Neutronics study for the installation of the LBVM in the medium flux area of IFMIF

IFMIF (International Fusion Materials Irradiation Facility) will be a fusion dedicated facility producing a large amount of neutrons with the appropriate energy spectrum to test materials and subcomponents for DEMO and future Fusion Power Plants.While the high flux area of IFMIF will be devoted to reduced activation structural materials for first wall and blanket, the medium flux area will be dedicated to functional materials for breeder blankets. In particular, the Liquid Breeder Validation Module (LBVM), will host experiments related with functional materials for liquid breeder blankets. Since IFMIF neutron spectra have been intended to fit the most irradiated areas of a fusion reactor in the high flux area, the irradiation conditions in the LBVM placed in the medium flux area of IFMIF have been assessed. The effect of some neutron shifter/reflector components to optimize the neutron spectra have been evaluated in order to find out the proper irradiation conditions for functional materials for liquid breeder blankets.Therefore, the objective of this report is to summarize the neutronic calculations developed to evaluate the viability of IFMIF neutron source to perform relevant irradiation experiments on functional materials for liquid breeder blanket concept for future nuclear fusion power reactors (ITER, DEMO). The irradiation parameters evaluated for this purpose are: the tritium production for liquid breeder material (Pb–17Li) and the damage dose (dpa) and gas production to damage dose ratios for Al₂O₃ and SiC functional materials.The main conclusion is that, it is possible to perform relevant irradiation experiments on functional materials for liquid breeder blanket concept for the future nuclear fusion reactor DEMO. Nevertless, the use of some shifter components will be needed to optimize some irradiation parameters.     

Issues from activation/dose rates calculation for full W ITER machine and increased fluence

The paper presents and discusses the results of the activation calculations for ITER to investigate the effect of a possible increase of the neutron fluence on the First Wall (FW) and of the use of tungsten instead of beryllium as inboard and outboard FW protective layer. The new analyses based on the recent ITER design have been performed using the most recent, reliable and validated nuclear data and codes. Three situations have been considered: (1) average FW neutron fluence of 1.0 MWa/m² with beryllium FW protective layers (PLs); (2) average FW neutron fluence of 0.5 MWa/m² with tungsten FW PLs and (3) average FW neutron fluence of 1.0 MWa/m² with tungsten FW PLs.The used approach considers the Scalenea-1 radiation transport sequence for obtaining the 175 groups neutron fluxes in all the materials/zones on the radial direction of the ITER equatorial midplane and the EASY2007 code package for the material activation analysis. For tungsten PLs, calculations have also been performed using a MCNP approach in a 1D geometry. In this way the effect of the multi-group treatment of cross-sections is compared versus a continuous energy treatment and the deriving self-shielding effect is determined in the case of W where many resonances are present.     

Measurement and calculations of long-lived radionuclide activity forming in the fast neutron field in some ITER construction steels

Measurement and calculations of long-lived gamma-emitting radionuclide activity forming in the fission reactor fast neutron field were done, for some ITER construction steels. The activation was conducted in fast neutron irradiation channel of the MARIA research fission reactor (Poland). The dimensions of steel samples were 10 mm × 10 mm × 1 mm and mass was approximately 0.8 g. The neutron flux density was measured by means of activation foil method and unfolding technique; fraction of neutrons above 1 keV was 95%. The activation lasted 242 h and cooling took 100 days; the mean neutron flux density was 2.9E12 n/(cm² s) (neutrons above 500 keV are 53% of total) whereas total fluency 2.53E18 cm⁻². The activity measurements were done by means of gamma-ray spectrometry. Activity calculations were done by means of FISPACT-II code using the activation libraries EAF-2010 and TENDL-2011 and experimentally determined neutron flux. Measured activity of long-lived gamma emitting radionuclides was, in average, about 6.3 MBq/g 100 days after activation; the dominant radionuclides were ⁵⁸Co and ⁵⁴Mn (about 81% and 14% of total activity respectively). The C/E ratio differs for particular radionuclides and is in the range 0.86–0.92 for ⁵¹Cr, 0.93–1.21 for ⁵⁴Mn, 0.77–0.98 for ⁵⁷Co, 0.91–1.21 for ⁵⁸Co, 1.17–1.27 for ⁵⁹Fe, and 1.75–2.44 for ⁶⁰Co.     

