Tritium Production Rate Research Articles

Simultaneous measurements for fast neutron flux and tritium production rate using pulse shape discrimination and single crystal CVD diamond detector

This paper presents the development of a simultaneous measurement method for fast neutron energy spectra and tritium production rates within mixed radiation fields using a single crystal chemical vapour deposition diamond detector combined with a lithium fluoride (LiF) foil. The method involves the separation of pulses with rectangular shapes and the determination of the depth position within the single crystal diamond (SCD) struck by fast neutrons or nuclear reaction products including recoil tritons from the LiF foil based on pulse width, extracting pulse events occurred at the specific bulk region and the surface region of the SCD. Subsequently, unfolding techniques were employed to analyse the energy deposition spectrum of pulses at the specific bulk region which are induced only by fast neutrons, allowing the deduction of the fast neutron energy spectrum. To evaluate the tritium production rate, the energy deposition spectrum of pulses from events occurring at the SCD surface facing the LiF foil was analysed. By estimating the energy deposition spectrum solely induced by fast neutrons striking the SCD surface and subtracting it from the energy deposition spectrum of events at the SCD surface, the contribution of energetic ions, such as recoil tritons generated by the 6Li(n,α)3H reaction in the LiF foil, was determined. The fast neutron flux and tritium production rate obtained through this study were consistent with particle transport calculations, demonstrating the successful development of a method suitable for performance testing of fusion reactor blankets.

OpenMC Interpretation of FNS SINBAD Shielding Benchmark Experiments

The Fusion Neutron Source (FNS) clean benchmark experiments on tungsten, vanadium, and beryllium assemblies from the SINBAD (Shielding Integral Benchmark Archive and Database) are analyzed to experimentally validate OpenMC (version 0.14.1-dev) fusion neutronics capabilities. The assemblies were irradiated with a 14-MeV deuterium-tritium neutron source. Neutron spectra, photon spectra, reaction rates, gamma heating rates (GHRs), and tritium production rates (TPRs) are compared to measured data in the experimental assemblies and MCNP-6.2 results. In the tungsten case, slight overestimations of the experimental data were observed in the neutron spectra, and the photon spectra agreed well with the experiments. Most of the GHRs agreed with the measured data within the range of experimental uncertainty in the tungsten and vanadium assemblies. In the vanadium assembly, the calculated neutron spectra underestimated the experiments in the low energy region while the photon spectra were well calculated when compared to experiments. The most noticeable discrepancies with experimental data in the gamma heating were observed at detector positions closest to the source. For the reaction rates, notable discrepancies with experimental data were seen at the front and rear of the assemblies. Compared to experiments, the OpenMC neutron spectra were well predicted in the beryllium assembly, whereas the calculated fission reaction rate and TPRs overestimated the experiments, an observation similar to that which has been reported by other authors. The average, overall calculation-to-experiment ratio (C/E) over nine TPR and seven GHR measurements were 1.03 ± 0.20 and 0.95 ± 0.14, respectively. In the case of verification, the OpenMC results of the benchmark calculations indicated comparable accuracy to MCNP-6.2. In general, the validation exercise showed that OpenMC can be used to analyze the fusion neutronics shielding benchmark problems.

Study on tritium permeation behavior in primary and second circuits of the hydrogen production system by methane steam reforming using HTGR

Hydrogen production by methane steam reforming (MSR) using the process heat from High-temperature gas-cooled reactor (HTGR) is one of the most promising technology for hydrogen production using nuclear energy in the near future. The permeation behavior of tritium produced by the reactor is extremely important when analyzing the safety of this hydrogen production system. There are two methods of connection between a nuclear reactor and a hydrogen production system. One is direct coupling that the steam reformer of the hydrogen production system is directly connected to the first circuit of the nuclear reactor. The other is indirect coupling via an intermediate heat exchanger (IHX) to connect nuclear reactor and hydrogen production system. Hence, this work proposed a model to analyze the difference in tritium permeation between these two coupling ways and to compare the tritium permeation behavior for different IHX alloys at various operating conditions. This model includes tritium's production rate, release fraction, and decay reduction in the system. The results show that the pre-exponential factor of the IHX alloy mainly limits the tritium permeation rate. In addition, the tritium permeation rate increases about two times when the temperature rises from 750 °C to 950 °C at a steady state. Further, the decrease of tritium permeation rate is seen from the calculation results for the system with IHX, e.g., for T = 750 °C, the mean permeation rate of tritium into the product gas is 9.31 × 10−12 mol/s at the whole year for indirect coupling with an IHX, compared to 3.86 × 10−12 mol/s for direct coupling, which reduced 58.54 % (This means that with IHX the permeation rate is lower than the system without IHX). These results will contribute to designing an effective and safe hydrogen production system by MSR using HTGR process heat.

