Neutron Transport Calculations Research Articles

Multiblock Adaptive Mesh Refinement for the SN Transport Equation Based on Lattice Boltzmann Method

Computational accuracy and resource consumption are two sides of the mesh-based neutron transport calculation, whose balance is a common concern in engineering application. To overcome the inflexibility of the multiblock (MB) refinement technique and the complexity of the adaptive-mesh-refinement (AMR) technique, this paper presents a MB-AMR–based neutron transport lattice Boltzmann method (LBM) for the first time, which is a development and in-depth study for the current nonuniform mesh technique. The neutron transport problems are solved using the LBM with finite Boltzmann scheme, and the mesh configuration is adaptively adjusted using the MB-AMR technique. The MB-AMR technique combines the simplicity of the MB technique and the flexibility of the AMR technique and overcomes their shortcomings. By using invariant blocks, the complicated tree structure used in the traditional AMR technique is eliminated. By adjusting the mesh configuration according to the calculation results adaptively, the inflexible of the MB technique is overcome. By using the finite Boltzmann scheme instead of the traditional LBM, the implementation is further simplified and the interface treatment between different blocks can be solved as inner nodes using streaming process. Based on the above advantages and the simplicity of the LBM, the difficulty of the AMR technique in neutron transport calculation has been greatly reduced. To verify the accuracy and flexibility of the proposed MB-AMR–based neutron transport LBM, five benchmark problems are simulated. Results show that the proposed neutron transport LBM can simulate the multigroup transient and steady-state neutron transport problems accurately and that the MB-AMR technique can adaptively adjust the mesh configuration flexibly and simply. This paper may provide some alternative perspectives to realize the nonuniform mesh–based neutron transport solution and a powerful technique for large-scale engineering.

Three-dimensional variational nodal method parallelization for pin resolved neutron transport calculations

Hybrid parallelization of the three-dimensional variational nodal method (VNM) for pin-resolved neutron transport calculations is presented. Sets of node-energy response and associated matrices are assigned evenly to the available set of MPI processors and OpenMP is further employed to parallelize integrations over angle in the response matrix formation. Parallelization of the solution algorithm is accomplished using non-overlapping domain decomposition with one MPI processor assigned to each subdomain. Additionally, the red and black subdomain iterations of the within-group equations are parallelized by OpenMP. Only the pre-calculated matrix sets needed in a subdomain are stored in the corresponding processor. This parallel solution strategy preserves the block Gauss-Seidel red-black iterations employed in serial calculation, with data transfer between processors only after each red and black iteration. Partitioned matrix acceleration is implemented in a similar manner.These parallel algorithms are implemented in the code PANX allowing both higher-order space-angle approximations to be employed and larger spatial domains to be treated than was the case with the serial form of the code. The KAIST problem is used to both evaluate the parallel performance and demonstrate the increased accuracy of the higher-order angular approximations. With 27 MPI processors and 4 OpenMP threads, speedups of nearly 39 are obtained. To demonstrate the applicability of parallel algorithms to larger spatial domains, a pin-resolved, three-dimensional seven-group model of the NuScale modular reactor core with sufficiently high order space-angle discretization is employed using 49 MPI processors and 4 OpenMP threads, and completed in 4 h with P3 approximation.

Numerical analysis of criticality of fuel debris falling in water

The dynamic behavior of fuel debris in water raises the possibility of criticality accidents during debris removal. Hence, accurate evaluations of criticality that consider the dynamic behavior of fuel debris are an essential part of the decommissioning process. In this study, we develop a calculation system that combines the moving particle semi-implicit method and Monte Carlo neutron transport calculation code. We then clarify that it is possible to evaluate criticality safety using the actual dynamic behavior of fuel debris in water with this calculation system. We first carry out validation experiments to compare the experimental and simulation results of solid-fluid multiphase flows calculation. We demonstrate that the MPS method could calculate the dynamic behavior of fuel debris in water, mostly within the 95% confidence interval of experimental results. We then show it could also evaluate criticality during the sedimentation of cubic fuel debris. Moreover, in the case of fundamental conditions such as a large amount of cubic debris falling in water, it was possible that the effective multiplication factor was maximized in the state of falling in water rather than with complete sedimentation.

