Shutdown Dose Rate Research Articles

Particle-transport/depletion/activation/source-term coupling analysis for CFETR with NECP-MCX

As a bridge between ITER and Chinese DEMO, the design of Chinese Fusion Engineering Test Reactor (CFETR) is being carried out. The tritium breeding ratio (TBR) and the ShutDown-Dose-Rate (SDDR) are two significant neutronics parameters for CFETR and other fusion reactors. However, in the previous researches, the variation of the TBR with time was analyzed without consideration of the activation of beryllium. In the analysis of the SDDR, conventional methods assumes that the neutron spectra are not changed in the whole operation duration. To analyze the TBR and the SDDR with higher precision, the functionalities of the particle-transport/depletion/activation/source-term coupling analysis have been developed based on NECP-MCX, which is a Monte-Carlo-Deterministic particle-transport code developed by the Nuclear Engineering Computational Physics (NECP) Lab at Xi'an Jiaotong University. The transmutation solver NECP-Erica, developed also by the NECP Lab, is coupled with NECP-MCX internally. The integrated analysis code, which inherits the name NECP-MCX, is verified by several fusion-related experimental benchmarks. The verification results show the reliability of NECP-MCX. After the verification, the particle-transport/depletion/activation/source-term coupling analysis of CFETR is performed with NECP-MCX. Numerical results show that ignorance of the activation of beryllium overestimates the TBR of CFETR by 1.42% after an operation time of 16 years with a capability factor of 0.5. In the analysis of the SDDR, the ignorance of spectrum variation brings ∼2.0% discrepancy in the activity and decay heat of the structure materials.

Status of Scoping Nuclear Analyses for the Evolving Design of ITER TBM Port Cells

ITER is an international collaborative effort towards the realization of fusion energy via the magnetic confinement concept. Two of the equatorial ports in the facility are dedicated to the testing of tritium breeding concepts, which is essential for the tritium self-sufficiency of future fusion reactors. The concerned Test Blanket System (TBS) consists of a Test Blanket Module (TBM) residing inside the TBM–Port Plug (TBM-PP) and its associated ancillary systems in the Tokamak facility. In this paper, the results of a full suite of nuclear analyses concerning the shielding performance of the Pipe Forest (PF) and Bioshield Plug (BP), to reflect on the evolution of their designs, are discussed. On the BP side, the design of the peripheral part has been reviewed considering the ventilation openings and butterfly doors, to assure the design compliance with the Radiation Map (RadMap) requirements for the neutron flux in the Port Cell (PC), behind the BP. On the PF side, the pipes routing and maintenance corridor door have been redesigned, by taking into account results from previously concluded nuclear analyses. The neutronics model was developed from CAD and was used to perform transport simulations in two plasma modes: on and off. For plasma-on mode, the plasma neutron field in the Port Interspace (PI) as well as behind the BP was assessed and few shielding options were explored. The responses due to decay neutrons from 17N in activated cooling water were also considered. For the plasma-off mode, the focus was shifted to further refine the ShutDown Dose Rate (SDDR) maps, which is of importance for maintenance operations that are foreseen to take place at various stages of ITER operation, in particular following the FPO-1, FPO-2, and Short operation scenarios. In addition, detailed activation analyses were carried out to provide a provisional waste classification.

Effect of non-nominal port gaps on nuclear responses in ITER

The size of the gaps around the ITER equatorial and upper ports has associated a certain tolerance due to manufacturing and installation. This variability has an impact on the nuclear responses at magnets and port interspace maintenance locations. This effect is not usually considered as the ITER neutronic reference models represent the nominal gaps configuration. This paper aims to estimate the effect of the gap size variability on nuclear heating in the magnets during plasma operation and shut-down dose rate at the port interspace maintenance location taking advantage of a set of MCNP simulations with different gap configurations. The effect has been shown to produce increases of up to 30% in the worst-case configurations. The novel modelling approach, which was developed ad-hoc and used to manipulate the size of the port gaps via displacements of the geometry universes, is also described here. This approach led to important time savings, and it may be used in future studies with similar requirements.

