Neutron inelastic scattering cross sections for molybdenum isotopes are calculated with the DWBA and the coupled-channels methods. Anomalous enhancement of the DWBA cross sections near the threshold energy appears when the adopted optical potential has a shallow imaginary part. Calculations with some simplified optical potentials indicate that the enhancement can be related with the p-wave strength, and it is found that the problem comes from the optical potential used.
When an adopted optical potential to the DWBA calculation is physically reasonable, differences between the calculated cross sections with the DWBA and those with the coupled-channels theory are small. Experimental data of ⁹²Mo, ⁹⁸Mo and ¹⁰⁰Mo are well reproduced by the calculated cross sections with the DWBA and the Hauser-Feshbach-Moldauer statistical model, and it is concluded that the DWBA is an appropriate method to evaluate cross sections of inelastic scattering from the molybdenum isotopes.

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A new fission product (FP) chain model has been studied to be used in a BWR lattice calculation. In attempting to establish the model, two requirements, i.e. the accuracy in predicting burnup reactivity and the easiness in practical application, are simultaneously considered. The resultant FP model consists of 81 explicit FP nuclides and two lumped pseudo nuclides having the absorption cross sections independent of burnup history and fuel composition. For the verification, extensive numerical tests covering over a wide range of operational conditions and fuel compositions have been carried out. The results indicate that the estimated errors in burnup reactivity are within 0.1%Δk for exposures up to 100GWd/t. It is concluded that the present model can offer a high degree of accuracy for FP representation in BWR lattice calculation.

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In-phase and out-of-phase instabilities have been observed in BWR reactors. To improve the safety systems of these reactors, it is necessary to be able to detect in a reliable way these oscillations from the neutronic signals. In this paper a methodology to decompose the neutronic signals in its modal amplitudes is proposed. This decomposition is based on the normal and the adjoint dominant Lambda modes of a static configuration of the reactor core. The calculation of these eigenmodes for a realistic problem is reviewed and the oscillation parameters for the modal decompositions of the neutronic signals from Ringhals reactor have been calculated using the proposed methodology.

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Optical fibers have advantages like flexible configuration, intrinsic immunity for electromagnetic fields etc., and they have been used as optical fiber sensors. By some of these techniques, continuous or discrete distribution of physical parameters can be measured. Here, in order to apply Raman distributed temperature sensor (RDTS) to the monitoring of nuclear facilities, some correction techniques for radiation induced errors were investigated. It has been shown that, when uniform loss distribution can be assumed, simple correction technique with two thermocouples can be applied. Moreover, if loop type arrangement is applied, even when the loss distribution is not uniform, radiation induced errors can be canceled.
For the demonstration of the feasibility of this technique, measurements using a commercial RDTS system were carried out along the primary piping system of the experimental fast reactor: JOYO. During the continuous measurements with the total dose of more than 10⁷R, the radiation induced errors showed a saturating tendency. The correction technique with two thermocouples was applied and its feasibility has been demonstrated. Although the time response of the system should be improved, the RDTS can be expected as a noble temperature monitor in nuclear facilities.

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An advanced startup procedure for the PIUS-type reactor has been developed. The procedure is related to the way to isolate the primary loops from the borated reactor pool by establishing stable hot/cold water interfaces in the so-called density lock sections. The procedure starts with accumulating preheated water in the high points of the steam-generator-side legs. Then, by restarting the reactor coolant pumps, the primary loops can be isolated from the pool as the primary loops reaches a uniformly higher temperature than the pool water. The additional components required for this procedure are only a low-pressure grade heater and a pump of small capacities. Since the isolation is achieved with the density locks left open, the core shutdown and cooling capabilities by means of the natural circulation of borated water are maintained in case of any abnormal events during startup. The feasibility and the predictability of this procedure were investigated by running an experiment in a scaled single-loop facility and conducting an analysis using a one-dimensional model. Both in the experiment and in the analysis, the primary loop was successfully isolated from the pool.

