A new approach to establish the long-term fuel management scheme is presented in this paper. The point reactivity model is used to predict the core average reactivity. An attempt to calculate batchwise power fraction is introduced through the two-dimensional nodal power algorithm based on the modified one-group diffusion equation and the number of fuel assemblies on the core periphery. Suggested is an empirical formula to estimate the radial leakage reactivity with ripe core design experience reflected.
This approach predicts the cycle lengths and the discharge burnups of individual fuel batches up to an equilibrium core when the proper input data such as batch enrichment, batch size, type and content of burnable poison and reloading strategies are given. Eight benchmark calculations demonstrate that the new approach used in this study is reasonably accurate and highly efficient for the purpose of scoping calculation when compared with design code predictions.

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The main objective of the experiments Mol 7C/3 to 6 was to investigate the role of burn-0up in the propagation of local faults in LMR fuel subassemblies. The test sections contained bundles of 30 fuel pins with burn-ups between 0.3 and 10 at% and porous blockages as local fault initiators. The main result of the experiments is that with rising burn-up there is an increasing potential for damage progression from the blockage to the surrounding bundle and structures. Only at low burn-up, faults of the kind and type investigated are self-limiting. At medium and high burn-ups, active protection systems, in particular the delayed neutron detection system, are required to prevent whole core involvement.

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A diagnostic system for BWR water chemistry has been developed incorporating high-speed fuzzy reasoning, holding in view preventive maintenance as principal objective.
The system combines two expert systems-one for detecting an anomaly in the plant operating condition and quickly inferring its cause in a preliminary analysis, and the other performing a detailed diagnosis of the detected anomaly.
The system for anomaly detection and quick preliminary diagnosis infers the occurrence of various anomalies and their causes from the fuzzy relation existing between the fuzzy causes and symptoms, applying a high-speed method for solving the inverse problem of fuzzy correspondence. The system can be run on a small size computer, and is thus conveniently adapted to practical application at machine side.
The system for detailed diagnosis incorporates the expertise of water chemistry engineers, and outputs pertinent information and data to guide operators.
Trial application of the system using data from actual plant operation conclusively demonstrated its effectiveness as a practical preventive maintenance system for actual plants.

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The emission probabilities for the γ-rays of ⁷⁵Ge were accurately measured by a 4πβ counter, a high purity germanium γ-ray detector and a coincidence apparatus using a live-timed bi-dimensional data acquisition system. Two series of measurements were carried out ; in the first a 4πβ counter with pressurized Ar 90% plus CH₄ 10% mixture was used, and in the second that with CF₄ gas at atmospheric pressure was used. The emission probability of the main 265 keV γ-rays was 0.1020±0.0017, that was smaller than the evaluated value, 0.113±0.011 (Nucl. Data Sheets, 1981), by about 10%, and the certainty was improved by about six times.

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To evaluate the heat removal capability of a water wall type cooling system, which is one passive containment cooling system (PCCS), the thermal hydraulic behavior in the suppression pool (S/P) and the outer pool (O/P. flat plate water wall) have been investigated experimentally. The following results were obtained. (1) A thermal stratification boundary, which separates the pools into the upper high temperature and lower low temperature regions, was formed just below the vent tube outlet. (2) Convection heat transfer characteristics in the S/P and O/P along the primary containment vessel (PCV) wall had no significant differences and were those of natural convection. Correlation of the natural convection heat transfer up to the Ra number of 2×10¹⁴ was obtained. (3) Vertical variations of local condensation heat transfer coefficients under a noncondensable gas presence were within ±10% of the average value for the 4.7 m heat transfer length. An experimental correlation for the average condensation heat transfer coefficients was obtained as a function of steam and noncondensable gas mass ratio. (4) An analytical model to evaluate the system performance of the water wall type PCCS was verified. (5) A baffle plate concept to mitigate thermal stratification at the vent outlet and to enlarge the high temperature region in the S/P was considered as a means to improve heat release capability. Thermal hydraulics with a baffle plate were examined, and effectiveness of the baffle plate to improve the heat release capability was confirmed.