A digital twin for 64Cu production with cyclotron and solid target system

One method for finding reliable and cost-effective solutions for designing radioisotope production systems is represented by the “digital twin” philosophy of design. Looking at cyclotron solid targets, uncertainties of the particle beam, material composition and geometry play a crucial role in determining the results. The difference between what has been designed and what can be effectively manufactured, where processes such as electroplating are poorly controllable and generate large non-uniformities in deposition, must also be considered. A digital twin, where the target geometry is 3D scanned from real models, can represent a good compromise for connecting “ideal” and “real” worlds. Looking at the 64Ni(p,n)64Cu reaction, different Unstructured-Mesh MCNP6 models have been built starting from the 3D solid target system designed and put into operation by COMECER. A characterization has been performed considering the designed ideal target and a 3D scan of a real manufactured target measured with a ZEISS contact probe. Libraries and physics models have been also tested due to limited cross-section data. Proton spectra in the target volume, 3D proton-neutron-photon flux maps, average energies, power to be dissipated, shut-down dose-rate, 64Cu yield compared with various sources of experimental data and beam axial shifting impact, have been estimated. A digital twin of the 64Ni(p,n)64Cu production device has been characterized, considering the real measured target geometry, paving the way for a fully integrated model suitable also for thermal, structural or fluid-dynamic analyses.

Nuclear Analysis for the Upper Ports in the NB Cell in ITER

In International Thermonuclear Experimental Reactor (ITER), 14 upper ports (UPs) are allocated for diagnostic systems mounted into the port plug. The upper port plug (UPP) consists of the port plug (PP) structure, diagnostic shield module (DSM), and diagnostic first wall (DFW). One of the main functions of the UPP is to provide the neutron shielding to reduce the radiation level in port interspace and the nuclear loads on the superconducting coils. In this work, the nuclear analysis of the UPs #4–#7 that are located in the neutral beam cell is presented. Two configurations have been analyzed with Monte Carlo N-Particle (MCNP) code: the generic UPP design of full shielding (22 tons) and a light version (18 tons) with reduced shielding. The UPP MCNP models have been developed on the basis of latest design files and integrated in a 120° neutral beam injection (NBI) sector extracted from E-lite. Calculations have been performed to evaluate the neutron flux at the port interspace, the nuclear heating of the PP and toroidal field (TF)/poloidal field (PF) coils (TFC/PFC), the neutron flux, and damage for the electrical feedthrough installed on the closure plate with D1SUNED v3.1.4 code. The impact of crosstalk with NBI and neighboring ports has been assessed for neutron flux and shutdown dose rate (SDDR) at 12 days after shutdown. The results of this analysis showed that the impact of the additional shielding of the 22 tons model compared to the 18 tons model is small and limited in space: the fluxes and damage at the electrical feedthrough are about 40%–50% smaller with additional shielding. Analyses enhanced a large radiation crosstalk with NBI ports and a large streaming through the gaps of UPPs.

Dose Rate Measurements During the Tritium Campaign at JET and Diagnostic Improvements for the Deuterium–Tritium Experiments

The technological exploitation of the deuterium–tritium campaign at JET is the objective of the EUROfusion work package JET3 and, through the NEXP subproject, it intends to take advantage of the expected significant neutron production to validate the numerical tools used for shutdown dose rate (SDR) assessment and to study the reasons of discrepancies between calculations and measurements. The tritium campaign C40 was held in the first half of 2021 and two active gamma dosimeters based on spherical 1-L air-vented ionization chambers (ICs) were installed in some ex-vessel positions close to the horizontal ports of the tokamak in octants 1 and 2, have been employed to measure the dose rate. Moreover, for the deuterium–tritium campaign C41 (DTE2), a third IC suitable for higher dose rates was installed in octant 1. The present work is dedicated to the analysis of dose rate measurements carried out during the tritium campaign during the inter-pulse periods and the shutdown. Influence quantities and error sources are analyzed in order to calculate the dose rate from the raw signal and experimental uncertainty. Some improvements in the experimental equipment are described as well, as the assessment of the sensitivity of the ICs to the variation of oxygen concentration in the air (to simulate the nitrogen-enriched atmosphere in the JET torus hall during tritium operation) and the upgrade of the acquisition software to interface it with CODAS data store.