A deterministic-stochastic energy-hybrid method for neutron-transport calculation

To solve the neutron-transport equation, the deterministic method discretizes the space-angle-energy phase-space to obtain the corresponding algebraic equations. It is superior in computing speed, but short at resonance energy range self-shielding treatment due to the multi-group approximation. In contrast, the stochastic or Monte Carlo method transforms the differential-integral equation into an integral form to construct the sampling space for continuous phase-space simulation. It can guarantee the computing accuracy with sufficient samples, but requires large computing time to simulate these samples. To combine the advantages of these two methods, a hybrid deterministic-stochastic method in incident-neutron energy is investigated in this paper. Considering the geometry handling ability, Method Of Characteristics (MOC) was employed to treat the fast and thermal energy ranges in which cross sections slowly vary along with the incident-neutron energy. Meanwhile, Monte Carlo method was employed to deal with the epithermal energy range where severe resonance effect appears. Different energy ranges are coupled through the neutron scattering and fission contributions between each other. Encouraging conclusions can be demonstrated by the numerical results. (1) The multi-group data library for deterministic calculation in fast and thermal energy ranges and the ACE format data library for stochastic calculation in resonance energy range are consistent if they are made from the same evaluated nuclear data library. (2) The transport correction employed by the deterministic calculation in fast energy range can handle the strong anisotropic scattering effect in the investigated hybrid method in this paper. (3) The hybrid method can combine the advantages of the deterministic and stochastic methods.

On the optimal radial build of a normal aspect ratio tokamak fusion system

The optimal radial build and the system parameters of a tokamak fusion system with a normal aspect ratio were found utilizing a tokamak systems analysis coupled with the neutron transport calculation. The optimum build was determined by the requirements on the shielding, the tritium breeding, and the magnetic flux density at the toroidal field (TF) coil. It was shown that for the given fusion power, a thick outboard blanket with a thin inboard blanket allowed for smaller system size. The case with a smaller aspect ratio allowed for a smaller major radius, R0 and a smaller system size than the case with a larger aspect ratio. With a confinement enhancement factor, H = 1.3, Q >30 was possible for a fusion power larger than 2200 MW for the case with the aspect ratio, A = 3.0; however, Q > 30 was impossible for the case with A = 4.0. The tritium breeding capabilities of the five blanket concepts to be tested as test blanket modules (TBMs) in the International Thermonuclear Experimental Reactor (ITER) were evaluated by varying the blanket thickness and the lithium-6 (Li-6) enrichment. A helium-cooled solid breeder (HCSB) concept showed the best tritium breeding capability while the shielding effect of Pb was noticeable for a helium-cooled lithium lead (HCLL) concept.

Evaluation of Spatial Resolution of Neutron Profile Monitor in LHD

The vertical neutron camera (VNC) has been developed to measure neutron emission profile in deuterium plasmas of the large helical device (LHD). The experiment on spatial resolution evaluation for VNC was carried out in November 2016 by using a 252 Cf neutron source of 800 MBq. The neutron source was introduced into the LHD vacuum vessel through an aluminum pipe from an upper diagnostics port to the equatorial plane. The source was placed at the following two positions. The first position is just on the cylindrical collimator axis, and the second position is in the middle of two neighboring collimator axes. The stilbene scintillation detectors are aligned radially at the collimator end to detect unscattered fast-neutrons generated in a deuterium plasma. The 3-D neutron transport calculations by using a general-purpose Monte Carlo N-particle code 6 reproduced the experimentally obtained neutron counting rate, indicating that VNC has sufficient spatial resolution and the crosstalk between two neighboring collimators is a fairly small fraction. It is expected that the VNC on LHD will function as a tool for the study on the radial transport of energetic ions.

