To obtain fundamental data for research on the transmutation of nuclear waste, the thermal neutron cross section and the resonance integral of the reaction ⁹⁹Tc(n, γ)¹⁰⁰Tc have been measured using an activation method.
Four ammonium pertechnetate targets containing 371375 kBq of ⁹⁹Tc were irradiated for 2min with reactor neutrons. Activation detectors of Co/Al and Au/Al alloy wires were irradiated for 10min to monitor the neutron flux and the fraction of the epithermal part (Westcott' sepithermal index). The Tc samples and flux monitors were irradiated with and without a Cd capsule. The γ-ray spectra from the irradiated samples were measured using a high purity Ge detector.
The thermal neutron cross section (2, 200m/s neutron cross section) and the resonance integral of the ⁹⁹Tc(n, γ)¹⁰⁰Tc reaction were found to be 22.9±1.3b and 398±38b, respectively.
The thermal neutron cross section obtained agrees with the previously reported values (20±2b by Lucas, 24.8 by Pattenden, 24±4 by Ovechkin) within the limits of error. On the other hand, the resonance integral is twice the value reported by Lucas (186±16b).

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Kinetic energy and neutron multiplicity as a function of fragment mass were measured for the thermal neutron induced fission of ²³⁹Pu. By measuring the velocities and energies of two fission fragments simultaneously, both of the pre-neutron emission fragment mass m^* and the post-neutron mass m were obtained. The fragment mass dependent neutron multiplicity ν(m^*)was deduced by subtracting m from m^*. The fragment mass dependent total kinetic energy TKE(m^*) was also obtained from this data. The fragment velocity was measured by time-of-flight (TOF) method, for which the start signal was triggered by a very thin plastic scintillation film detector (TFD) and the stop signal was obtained by a silicon surface barrier detector (SSBD) which was also used for the measurement of the fragment kinetic energy.
The present result of ν(m^*) is in good agreement with that of Apalin et al. and that in the light fragment region of Fraser et al. The obtained TKE(m^*) agrees well with the data of Wagemans et al. A calculation was carried out with the model proposed by Brosa et al. who assumed multichannel fission paths and a random neck rupture. It is seen that the calculated results of both fragment mass dependent kinetic energy and neutron multiplicity represent experimental data well.

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By analysing the neutron noise signals of the in-core detectors, the two-phase flow transit times have been measured for unit 1 of Tarapur Atomic Power Station (General Electric Boiling Water Reactor). Our study has shown that it is reasonable to identify the fluid velocity derived from the measured transit time with the volumetric flux. This interpretation offers the advantage of determining the mass flow rate by using the measured transit time and the steam quality obtained by the heat balance. Further, it is feasible to infer gross flow rate changes by combining the calculated value of the transit time as a function of flow with the periodic measurements of transit time over the operating cycle of the reactor.

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The selective removal of radiocesium was studied using a column filled with resorcinol-formaldehyde polycondensate resin (RFPR) prepared in the laboratory. The feed to the column was a simulated alkaline radioactive waste solution with high salt load of spent fuel reprocessing plant origin. Repeated loading-elution-regeneration cycles were conducted to determine reproducibility of breakthrough behaviour. The breakthrough data were correlated with batch equilibration data on cesium uptake. The quantitative elution of sorbed radiocesium was also confirmed.

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To aim at a better understanding of the uranium isotope exchange reaction between gaseous UF₆ and solid UF5. experiments were done with natural UF₆ gas and solid UF₅ containing 3% ²³⁵U under different pressures of UF₆. The experimental results suggest a two-process reaction with an initial rapid increase of ²³⁵UF₆ in the gas phase followed by its slight and gradual increase. A rate equation based on a collision model is given for the two-process reaction which includes a primary exchange reaction on the solid surface and a secondary reaction participated by underlying UF₅ molecules. An analytical solution is provided for both of ²³⁵UF₆ concentration in the gas phase and ²³⁵UF₅ concentration on the solid surface, which is useful for determining the parameters characterizing the exchange reaction. A numerical analysis is also made to evaluate the influence of gas samplings. A remarkable agreement is found between the particle sizes of UF₅ estimated from the reaction parameter and from the direct observation with an electron microscope. The depletion of ²³⁵UF₅ concentration by the exchange reaction is very small when averaged over the whole solid UF₅, because the depletion is virtually limited to the solid surface due to the small reaction probability of underlying UF₅ molecules.

