Verification and validation of the "ARKAS" code, using problems defined in the IWGFR Coordinated Research Programme (CRP) for the comparison of LMFBR Core Mechanics Codes, are discussed. The problems to be used in the verification (code against code) and validation (code against experiment) were defined and calculated by 11 core mechanics codes from 9 countries. The solutions obtained by these codes were compared in two steps, the first step consisting of Stage 1 "Verification", and the second step including Stage 1A "Advanced verification" and Stage 2 "Validation". The results of the first step for ARKAS, a core structural mechanics code, were presented and discussed in a previous paper. This paper describes some typical ARKAS results in detail, and compares them with the average values taken from solutions obtained by participants in the CRP. This corresponds to the second step ; Stage 1A and Stage 2. All the problems have been completed and were solved using ARKAS. Predictions made by ARKAS agreed very well with other solutions to the advanced verification problems, except in the case of problems involving high friction. As for validation problems based on Japanese ex-reactor 2-D thermo-elastic experiments, the agreement between measured and calculated values by ARKAS was fairly good.

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Steady-state and transient critical heat flux (CHF) experiments were performed using triangular pitched 7-rod assemblies with non-uniform axial power distributions under the maximum pressure of 15.5 MPa. The onset of steady-state CHF was predicted within the uncertainty of 10% with the KfK correlation using the local flow conditions calculated by the subchannel analysis code COBRA-IV-I. On the other hand, various mechanistic CHF models did not agree with the steady-state CHF data. The transient CHFs under the conditions of flow reduction, power increase or flow and power simultaneous variation were predicted with the quasi-steady-state method within approximately the same uncertainty as the steady-state CHF experiments.The predictive capability did not depend on the transient speed within 30%/s of the flow reduction rate and within 120%/s of the power increase rate. It was also revealed that there exists large CHF margins under the thermal-hydraulic conditions simulating the locked rotor accident and the control rod cluster ejection accident of the double-flat-core type high conversion pressurized water reactor (HCPWR).

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An accurate volume measurement system for a large amount of plutonium nitrate solution stored in a reprocessing or a conversion plant has been developed at the Plutonium Conversion Development Facility (PCDF) in the Power Reactor and Nuclear Fuel Development Corp. (PNC) Tokai Works. A pair of differential digital quartz pressure transducers is utilized in the volume measurement system. To obtain high accuracy, it is important that the non-linearity of the transducer is minimized within the measurement range, the zero point is stabilized, and the damping property of the pneumatic line is designed to minimize pressure oscillation. The accuracy of the pressure measurement can always be within 2 Pa with re-calibration once a year. In the PCDF, the overall uncertainty of the volume measurement has been evaluated to be within 0.2%. This system has been successfully applied to the Japanese government's and IAEA's routine inspection since 1984.

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Adsorption behavior of Cs and Sr on a synthetic zeolite P has been studied. Natural zeolites, clinoptilolite and mordenite, were converted into the zeolite P having a high crystallinity through hydrothermal treatment with NaOH solutions. The distribution coefficient (K_d) for Cs⁺ and Sr²⁺ depended on pH in acidic region, while attained nearly constant values around 10³ cm³/g in neutral and alkaline regions. This value is almost the same as that for Cs⁺ and about ten times that for Sr²⁺ on the original clinoptilolite. The adsorption obeys Langmuir-type isotherms, and the saturated amounts of Cs and Sr were respectively estimated to be 1.76 and 1.84 mmol/g. The zeolite P containing these cations was readily converted into a stable solid form consisting of pollucite and Sr-feldspar by calcination above 1, 000°C.

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The body burden of ¹³⁷Cs in individuals from the USSR who visited Japan and Japanese who returned from the USSR after various periods of stay there, was measured with a whole-body counter at the National Institute of Radiological Sciences. A compartment model was used to describe the temporal change in radiocesium body burden following the continuous ingestion of radiocesium. The daily amount of radiocesium intake in mBq per kg of body weight was estimated to have decreased with a half-time ranging from 391 to 920 days for adults in Kiev Ukraine after the Chernobyl nuclear power plant accident. Countermeasures against the consequences of the accident appear to have exempted the inhabitants from aggregation of radiocesium in their bodies. Differences in foodstuffs consumed should be taken into account when radiation health risks are compared with different age groups.

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Hexagonal assemblies are used in a high conversion light water reactor (HCLWR) and a nuclear reactor for cancer therapy. A cross section generation code for hexagonal assemblies is needed to design these cores precisely. For this purpose, a two-dimensional spectrum calculation code HARPY which can treat a hexagonal assembly has been developed. The HARPY code consists of a nodal code for pin-cell coupling in hexagonal geometry and a hexagonal mesh S_N transport code and calculates detailed neutron spectra of fuel cells in the assembly.
Verification calculations have been performed on a small hexagonal assembly by using SRAC and a Monte Carlo code GMVP to check the calculational accuracy of HARPY. The comparisons of k_∞'s, rod worths and pin power distributions have shown good agreement. This shows the adequacy of HARPY calculation.

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The control rods in the reactor of the nuclear ship MUTSU are classified into four groups: groups G1 and G2 are located in the central part of the core, while groups G3 and G4 are in the peripheral zone of the core. Several types of mutual interference effects among these control-rod groups were observed during reactor physics experiments with this reactor. During normal hot operations, positive shadowing was dominant between the GI and G2 groups; the degree of the shadowing effect of one rod group depended on the position of the other rod group. Both positive and negative shadowing effects occurred between an inner rod group (G1 or G2) and an outer group (G3 or G4) depending on the three-dimensional arrangement of the control rods. The rod worths of GI and G2 increased as a result of slight core burnup, about 1, 400 MWd/t, mainly due to the decrease in shadowing effects resulting from a change in control-rod pattern. A three-dimensional diffusion calculation with internal control-rod boundary conditions has proved to be useful for analyzing these various interaction effects.

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