A new treatment of sodium void coefficient in the fast reactor transient analysis is proposed. The method is based on the multigroup exact perturbation theory and utilizes the void reactivity map constructed by the multigroup first order perturbation theory. The change in the spatial distribution of neutron flux is corrected by the flux ratio of the fluxes in the transient and the steady states calculated with few-energy groups. The void reactivities for a homogeneous and a heterogeneous reactor are calculated by the new method with 4 and 6 energy groups and are compared with standard values obtained by the 25 group exact perturbation calculation. It is demonstrated that the new method gives almost the same values as the standard values while the ordinary calculation with 4 and 6 groups gives considerable deviation from the standard calculation. The present method has been incorporated to the two-dimensional quasistatic transient analysis code QUASAR and the code has been applied to an analysis of a control rod withdrawal accident. The results are compared with those calculated by the ordinary quasi-static method.

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Two-phase flow instability in a system including downcomers is studied experimentally. The system used for experiments consists of two parallel channels each composed of a "first riser", a downcomer and a "second riser" in order. Four series of experiments are carried out with a first-riser-heated single channel system, a downcomer-heated single channel system, a first-riser-heated parallel channel system and a down-corner-heated parallel channel system.
A dynamic instability is observed in the experiments with the single channel system, while a kind of static instability, "one-sided flow" as well as a dynamic instability, are observed in the experiments with the parallel channel system. As it is proved that the static instability, in this case, results from the negative slope in W-ΔP characteristics which is mainly due to the term of static pressure loss in the downcomer, it is concluded that this instability is peculiar to systems including downcomers. Another instability associated with condensation of vapor trapped in the top of the downcomer is also observed.

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The system pressure is one of the most important parameters for the thermal-hy-draulic behavior during the reflood phase of a PWR LOCA. System reflood tests are performed using the Cylindrical Core Test Facility (CCTF) to study the system pressure effect during the reflood with the same ECC water injection method as the actual PWRs. Through the comparisons with the FLECHT SET results, the qualitative differences of the system pressure effect are identified for the downcomer water head, core inlet subcooling and differential pressure through primary loops between two facilities. The differences of the system pressure effect between the FLECHT SET and the CCTF are caused by the differences of the flow area scaling of the broken cold leg and the ECC water injection method. The higher system pressure resulted in the better core cooling in the CCTF tests are observed in the FLECHT SET tests. It was confirmed quantitatively that the predominant factor for the core heat transfer enhancement with the system pressure is the steam density increase with the system pressure, The CCTF results support the assumption that the lower system pressure should give conservative core cooling behavior during the reflood phase of a PWR LOCA.

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A possible accumulation mechanism of Ni and radioactive ⁵⁸Co ions in the reactor water of a low iron crud concentration plant was modeled based on a direct deposition of Ni ion on fuel cladding surfaces as a form of NiO, followed by the solid phase reaction between NiO and iron crud (α-Fe₂O₃) to produce NiFe₂O₄.
The calculated results agreed satisfactorily with measurements in the first Japanese standard BWR, Fukushima 2-2 (1, 100 MWe), These calculations showed that controlling the Fe/Ni corrosion products ratio in the feed water to a value larger than 2 was effective to suppress ⁵⁸Co activity in the reactor water as well as Ni ion concentration.

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Characterization of crud on surfaces of the channel box in JPDR has been carried out by means of chemical, radiochemical, X-ray diffraction and infrared spectrum analyses. The main cations in the crud are Fe and Ni: The sum of their weights amounts to more than 90% of the total weight of the cations found. The results of X-ray diffraction and infrared spectrum analyses revealed that the crud consisted of Ni_0.65Fe_2.35O₄, NiO and γ-FeOOH.
Based on the neutron flux calculated from the burn-up of ²³⁵U and ²³⁸U in the spent fuel, the apparent residence time of elements on the surface of the channel box was calculated to be 230 d for Co, 260 d for Ni and 70 d for Fe. The value for Fe should be taken as a minimum value, because of the presence of γ-FeOOH in the crud, which has been formed during the storage in a pond.
The present data are discussed in correlation with the one in the reactor water.