Nuclear analysis of the IFMIF European lithium target assembly system

In the framework of the Engineering Validation and Engineering Design Activities (EVEDA) phase of the International Fusion Materials Irradiation Facility (IFMIF) project, ENEA was in charge of the design of the European version of the target assembly (TA) system which employs a removable bayonet backplate (BP) concept. With the aim of assessing the nuclear behaviour of the system and supplying the necessary input data to the thermomechanical analysis, coupled neutron-gamma transport calculations have been carried out for the whole TA + BP system, using the MCNP5 1.6 Monte Carlo transport code integrated with the McDeLicious-11 neutron source code provided by KIT. Neutron activation calculations have been performed by means of the EASY-2010 activation system in order to provide radioactive inventories useful for thermomechanical analysis and safety purposes. This paper summarizes the results obtained by the neutronic and activation calculations for the most irradiated components of the TA, such as backplate, frame, nozzle and target chamber.     

Tritium permeation in EUROFER in EXOTIC and LIBRETTO irradiation experiments

The reduced activation martensitic steel (RAFM) EUROFER is foreseen as a structural material in test breeder module (TBM) in ITER and breeder blanket in DEMO design. In a number of irradiation experiments conducted in high flux reactor (HFR) in Petten EUROFER was used as a containment wall of the breeder material, through which tritium permeation was monitored on line. Thus in EXOTIC-9/1 (EXtraction Of Tritium In Ceramics) experiment where Li₂TiO₃ pebbles were the breeder material, EUROFER was irradiated up to 1.3 dpa at 340–580 °C. In LIBRETTO experiments (LIBRETTO-4/1, -4/2 and -5) the breeder material was lead lithium eutectic which was in direct contact with the EUROFER containment wall. The neutron damage in steel achieved in the LIBRETTO experiments varied from 2 to 3.5 dpa. The irradiation temperature was 350 °C (LIBRETTO-4/1), 550 °C (LIBRETTO-4/2), and 300–500 °C (LIBRETTO-5).Tritium permeability was studied by varying the irradiation temperature and hydrogen concentration in the purge gas. From the analysis of the temperature transients performed in all four experiments yielded the tritium diffusion coefficients were derived, which appear to be factor ten lower than the literature data obtained in the gas driven permeation experiments.     

Neutron shielding and activation of the MASTU device and surrounds

A significant functional upgrade is planned for the Mega Ampere Spherical Tokamak (MAST) device, located at Culham in the UK, including the implementation of a notably greater neutral beam injection power. This upgrade will cause the emission of a substantially increased intensity of neutron radiation for a substantially increased amount of time upon operation of the device. Existing shielding and activation precautions are shown to prove insufficient in some regards, and recommendations for improvements are made, including the following areas: shielding doors to MAST shielded facility enclosure (known as “the blockhouse”); north access tunnel; blockhouse roof; west cabling duct. In addition, some specific neutronic dose rate questions are addressed and answered; those discussed here relate to shielding penetrations and dose rate reflected from the air above the device (“skyshine”). It is shown that the alterations to shielding and area access reduce the dose rate in unrestricted areas from greater than 100 μSv/h to less than 2 μSv/h averaged over the working day.The tools used for this analysis are the MCNP (Monte Carlo N-particle) code, used to calculate the three-dimensional spatial distribution of neutron and photon dose rates in and around the device and its shields, and the nuclear inventory code FISPACT, run under the umbrella code MCR2S, used to calculate the time-dependent shutdown dose rate in the region of the device at several decay times.     

Concept of curved magnetically guided liquid lithium target without a back plate for IFMIF

Curved magnetically guided lithium target (MGLT) without a back plate was newly proposed in light of simplified structure, easy maintenance and enhanced availability and performance for international fusion materials irradiation facility (IFMIF). It can replace conventional lithium target with a curved material back plate under the most severe condition on neutron irradiation. Magnetic field suited for the curved MGLT is produced in combination of a couple of radiation-proof resistive coils and reduced activation ferritic/martensitic steel (F82H) parts (yokes, ducts/nozzles and high flux test module (HFTM)). Shape of the magnetic field becomes curved automatically in the target region by setting HFTM closely to MGLT. Characteristics of the lithium flow on MGLT was analyzed in detail by two dimensional equations of motion with the magnetic field calculated by the Poisson Superfish code. The necessary magnetic flux density at the target region was found to be about 0.5 T to fulfill the IFMIF target conditions, i.e., lithium flow speed of 15 m/s, curvature radius of 1–1.6 m and flow thickness of 0.025 m. A narrow gap (a few mm) between MGLT and HFTM could be controlled by adjusting the coil current. Future subjects for further development of this concept were identified.     