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Intergranular cracks of cladding tubes had been observed at the tips of the rodlets of PWR rod cluster control assemblies (RCCAs). Because RCCAs are important core components, an investigation was carried out to estimate their service life time.
(1) As it is essential to know the effect of slumping of the neutron absorber for the life time estimation, tests on absorber material were carried out. Both the dynamic and static stresses of the absorber are sufficiently small compared with its mechanical characteristics and it is concluded that slumping does not occur.
(2) The crack was initiated at the inner surface of the cladding tubes and Sipush et al. obtained an intergranular fracture surface in tensile tests of similar material at a very slow strain rate in an argon gas atmosphere. Therefore the mechanism of the intergranular crack of the cladding tube is not IASCC but irradiation assisted cracking (IAC) caused by an increase in hoop strain due to the swelling of the absorber and a decrease in elongation due to neutron irradiation.
(3) The crack initiation limit of cylindrical shells made of low ductile material and subjected to internal pressure is determined in relation to the uniform strain of the material and is in accordance with that of the RCCA rodlets in an actual plant.
From the above investigation, the method of estimating the life time and countermeasures for its extension are obtained.

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The apparent stability constant, β, for the complexation of Sr(II) with humic acid (HA) was determined by a dialysis method in the pH range from 4 to 10 at the ionic strength of 0.1mol/dm³ (NaClO₄) at 298K under N2 atmosphere. It was found that log β increased with pH from 2.5 at pH 4.0 to 4.1 at pH 10.0. The log β tended to level off at higher pHs, indicating that saturation was being reached. Further, the apparent ligand to metal ratio was found to be 1.15±0.10. Dissociation of carboxylic and phenolic groups of HA with pH is probably the main factor that gives rise to the changes in stability constants.

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The energy-and-angle dependent doubly differential γ-ray albedos for homogeneous semi-infinite medium have been calculated for water, ordinary concrete, soil, heavy concrete, iron, tin and lead. The procedure of calculation employed in the present paper is the invariant embedding method which is being developed to solve neutral particle transport problems in homogeneous one dimensional medium. The calculated γ-ray albedos are stored in the data base. One can easily obtain from this data base the various kinds of albedos (number, dose, energy, etc.) with a simple interface program. The accuracy of the calculated γ-ray albedos is ascertained by comparing with the Monte Carlo calculations (MCNP4A and EGS4).

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Radiation shielding analysis around the core of BWR plant was performed with a three-dimensional Sn code TORT to prepare detailed neutron flux distribution for the recent design needs of a BWR plant such as a life extension and a preventive maintenance. Two-dimensional R-Z, R-θ and three-dimensional R-θ-Z models, and JSSTDL cross section data based on JENDL-3 library were applied to the analysis.
Consequently, detailed spatial distribution of neutron flux in the reactor pressure vessel (RPV) was obtained with the energy range of MeV to thermal neutron. The analysis method was validated by comparing the calculated results with the measured data in a surveillance test program. It was found out that both the effects of rectangular shaped geometry of the core and the neutron leakage from the upper core region are considerably high especially in the location near the core such as the shroud. In addition, the effects of the parameters; the order of P_l, the number of Sn quadrature, mesh width and collapsing of cross section data on calculated results were also examined.

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TRACG is a new version of the best estimate BWR transient analysis code, which utilizes a multi-dimensional two-fluid model for the thermal hydraulics and a three-dimensional neutron kinetics model. A three-dimensional neutronics, a fully implicit integration scheme and models for advanced BWR components have been implemented in the code upon TRAC-BF1.
Assessment of TRACG has been performed in this study for the predictive capability of plant transients, which include thermal-hydraulic and neutronic interactions, as affected by responses of the plant control system. Simulations were presented for BWR representative transient tests, which were done as part of a series of BWR5 startup tests. As for the capability to predict thermal hydraulics during the design basis LOCAs, simulations were presented for the LOCA integral tests conducted in the ROSA-III at JAERI and the Hitachi TBL, which had been used for assessment of the TRAC former version.
Consequently, (1) the space-dependent power flow transitions in a BWR were confirmed by TRACG simulations in which the module coupled with neutronics and thermal hydraulics during transients has been newly introduced, and (2) the characteristic thermal-hydraulic phenomena including multi-channel effects during the design basis LOCAs were confirmed, as well as the TRAC former version, by TRACG simulations on which the influence due to a fully implicit integration scheme has not extended. Capability of TRACG to predict BWR transients ranging from simple plant operational transients to design basis LOCAs was successfully demonstrated.

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