolution and operation condition of the K-DEMO, the neutron flux environment of it will be changing continuously. As such, the neutronic response analysis is continuously required as the design of the K-DEMO evolves. Through continued studies, the optimization of the design to freeze the K-DEMO conceptual design will be continued. Accordingly, the experimental results from the Korea superconducting tokamak advanced research (KSTAR) and construction experiences of the KSTAR will be incorporated into the K-DEMO design as well.