The coupled neutron transport calculation of Monte Carlo multi-group and continuous cross section

It is well known that Monte Carlo neutron transport calculation based on the continuous cross section is with high precision, but the computing time is too long and memory consumption is large relative to the multi-group calculation. In contrast, the multi-group calculation is good both in terms of speed and memory relative the continuous calculation, but the short appears on S(α,β) and resonance energy region. For combining the advantages of these two models, we firstly develop the collision mechanism based on material. Then the coupled model is investigated based on the multi-group and continuous. The continuous model was employed to treat the thermal and resonance energy ranges, the multi-group model uses in the other energy ranges. Different energy ranges are coupled through the neutron scattering and fission contributions between each other. We chose the ITER 1-D sphere model for test example. This example contains a great amount of hydrogen, deep-subcritical and involves thermal scattering and resonance. The result of continuous model is as standard. The cases of criticality fission source and fixed source are considered. The comparison shows good agreement between coupled and continuous model. The speedup of coupled model increases about two times relative continuous model. The validity of coupled model has been proved by numerical simulation.

On the necessity and the role of descriptors of neutron activated structural and shielding materials of nuclear installations for future decommissioning

Existing situation in nuclear industry is characterized with simultaneous development of the following two processes: design and construction of new generation of nuclear installations and decommissioning of installations of older generations. Significant amounts of radioactive wastes generated during the decommissioning phase are determined both for the first and the second types of installations by the induced activity of neutron irradiated structural and shielding materials. Concentration of the so-called radioactivity-hazardous nuclides in primary building and construction materials is the most important characteristics determining the resulting levels of induced activity. Values of these concentrations for the same type of material extracted from different geological deposits may differ by one or two orders of magnitude. Information about concentrations of radiation-hazardous elements in radiation shielding materials is fragmented and, as a rule, unsuitable for practical application. The purpose of the present study was to substantiate the necessity of compiling and recording the data on the concentrations of radioactivity-hazardous nuclides for building and structural materials for nuclear installations during the phases of design, operation and decommissioning. Three types of shielding concrete compositions were selected for the investigation. Concentrations of radioactivity-hazardous nuclides were mainly obtained by neutron activation technique. Neutron transport calculations were performed in one-dimensional cylindrical geometry at the core mid-plane according to usual core-vessel-shielding model of modern VVER reactor unit including 2-m thick concrete shield. Both transport and activation calculations were undertaken using modules of SCALE system. The obtained results allow estimating neutron-induced activation levels in the material as the function of irradiation time, amounts and categories of radioactive waste and their evolution during the decay time from 1 to 100 years. It was established that neutron-induced activity of shielding concrete strongly depends on the actual concentrations of radioactivity-hazardous nuclides in the concrete including ‘trace’ concentrations (other factors being the same). It was also shown that failure to take such concentrations into account may lead to the underestimation of neutron-induced activation levels and amounts of radioactive wastes and their category. The obtained results confirmed the necessity of compiling and maintaining data records on the concentrations of radioactivity-hazardous nuclides for materials used in structural and shielding materials of nuclear installations. Proposals were formulated on the potential accumulation of information, composition and formatting of descriptors of chemical composition of shielding and structural materials of nuclear installations.

The MOCP code: method of characteristics code for pebble-bed HTR

The method of characteristics which is an effective method for neutron transport calculation with high flexibility in complex geometry has been applied to high temperature gas-cooled pebble-bed reactor geometry. The MOCP (the Method of Characteristics for Pebble-bed high temperature gas-cooled reactor) code was developed in Institute of Nuclear and New Energy Technology, Tsinghua University to achieve this goal. The first stone standing ahead is the geometry modeling for pebble-bed district. In this work, a solid modeling method called constructive solid geometry (CSG) is introduced to model the pebble-bed district. A universal ray tracing technique is developed to arrange tracks in 3-D cylinder geometry. Furthermore, tree search method is implemented into MOCP to improve the ray tracing efficiency. Automatic meshing is implemented to improve the geometry modeling efficiency. A redundant tree was constructed to mesh the interspace among pebbles. The geometry modeling capability of MOCP is verified by fast converging single group benchmarks.