Cosmogenic activation of sodium iodide

The production of radioactive isotopes by interactions of cosmic-ray particles with sodium iodide (NaI) crystals can produce radioactive backgrounds in detectors used to search for rare events. Through controlled irradiation of NaI crystals with a neutron beam that matches the cosmic-ray neutron spectrum, followed by direct counting and fitting the resulting spectrum across a broad range of energies, we determined the integrated production rate of several long-lived radioisotopes. The measurements were then extrapolated to determine the sea-level cosmogenic neutron activation rate, including the first experimental determination of the tritium production rate: $(80\ifmmode\pm\else\textpm\fi{}21)\text{ }\text{ }\mathrm{atoms}/\mathrm{kg}/\mathrm{day}$. These results will help constrain background estimates and determine the maximum time that NaI-based detectors can remain unshielded above ground before cosmogenic backgrounds impact the sensitivity of next-generation experiments.

Space Dependent Study of Fast Neutron Spectra and Tritium Production Rate in a Fusion Reactor Blanket of Li2O

Study of spatial spectra and tritium breeding ratio (TBR) in a lithium blanket play a crucial role for the successful operation of a Fusion reactor. In this paper, we have reported the results for the space-dependent fast neutron spectra and tritium production rate (TPR) in cylindrical assemblies of Li₂O. The multigroup diffusion equation has been solved using eigenfunction expansion method. The obtained results for scalar spectra, total flux and TPR in Li₂O are compared with corresponding values for natural Li. The space dependent scalar spectrum in Li₂O has been compared for different configurations (slab, sphere and cylindrical). We find that the shapes of the curves are almost similar, though the neutron flux is found to be highest in the case of cylindrical assembly. This suggests a larger value of total TBR for a cylindrical geometry. Further, the study of space variation of total flux for Li₂O and natural Li shows that there is not much change in the shapes and numerical values, however, it falls sharply near the boundary end surface of the assembly for Li₂O as compared to natural Li. Similarly, the fall in the TPR curves becomes more rapid for the case of Li₂O as compared to natural Li near the boundary end.. The calculated results of spatial distributions of TPR for 6Li (n,α)t and ⁷Li (n,n'α)t reactions are compared with experimentally obtained results for cylindrical slab assemblies. and it is found that agreement is fairly good, both in the case of 7Li and 6Li .

Benchmarking of emergent radiation transport codes for fusion neutronics applications

The accurate and efficient mapping of the radiation environment in a nuclear fusion reactor requires the most advanced radiation transport tools. The Monte Carlo method has long been deployed to deal with the complexity of fusion relevant geometries, with MCNP the adopted industry standard code among the European and wider international community. However, reliance on a single code has driven explorations into alternatives to establish their capabilities and maturity for fusion analyses. These alternatives are being actively developed and increasingly adopted in some types of applied analysis by the user community. In this paper, we explore both experimental and computational benchmarks in order to examine the capabilities of Serpent, OpenMC as well as the framework for allowing CAD based particle transport, DAGMC, over a broad range of fusion relevant nuclear responses and geometries. This spans from more simple parametric models adopted in reactor scoping studies to the current ITER reference model which has been successfully translated to Monte Carlo codes other than MCNP using an open source utility, csg2csg. An assessment for both CSG and CAD based workflows has been conducted as well as a hybrid approach combining the two. The FNG HCPB, Cu and a subset of the FNS experiments were also converted to Serpent and OpenMC input files for comparison of calculation to available experimental data. Good agreement was observed across all codes for the determined tritium production rates and activation foil measurements. Potentially more efficient workflows for complex tokamak models are detailed. For a heterogeneous model of JET octant 1, an optimised CAD based model in Serpent is over 50% faster than the MCNP CSG equivalent model. In the case of the generated OpenMC model of ITER, an order of magnitude reduction in simulation time, including a model loading time of the order of minutes, is reported. Such validation and benchmarking activities should in the future be integrated as part of a testing suite for which a basic framework has been demonstrated here. To conclude, the current limitations and required development are outlined as well as identifying where each code may specialise for a particular application. Based on this future work relevant to both the developers and user community is briefly discussed.