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It is well known that hydrous titanium oxides are useful as an inorganic adsorbent for uranium resources in seawater. In this work, it was found that a characteristic layer structure appears when the Ti-U coprecipitates formed by the addition of aqueous ammonia to the nitric acid/sulfuric acid solution are heated in an 8%H₂-92%Ar reducing atmosphere at a relatively low temperature (around 350°C). The formation of this layer structure and the phase transformation which proceeds in air with time were examined by means of X-ray diffraction analysis, magnetic susceptibility measurement, X-ray photoelectron spectroscopy, thermogravimetry etc. These results are considered to suggest that the layer structure is formed by the occupation of uranium (probably as uranyl units) between the layers in the titanium oxide crystal.

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Carbon fiber reinforced SiC (C/SiC) and SiC fiber reinforced SiC (SiC/SiC) composites were fabricated by a chemical vapor infiltration (CVI) method to obtain the composites with minimized impurity contamination. The neutron activation analysis showed that 30 elements were detected as impurities. However total impurity concentrations in the C/SiC and SiC/SiC were lower than 14.5 wt ppm and 9.80 wt ppm, respectively.
The evaluation of induced activity under the simulation of fusion neutron irradiation of 10 MW•y/m² was made based on the impurity concentration. The dose rates of both composites were assumed to decrease by about 6 orders of magnitude in one day cooling and by 9 orders of magnitude in several ten years. The results indicate that the present composites behave excellent in terms of low activation and the CVI is a potential process to obtain high-purity composites.

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Changes of the extraction behaviour and cation-transport of U(VI) and Pu(IV) nitrates with γ-irradiated DC18C6 in toluene have been investigated. The effect of radiation damage to DC18C6 for extraction and permeation at excessively higher doses (70 Mrads) has been studied systematically. No deterioration in its performance is noted at lower doses. Hydrolytic stability of the macrocycle/diluent system in the presence of nitric acid and radiolytic stability of immobilized liquid membrane using 'Enka' Accurel flat sheet polypropylene membrane films as solid support are also examined.

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Iodine separation technology using an inorganic adsorption material has been investigated in order to apply the technology to the off-gas treatment systems of nuclear facilities. Iodine removal efficiencies were checked by laboratory experiments using simulated off-gas streams of various conditions and the developed adsorbent, silver impregnated alumina (AgA). Laboratory test results demonstrated effective iodine removal with high decontamination factors (DF's) at relatively high temperatures ( ?? 100°C). Then the removal efficiency were confirmed using actual off-gas streams sampled from the dissolver off-gas treatment system of the Karlsruhe reprocessing plant. The DF's were over 10³ with the AgA bed depth of 10 cm and showed little change during the adsorption period, which indicated applicability of the iodine removal technology with AgA to nuclear fuel reprocessing plants. Iodine adsorption capacity and its release property were also investigated using simulated off-gas streams. The former had a value of 0.22 g/g-AgA and this value could well predict the breakthrough property. The adsorbed iodine was judged stable during the storage of AgA saturated with iodine in air at temperatures below 500°C and in water at 20°C after changing the adsorbed iodine form from AgIO₃ to Agl. Thus, the separation technology provided effective and stable iodine separation from the off-gas of nuclear facilities.

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A highly simple and effective new methodology has been proposed to analyze the reaction kinetics of non-equilibrium mass transport. Interphase mass transfer in an evidently non-ideal liquid system has been described as trajectories on a reaction plane by introducing a logarithmic driving force L_G which equals to A⁺/RT, where A+ is newly defined chemical affinity of a solute. The affinity A⁺ is referenced not only to the reaction state but to the equilibrium state, and is different from the De Donder's affinity defined solely by a reaction state.
The affinity A⁺ is always smaller than the conventional one. It has also been concluded that extractive transfer might occur in a region A⁺>0. This description of transfer phenomena enables us to treat the reactions directly without the knowledge of rigorous activities of solutes, and is independent to the selection of concentration scale as well as the reference systems for chemical potentials.
The affinity A⁺, i.e. the upper limit of the reaction trajectories, and the rate of the decrease in the logarithmic driving force have been represented by A⁺ ?? -RT ln ξ and -dL_G/dξ=1/ξ using the degree of advancement ξ. Consequently, the thermodynamic analysis of reaction trajectories gave insights about the reaction kinetics and its characteristics for two phase systems containing multi-solutes.

