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Session Chair: Jorge Gonzalez-Amoros, Seoul National University, Korea, Republic of (South Korea)
Validation of a Multi-Group Pin Homogenized SP3 Code SPHINCS through BEAVRS Benchmark Analyses
Jorge Gonzalez-Amoros, Hyunsik Hong, Hyun Ho Cho, Han Gyu Joo
Seoul National University
The achieved high computing capabilities have eased the computational burden of Direct Whole Core (DWC) calculations. However, the computational resources required for DWC generalization for core design and analysis purposes are still limited and the two-step method with spatial and energy refinement is still essential in industry practical applications.
In this regard, the NTRACER/SPHINCS two-step calculation system that involves the multi-group, pin-by-pin, Simplified P3 (SP3) SPHINCS code is developed at Seoul National University. The finite difference method (FDM) and the superhomogenization factors (SPH) are introduced in SPHINCS to account for the errors associated with the use of pin size FDM as well as cell homogenization. The pin homogenized MG group constants are generated by single assembly nTRACER calculations. In order to correct the leakage effect in pin homogenized cross sections in the core calculation, the 3-group pinwise Leakage Feedback Method (LFM) is applied in SPHINCS.
For the validation of the NTRACER/SPHINCS code system and the associated calculation methodology, the BEAVRS benchmark is solved and the results are compared with the DWC results obtained by nTRACER for the whole core geometry. The comparisons show good agreement between the two results in that a few pcm difference of reactivity and a pin power RMS error less than 0.5% are observed.
In addition, the effectiveness of the two correction methods, namely the SPH factor method and the three group pinwise LFM is investigated. It is demonstrated that the SPHINCS calculation for an assembly with SPH factors produces exactly the same reaction rate distribution as the reference nTRACER code for assembly problems. The 3G pinwise LFM is proved to reduce effectively the error of SP3 pin-by-pin core calculation results.
Implementation and Verification of Explicit Neutron/Photon Heating in nTRACER
Seungug Jae, Namjae Choi, Han Gyu Joo
Seoul National University
In the conventional reactor analysis, the heat generated from fission has been considered to be deposited in the very location of the reaction. In reality, the fission energy released is transferred to kinetic energies of fission products and the energies of the particles are deposited in other places of collision and scattering. Therefore, the approximate heat model would be valid only for infinite fuel cell geometries. It is not valid in actual core calculations involving structural materials and burnable absorbers. In particular, photons which have a smaller energy portion than neutrons, yet longer mean free paths, could induce non-negligible impacts on the low power pins such as gadolinia fuels. In the nTRACER, a gamma transport scheme is applied to resolve this issue and the explicit heat model is adopted, which considers the actual heat deposition by neutrons and photons. As photon induced fissions are neglected in the calculation, one-way relationship between photon and neutron exists where photons are generated by neutron-induced reactions, but not vice versa. The photon production cross sections and matrices are generated by the GROUPR module of NJOY just like the reaction cross sections and the scattering matrices for neutrons. The photon reaction cross sections and scattering matrices are, however, generated by the GAMINR module. The heat deposition of particles is represented with KERMA factors, which are generated with the HEATR and GAMINR module for neutrons and photons, respectively. The transport of photons in nTRACER is solved by using a method similar to that of neutrons, namely, the planar MOC. This neutron/photon heat deposition method is verified with the results of stochastic neutronics code McCARD for an extensive set of problems spanning pin cells and cores. It turned out that the accuracy of the combined heat distributions is similar to that of fission only power distributions.
Oskarshamn-2 instability modeling in TRACE/PARCS with nuclear libraries generated from CASMO: Comparison with reference plant data and SIMULATE
Antonella Labarile, Teresa Barrachina, Rafael Miró, Gumersindo Verdú
Universitat Politècnica de València
The framework of nuclear safety and reactor design requires computational tools to analyze the multi-scale nature of the system. The reactor core analysis involves multi-scale phenomena that are affected by heterogeneities, where a local perturbation can affect the properties of the core and its behaviour. Consequently, one should take into account the multi-scale heterogeneity even in the one-phase calculation.
Some computational tools have been developed in the past to carry out a multi-scale neutronic calculation, in both commercial and research fields. With these tools, from the single-assembly lattice calculation, one can generate reliable data for a nodal code simulation to study the parameters related to nuclear safety at the core level.
The purpose of this work is to obtain an accurate power distribution of an LWR using the 3D nodal code PARCS and neutronic libraries generated from Polaris lattice code. The methodology is applied to the Oskarshamn-2 BWR in the framework of the OECD/NEA Stability Benchmark, for which reference data from CASMO and SIMULATE codes are available to the participants.
The single-assembly calculation is carried out in Polaris including history and branches calculation along the burnup, to obtain neutronics and kinetics data at lattice level. Large buffer zones surrounding each fuel assembly are considered to obtain data that permits accurate nodal calculation of core power distributions.
The multi-group diffusion constants are generated in the PMAXS (Purdue Macroscopic Cross Section) files by the GENPMAXS code. These files are required by the neutron diffusion code PARCS for a whole core simulation, both steady-state and transient. The results are finally verified with reference data to study the main differences due to the heterogeneities and cross-sections parametrization in the generated neutronic libraries in PARCS.
Development and Assessment of ESCOT Pin-wise Thermal-Hydraulics Coupling in a Direct Whole Calculation Code nTER
Alberto Facchini1, Jaejin Lee1, Jin Young Cho2, Han Gyu Joo1
1Seoul National University; 2Korea Atomic Energy Research Institute
High-fidelity multi-physics simulation with coupled neutronics and thermal-hydraulics (T/H) codes for the whole core of light water reactors has become a critical issue when performing thorough core design analyses. Considering the compensating relationship between accuracy and computational time, a drift-flux model based pin-level core T/H code ESCOT (Efficient Simulator of Core Thermal-Hydraulics) has been coupled to a direct whole core calculation code nTER (Neutron Transport Evaluator for Reactors). The nTER code of KAERI employs the planar method of characteristics (MOC) solution based transport calculation scheme where the axial simplified PN solver is implemented within the 3D CMFD framework. nTER has its own simple internal T/H solver which is for closed channels experiencing no pressure drop. The ESCOT code of SNU employs the four-equation drift flux model and the solution is parallelized by using MPI with axial as well as radial domain decomposition. ESCOT is capable of accurate prediction of cross-flow, spacer-grid effect and fuel heat conduction in contrast to the simple internal solver. In addition, the ESCOT fuel conduction model accepts subpin level power, burnup and gadolinium fractions for a realistic estimation of fuel temperature. The coupling scheme and the solution accuracy of the nTER/ESCOT code is described in this paper. The assessment of the coupled code has been carried out through two categories of calculations: feedback and burnup analyses. The nTER’s standalone solution and the nTER/ESCOT coupled solution are compared with the OPR-1000 core model also with the VERA benchmark models (single assembly, checkerboard and full core problems). It turned out the critical boron concentration difference is as high as 20 ppm for burned cores. Nonetheless, the RMS power error is within 1.5%.