Solar activity cycles recorded in long-term data on tritium activity concentration in precipitation at Zagreb, Croatia

It has been expected that the solar cycle may have an influence on the production rate of cosmogenic tritium. However, prevailing anthropogenic bomb tritium in precipitation in the second half of the 20th century prevented such studies. We analyzed long-term data (1976–2020) of tritium activity concentration in precipitation at Zagreb, Croatia, by various statistical methods looking for the solar cycle periodicity. Frequency and Wavelet analyses were performed on the whole data set, while sinusoidal fitting was performed on data since 1996, when bomb tritium is negligible. All methods resulted in two distinct periods. The shorter 1-year period corresponds to the annual variation of tritium in precipitation typical for the Northern Hemisphere stations. The longer period of 10.21 yr, ≈11 yr and 12.17 yr was obtained by frequency analysis, wavelet analysis and sinusoidal fitting, respectively. These values are close to the solar cycle with 11 yr period on the average (ranging from 9 yr to 13.6 yr). Additionally, we compared the maxima in the smoothed monthly sunspot number in the 23rd and 24th solar cycles and the minima in tritium activity concentration in precipitation. A delay of about 30 months between them was observed and plausibly explained by the residence time of cosmogenic tritium in the stratosphere before entering the troposphere.

Neutronic experiment and analyses of a hybrid tritium breeding blanket mockup for CFETR

A neutronic experiment has been performed using DT neutrons on a blanket mockup of the hybrid tritium breeding blanket (HTBB) at CAEP. As a backup blanket configuration for CFETR, HTBB design which uses uranium alloy, water coolant and ceramic breeder pebble bed has been proposed to obtain high tritium breeding ratio. The experiment has provided validation not only to the preliminary design of HTBB, but also to calculation code and nuclear data used in the simulations. Various neutron responses such as fission reaction rate, plutonium production rate, activation reaction rate and tritium production rate (TPR) have been measured. Numerical analyses were carried out with MCNP-4C code using ENDF/B-VII.1, TENDL-2017, IRDFF-v1.05 and FENDL3.1 nuclear data libraries. The comparisons between the calculation and experimental results show an agreement within about 10% for most of the measured quantities. The ratios of calculated to measured results (C/E) of tritium production rate are in 0.98–1.04 range which shows good agreement.

Experimental study on tritium breeding in water-cooled ceramic breeder blanket mock-up under D–T neutron irradiation conditions

The accurate assessment of tritium breeding parameters within fusion blankets is crucial for future magnet-confined fusion machines to realize fuel self-sufficiency. Such an assessment can be conducted using the simulation approach with nuclear data of high-fidelity and, most importantly, validated against experimental data. In this paper, we report neutronics experiment studies carried out on a mock-up of the water-cooled ceramic breeder blanket of the China fusion engineering test reactor (CFETR), irradiated at a deuterium–tritium (D–T) neutron source. The mock-up’s nuclear responses to 14 MeV neutrons, including tritium production rates (TPR) and neutron-induced reaction rates, are simulated and validated by experimental results. Redundant measurement techniques are used, including Li2TiO3 pellets for offline TPR measurements and the developed lithium glass detector with a significantly reduced size for online TPR measurements. The validation of the TPR value is complemented by the experimental evaluation of the neutron-induced reaction rates for Au and Zr foils. All experimental results are analyzed using the MCNP-4C code and FENDL-3.0 nuclear data library. The source term for the Monte Carlo simulation is built using a newly-developed method based upon the modeling of the depth profiling of tritium in the tritiated target. The experiments are in good agreement with the simulations; ratios of the calculation to experimental results (C/E) on TPR are found to be 0.97–1.08. The influence of the first wall tungsten armor on the mock-up nuclear responses is also studied.