MCNP5 modeling of the IPR-R1 TRIGA reactor for criticality calculation and reactivity determination

The IPR-R1 TRIGA is a research nuclear reactor managed and located at the Nuclear Technology Development Center (CDTN) a research institute of the Brazilian Nuclear Energy Commission (CNEN). It is mainly used to radioisotopes production, scientific experiments, training of nuclear engineers for research and nuclear power plant reactor operation, experiments with materials and minerals and neutron activation analysis. In this work, criticality calculation and reactivity changes are presented and discussed using two modelings of the IPR-R1 TRIGA in the MCNP5 code. The first model (Model 1) analyzes the criticality over the reactor. On the other hand, the second model (Model 2) includes the possibility of radial and axial neutron flux evaluation with different operation conditions. The calculated results are compared with experimental data in different situations. For the two models, the standard deviation and relative error presented values of around 4.9 × 10^?4. Both models present good agreement with respect to the experimental data. The goal is to validate the models that could be used to determine the neutron flux profiles to optimize the irradiation conditions, as well as to study reactivity insertion experiments and also to evaluate the fuel composition.     

Preliminary study on capsule material for ITER neutron activation system

A neutron activation system (NAS) measures neutron fluence at the first wall and the total neutron flux from plasma, providing the fusion power evaluation. A pneumatic transfer method is conventionally utilized to transfer encapsulated activation samples between the irradiation stations and counting station. The temperature of the irradiation station, near the first wall could reach too high for the conventional polymer-based materials, such as polyethylene, to be used as a capsule material for the ITER NAS. Considering the environment of the irradiation station of the ITER NAS, the candidate capsule materials are chosen as four materials: RAFM (reduced activation ferritic martensitic) steel, SiC_f/SiC composite, tungsten, and CFC (carbon fiber-reinforced carbon). Preliminary investigation reveals that the CFC is the most promising capsule material for ITER NAS due to its good thermal and magnetic properties as well as low activation by neutron irradiation. Various kinds of mock-up capsules are fabricated using CFC with the consideration of the volume of inner space accommodating activation samples. Preliminary pneumatic transfer experiments carried out in the small-scale test-bed suggest that the transfer speed of capsule should be slower than 10 m/s and the wall thickness of the capsule should be thicker than 2 mm so as not to be broken by impact damage. The present study shows the feasibility of using CFC as a capsule material for the ITER NAS.     

Effect of He-pre-injection on dislocation loop formation and irradiation-induced segregation of Fe–12%Cr–15%Mn austenitic steel

The effect of He-injection on irradiation-induced segregation of aging treated Fe–12%Cr–15%Mn austenitic steels, which are candidate materials as the reduced radio-activation of structure material for nuclear and/or fusion reactors was investigated by using the 1250 kV high voltage electron microscope (HVEM) connected with an ion accelerator. The Fe–Mn–Cr steel has been irradiated at 573 K by three irradiation modes of single electron-beam irradiation, electron-beam irradiation after He-injection and electron/He⁺-ion dual-beam irradiation in a HVEM. Irradiation-induced segregation analyses were carried out by an energy dispersive X-ray analyzer (EDX) in a 200 kV FE-TEM with beam diameter of about 0.5 nm. Dislocation loops with strain contrast were formed during irradiation and the loop numbers density increased rapidly with irradiation dose for He-pre-injected specimens. Voids were not observed after irradiations with three irradiation modes up to 5.4 dpa at 573 K. Irradiation-induced segregations of Cr and Mn near grain boundary were observed in each irradiation condition, but the amounts of Mn segregation decreased in the cases of electron/He⁺-ion dual-beam irradiation compared with single electron-beam and electron-beam irradiation after He-injection conditions.     