Uncertainty Quantification of the Decay Gamma Source in Mesh-Based Shutdown Dose Rate Calculations

In fusion energy systems, part of the design effort is dedicated to the assessment of the shutdown dose rate (SDR) due to the decay photons that will be emitted from activated components. Monte Carlo transport codes are often used to obtain the neutron flux distribution in the problem domain. The neutron flux distribution is used in the rigorous 2-step (R2S) workflow to obtain the photon emission density distribution of decaying radionuclides. The photon emission density is then used as an input for a dedicated photon transport step to calculate the SDR. In this paper, the uncertainty of the decay gamma source due to the uncertainty of the neutron flux distribution in the R2S workflow is investigated. A scheme is developed to estimate the uncertainty of the decay gamma source, building on the concept of groupwise transmutation and using standard error propagation techniques. The applicability of the newly developed scheme is then demonstrated on one of the conceptual designs of the fusion nuclear science facility.

Scoping Nuclear Analyses of Shielding Options and Shutdown Dose Rate Contributions in ITER TBSs

One of the advances in the test blanket module program within the ITER project in the last few years concerned the evolution of the pipe forest (PF) and bioshield plug (BP) designs. In support of the design phase, nuclear analyses to assess several responses in the fusion neutronics environment inside the port interspace (PI) with the existence of the evolved PF and BP are deemed essential. Nuclear analyses were commenced using the new PF and BP with developing the neutronics models and performing preliminary assessment of the radiation fields and shutdown dose rate (SDDR) in the PI. In this paper, the results of a full suite of nuclear analyses are discussed, which covers more configurations and radiation sources, in two plasma operational modes: on and off. For the plasma-on mode, different shielding options were examined. The results show a clear benefit of combining the installation of shielding panels on the PF enclosure with those in the BP “dogleg,” through which the pipes penetrate to the port cell area. For the plasma-off mode, the SDDR was assessed from different sources: activated components and residual LiPb layers in pipes after drainage. As maintenance operations are foreseen during the lifetime of the facility, the SDDR was also assessed for access conditions, open BP doors, and transport conditions, with PF extracted in the gallery.

Analysis of discrepancies between D1S and R2S results of 2016 DD JET Campaign

In the frame of the WPJET3-DT Technology project within the EUROfusion Consortium program, neutronics experiments are being carried out in preparation for the current deuterium–tritium campaign on JET (DTE2). The experiments are conducted to validate the neutronics codes and tools used in ITER, thus reducing the related uncertainties and the associated risks in the machine operation. Recently, the shutdown dose rate (SDR) results of the 2016 DD campaign have been analyzed using the D1S and R2S computational methodologies showing relevant differences between the results provided by both methods. The greatest discrepancies, found in the SDR calculated in the second octant at 6 h after shutdown, are a factor 2 of difference in the total decay photon flux and a different peak location close to 0.511 MeV. This article presents a detailed analysis of the D1S and R2S simulations performed to find the causes of these discrepancies. The conclusions of the analysis determine that the use of the typical neutron and gamma energy structures used in common R2S simulations can tend to a photon flux significant overestimation or a shift in the output spectrum, respectively. In addition to alerting about this important problem in R2S calculations, this article also proposes a solution to mitigate each effect that can be implemented in R2S codes. The application of the proposed solutions significantly reduces the differences between the photon spectrum calculated with D1S and R2S methods.

Implementation and application of PyNE sub-voxel R2S for shutdown dose rate analysis

PyNE R2S is a mesh-based R2S implementation with the capability of performing shutdown dose rate (SDR) analysis directly on CAD geometry with Cartesian or tetrahedral meshes. It supports advanced variance reduction for fusion energy systems. However, the assumption of homogenized materials of PyNE R2S with a Cartesian mesh throughout a mesh voxel introduces an approximation in the case where a voxel covers multiple non-void cells. This work implements a sub-voxel method to add fidelity to PyNE R2S with a Cartesian mesh during the process of activation and photon source sampling by performing independent inventory calculations for each cell within a mesh voxel and using the results of those independent calculations to sample the photon source more precisely. PyNE sub-voxel R2S has been verified with the Frascati Neutron Generator (FNG)-ITER and ITER computational shutdown dose rate benchmark problems. The results for sub-voxel R2S show satisfactory agreement with the experimental values or reference results. PyNE sub-voxel R2S has been applied to the shutdown dose rate calculation of the Chinese Fusion Engineering Testing Reactor (CFETR). In conclusion, sub-voxel R2S is a reliable tool for SDR calculation and obtains more accurate results with the same voxel size than voxel R2S.