Assessment of tritium effluent from Prototype Generation IV Sodium-cooled Fast Reactor

A tritium transport model for a pool-type sodium cooled fast reactor was developed for estimation of tritium effluents from the Prototype Generation IV Sodium-cooled Fast Reactor (PGSFR). For this purpose, the transport model developed for Experimental Breeder Reactor II and Fast Flux Test Facility was updated and validated. The tritium production rate from ternary fission and activations in PGSFR were calculated by the Argonne Reactor Computation (ARC) code suite for fast reactor analysis, and due to the potential uncertainties in source terms and variations of cold trap performance and material permeability during reactor operation, the tritium distributions and tritium effluents were estimated for several bounding cases. The upper bounding gaseous and liquid tritium release rates from PGSFR are ~ 440.7 Ci/year and 14.7 Ci/year, respectively. The gaseous tritium effluent is about few orders of magnitude smaller than the regulatory constraint of the tritium concentration in the air. The peak tritium concentration in the steam-generator feedwater is ~ 1.1x10-3 μCi/g, which is required to be diluted prior to releasing as tritiated water. The accumulated tritium in the cold trap after five cycle operations is ~ 52,000 Ci. Because of significant conservatism in these upper bounding estimations, it is expected that the tritium effluent at the nominal case with plausible source terms would be about 2 orders of magnitude smaller than that of the upper bounding case.

Study on the performance of hydrogen isotopes permeation sensor in liquid Li–Pb

Liquid Li–Pb alloy is served as neutron multiplier, tritium breeder, and coolant in Dual Coolant Lithium–Lead (DCLL) breeding blankets. In order to monitor the tritium transport continuously and measure the tritium production rate accurately, the concentration of hydrogen isotopes in liquid breeder is an essential factor for tritium measurement. The hydrogen isotopes permeation-based capsule seems to be the most reliable sensor and is simple from the fabrication point of view. To verify the stability and repeatability of hydrogen isotopes concentration measured by permeation capsule in Li–Pb, a system based on a Nb-capsule and a quadrupole mass spectrometry was devised. The sensor would be applied to determine the variation rate of total pressure while a quadrupole mass spectrometry to measure partial pressure values of different hydrogen isotopes. The response time of the sensor is less than 100 s in liquid Li–Pb alloy and the growth rate of pressure has good repeatability in dynamic mode. Besides, calculations based on numerical simulation of the hydrogen isotopes sensor are presented.

Evaluation of tritium production rate in a blanket mock-up using a compact fusion neutron source

We report a neutronics study of a blanket mock-up using a discharge-type compact fusion neutron source. Deuterium–deuterium fusion neutrons were irradiated to the mock-ups composed of tritium breeder and neutron reflector/moderator. The tritium production rate (TPR) per source neutron was measured by a single-crystal diamond detector with a 6Li-enriched lithium fluoride film convertor after the calibration process. Despite the low neutron yield, energetic alpha and triton particles via 6Li(n, t)α neutron capture as well as 12C via elastic scattering were successfully detected by the SDD with high signal to noise ratios. The TPRs were experimentally evaluated with errors of 8.4%–8.5% at the 1σ level at the positions with high thermal neutron fluxes where the errors were dominantly introduced by uncertainties in the monitoring of the neutron production rate. The calculated to experimental (C/E) values of TPR were evaluated to be 0.91–1.27 (FENDL-2.1) and 0.94–1.28 (FENDL-3.1). As the neutron source can generate 14 MeV neutrons using a mixed gas of deuterium and tritium, this approach provides more opportunities for blanket neutronics experiments.

NUCLEAR DATA SENSITIVITY/UNCERTAINTY PRE-ANALYSIS OF FNG WCLL FUSION BENCHMARK

To assure tritium self-sufficiency in future fusion reactors such as DEMO the accuracy of TRP calculations has to be demonstrated within the design uncertainties. A new neutronics experiment representing a mock-up of the Water Cooled Lithium Lead (WCLL) Test Blanket Module (TBM) is under preparation at the Frascati neutron generator (FNG) with the objective to provide an experimental validation of accuracy of nuclear data and neutron transport codes for the tritium production rate (TPR) calculations. The mock-up will consist of LiPb bricks, EUROFER plates and Perspex substituting water. The mock-up will be irradiated by 14 MeV neutrons at the FNG facility, and the TPR and detector reaction rates will be measured using Li2CO3 pellets and activation foils placed at different positions up to about 55 cm inside the mock-up. Computational pre-analyses for the design of the WCLL neutronics experiment using the SUSD3D sensitivity/uncertainty (S/U) code system is described and compared with the results of some similar FNG experiments performed in the past, in particular the FNG HCPB Tritium Breeder Module Mock-up (2005) and FNG-HCLL Tritium Breeder Module Mock-up (2009). The objective of the pre-analysis is to provide the calculated nuclear responses including the uncertainties due to the uncertainties in nuclear data and thus contributes to the optimisation of the design of the experimental set-up.