Nuclear Analysis of an ITER Blanket Module

ITER blanket system is the reactor’s plasma-facing component, it is mainly devoted to provide the thermal and nuclear shielding of the Vacuum Vessel and external ITER components, being intended also to act as plasma limiter. It consists of 440 individual modules which are located in the inboard, upper and outboard regions of the reactor. In this paper attention has been focused on to a single outboard blanket module located in the equatorial zone, whose nuclear response under irradiation has been investigated following a numerical approach based on the Monte Carlo method and adopting the MCNP5 code. The main features of this blanket module nuclear behaviour have been determined, paying particular attention to energy and spatial distribution of the neutron flux and deposited nuclear power together with the spatial distribution of its volumetric density. Moreover, the neutronic damage of the structural material has also been investigated through the evaluation of displacement per atom and helium and hydrogen production rates. Finally, an activation analysis has been performed with FISPACT inventory code using, as input, the evaluated neutron spectrum to assess the module specific activity and contact dose rate after irradiation under a specific operating scenario.     

Nondestructive Magnetic Adaptive Testing of nuclear reactor pressure vessel steel degradation

Inspection of neutron-irradiation-generated degradation of nuclear reactor pressure vessel steel (RPVS) is a very important task. In ferromagnetic materials, such as RPVS, the structural degradation is connected with a change of their magnetic properties. In this work, applicability of a novel magnetic nondestructive method (Magnetic Adaptive Testing, MAT), based on systematic measurement and evaluation of minor magnetic hysteresis loops, is shown for inspection of neutron irradiation embrittlement in RPVS. Three series of samples, made of JRQ, 15CH2MFA and 10ChMFT type steels were measured by MAT. The samples were irradiated by E > 1 MeV energy neutrons with total neutron fluence of 1.58 × 10¹⁹–11.9 × 10¹⁹ n/cm². Regular correlation was found between the optimally chosen MAT degradation functions and the neutron fluence in all three types of the materials. Shift of the ductile–brittle transition temperature, ΔDBTT, independently determined as a function of the neutron fluence for the 15CH2MFA material, was also evaluated. A sensitive, linear correlation was found between the ΔDBTT and values of the relevant MAT degradation function. Based on these results, MAT is shown to be a promising (at least) complimentary tool of the destructive tests within the surveillance programs, which are presently used for inspection of neutron-irradiation-generated embrittlement of RPVS.     

Waste management strategy focused on maintenance,storage and recycling

Waste is generated at the moment when the operation of a fusion reactor is halted and maintenance is started for periodic replacement of blanket modules and divertor. Used blanket and divertor need to be replaced shortly after the shutdown for high plant availability, as long as high surface dose rate and decay heat of the blanket and divertor can be handled. In this sense, nuclear characteristics of the blanket and divertor need to be understood for a reasonable maintenance scheme. For the purpose, neutronic calculations were carried out on the blanket and divertor using a THIDA-2 code with FENDL-2.0. For a SlimCS DEMO reactor, the calculated decay heat for each 1/12-sector was as high as 5 MW just after the shutdown and 0.3 MW one month later. For the maintenance, a cooled shielding structure (CSS) was proposed to remove the decay heat and to shield gamma-rays from the sector. When maintenance is done one month after the shutdown, the sector temperature is maintained to be 550 °C or lower with the cooling by the CSS of 50 °C. In order to avoid tritium release from the sector during the maintenance, a cask should be used to transport the sector. For efficient use of resources, breeding and neutron multiplying materials should be reused or recycled. A possible strategy for reuse or recycle is also presented.     

Development of self-powered neutron detectors for neutron flux monitoring in HCLL and HCPB ITER-TBM

Self powered neutron detectors (SPND) have a number of interesting properties (e.g. small dimensions, capability to operate in harsh environments, absence of external bias), so they are attractive neutron monitors for TBM in ITER. However, commercially available SPNDs are optimized for operation in a thermal nuclear reactor where the neutron spectrum is much softer than that expected in a TBM. This fact can limit the use of SPND in a TBM since the effective cross sections for the production of beta emitters are much lower in a fast neutron spectrum.This work represents the first attempt to study SPNDs as neutron flux monitors for TBM. Three state-of-the-art SPND available on the market were bought and tested using fast neutrons at TAPIRO fast neutron source of ENEA Casaccia and with 14 MeV neutrons at the Frascati neutron generator (FNG).The results clearly indicate that in fast neutron spectra, the response of SPNDs is much lower than in thermal neutron flux. Activation calculations were performed using the FISPACT code to find out possible material candidates for SPND suitable for operation in TBM neutron spectra.     

