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The neutron flux distribution in the core region of the JRTR was measured through the neutron activation analysis method. Sample mount and a tool for withdrawing the sample mount were devised to irradiate the Au wires and foils. The measured data were compared with calculated values to confirm the reliability of the power distribution calculation. The reactor power during irradiation was estimated to be 9% lower than indications of power monitoring systems calibrated at around 10 μW. The measured transversal and axial neutron reaction rate distributions showed agreement within 1.32% difference with the calculated results.

Two bipolar host materials were designed and synthesized for blue thermally activated delayed fluorescence (TADF) OLEDs. Both bipolar materials, 5-(4-(dibenzo[b,d]furan-2-yl)-pyridin-2-yl)-5H-benzo[d]benzo-[4,5]imidazo[1,2-a]imidazole (4-DBFBI) and 5-(5-(dibenzo[b,d]furan-2-yl)-pyridin-2-yl)-5H-benzo[d]benzo-[4,5]imidazo[1,2-a]imidazole (5-DBFBI), were constructed with a hole-transporting type of benzimidazole moiety and an electron-transporting type of dibenzofuran moiety, and these derivatives exhibit high triplet energy over 3.10 eV. To achieve a high triplet energy level, a pyridine linker was placed between the benzimidazole and dibenzofuran moieties. As a result, 4-DBFBI and 5-DBFBI showed a high triplet energy level of 3.06 eV and 2.96 eV, respectively. Further, blue TADF devices were fabricated with our synthesized bipolar host materials. For our current study, we adopted 5,10-diphenyl-15-(10-(2,4,6-triisopropylphenyl)-10H-dibenzo[b,e][1,4]oxaborinin-3-yl)-10,15-dihydro-5H-diindolo[3,2-a:3′,2′-c]carbazole (PXB-DI) as the blue TADF dopant. The maximum external quantum efficiencies of the 20 wt% blue TADF devices were 31.8 and 32.5% for 4-DBFBI and 5-DBFBI, respectively. Moreover, the 4-DBFBI based host device exhibited a very low efficiency roll-off of 0.93 up to 1000 cd m−2, which is attributed to its well-balanced charge transporting ability.

The rotational plane slicing (RPS) method is proposed for 3-D whole-core neutron transport calculation. This method takes a view of the 3-D structure with a set of vertical planes that are obtained by slicing along the characteristic lines. In a 3-D structure with a simple axial configuration, the sliced planes consist of rectangular cells. Each sliced plane is then analyzed quite easily with a 2-D transport discrete ordinate solver. Several numerical tests are performed and the results indicate that the RPS method reproduces the reference solution with good accuracy.

In this paper, the overlapping local/global (OLG) iteration method based on deterministic SN transport/p-CMFD calculations was extended to hexagonal core geometry. For this framework, a deterministic SN transport code and p-CMFD equation solver was implemented on triangular cell base. These two calculations are iteratively performed with proper boundary condition update until it reaches converged solution, which is then compared with the direct whole-core transport solution. Two variations from the original overlapping local/global iterative scheme are also introduced and tested: (1) zeroth local calculation of fixed-k problem type with initially guessed incoming angular flux and multiplication factor, and (2) local calculation with non-overlapping region (nonoverlapping local/global iteration). Several numerical test results show OLG system has been successfully implemented on hexagonal core geometry and its variations improve the computational efficiency.

In current practice of nuclear reactor design analysis, the whole-core diffusion nodal method is used in which nodal parameters are provided by a single-assembly lattice physics calculation with th...