Numerical investigation of purge gas flow through binary-sized pebble beds using discrete element method and computational fluid dynamics

Abstract The main function of the breeder materials, such as Li2TiO3, Li4SiO4, and Li2ZrO3, in a solid type breeding blanket is to produce tritium, which is transferred to the fuel cycle system by helium purge gas and is used as a fuel for nuclear fusion reactions. Since the configuration of the breeder should be designed considering the purge gas flow, the form of pebble beds is mainly adopted, thereby leading to the smooth flow inside a blanket. The flow characteristics can be important design drivers because they also affect the design of the fuel cycle system. In order to increase a tritium production rate, a large number of breeders should be placed in the same space and various sizes of pebbles instead of one size of pebbles can be thus applied to increase the packing fraction of pebble beds. In this case, however, the flow resistance may increase and hence the relationship between the pebble bed configuration and the purge gas flow should be carefully investigated. In this study, a computational framework for one-way coupling analysis of the discrete element method and computational fluid dynamics is established to simulate the purge gas flow through binary-sized pebble beds. The configurations of pebble beds are obtained using the discrete element method and their geometries are modified by smoothing the sharp circular edges at the contact points of pebbles to construct proper meshes for computational fluid dynamics simulations. The flow resistance of binary-sized pebble beds is then investigated in terms of the pressure drop in the breeding zone.

Nuclear responses of WCCB TBM with different container designs

Abstract In test blanket module (TBM) program blanket major functions, which are tritium production, heat extraction for electric power production, and neutron shielding, will be demonstrated under fusion reactor environments in ITER. National Institutes for Quantum and Radiological Science and Technology (QST) has performed developments of water cooled ceramic breeder (WCCB) TBM as a primary concept in Japan. The container design of WCCB TBM was changed from the Box-shaped to a cylindrical structure to increase tritium breeding capability while maintaining pressure resistance under In-Box LOCA caused by water ingress into the container. The nuclear responses related to the blanket major functions would be changed with the container design change. The nuclear responses of WCCB TBM with the container design change were evaluated and the effects of the changed nuclear responses were described in this study. Tritium production rate increased by about 2 times with the container design changes. In contrast neutron shielding performance decreased due to the container design change. Degradation of neutron shielding in the cylindrical TBM led to an increase in nuclear heating of the TBM frame as well as an average of neuron flux behind TBM-set.

Conceptual design of test modules for DEMO blanket, diagnostic device, and RI production for A-FNS

In the conceptual design activity of advanced fusion neutron source A-FNS, a variety of test modules for fusion DEMO reactor are planned. In these modules, progresses of design activities on Blanket Nuclear Property Test Module (BNPTM) and Diagnostic and Control Device Test Module (DCDTM) were reported. The BNPTM is a module in order to evaluate accuracies of nuclear analyses of the DEMO blanket such as tritium production rate. The influences of the cooling water and test cell wall were evaluated in order to decide its design. The DCDTM is one to achieve irradiation data of functional materials on diagnostic and control devices such as mirror and window. It was clarified from its nuclear analysis that the neutron fluence obtained in the DCDTM was enough to investigate the accumulated irradiation effects on the materials. In addition to the irradiation tests on the fusion reactor materials, various neutron applications are planned at A-FNS. One of the applications is medical isotope production. We clarified that enough amount of the medical isotope molybdenum-99 could be produced in comparison with the demand in Japan by using Radio-Isotope Production Module (RIPM). The RIPM affects little influence on the fusion reactor material irradiation tests.

Pre-analysis of the WCLL breeding blanket mock-up neutronics experiment at the frascati neutron generator

Abstract The Water Cooled Lithium Lead (WCLL) is one of the breeding blanket (BB) concept candidate for the future European Demonstration Fusion Power Reactor (DEMO). In the frame of the EUROfusion Consortium, a WCLL BB mock-up neutronics experiment will be performed at the 14 MeV Frascati Neutron Generator (FNG) irradiation facility, devoted to the validation of the Tritium Breeding Ratio (TBR) and the shielding capability predicted by neutronics design calculations. It will include the assessment of the uncertainties on the calculated Tritium Production Rate (TPR) and the neutron/gamma attenuation which will be useful to evaluate design margins required to guarantee tritium self-sufficiency. In preparation of the future experiment, a detailed pre-analysis has been performed with the MCNP Monte Carlo code for the definition of the mock-up assembly and for the optimisation of the experimental setup. The WCLL mock-up has been designed in such a way that the essential nuclear features of the WCLL BB in DEMO can be represented in the experiment and the relevant nuclear quantities can be measurable with sufficient accuracy under FNG irradiation conditions, with several experimental techniques. This study reports the results of the pre-analysis: the optimised experimental layout and its representativeness of WCLL DEMO blanket.