Irradiation dose and temperature dependence of fracture toughness in high dose HT9 steel from the fuel duct of FFTF

To expand the knowledge base for fast reactor core materials, fracture toughness has been evaluated for high dose HT9 steel using miniature disk compact tension (DCT) specimens. The HT9 steel DCT specimens were machined from the ACO-3 fuel duct of the Fast Flux Test Facility (FFTF), which achieved high doses in the range of 3–148 dpa at 378–504 °C. The static fracture resistance (J-R) tests have been performed in a servohydraulic testing machine in vacuum at selected temperatures including room temperature, 200 °C, and each irradiation temperature. Brittle fracture with a low toughness less than 50 MPa √m occurred in room temperature tests when irradiation temperature was below 400 °C, while ductile fracture with stable crack growth was observed when irradiation temperature was higher. No fracture toughness less than 100 MPa √m was measured when the irradiation temperature was above 430 °C. It was shown that the influence of irradiation temperature was dominant in fracture toughness while the irradiation dose has only limited influence over the wide dose range 3–148 dpa. A slow decrease of fracture toughness with test temperature above room temperature was observed for the nonirradiated and high temperature (>430 °C) irradiation cases, which indicates that the ductile–brittle transition temperatures (DBTTs) in those conditions are lower than room temperature. A comparison with the collection of existing data confirmed the dominance of irradiation temperature in the fracture toughness of HT9 steels.     

The impact of time dependant spectra on fusion blanket burn-up

Knowledge of nuclide burn-up within tritium breeding blankets has a crucial part to play in the safety, reliability and efficiency of fusion reactors. The modelling of burn-up requires a series of neutron transport calculations which can compute the reaction rate either directly, via Monte-Carlo estimators, or by implementing the multi-group method. These reaction rates can then be directly substituted into the burn-up equations, which can calculate nuclide number densities after a specified period of burn-up. The material burn-up will change the neutron spectra and the rate of nuclear reactions. Hence, a new neutron transport calculation needs to be performed after burn-up and the sequence is repeated for several time-steps. Radiation transport calculations are computationally expensive, therefore the minimisation of reaction rate calculations via Monte-Carlo simulations is desirable. Thus, time-intervals between Monte-Carlo simulations should be as large as possible. This paper addresses the effect of neutron spectra on the burn-up of parent and daughter nuclides found in EUROFER steel and the tritium self-sufficiency time.Using a spherical reactor geometry with lithium–lead tritium breeding material, a neutron spectrum is computed at time = 0 and time = 2 years after a detailed depletion calculation using 1 day time intervals. These two spectra are then used to calculate reaction rates for every isotope listed within the EAF2005 database using the FISPACT code. The results show that the difference in nuclide number densities are less then 11% for all nuclides within the database and less than 4% for all fusion relevant nuclides.Using the same methodology as the first model, EUROFER parent and daughter nuclide number densities produced by each neutron spectra are compared. This study found that the change in burn-up for parent nuclides is statistically insignificant. However, the difference in daughter nuclide production is significant, especially for Rh, Ru, Re, Os, Pt and Ir where the differences range from 20% to 504%. Thus, in order to model the metallurgical properties of steels within fusion blankets over time, a multiple transport-burnup depletion code (such as used by FATI, VESTA or MONTEBURNS) must be implemented.The final part of this work studied the effect of time-step interval (used to update neutron spectra) on the tritium self-sufficiency time of a blanket. The FATI depletion code modelled the same geometry as in part 1 of the study, however time-steps ranging from 1 day to approximately 800 days were used to predict when the blanket would cease to be able to breed enough tritium to sustain the fusion reactor. The single time-step model (i.e. where a constant neutron spectrum is used for the entire simulation) underestimated the tritium self-sufficiency time of the blanket by approximately 70%. Only time-steps less than 1 month produce a self-sufficiency time which is within 5% accuracy. Hence, this work suggests that spectra time-stepping is important in the modelling of tritium production with solid breeders.     