Analysis of a loss-of-flow accident resulting from the primary pump shaft break transient of the NUR research reactor

ABSTRACT This paper presents a combined experimental and numerical simulation approach to investigating safety issues related to postulated loss-of-flow accident (LOFA) cases, which are more likely to occur in the NUR Research Reactor (Algeria). The transients investigated at nominal-power operating conditions are related to the loss of flow resulting from an instantaneous shaft break in the main cooling pump of the NUR reactor. The investigations are based on hydrodynamic and thermal hydraulic experiments to assess the reactor cooling system’s behavior. 3D Monte Carlo neutron transport calculations were performed with the (MCNP) code to determine the resulting neutronic properties of the core. In the accident analysis, a model of the primary cooling system was applied via the RELAP5 code. The experimental data and RELAP5 predictions showed good agreement. Additionally, the LOFA due to the transient scenario of the pump shaft break was compared with the LOFA due to normal loss of the coolant pump power transient. The results obtained from the transient (LOFA) studies revealed that in both cases, the lower limit of the minimum critical heat flux ratio and minimum onset of flow instability ratio for NUR is satisfied with a sufficient margin.

The preliminary neutronics analysis for a 10 MW molten salt reactor with solid fuel

The 10 MW MSR-SF is a type of Pebble-bed Fluoride-salt-cooled High-temperature Reactors (PB-FHRs) and was designed by the Shanghai Institute of Applied Physics (SINAP), Chinese Academy of Sciences (CAS). The CAS planned to construct it around 2020 as the initial step with the expectation of a larger 100 MW PB-FHR. The PB-FHRs are composed of randomly distributed fuel pebbles consisting of randomly distributed TRISO coated particles, which is known as the feature of stochastic double heterogeneity. In this work, an explicit modeling method is adopted to consider the stochastic double heterogeneity and the code Serpent is used to execute the Monte Carlo neutron transport calculation. Moreover, the regular lattice method will introduce non-negligible bias in the neutronics analysis for the PB-FHR, while the exact explicit method can provide more reliable results. Therefore, the accurate three-dimensional power distribution of the 10 MW MSR-SF is obtained with the code Serpent based on the explicit modeling approach. The power peaking factor of the 10 MW MSR-SF is small and reasonable.

Conceptual design study for heat exhaust management in the ARC fusion pilot plant

The ARC pilot plant conceptual design study has been extended beyond its initial scope [B. N. Sorbom et al., FED 100 (2015) 378] to explore options for managing ∼525 MW of fusion power generated in a compact, high field (B0 = 9.2 T) tokamak that is approximately the size of JET (R0 = 3.3 m). Taking advantage of ARC’s novel design – demountable high temperature superconductor toroidal field (TF) magnets, poloidal magnetic field coils located inside the TF, and vacuum vessel (VV) immersed in molten salt FLiBe blanket – this follow-on study has identified innovative and potentially robust power exhaust management solutions. The superconducting poloidal field coil set has been reconfigured to produce double-null plasma equilibria with a long-leg X-point target divertor geometry. This design choice is motivated by recent modeling which indicates that such configurations enhance power handling and may attain a passively-stable detachment front that stays in the divertor leg over a wide power exhaust window. A modified VV accommodates the divertor legs while retaining the original core plasma volume and TF magnet size. The molten salt FLiBe blanket adequately shields all superconductors, functions as an efficient tritium breeder, and, with augmented forced flow loops, serves as an effective single-phase, low-pressure coolant for the divertor, VV, and breeding blanket. Advanced neutron transport calculations (MCNP) indicate a tritium breeding ratio of ∼1.08. The neutron damage rate (DPA/year) of the remote divertor targets is ∼3–30 times lower than that of the first wall. The entire VV (including divertor and first wall) can tolerate high damage rates since the demountable TF magnets allow the VV to be replaced every 1–2 years as a single unit, employing a vertical maintenance scheme. A tungsten swirl tube FLiBe coolant channel design, similar in geometry to that used by ITER, is considered for the divertor heat removal and shown capable of exhausting divertor heat flux levels of up to 12 MW/m2. Several novel, neutron tolerant diagnostics are explored for sensing power exhaust and for providing feedback control of divertor conditions over long time scales. These include measurement of Cherenkov radiation emitted in FLiBe to infer DT fusion reaction rate, measurement of divertor detachment front locations in the divertor legs with microwave interferometry, and monitoring “hotspots” on the divertor chamber walls via IR imaging through the FLiBe blanket.