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Simulated high-level waste glass was leached at 70°C for 28 d in one type of synthesized groundwater in Japan and in deionized water. The MCC-4 low-flow-rate leach test method was used for producing continuous-flow leach conditions with flow rates of 0.1, 0.01 and 0.001 m//min. The MCC-1 static leach tests were also carried out for a comparison.
First, the increase of the normalized elemental mass losses (NL) for Si and B was observed for all flow rates studied using synthesized groundwater and for flow rates of more than 0.01 ml/min in deionized water. This was in line with the results of several other previous studies.
Second, in the deionized water, the values of NL by the MCC-1 method were larger than those obtained by the MCC-4 method with flow rates of 0.001 and 0.01 ml/min. Since the deionized water does not have buffer capacity, the pH-value in the MCC-1 leach tests was higher than those in the MCC-4 leach tests with flow rates of 0.001 and 0.01 ml/min, thereby helping the glass to dissolve progressively in the MCC-1 static conditions.

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Three kinds of granulated hydrous titanium oxides (HTiOs) with different particle size (100120, 4260 and 3235 mesh size) were prepared by using poly acrylic hydrazide (PAH) as a binder. Rates of U adsorption on native HTiO powder and the granulated HTiOs were measured from U-enriched sea water by batchwise method. The intraparticle diffusivity (D_p) of U was evaluated by using the model of pore.diffusion with Freundlich-type adsorption isotherm. The estimated diffusivity (D_p) was 1.07 × 10^-6 cm²/s for the HTiO powder. Its tortuosity factor calculated by parallel plate pore model was 1.6.
Although the granulated HTiOs had macropores around 2, 000 nm in radius in addition to the mesopores originated from HTiO powder, they showed smaller D_p values (3.0 × 10^-78.6 × 10^-7 cm²/s) than the native HTiO. The smaller D_p values of the granulated HTiOs may be due to the influence of the PAH which adheres on the HTiO particles.

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Mixture of cylindrical steel pellets and Al₂O₃ balls, which simulated the intact fuel pellets and fragmented claddings, respectively, was inductively heated in a 50 mm I.D. pyrex glass cylinder filled with water, to investigate the coolability of TMI-2 type degraded core debris bed. The size of steel pellets was 11 mm dia. × 11 mm for BWR, 8 mm dia. × 12 mm for PWR and 5.5 mm dia. × 9 mm for FBR and Al₂O₃ balls were about 2 mm in diameter. The height of the debris bed was 25 cm or lower.
The dryout heat flux does not level off up to a bed height of 25 cm or over for the TMI-2 type bed while 8 cm or so in the bed of only steel balls. The dependence of dryout heat flux on the system pressure agrees with the Lipinski's 0-D model by adopting a proper equivalent diameter. When a simple number-weighted average is used as the equivalent diameter, the prediction gives a fairly good agreement with the experiment for FBR type bed but underestimations for the PWR and BWR type beds. It should be noted that the small balls of less fraction, not the large pellets, substantially govern the dryout. When the coolant flow is allowed from the bottom, however, the dryout heat flux is enhanced up to the level for the complete vaporization of coolant, and small amount of mass flux or circulation head can greatly improve the coolability.

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Both multi-group and point (continuous energy) Monte Carlo calculations have been performed on two types of benchmark problems, and the results have been compared to investigate the accuracy of the multi-group transport calculations in D-T fusion neutronics. In the multi-group calculations, space dependent self-shielding factors were used. As the first benchmark problem, spheres of four materials were analyzed. The discrepancy in the total fluxes obtained by the multi-group and point calculations was small in the lithium sphere, but it was great in the regions composed of iron. Nuclear design calculations of two types of fusion reactor blankets were performed as the second benchmark problem. The tritium breeding ratios obtained by the multi-group calculations agreed well with those obtained by the point calculations. However, the total fluxes in the reflectors or in the shields and the fast neutron leakage fluxes from the inboard shield were greatly underestimated by the multi-group calculations.

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