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This paper summarizes results of sensitivity analysis of a generic geologic disposal system for HLW, using a GSRW code and an automated sensitivity analysis methodology based on the Differential Algebra. The results of sensitivity analyses indicate that parameters related to a homogeneous rock surrounding a disposal facility have higher sensitivities to the performance measure analyzed here than those of a fractured zone and engineered barriers.
This methodology permits sensitivity analyses of a single parameter with changing values of other parameters simultaneously, and thus gives quantitative information on the interrelation-ship between the parameters; the parameters for engineered barriers are generally insensitive to the output, while they are somewhat sensitive to the output only in a case of the low solubility condition. The methodology also provides technical information which might be basis for the optimization of design of the disposal facility.

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The coextraction model of Tc(VII) and U(VI) from nitric acid solution was developed on a tributyl phosphate (TBP)/n-dodecane system involving n-octyl(pheny1)-N, N-diisobutylcarbamoyl-methylphosphine oxide (CMPO). Technetium was considered to be extracted by the formation of Tc-CMPO and U-Tc-CMPO complexes in the presence of uranium. The distribution coefficient of technetium in the presence of uranium was defined by taking these complexes into account, and expressed as a function of nitric acid concentration, free CMPO concentration, and distribution coefficient of uranium. The equation of the distribution coefficient derived was applied to the previous data obtained in the mixed system of CMPO-TBP-dodecane at the low nitric acid concentration. It was found that the distribution behavior of technetium in the presence of uranium could be successfully simulated by the present coextraction model.

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Critical experiments of several thermal reactor cores are analyzed with ENDF/B-VI to assess its applicability to light water reactor analyses. The benchmark tests chosen use the continuous energy Monte Carlo code MCNP with exact treatment of a three dimensional geometry, and a cross section library is prepared by using the NJOY91 code. Calculated keff's underestimate experimental values from 0.3 to 0.9%Δk, and differences between calculated and measured keff's linearly increase with ²³⁸U effective capture rate. With a view to investigating the differences, sensitivity analyses are performed by minor modification of the ²³⁸U capture cross section at resonance as well as thermal energy regions. As a result, calculated keff's agree well with experimental results by changing the ²³⁸U resonance parameters to those of Liou and Chrien below 120eV and decreasing the thermal capture cross section value by 2.3%. These modifications improve the keff's of the TRX-1 and TCA 3.00U cores by 0.8 and 0.3%Δk, respectively.

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High pressure two-phase discharge flow from a vessel through a short pipe into an open space at atmospheric pressure was numerically analyzed using one- and two-dimensional non-equilibrium two-fluid models, in order to investigate effects of two-dimensional flow structure on thermal hydraulic phenomena in critical two-phase flows. Two-dimensional calculations revealed that flow contraction at the pipe inlet caused a liquid core in the central region of the pipe cross section, while void fraction in the vicinity of the pipe inner wall was large. This uneven phase distribution, which cannot be treated with one-dimensional calculations, seemed to result in larger discharge flow rate predictions because of phase slip promotion in two-dimensional calculations than that in one-dimensional calculations. Barrel-shaped shock waves, which are commonly observed in critical gas flow, did not appear in the calculated two-phase jets because large volumetric expansion due to flashing prevented the discharged two-phase mixture from over-depressurization.

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If an accident resulting in an emergency occurs at a nuclear facility, the emergency technical advisory body (ETAB) is asked to make technical advice to the national government for emergency response. In order for the ETAB to carry out its role, a significant amount of materials containing information on nuclear power plants is prepared in the ETAB. These materials are all paper copies and very voluminous. Therefore it is required that they should be computerized for more effective use. Further, methods should be developed which aid the ETAB to understand the status of the affected facility and to predict potential radioactivity release to the environment.
To support the ETAB's activities in an emergency, the materials have been computerized as databases for more effective and quick use of them, and an expert system has been developed to quickly predict accident progression and environmental fission product release with the multivariate analysis method. The databases and the expert system have been unified to be a system of COSTA. The first phase of developing COSTA has been completed in FY1992. This paper describes the present status of development of COSTA.

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Atomic fractions of each isotope in ³⁶Ar-³⁸Ar-⁴⁰Ar gas mixture near natural abundance were measured by using a general-purpose double-focused mass spectrometer HITACHI M-80. Because an amount of gas in a sample bulb and consequently a height of output peak decreases while taking measurements, 30 sets of peak height and time elapsed for each isotope were measured for calculation of atomic fractions through a least squares method under the assumptions (1) the reduction in heights follows an exponential function of time and (2) the rates of reduction are the same among argon isotopes. Application of this method to argon gas samples extracted from the top and bottom of a thermal diffusion column for isotope separation have proved that the method gives data with small dispersion even for ³⁸Ar, and is satisfactory for studying separative performances of thermal diffusion column.

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