In-Pile Qualification of a Fast-Neutron-Detection-System

In order to ensure the quality and the relevance of irradiation programs in the future Jules Horowitz Reactor (JHR), the French Alternative Energies and Atomic Energy Commission (CEA) has significantly increased its research and development effort in the field of in-pile instrumentation during the last decade. Major progress has been achieved in the capability to perform accurate in-pile measurements using reliable and updated techniques. A significant part of this effort has been conducted in the framework of the Joint Instrumentation Laboratory between the CEA and the Belgian Nuclear Research Centre (SCK $\cdot$ CEN). In order to improve measurement techniques for neutron flux assessment, a unique system for online measurement of fast neutron flux has been developed and recently qualified in-pile. The fast-neutron-detection-system (FNDS) has been designed to monitor accurately high-energy neutron flux (E > 1 MeV) in typical material testing reactor (MTR) conditions, where neutron flux levels can be as high as $10^{15}\,\,\text {N}\cdot \text {cm}^{-2}\cdot \text {s}^{-1}$ and are generally dominated by thermal neutrons. Moreover, the neutron flux is coupled with a high gamma flux of typically a few $10^{15} \gamma \cdot \text {cm}^{-2}\cdot \text {s}^{-1}$ , which can be highly disturbing for the online measurement of neutron fluxes. The patented FNDS system is based on two detectors allowing the simultaneous detection of both thermal- and fast-neutron flux. Thermal neutrons can be measured using a self-powered neutron detector or a 235U miniature fission chamber, while fast neutron detection requires a miniature fission chamber with a special fissile material presenting an energy threshold near 1 MeV, which can be 242Pu for MTR conditions. Fission chambers are operated in Campbelling mode for an efficient gamma rejection. FNDS also includes a specific software that processes measurements to compensate online the fissile material depletion and to adjust the sensitivity of the detectors, in order to produce a precise evaluation of both thermal and fast neutron flux even after long-term irradiation. FNDS has been validated through a two-step experimental program. A first set of tests was performed at BR2 reactor operated by SCK $\cdot$ CEN in Belgium. Two FNDS prototypes were operated in-pile during nearly 1000 h. These tests exhibited the consistency of the measurement of thermal to fast neutron flux ratio with Monte Carlo neutron and photon transport calculations, as well as the right compensation of fissile material depletion. This paper describes a second test recently completed at ISIS reactor operated by CEA in France. For this irradiation, FNDS signal was compared to reference thermal- and fast-neutron flux measurements using activation dosimeters analyzed under COFRAC Quality Certification. During this latter test, FNDS proved its ability to measure online the fast neutron flux with an overall accuracy better than 5%. FNDS is now operational and is assumed to be the first and unique acquisition system able to provide an online measurement of the fast neutron flux in MTR conditions. This system will be used to perform spectral neutron characterization of JHR channels, but it may also be implemented in future irradiation experiments, for a better and real-time evaluation of the fast neutron flux received by material and fuel samples.

