[en] Highlights: • New data libraries are generated to quantify the thermal neutron scattering effect. • DBRC is implemented in MCNP to quantify the resonance elastic scattering effect. • The impacts of the two negative effects does not interfere each other. • A significant improvement of critical calculation has been made for the FHRs. - Abstract: Two effects in Fluoride-salt-cooled High-temperature Reactors (FHRs) are analyzed based on Monte Carlo simulation. Firstly, the thermal neutron scattering effect of fluoride salt (2LiF-BeF_2) is considered by utilization of the newly generated Thermal neutron Scattering Library (TSL) files. It is found that the neutron spectrum becomes harder and the fission reaction rate of "2"3"5U decreases at thermal energy range due to the up-scattering introduced by thermal neutron scattering effect. Secondly, the resonance elastic scattering effect of heavy nuclides in epithermal energy range is covered by implementation of the Doppler Broadening Rejection Correction (DBRC) method in MCNP. It is shown that neutron up-scattering is enhanced in the low energy wing of the resonance peak and neutron down-scattering is increased in the high energy wing of the resonance peak. This phenomenon leads to an increase in the neutron capture rate of "2"3"8U by about 1.0%. For the analyzed FHR pebble unit cells at 1200 K depending on TRISO packing factor, the thermal neutron scattering effect of 2LiF-BeF_2 and the resonance elastic scattering effect result in a decrease in k_i_n_f of 93–290 pcm and 131–591 pcm, respectively. By taking into account the two effects simultaneously, the k_i_n_f of FHR pebble unit cells decreases by 248–881 pcm.

[en] Highlights: • Improved resonance calculation of FHR based on subgroup method. • A fast resonance interference factor method to treat resonance interference. • Consideration of resonance elastic scattering effect in FHR. - Abstract: The subgroup method is improved in several aspects to address challenges brought on by design features of the Fluoride salt-cooled High-temperature Reactor (FHR). Firstly, the Dancoff correction is applied to resolve the double heterogeneity arising from embedding TRISO fuel particles in the matrix of pebbles. Secondly, a fast Resonance Interference Factor (RIF) scheme is proposed to treat the resonance interference effect in the FHR. In this scheme, the heterogeneous system is converted into a homogeneous one according to self-shielded cross section conservation of the dominant resonant nuclide. The resonance interference effect is considered in the equivalent homogenous system by correcting the non-interfered self-shielded cross sections with RIFs which are obtained by solving the slowing down equation in hyper-fine energy group (∼1M number of energy groups). Finally, the resonance elastic scattering effect becomes considerable due to high temperatures in the FHR. This effect is considered by substitution of the conventional Resonance Integral (RI) table with that generated by the Monte Carlo method. The Monte Carlo method is modified via the Doppler Broadening Correction Rejection (DBRC) method to implement the Doppler broadened scattering kernel. The numerical results show that the Dancoff correction can significantly reduce errors brought about by the double heterogeneity. The fast RIF scheme provides more accurate effective self-shielded cross sections than the conventional iteration scheme. In addition, the speedup ratio of the fast RIF scheme is ∼3.3 compared with the conventional on-the-fly RIF schemes for TRU TRISO. The scheme to generate RI table can resolve the resonance elastic scattering effect encountered by the conventional scheme.

[en] To meet the challenges faced in high-fidelity resonance self-shielding calculations, the global-local self-shielding calculation method is proposed. The resonance self-shielding and correlated effects are classified into global and local effects. The global effects are weak or independent of energy, while the local effects are strong. Therefore, the resonance self-shielding calculation is split into global, coupling and local calculations. Coarse model is built for global calculation and the neutron current method is employed to compute Dancoff correction factors, where the global effects are considered. The coupling calculation is based on preservation of the Dancoff correction factors and the equivalent 1-D models of the fuel rods are obtained. The pseudo-resonant-nuclide subgroup method is employed to perform the local calculation to treat the local effects. This method is realized in NECP-X. The numerical results show that the efficiency of this method is increased by one order of magnitude compared with the conventional method. Besides, the precision of infinite medium multiplication factor is increased by 100 ∼ 300 pcm. (authors)

[en] The CSG (Constructive Solid Geometry) method is widely used in the MC (Monte Carlo) codes because of its flexibility in geometry. In this paper, the CSG method is implemented in NECP-X, which is a newly developed high-fidelity neutronics code, to improve its geometric flexibility. To improve the user experience of using CSG, meshes for the multi-group transport calculation are automatically generated with only several parameters. Several problems are tested. The numerical results show that complicated reactor cores can be conveniently modeled and the results of NECP-X agreed well with those of the MC code. (author)

[en] A new task of using the Jacobian-Free-Newton-Krylov (JFNK) method for the PWR core transient simulations involving neutronics, thermal hydraulics and mechanics is conducted. For the transient scenario of PWR, normally the Picard iteration of the coupled coarse-mesh nodal equations and parallel channel TH equations is performed to get the transient solution. In order to solve the coupled equations faster and more stable, the Newton Krylov (NK) method based on the explicit matrix was studied. However, the NK method is hard to be extended to the cases with more physics phenomenon coupled, thus the JFNK based iteration scheme is developed for the nodal method and parallel-channel TH method. The local gap conductance is sensitive to the gap width and will influence the temperature distribution in the fuel rod significantly. To further consider the local gap conductance during the transient scenario, a 1D mechanics model is coupled into the JFNK scheme to account for the fuel thermal expansion effect. To improve the efficiency, the physics-based precondition and scaling technique are developed for the JFNK iteration. Numerical tests show good convergence behavior of the iterations and demonstrate the influence of the fuel thermal expansion effect during the rod ejection problems.

