The burnup uncertainties of a fast-breeder reactor have been evaluated based on the cross section adjustment technique.
A group cross section set, obtained using JENDL-2, was adjusted using experimental data of fast critical assemblies ZPPR and FCA. The adjusted cross section was applied to the neutronics characteristic calculation of a 1, 000 MWe-class fast reactor, and the changes in neutronics characteristics and their uncertainties due to the cross section adjustment were evaluated. The burnup reactivity loss was found to increase by 23% when the adjusted cross section library was used. This value corresponds to a reactivity decrease of about 0.7%Δk at the end of the equilibrium cycle for a typical 1, 000 MWe LMFBR of 1 yr operation cycle.
The uncertainty in burnup reactivity due to the cross section error was found to decrease from 21% to 12% if the cross section ajustment method was employed. This brings the benefit of a reduction in control rod worth requirement by about 0.3%Δk.

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A nodal method has been developed which accurately and rapidly solves the three-dimensional neutron diffusion equation in both Cartesian and hexagonal geometries.
To reduce the number of unknowns in comparison with the interface current method in favor of short computation time and a small computer memory requirement, this method employs the finite difference method (FDM) as its global neutron balance solution method. In the global neutron balance solution, the coupling coefficients are modified in such a way as to conserve the nodal interface neutron currents which are obtained by a local neutron balance solution method fit to each geometry.
To validate the method developed here, it has been applied to neutron diffusion calculations for reactor cores, typical of Cartesian and hexagonal geometries, with different core sizes, assembly pitches, core configurations, control rod patterns, etc. For a large FBR (hexagonal geometry), for example, errors in power distribution and control rod worth by the method are less than 1/2 and 1/10, respectively, compared to an ordinary FBR core design method, while the computation time is 1/8 that of the latter method.

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In pressurized water reactors (PWRs) of nuclear power plants, it is periodically required to recalibrate the correlation between the incore and excore nuclear instruments, where the recalibration guarantees the accurate monitoring of incore axial power distributions by excore instrumentation. In order to obtain the correlation between incore power distribution and excore detector response, however, it is necessary to perturb the axial power distribution of the reactor by control rods or to introduce a xenon spatial oscillation.
In this paper a simplified analytical method that eliminates such perturbations is proposed to evaluate the correlation between incore power distribution and excore detector currents which is utilized to recalibrate the nuclear instrumentation system. This method provides as good an accuracy as that obtained by conventional methods.

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A new counting method of CR-39 etchpit has been developed by image processing technique using multivariate analysis. This method consists of two parts of image processing and multi-variate analysis. Main features of this method are the flexibility for lighting condition, etchpit density and so on because standard image data analyzed by each observer is applied as a reference in his counting process.

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Deposition behavior and solubility of sodium iodide in sodium-stainless steel system was studied with using capsules placed vertically under a large temperature gradient.
The stainless steel capsule loaded with 20 g of reactor grade sodium containing sodium iodide was heated at its upper part and cooled at its bottom to establish the temperature gradient along the capsule tube and to precipitate and accumulate the iodide onto the capsule bottom, and was quenched to fix the iodide in solidified sodium after the diffusing run. Distributions of the iodide in the sodium in radial and axial directions were obtained as a function of temperature during the run. A partition value of the iodide between the core sodium and its periphery was calculated from the radial distribution in the sample in unsaturated regions. The solubility of the iodide in sodium was determined from the axial distribution in the core sodium samples.
Heat of adsorption obtained from the partition value and the solubility determined with this method are in reasonable agreement respectively with the data reported by Allan.

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The effects of r-irradiation on a simulated nuclear waste glass were studied by electron spin resonance spectroscopy (ESR), and were compared with the results on silica glass and Pyrex glass. Three kinds of glasses were r-irradiated up to the dose of 1.22 MGy and the ESR spectra were obtained. The intensity of ESR spectra were obtained as a function of irradiation dose and annealing temperature.
The spectrum of the waste glass was characteristic of two typical peaks, Peak 1 was the strong resonance at g=4.3 showing the existence of four coordinated Fe³⁺ and Peak 2 was the weak and broad resonance at g=2.0 showing the existence of six coordinated Fe³⁺. The ESR spectra of the waste glass before and after r-irradiation were almost overlapped and a little difference only in the intensity was observed. While in silaca glass and Pyrex glass, the peaks from E'r center and boron-oxygen hole center (BOHC) were observed to arise after irradiation. The absolute intensity of Peaks 1 and 2 described above changed in complicated way depending on the dose. The result suggests oxidation or reduction of iron takes place in the waste glass depending on the dose. The isochronal annealing of irradiated glasses shows most of r-ray-induced damages in the waste glass are restored even at room temperature, although most of the damages in silica glass and Pyrex glass are disappeared at the temperature from 550 to 600 K. The results show that the waste glass with a few weight percent of iron is resistent to radiation than other commercial glasses.

