This paper presents a number of fast iterative methods for solving systems of linear equations appearing in fixed source problems for neutron diffusion. We employed the conjugate gradient and conjugate residual methods. In order to accelerate the conjugate residual method, we proposed the conjugate residual squared method by transforming the residual polynomial of the conjugate residual method. Since the convergence of these methods depends on the spectrum of coefficient matrix, we employed the incomplete Choleski (IC) factorization and the modified IC (MIC) factorization as preconditioners. These methods were applied to some neutron diffusion problems and compared with the successive overrelaxation (SOR) method. The results of these numerical experiments showed superior convergence characteristics of the conjugate gradient like method with MIC factorization to the SOR method, especially for a problem involving void region. The CPU time of the MICCG, MICCR and MICCRS methods showed no great difference. In order to vectorize the conjugate gradient like methods based on (M)IC factorization, the hyperplane method was used and implemented on the vector computers, the HITAC S-820/80 and ETA10-E (one processor mode). Significant decrease of the CPU times was observed on the S-820/80. Since the scaled conjugate gradient (SCG) method can be vectorized with no manipulation, it was also compared with the above methods. It turned out the SCG method was the fastest with respect to the CPU times on the ETA10-E. These results suggest that one should implement suitable algorithm for different vector computers.

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An analytical method for the seismic response of the two-dimensional horizontal slice core model of a high temperature gas-cooled reactor core with block-type fuel has been developed. In the analytical method, blocks are modeled as rigid bodies and a spring dashpot model is used for the collision process between blocks. Analytical results are compared with experimental ones and both are found to be in good agreement. The analytical method can be used to predict the behavior of the high temperature gas-cooled reactor core under seismic excitation.

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The release behavior of delayed neutron (DN) precursors from the breached fuel was analyzed, based on the diffusional release model. In this analysis, spherical co-ordinate simple approach as well as two-dimensional finite element method were used for the predomiant DN precursor ¹³⁷I.
It was shown that the fuel sodium reaction product (FSRP) layer at the pellet outer surface provided a predominant role for the accelerated release of DN precursors. The macroscopic diffusion coefficient for DN precursors in the FSRP layer was estimated to be 10^-9 to 10^-10m²/s, which is greater than that in unreacted fuel by more than about 10⁶. It was demonstrated that the DN signal response correlated to the growth of the FSRP layer, and did not depend on temperature change of interior fuel itself.

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The behavior of impurity sulphate, fed from source Th(NO₃)₄, in a sol-gel process for fabricating (Th, U)O₂ microspheres is studied. In sols, almost all the sulphates are adsorbed on colloid particles regardless of U content U/M (M : Th+U). However, the sulphate in gel-microspheres is easily washed out for U/M higher than 10 mol/0, while not for lower U/M ; especially, the sulphate in ThO₂ gel-microspheres is difficult to be removed even by long-time washing. The sulphur remaining in the gel-microspheres is able to be removed out by heating in air or steam at 1, 000°C for 3 h ; as for ThO2, slight residue is found but not after 1, 300°C heating. On the other hand, heating them in Ar-4%H₂ atmosphere, even at 1, 500°C, allows most of the sulphur to remain as an unidentified M-O-S compound ; during the heating, the sulphur is gradually released by the reaction with H₂ to form H₂S.
The study finds optimum conditions for removal of the impurity sulphate in the sol-gel process.

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Lead-iron phosphate (LIP) glasses loaded with a simulated high-level nuclear waste wen. studied on their leach rates and thermal properties.
The obtained results showed that the phosphate glass matrix consisting of lead monoxide, phosphorus pentoxide and ferric oxide of 56:35:9w/0 is able to vitrify the waste, pretreated with formic acid to remove Zr, to about 15w/0 at 950°C. The leach rate of the vitrified waste glass was in the order of 10^-7 g/cm²•d at 110°C, which is low compared with that of the borosilicate glass waste form. Increasing the phosphorus pentoxide content of the matrix to higher than 35w/0 enabled it to produce the glass form with the waste near 20w/0 at 950°C, but this increase rendered the glass waste form more soluble than the former. Thermal properties such as thermal expansion coefficient, critical cooling rate for vitrification and temperatures of glass transition, softening and maximum rate of crystallization were measured and discussed.
Removing Na ions from wastes improves considerably both the leach rate and the thermal stability of the LIP glass waste form.

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Neutron-induced reaction cross sections of ⁵⁵Mn are calculated for the evaluated nuclear data libraries, ENDF/B-VI and JENDL-3. Simultaneously calculated are the inelastic scattering, (n, 2n), (n, p), (n, α), (n, np), (n, nα) and (n, γ) reaction cross sections, the angular distributions of emitted neutrons and protons, and the energy distributions of emitted particles and γ-rays. A unified Hauser-Feshbach code is applied to calculate these quantities. Precompound and direct-reaction processes are taken into consideration, in addition to the compound process. The calcu-lated results reproduce the experimental data very well. Forward-peaked angular distributions of continuous neutrons are obtained from the calculation, and found to be consistent with the measurements at 14 MeV.

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