Application of Krylov Acceleration Technique in Method of Characteristics–Based Neutron Transport Code

In our earlier work, a computer code based on Method of Characteristics (MOC) was developed to solve the neutron transport equation for mainly assembly-level lattice calculations with reflective and periodic boundary conditions and to some extent core-level calculation with a vacuum boundary condition. Performance of the MOC code was also demonstrated with flat and linear flux approximations. Since neutron transport calculations involve extensive computation, an attempt is made to develop an efficient numerical recipe that will reduce the computation time. First, a conventional MOC solution of the neutron transport equation is transformed into a matrix equation to apply the Krylov subspace iteration method for accelerating the solution. It is found that even in the most sophisticated and compact formats, forming the matrix equation explicitly by storing its nonzero elements requires extremely large computer memory. Hence, an alternate way to apply the Krylov iteration is demonstrated by incorporating the effect of the matrix-based approach into the solution without storing the matrix elements. This computationally viable and novel acceleration technique is used in combination with the existing formalism of flat as well as linear flux approximation to solve a number of benchmark problems. Results show significant improvement in terms of faster convergence of the solution over the conventional inner-outer iteration without compromising accuracy.

A Coarse-Mesh Diffusion Synthetic Acceleration Method with Local hp Adaptation for Neutron Transport Calculations

We present a local adaptive diffusion synthetic acceleration (DSA) method for neutron transport calculations. This new DSA method, called DG-DSA, solves the diffusion equation on a coarse mesh using the interior penalty discontinuous Galerkin (DG) methods. We investigate various numerical aspects of the DG-DSA method such as convergence performance and local adaptation. We demonstrate that our DG-DSA method can effectively and efficiently accelerate transport source iterations.

Monte Carlo method for solving a B1 equation with complex-valued buckling in asymmetric geometries and generation of directional diffusion coefficients

Direction-dependent diffusion coefficients generated using a Monte Carlo method that explicitly solves a heterogeneous B1 equation are validated in this paper. The validation is conducted by comparing the direction-dependent neutron leakage using directional diffusion coefficients with the results of a neutron transport calculation. This paper develops a new method for critical buckling search calculation to solve a heterogeneous B1 equation for an asymmetric unit cell in which the critical bucking has a complex value. A complex-valued critical buckling can be searched using the differential operator sampling method for estimating the sensitivity coefficients of keff with respect to the buckling. In an asymmetric unit cell, the neutron flux and the neutron current also have complex values. The real part of the neutron flux obtained by a critical buckling search calculation represents the neutron spectrum in the center portion of the finite geometry composed of the unit cells.

Experimental analyses of bismuth sample reactivity worth at Kyoto University Critical Assembly

ABSTRACT Sample reactivity worth experiments are carried out by substituting aluminum (Al) plates for bismuth (Bi) ones at the Kyoto University Critical Assembly. At the beginning, uncertainty quantification of bismuth isotope is conducted by deterministic calculations with nuclear data library JENDL-4.0, with the use of experimental results of sample reactivity worth. Then, with the combined use of current (Bi) and previous (Pb) experimental results that demonstrate the comparative difference in the sensitivity and uncertainty of Bi and Pb isotopes, experimental results of cross-section uncertainties of Bi isotope are available for examination of neutron characteristics of Pb–Bi coolant material in the accelerator-driven system. From the experimental analyses, further uncertainty analyses by neutron transport calculations are needed for several reactions of Bi isotope, especially with the use of the covariance data of capture, elastic scattering and inelastic scattering reactions in another nuclear data library.