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Alumina was studied on its applicability for a host material to incorporate lanthanide and actinide elements present in HLW. Alumina-based waste forms containing about 8 mol% in all of Ce₂O₃ and/or Nd₂O₃ were prepared by firing at 1, 500°C for 25 h in a reducing atmosphere, after drying and denitrating the mixed solutions of aluminum, cerium and/or neodymium nitrates, and then pelletizing the powder mixtures under 36 MPa.
The obtained ceramic assemblages were analyzed by X-ray diffraction to determine crystalline phases produced. Tested also was the leachability of the assemblages in a nitric acid solution of pH=1 and deionized water at 150°C. These observations revealed that Ce prefers to be stabilized in magnetoplumbites, while Nd in perovskites. To obtain even denser assemblages, it seems necessary to suppress the formation of magnetoplumbites as much as possible. Leaching study showed that the relatively high dissolution rates of the forms in the solution of pH =1(10^-6g/cm²•d) were probably due to the innate reactivity of alumina with acid, and perovskites were more prone to leaching than magnetoplumbites. In the deionized water, the elemental release rates from the forms are lower than 10^-7 g/cm²•d, and not influenced practically by the presence of perovskites.

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The resonance interaction effect between hot sample and adjacent cold core was evaluated in a small sample ²³⁸U Doppler effect measuremnt. For this analysis a collision probability code-with an ultra-fine group structure PEACO-X was developed. The previously measuredDoppler effect reactivity confficients at FCA are systematically analyzed by using the PEACO-Xcode. The results were compared with the conventional calculation results using the SLAROM code. From these results, it was shown that this effect depens on the background cross section for ²³⁸U in a core, and also depends on the radius and density of the sample. When this resonance interaction effect is taken into account, the calculated Doppler reactivity woths becomes larger than those using the conventional method and the sample dependency of calculation disappeared. The reliability of Doppler effect calculation was significantly improved.

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Post irradiation examination (PIE) of a high burnt lead fuel assembly. which was irradiated to demonstrate fuel integrity at high burnup, was performed before the start of the full batch loading of high burnup fuel of 48GVid/t maximum fuel assembly burnup.
The lead fuel assembly was 17×17 B-type PWR fuel which was supplied by Nuclear Fuel Industries, Ltd. (NEI) and achieved the maximum burnup of 45 GWd/t after 4 cycles of irradiation in Ohi Unit 1 of the Kansai Electric Power Co. Inc. (Kansai).
Twelve fuel rods extracted from the lead fuel assembly at the reactor site were examined at the hot-cell facility of Japan Atomic Energy Research Institute (JAERI) in Tokai-mura.
Visually, the fuel rods appeared to be in good conditions, but some small spallings were observed at the second span from the top where oxide film was relatively thicker than other spans. Even in this span. the maximum oxide film thickness was less than 50 μm. Fission gas release rate was less than 1%. which caused only a small increase in fuel rod internal pressure. Mechanical properties of the fuel cladding were evaluated by tensile tests.
These PIE results were within the range of other PIE data previously obtained from domestic and foreign PWR fuel rods. The data confirmed that the integrity of B-type fuel would be maintained at least up to 48GWd/t.

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Destructive analyses for five spent fuel samples taken from a Gd bearing fuel assembly were done. The measured amounts of actinides of ^234-238U, ²³⁷Np, ^238-242Pu, ^{241, 242m, 243}Am and ^{242, 244}Cm, and fission products of ¹³⁴Cs and ¹⁵⁴Eu were used for evaluating the accuracy of calculation made by CASMO-MICBURN and ORIGEN-2 codes. The effect of Gd on the neutron spectrum was taken into account in the CASMO-MICBURN calculation.
The amounts of ²³⁵U, ²³⁹Pu and ²⁴¹Pu calculated by CASMO-MICBURN agreed well with the observed values within about 3%. On the other hand, the amounts obtained from ORIGEN-2 calculation showed lower values than those observed, especially by -12% in average in ²³⁵U for Gd₂O₃-UO₂ fuel. The main cause of this large difference may be attributed to the effect of Gd on the neutron spectrum. The amounts of the other actinides by both calculation codes revealed no significant difference in nearly 10% except for _242mAm, in which a large fluctuation among the samples was observed. About 10 % difference between the measured values and the calculated values was also observed for ₁₃₄Cs, but the calculated values for ₁₅₄Eu showed a significant difference from measured values.

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