Transient analysis of tritium transport characteristics of thorium molten salt reactor with solid fuel

Tritium management is a vital factor that needs to be considered before a molten salt reactor (MSR) or fluoride-salt-cooled high-temperature reactor (FHR) is constructed. The analysis of tritium transport characteristics under steady and transient conditions is essential in the conceptual design of MSRs and FHRs. In this study, tritium transport characteristics analysis is conducted for a fluoride-salt-cooled pebble bed reactor, i.e., thorium molten salt reactor with solid fuel (TMSR-SF) under accident conditions. The tritium transport characteristics analysis (TAPAS) code is improved with a power model and heat-and-mass transfer model to adapt to TMSR-SF. The transient responses of the tritium transport characteristics of TMSR-SF under the unprotected reactivity insertion accident (URIA) and unprotected overcooling accident (UOC) are determined. It is found that the amount of tritium fluoride (TF) adsorbed on graphite varies, indicating the dynamic equilibrium between the TF adsorption and desorption on graphite. Due to the nuclear density variations of Li-6 and Be-9 in the fluoride salt under the accident conditions, the tritium production rate increases and decreases under the UOC and URIA conditions, respectively. Generally, under these two accident conditions, the tritium permeation rates are lower than that of the normal operation condition. In these cases, it is unnecessary to develop extra safety facilities for tritium control. This study contributes to the tritium management of TMSR-SF.

Evaluation of lithium glass scintillation detector responses for tritium production rate measurement in blanket neutronics experiments using the PHITS code

Lithium (Li) glass scintillation detectors were utilized by the neutron response of a 7Li glass subtracted from the neutron response of a 6Li glass to evaluate the tritium production rate (TPR) in the fusion blanket neutronics experiments, which is a so called Li glass method. TPRs in the Li2O blanket and the Li/V-alloy blanket were measured by a 6Li2O and a 6Li2CO3 pellet method, and by the Li glass method in the benchmark neutronics experiment with D-T neutrons at the Fusion Neutronics facility (FNS) of JAERI. TPRs measured by the Li glass method are lower than TPRs measured by the pellet method. The neutron responses of Li glass detectors have been calculated to study the discrepancies of two different methods by using the PHITS code. In the blanket, the neutron spectra change to harder from the backside to the front side of blanket. The reaction rate of several reactions, such as 6Li(n, n’α)D, 7Li(n, n’α)T, 7Li(n, 2n)6Li, and the elastic scattering of 6Li and 7Li increase with the neutron energy. The low-energy neutron peak was considered only for the calculation of TPR from Li glass, which loses the information of the 6Li(n, α)T reaction from high-energy neutron range. In addition, the reaction rate of this peak evaluated by the Li glass method is lower than real TPR of this peak for 6Li glass caused by weakened effect from other reactions. Therefore, the ratios of this peak reaction rate from the Li glass method against the total TPR of 6Li glass were evaluated to be a correction factor to correct Li glass method. After the correction, TPRs measured by Li glass method agree well with TPRs measured by pellet method.

Numerical benchmark of SuperMC3.2 with HCPB mock-up experiment

In fusion reactors, the blanket is one of the crucial components which play the key role of breeding tritium. This is necessary because there sufficient external tritium supply is not available for fusion reactors. Furthermore, blankets have to provide shielding for ex-vessel components, which are sensitive to the radiation (e.g., Toroidal Field (TF) superconductor coils). To support the development of blanket in EU, two mock-up neutronics experiments based on two European blanket conceptual designs: the Helium Cooled Pebble Bed (HCBP), and the Helium Cooled Lithium Lead (HCLL) have been conducted to validate computational tools and qualify nuclear data. In this work, SuperMC-a nuclear design and safety evaluation software system, developed by FDS Team was validated with HCBP mock-up experiment. SuperMC calculations results are compared with experimental results from SINBAD 2016. The results showed good agreements of reaction rates for the activated foils and tritium production rate (TPR) of the mock-up experiment. This result indicates that SuperMC is a valid and promising 3D neutronics simulation tool for supporting design and safety analysis of fusion reactors such as DEMO.