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By using the two-dimensional rigorous numerical solution of flow and convection-diffusion equations, the H₂-HT separative performances of thermal diffusion column with 1, 500 mm-height, 15 mm-radius and 77.35 K cold-wall was analyzed with pressure P, temperature difference between hot-wire and cold wall ΔT and feed rate F being parametrically changed. Flow analysis has revealed that the magnitude of the free convection is almost proportional to the pressure, and the secondary circulation is observed near the bottom of the column as the pressure increases. Moreover, laminar solution of free convection could not be obtained at the pressure more than 0.065 MPa. Separative analysis for the "cryogenic-wall" thermal diffusion column with 77.35 K cold-wall has made it clear in comparison with the separative performances of the "ordinary cold-wall" column with the same condition except the temperature of the cold wall that :
(1) The optimum pressure, being proportional approximately to the cold wall temperature, maximizing separation factor and/or the separative power is fairly smaller.
(2) The degree of separation is remarkably larger and the heads separation factor α is nearly equal to the limited value (an inverse of cut θ). The total separation factor αβ exceeds 10⁴ when ΔT=1, 200 K, F=1 cm³ (at 288.15 K, 0.1 MPa)/min and θ=0.1.
(3) Even for the higher feed rate F of 100 cm3 (at 288.15 K, 0.1 MPa)/min, αβ remains 100. In addition, the optimum pressure increases with ΔT and F.

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Outdoor ²²²Rn concentrations were measured with electrostatic integrating radon monitors (EIRM) at 40 points around Nagoya, in which 15 sets of 2-month-exposure data over 2.5 yr were obtained. Seasonal variation of ²²²Rn concentration showed a clear pattern which had a spring-to-summer minimum and an autumn-to-winter maximum. Annual means 222Rn concentration ranged 3.511.7 Bq•m^-3 depending on locations. Higher concentrations were obtained in mountainous regions while lower concentrations were obtained near the sea. The relation between atmospheric ²²²Rn concentration and uranium content of granitic rocks was also discussed. High level ²²²Rn concentrations were obtained over areas of granitic rocks which were comparatively uranium-rich. As an almost linear relationship was recognized between the ²²²Rn concentration and exposure rate due to external natural radiation, a close correlation was anticipated between the concentration and ²²⁶Ra content of soil. From the obtained regional distribution of 222Rn concentration, per caput lung dose in the area was estimated to be about 0.44±0.11mSv•yr^-1

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Chemical behavior of nitrogen species in intense radiation fields was studied. Simulation models and a N/H/O reaction scheme were developed and verified with a series of experiments. The following results and suggestions were obtained :
(1) Ammonium formation rates in water pressurized with H₂ and N₂ gases at temperatures of 288473 K were measured and numerically simulated with the reaction scheme within an error factor of 2.
(2) Chemical behaviors of radioactive nitrogen (¹³N and ¹⁶N) were also theoretically analyzed. Computer simulations demonstrated that the major species of radioactive nitrogen released to the main steam was nitric oxide (NO). This prediction was partly supported by a ¹³N analysis in the main steam of an actual BWR: About 85% of the ¹³N was found to be in anionic species whose precursor was supposed to be NO.
(3) Calculated results showed that ¹³N and ¹⁶N in the main steam under hydrogen water chemistry increased similarly but quantitatively not identical with each other : Nitrogen-13 seemed less volatile than ¹⁶N.
(4) Numerical simulation showed ¹⁶N released to the main steam might be decreased by addition of hydrogen atom scavenger such as nitrous acid in the reactor water.

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There was few post irradiation examination data on the mechanical properties of domestic fuel cladding tubes used for light water reactors, then those data obtained abroad have been often used in the fuel design or fuel performance codes. Although, many reports discussed the deformation mechanism of the tube, almost all the data were not obtained from irradiated specimens but unirradiated ones. In recent years, systematic post irradiation examinations on domestic fuel elements used in Japanese light water reactors and the related studies were performed.
This report first summarizes briefly the crystallographic texture which characterizes the properties of Zircaloy fuel cladding tubes, followed by an explanation of basic properties such as elasticity, plasticity, creep and fatigue. Finally, the up-to-date results are introduced.

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