A new novel-1-step shutdown dose rate method combining benefits from the rigorous-2-step and direct-1-step methods

A new method for the calculation of Shutdown Dose Rates (SDDR) has been developed, the Novel-1-Step (N1S) method. The new method retains the benefits of only requiring a single radiation transport calculation, as in the use of the direct-1-step (D1S) method, while removing the need for pre-calculations to determine dominant nuclides and time correction factors. The N1S method uses a time dependent source and decay data for all nuclides. When reactions in the transport occur leading to unstable daughter nuclide, the correct contribution of photon radiation from all the decay products of a nuclide are calculated with no need for additional external activation calculations. Weights of these decay photons are calculated for each decay time of interest and are analytically determined based on the solutions to the Bateman equations.The N1S method has been implemented into MCNP and preliminary verification calculations performed. These calculations included the FNG ITER shutdown dose rate benchmark and the ITER SDDR cross comparison. For the FNG ITER SDDR benchmark the N1S method showed good agreement, within experimental error, for the first campaign apart from the first decay time where a C/E value of 1.34 was obtained. This overestimation was shown to be due to the decay of 64Cu inside the copper cup of the neutron generator. For the second campaign the N1S method showed an under prediction of up to 20% at short decay times and an over prediction up to 20% at longer decay times. These times are dominated by 56Mn and 58Co respectively and it is likely the difference is due to under and over predictions in the reaction rates leading to these nuclides. The ITER cross comparison showed good agreement between the N1S method and MCR2S (and by association other D1S and R2S codes). Differences seen in the results were shown to be due to difference in the calculated reaction rates using EAF2010, TENDL2019 and FENDL3.2.

Development of unstructured mesh tally capability in cosRMC and application to shutdown dose rate analysis

Accurate estimation of shutdown dose rate (SDDR) is one of the key issues in the neutronics calculation of fusion reactors. cosRMC R2S was developed based on the Rigorous two-step (R2S) method using a mesh tally for shutdown dose rate calculation, which requires higher transport calculation accuracy in complex geometry. However, the traditional method in this approach was based on the orthogonal structured mesh. The orthogonal structured mesh does not necessarily coincide with geometry and material boundaries, which can introduce large errors in activation calculation and reduce the calculation accuracy of shutdown dose rate. To solve this problem, the unstructured mesh tally capability was developed in the Monte Carlo code cosRMC. On this basis, cosRMC R2S was further developed, using unstructured meshes to replace structured meshes. Due to the geometry particularity of the unstructured tetrahedral mesh, the rejection sampling strategy was adopted to sample the position of the decay gamma source. In order to verify the reliability and validity of the code, the verification analyses were carried out on the ITER port plug benchmark and the FNG ITER shutdown dose rate benchmark respectively. The results obtained with unstructured meshes were compared with those obtained with structured meshes, and the reliability of the code was preliminarily verified. The results showed that the maximum discrepancy between experimental results was 10.1% with unstructured meshes, which was closer to experimental results than that with structured meshes, indicating that the computational accuracy of the result was improved effectively. It is demonstrated that the code developed in this paper is capable of accurate shutdown dose field calculations and has the potential to be applied to complex shielding devices.

Assessment of shutdown dose rates in the ITER Collective Thomson Scattering system and in equatorial port plug 12

The ITER Collective Thomson Scattering (CTS) system will be the main diagnostic responsible for measuring the velocity distribution function of fusion-born alpha particles in the plasma. As the CTS diagnostic is integrated in the equatorial port plug 12 (drawer 3), with direct apertures to the port interspace where maintenance hands-on operation will be carried out, it is essential to assess the shutdown dose rates (SDDR) in these maintenance areas. In this work, the D1S-UNED3.1.4 Monte-Carlo transport code, based on the implementation of the direct-one-step methodology in MCNP5 v1.60, was used to estimate the dose rate level 12 days (106 s) after shutdown in the port interspace. The results show that the CTS system does not contribute significantly to the SDDR in the area where hands-on maintenance is foreseen with contribution to dose rates less than 1 µSv/h. This is consistent with previous estimates, although with the most recent model of the CTS design there is a slight increase of the SDDR values. This deviation can be attributed to design changes and improved shielding modelling and/or most importantly, to statistical fluctuations of the D1S simulations. From a neutronics point of view, the increase in the SDDR falls within the range of the statistical fluctuations, and the design is still compliant with the radiation safety ALARA principle aiming at minimizing radiation doses, and there is no requirement for further design optimizations.