[en] The free gas model is adopted to consider the thermal scattering effect of the elastic collision between neutron and target. The conventional model assumes that the elastic scattering cross sections are constant at 0 K, which neglects the influence of resonance effect. In order to consider the resonance elastic scattering in the free gas model, the Doppler broadening rejection correction (DBRC) method was applied to correct the free gas model of MCNP. The Mosteller's Doppler defect benchmark for LWR pin cell was analyzed. The numerical results show that neglect of resonance elastic scattering effect contributes to overestimation of the infinite multiplicative factor to the extent of 40-100 pcm and 140-200 pcm for hot zero power and hot full power cases, respectively. The fuel temperature coefficients are also overestimated 7%-15%. The computational time of the newly developed sampling technique was studied and the influence of the resonance elastic scattering effect on the emergent energy distribution was analyzed. (authors)

[en] Considering that the conventional Bondarenko-Iteration Method (BIM) would introduce much error in treating resonance interference effect, the Resonance-Interference-Factor Method (RIFM) and the Heterogeneous-Pseudo-Resonant-Isotope Method (HPRIM) was developed by researchers. Though these two methods can give relatively precise pin-averaged self-shielded cross sections by considering resonance interference effect, they cannot give exact spatial dependent self-shielded cross sections for coarsely modeling spatial self-shielding effect. In order to overcome this problem, the Pseudo-Resonant-Nuclide Subgroup Method (PRNSM) is proposed in this paper. The PRNSM takes into account the resonance interference effect by making resonance cross section table of the pseudo resonant nuclide on-line. The subgroup parameters of the pseudo resonant nuclide and the partial resonant nuclide are obtained by fitting method. The spatial dependent self-shielded cross sections are obtained by condensing the subgroup cross sections of the partial resonant nuclide with the subgroup flux which is obtained by solving subgroup fixed source problem of the pseudo resonant nuclide. The numerous numerical results show that the PRNSM can give both precise pin-averaged self-shielded cross sections and precise spatial dependent self-shielded cross sections for UO₂ problems with different enrichments. (authors)

[en] Based on the ENDF/B-VII. 0 nuclear evaluation data library, an improved multi-group cross section data library NECL2.0 was produced for Bamboo-Lattice code by using the nuclear data processing code NJOY and LATTICE_-PRE. The results based on benchmark and numerical analysis show that the calculation results of infinite multiplication factor k_inf, fission rate distribution and few-group homogenized cross section by using NECL2.0 library agree well with the corresponding reference values. Considering the resonance of Ag-In-Cd can increase accuracy of k_inf by 1000 pcm. Compared with reference values, the maximum relative deviation of fission rate reduces from-0.97% to-0.53%. Considering the resonance of zirconium in cladding can improve the accuracy of k_inf by about 60 pcm. (authors)

[en] A new variance-reduction method, called the first-collision source (FCS) method, is proposed to optimize Monte-Carlo source sampling to reduce variance in deep-penetration problems. In the FCS method, Monte-Carlo samples source particles from the first-collision source distribution rather than the original source distribution. The first-collision source is generated from the original source through a transport process and a scatter process, thus it has wider phase space distribution compared with the original source. Sampling from the first-collision source is capable of obtaining source particles near the aim phase space (specific space area, energy segment, and angle segment which are most related to the aim response) to increase particle popularity in aim phase space. Moreover, a new method called the FCS-CADIS method has also been proposed to further improve the sampling efficiency and performance of variance reduction. The FCS-CADIS method combines the FCS method and the Consistent Adjoint Driven Importance Sampling (CADIS) method. Both of the FCS method and the FCS-CADIS method have been implemented in the hybrid Monte-Carlo-Deterministic particle-transport code NECP-MCX. Three problems are used to assess the performance of the FCS method and the FCS-CADIS method. The results show satisfactory improvement of Monte-Carlo calculation efficiency. (author)

[en] Fully ceramic micro-encapsulated (FCM) fuel is a type of dispersed particle fuels. Due to the double heterogeneity of dispersed particle fuels, the traditional deterministic method and Monte Carlo method are difficult to be applied to it. In order to overcome this problem, the multi-group effective cross section calculation method for FCM fuel was proposed. To describe the heterogeneity of TRISO particles, the hyperfine group disadvantage factors were calculated by solving a one-dimensional model of the TRISO particle with hyperfine group method in resonance energy range. The matrix and TRISO particles will be homogenized by correcting the hyperfine group cross section with hyperfine group disadvantage factors. Due to the strong self-shielding in thermal energy range, multi-group disadvantage factors were calculated by penetrating probability and collision probability equivalent, and then the matrix and TRISO particles were homogenized by correcting the multi-group cross section with multi-group disadvantage factors. The FCM fuel is able to be calculated as traditional PWR fuel after homogenized. The one-dimensional model of every fuel rod in fuel assembly will be got by Dancoff correction factor equivalent. And then the hyperfine group calculation is carried out based on the one-dimensional rod model to get the effective cross section in resonance energy range. The numerous numerical results show that the proposed method can deal with the double heterogeneity of FCM fuel and get precise effective self-shielding cross section. (authors)