Analysis of radiation environment at divertor in helical reactor FFHR-d1

Neutron transport calculations with a three-dimensional model of the helical reactor FFHR-d1 have been performed for the accurate analysis of neutronics environment in the divertor areas. Based on the obtained neutron spectra, magnitudes of irradiation damage, contact dose rates and decay heat have been evaluated mainly for Fe, W and Cu. Since divertors can be placed behind radiation shields in helical reactors, magnitudes of damage and radioactivation at the outboard divertors are almost two orders lower than those at blanket first walls. Cu materials could be used as a cooling channel material of the outboard divertors. In contrast, the magnitudes at the inboard divertors are only one order lower compared with those at the first walls due to the limited space at the inboard side. Damage on Cu is evaluated to be ∼10 dpa after 6 years operation. Further efforts in divertor development and reactor design could suppress the magnitude of damage to less than half for adoption of Cu materials for the inboard divertors.     

Analysis of dark current images of a CMOS camera during gamma irradiation

We report on the behaviour of the dark current images of the Event Detection Intelligent Camera (EDICAM) when placed into an irradiation field of gamma rays. EDICAM is an intelligent fast framing CMOS camera operating in the visible spectral range, which is designed for the video diagnostic system of the Wendelstein 7-X (W7-X) stellarator. Monte Carlo calculations were carried out in order to estimate the expected gamma spectrum and dose for an entire year of operation in W7-X. EDICAM was irradiated in a pure gamma field in the Training Reactor of BME with a dose of approximately 23.5 Gy in 1.16 h. During the irradiation, numerous frame series were taken with the camera with exposure times 20 μs, 50 μs, 100 μs, 1 ms, 10 ms, 100 ms. EDICAM withstood the irradiation, but suffered some dynamic range degradation. The behaviour of the dark current images during irradiation is described in detail. We found that the average brightness of dark current images depends on the total ionising dose that the camera is exposed to and the dose rate as well as on the applied exposure times.     

Gamma-ray dose analysis for ITER JA WCCB-TBM

To evaluate the nuclear properties of the International Thermonuclear Experimental Reactor (ITER) JA Water-Cooled Ceramic Breeder Test Blanket Module (WCCB-TBM) and to ensure its design conforms to nuclear licensing regulations, nuclear analyses have been performed for the WCCB-TBM's components, including its frame, shield, flange, port extension, pipe forest, bio-shield and Ancillary Equipment Unit (AEU). Utilising Monte Carlo code MCNP5.14, activation code ACT-4 and the Fusion Evaluated Nuclear Data Library FENDL-2.1, this paper focusses on the shutdown dose rate calculation for the WCCB-TBM. Monte Carlo N-Particle Transport Code (MCNP) geometry input data for the TBM are created from computer-aided design (CAD) data using the CAD/MCNP automatic conversion code GEOMIT, and other geometry input data are created manually. The ‘Direct 1-Step Monte Carlo’ method is adopted for the decay gamma-ray dose rate calculation. Behind the bio-shield, the effective dose rates 1 day after shutdown are about 0.2 μSv h⁻¹, which are much lower than 10 μSv h⁻¹, the upper limit for human access. Behind the flange, the effective dose rates 10⁶ s after shutdown are 50–80 μSv h⁻¹, which are lower than 100 μSv h⁻¹, the upper limit for human hands-on access for workers performing maintenance.     

Neutronics diagnostics for European ITER TBMs: Activation foil spectrometer for short measurement cycles

An important aim of neutronics Test Blanket Module (TBM) experiments in ITER will be to check the prediction accuracy of nuclear responses in an environment closer to a future fusion power reactor than so far provided by existing facilities. The development of measurement methods suitable for the harsh environment in an ITER TBM has been addressed in several recent R&D programs supported by Euratom. Within this framework, KIT is developing an activation foil spectrometer for the measurement of local neutron flux densities in the TBM. We intend to establish a measurement method which allows to record the induced activities in small packages of activation foils simultaneously and to calculate the corresponding spectral neutron flux densities with moderate time resolution of tens of seconds immediately after extraction from the TBM. In the present work we propose a candidate set of activation foil materials which cover the neutron energy range from thermal to 14 MeV. In order to assess their basic suitability for such measurements, we have computed induced gamma-ray activities in the foils using a calculated neutron spectrum in a representative position in the European HCPB TBM assuming a short irradiation time of 30 s. In a further step we have investigated pulse height spectra which would be obtained in a typical gamma-ray measurement arrangement in a HPGe detector and concluded that the proposed set of activation foils should be basically suitable for such a measurement system but require improvement of relevant cross sections uncertainties.