RayActive: An all-in-one R2S neutron induced activation analysis tool based on new intrinsic resolution methods.

Neutron production in nuclear facilities can induce a SDDR (Shutdown Dose Rate) even after months. Different methods exist to simulate the SDDR of neutron activated materials. The most used method is the mesh-based R2S (Rigorous-Two-Steps). This method couples a Monte-Carlo code and an isotopic inventory code to compute the SDDR. The Monte-Carlo code computes the neutron flux on a user-defined mesh to take into account the neutron flux gradients in the cells. The aim of this article is to present RayActive, a prototype tool elaborated for computing the SDDR. RayActive is based on a modified R2S method with an almost exact identification of materials enclosed in each voxel of a superimposed mesh. The CAD based geometry definition for the model makes possible this identification. Moreover, RayActive is designed as an ”all-in-one” package without any other code coupling. RayActive is designed to be as easy to use as possible and each part of the computation has been optimized to minimize the computational time. Verification and validation of RayActive has been done on shutdown dose rate benchmarks made with FLUKA code and classical R2S process coupling MCNP and FISPACT codes. Good agreement was found for these benchmarks even if some differences occur for the validation benchmark. These discrepancies show the impact on the results of the modified R2S method employed in RayActive.

Shutdown dose rate assessment of the interspace components of wide angle viewing system diagnostic for ITER

Shutdown dose rate (SDDR) calculations have been performed to give support to the design of the Interspace (IS) VIS/IR WAVS diagnostic components of the Equatorial Port 12 (EP12), in the framework of the F4E-FPA-407 SG04 project. The Vis/IR WAVS comprises the viewing systems in four Equatorial Ports: 3, 9, 12 and 17, reminding that the present results are relative to the EP12 specifically. The Vis/IR WAVS is an optical diagnostic aimed at monitoring in visible and infrared radiation the ITER plasma facing components for machine protection.This paper describes the methodology used and shows the results obtained. The objective of the assessment is to evaluate whether the design meets the ALARA criteria in the IS area. In this work, once the materials have been selected to contribute as less as possible to the activation, the main criterion is that the contribution of all diagnostics installed in the port should not overcome the target of the biological dose rate of 100 μSv/h in the IS area after 106s of decay time.The main conclusion of the analysis is that the contribution of the VIS/IR WAVS IS components to the SDDR meets with the ALARA criterion in the IS. The contribution of the VIS/IR WAVS IS components to the SDDR in the IS, concerning the contribution of the whole ITER model, is about 9.6% in the left corridor and 5.5% in the right corridor. Besides, the percentage of contribution of the SS316L structure with respect to the total SDDR induced by the IS VIS/IR WAVS components is 99.3% in the left corridor and 99.5% in the right corridor. However, it has to be taken into account that these values have been obtained using a local model, therefore these values will be lower than the realistic values. Therefore, these values should be taken only as relative values.

Development of shutdown dose rate calculation code based on cosRMC and application to benchmark analysis

Accurate estimation of shutdown dose rate (SDR) has important reference significance for the design, operation and maintenance of fusion devices. In order to obtain the decay gamma source distribution with higher spatial resolution, a shutdown dose rate calculation code was developed based on the Rigorous two-step (R2S) method. The Monte Carlo code cosRMC and the radioactivity inventory code ALARA were coupled to calculate SDR in complex geometry based on mesh-tally. The ray tracking method was used to obtain the voxel-averaged nuclide density, and a decay photon sampling module was developed to accurately describe the distribution of the activation source. In order to improve the computational efficiency in deep penetration problems, the global variance reduction technique based on iterative weight window was adopted. The developed code was firstly verified with rectangular parallelepiped model consists of steel and water, in which the effect of bremsstrahlung photon was also discussed. Then the code was applied to the calculation and analysis of the ITER port plug benchmark, by comparing with other SDR calculation codes. The influence of different activation libraries on the decay gamma spectra and the effect of mesh resolution on the calculation accuracy were investigated. The results showed that the SDR function in cosRMC code is able to calculate SDR accurately with high spatial resolution for deep penetration problems.