Development and validation of the code COUPLE3.0 for the coupled analysis of neutron transport and burnup in ADS

The analysis of the fuel depletion behavior is critical for maintaining the safety of accelerator-driven subcritical systems (ADSs). The code COUPLE2.0 coupling 3-D neutron transport and point burnup calculation was developed by Tsinghua University. A Monte Carlo method is used for the neutron transport analysis, and the burnup calculation is based on a deterministic method. The code can be used for the analysis of targets coupled with a reactor in ADSs. In response to additional ADS analysis requirements at the Institute of Modern Physics at the Chinese Academy of Sciences, the COUPLE3.0 version was developed to include the new functions of (1) a module for the calculation of proton irradiation for the analysis of cumulative behavior using the residual radionuclide operating history, (2) a fixed-flux radiation module for hazard assessment and analysis of the burnable poison, and (3) a module for multi-kernel parallel calculation, which improves the radionuclide replacement for the burnup analysis to balance the precision level and computational efficiency of the program. This paper introduces the validation of the COUPLE3.0 code using a fast reactor benchmark and ADS benchmark calculations. Moreover, the proton irradiation module was verified by a comparison with the analytic method of calculating the 210Po accumulation results. The results demonstrate that COUPLE3.0 is suitable for the analysis of neutron transport and the burnup of nuclides for ADSs.

Whole-core forward-adjoint neutron transport solutions with coupled 2-D MOC and 1-D SN and kinetics parameter calculation

The 2-D/1D whole-core transport method has been widely studied for 3-D fine flux or power distributions and implemented in many deterministic codes. KYCORE, a 2-D/1-D transport code, developed a new iteration strategy to calculate forward flux. In order to perform kinetics parameter computation and sensitivity analysis, an adjoint flux solution is required in modern reactor physics analysis. In this study, a new adjoint neutron transport solver, named KYADJ, was developed by utilizing the 2-D MOC and 1-D SN coupling method. A 3×3 lattice test problem and C5G7 OECD/NEA 3-D benchmarks were used to verify the forward-adjoint neutron transport calculation of KYADJ. Forward-adjoint multiplication factors, forward flux, adjoint flux and kinetics parameters were compared with reference results generated by the Reactor Monte Carlo code RMC. Results showed that KYADJ agreed well with RMC and KYADJ had the ability to provide fine pin-by-pin forward and adjoint flux distributions and accurately compute kinetics parameters.

Unstructured coarse mesh finite difference method to accelerate k-eigenvalue and fixed source neutron transport calculations

Unstructured coarse mesh finite difference (CMFD) method was implemented to accelerate the k-eigenvalue and fixed source neutron transport calculations in the nodal SN transport code DNTR, which was based on arbitrary triangular-z meshes. The same triangular-z meshes were adopted in the CMFD calculation to accommodate the arbitrary problem boundaries and simplify the coarse mesh generation and mapping processes. Two-level CMFD formulations were derived for the k-eigenvalue problems, and the few-group CMFD calculation was used to speed up the convergence of the multigroup CMFD calculation. For the fixed source problems in the neutronics analyses of subcritical reactors or transient neutron transport calculations, a special ks iteration method by introducing the source multiplication factor ks was employed in the SN transport calculation to remove the dependence of the convergence rate on the subcriticality. Multigroup CMFD formulations were established to accelerate the convergence of the ks iteration procedure in the fixed source neutron transport calculation. The acceleration effects of CMFD on the k-eigenvalue and fixed source calculations were verified using four representative neutron transport problems, i.e. the small fast breeder reactor with Cartesian boundaries, the BN-600 fast reactor with hexagonal boundaries, the space nuclear power reactor with cylindrical boundaries, and the accelerator driven subcritical reactor with hexagonal boundaries. The numerical results showed that for the first three eigenvalue problems, the multigroup CMFD achieved a speedup of total computation time by a factor of 2–5, and the two-level CMFD reduced the computation time by a factor of 3–7. It was also shown that the multigroup CMFD was able to accelerate the ks iteration procedure in the fixed source problems by a factor of about 6. The implementation of the unstructured CMFD method turns the DNTR code into an efficient neutron transport solver with flexible geometry treatment capability for both k-eigenvalue and fixed source problems.