Development of a methodology to estimate the statistical SDR uncertainty with R2S-UNED

The Rigorous-Two-Steps (R2S) is one of the most useful methods to estimate the Shutdown Dose Rate (SDR). The most advanced R2S tools couple neutron and photon transport, which are often simulated using Monte Carlo (MC) codes, through an activation simulation using mesh-based techniques to improve the spatial resolution of the neutron flux and the decay gamma source. One of the problems of the methodology is that the statistical uncertainty of the neutron flux due to the MC method used by the transport codes is not considered by most R2S implementations. Consequently, larger tolerance must be assumed affecting to the design of the nuclear facilities. This article describes a scheme allowing the calculation of the SDR statistical uncertainty without any additional assumptions than those used in the R2S methodology. The approach proposed in this article is suitable for cell- and mesh-based R2S implementations. In this work, the methodology was implemented in the R2S-UNED code. The accurate application of the methodology requires the full the neutron flux uncertainty (covariance matrix) as input data. MCNP was modified to calculate this matrix, although, it cannot be calculated for most of the realistic R2S simulations due to its size. If that is the situation, we propose a guideline to reduce the size of the covariance matrix to be calculated according to its element contribution to the SDR. When this guideline cannot be applied, the methodology still allows calculating the upper and lower SDR uncertainty bounds. In this article, the guideline is applied to the calculation of the SDR uncertainty in the computational benchmark of ITER. In addition, we also study the possible impact of the neutron flux correlation degree on the SDR uncertainty in this benchmark.

Nuclear data for D1SUNED for the study of ITER planned in-situ maintenance dose scenarios

ITER will be the largest Tokamak in the world, and it is expected to produce a total of up to 2.8•1027 neutrons during its operation. The Tokamak infrastructure will become radioactive as the machine operates, which is an aspect that needs to be considered in the design of the planed in-situ maintenance activities. The occupational radiation exposure (ORE) must be accounted for, and an ALARA program must be implemented to minimize it. It requires the determination and analysis of shutdown dose rates fields, the purpose of D1SUNED tool. Common to all Direct one-step-based approaches, this tool needs to be fed with a nuclear data set tailored for each problem and produced ex professo. In this paper, we characterize the radioactive inventory relevant for the ORE of ITER maintenance activities using ACAB activation code. The ITER reference irradiation scenario SA-2 and the cooling times of interest 105 s and 106 s are considered. A comprehensive list of typical materials has been studied with detailed impurities content. The nuclear data set covers 99% of the radioactive inventory of concern, and all the pathways responsible for 99% of its production. It has been prepared following the ITER recommendations of cross-sections libraries and it has been verified against Rigorous-two steps calculations. This batch of nuclear data complements D1SUNED for shutdown dose rate exhaustive analysis for ITER with the perspective of maintenance according to the latest ITER nuclear data standards.

Shutdown dose rate calculations with modified DEMO single sector model

A new approach for particle transport was used to calculate the Skyshine dose rates for the DEMO tokamak for the unsymmetrical case with one breeding blanket module removed for maintenance. Transport calculations for the large tokamaks, including DEMO and partly ITER, are usually performed in sector models postulating cyclic symmetry of geometry and source terms and applying reflecting boundary conditions on the sector edges. In the case of non-symmetric localized sources the sector models are not suitable.The calculation of shutdown dose rates, caused by one single tritium breeding blanket module moved from DEMO vacuum vessel into the upper maintenance hall, is an example of a localized radiation source and cannot be calculated in the standard way using a sector model. A new approach was followed for the calculations using the same sector model of DEMO, however with an according modification of the MCNP transport code. In this article the new approach is presented including its verification based on a simpler baseline model of DEMO. The results show that the obtained radiation dose distribution due to the one moved breeding blanket module differs by a factor of up to 2.8 from what would be obtained using a single sector model.

Novel concept of neutron shield for ITER diagnostic Lower Port Integration

Neutron shield for ITER Diagnostic Lower Port Integration in Vacuum Vessel consists of Diagnostic Rack that put tenant’s diagnostic on it and Backplates. This paper proposes a novel concept of Neutron shield that uses Boron carbide powder canned in a stainless-steel box. A prototype was made to demonstrate the novel concept of Neutron shield, and it confirmed the density of B4C for the prototyping reached 1915 kg/m3, 76 % of the actual density of 2520 kg/m3. Neutronics was performed by applying the density confirmed by the prototyping. As a result, 45 μSv/h of Shut Down Dose Rate (SDDR) was achieved and satisfied its target. And additionally, 9.35 ton of neutron shield weight satisfied its target.