The numerical solution of the transport equation has the errors caused by the approximations used in the computational method. In the past estimations of these errors have been performed experimentally. In the present study, formulas to estimate the errors have been derived on the basis of the perturbation theory. This method enables us to deterministically estimate the numerical errors due to the iteration, spatial discretization and Legendre polynomial expansion of scattering transfer cross sections.
Using the error estimation method developed in the present study, two examples of error analyses were carried out to confirm its validity and applicability to error estimation for a practical purpose. The errors of the calculated tritium breeding ratio for 7Li in a infinite slab geometry were estimated, and they agreed well with the values predicted from direct calculation. As the second example, error analysis was carried out for onedimensional nuclear calculations on two types of commercial fusion reactor blankets. In this analysis the tritium breeding ratio and the fast neutron leakage flux from the inboard shield were investigated, and the errors from different causes were quantitatively compared.

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A three-dimensional radiation transport code TRISTAN has been developed by applying the method of direct integration to the group-flux calculation in order to enhance the accuracy of shielding analysis. It uses group-angle transfer matrices derived from the double-differential cross section data instead of the Legendre polynomial expansion.
In order to improve the numerical accuracy, a technique to separate radiation into two components (the scattered and the unscattered) is adopted and the balance equation of radiation flux is applied to the solution method. Two new techniques, a stratification of the angular mesh and a kind of the forward-adjoint coupling, are introduced to enhance its applicability to the practical shielding analysis.
The validity of the TRISTAN is verified by making comparison with the results of the Monte Carlo code MCNP, and with the experimental values in the JRR-4 streaming experiments.

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A two-dimensional homogeneous equilibrium model is developed in order to evaluate impingement load caused by discharge of a two-phase mixture in postulated pipe rupture accidents of light water reactors. The present analysis differs from previous studies mainly in that a backward expansion around the pipe exit is taken into account. As a result: (1) it is confirmed that the backward expansion occurs around the pipe exit in a supersonic two-phase flow; (2) when the dimensionless position of an impingement wall z/D is larger than 2.0, the present calculations predict the pressure distribution on the impingement wall within an error of 10%, while the previous calculations, which did not take the backward expansion into account, overestimated the pressure by 25%; (3) existence of jet core and occurrence of shock waves in the two-phase jet are obtained and (4) a supersonic state of the jet is illustrated by comparing the velocity with the sonic velocity.

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A simple mechanistic model is presented to evaluate the subcooled void reactivity effect under a Reactivity Initiated Accident (RIA) at cold critical condition of BWR. This model consists of a drift flux model for vapor velocity and a vapor mass conservation model with a term of vapor source on a heated wall, and it was incorporated into a homogeneous and equilibrium thermal-hydraulic code EUREKA-JINS. A sample analysis by this model showed that the subcooled void reactivity effect leads to reduction of the maximum fuel enthalpy by about 20 cal/g UO₂ in the case of RIA at cold critical condition. Though the reduced value is dependent on the reactor core condition, this result indicates the significance of subcooled void reactivity effect in the accident, while the effect can be neglected in the hot stand-by case where, at most, only 4 cal/g UO₂ is reduced for the maximum fuel enthalpy.

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Thermal transient and isothermal tests on unirradiated Zircaloy-2 fuel cladding were conducted by determining short-term failure behaviors depending on temperature transients under constant differential internal pressure conditions. Failure data for constant heating rate transient tests were converted to Larson-Miller Parameter (LMP) values on the basis of the life-fraction rule. It was found that failure data for both the transient and isothermal tests were reduced to a single curve described by a correlation between LMP and initial hoop stress. Also, the failure strength during isothermal holdings after thermal transients was successfully evaluated by summing up life-fractions corresponding to each temperature process, using the LMP correlation. The LMP correlation approach could be used to predict the transient mechanical response of fuel cladding under high temperature conditions.

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Radiolytic, environments in the BWR primary system were evaluated by using a theoretical model. Reactor core was divided into two regions, boiling and by-pass (nonboiling) channels. The major findings are summarized as follows:
(1) Under normal operation without hydrogen addition, dissolved hydrogen and oxygen concentrations had their maximum values at the height where boiling started. Their concentrations in the boiling channel were lower than those in the by-pass channel because of hydrogen and oxygen release from liquid phase to gas phase.
(2) The most efficient oxygen suppression by hydrogen addition was expected in the non-boiling regions of the reactor core where injected hydrogen was confined in the liquid phase.
(3) When the bulk decomposition of water in the reactor core is represented by
H₂O→(α+β/2)H₂+α/2O₂+β/2H₂O₂ (α+β=1.0),
the coefficient α decreased as hydrogen injection rate increased, but β hardly changed at a low hydrogen addition rate.

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As a fundamental study for separating neptunium from high-level radioactive waste, the mechanism and the rate of extraction of pentavalent neptunium with DIDPA were investigated. From the analysis of oxidation state of neptunium in organic phase, it was proved that disproportionation reaction of Np(V) was concerned in the extraction.
The reaction order of the extraction with respect to neptunium concentration was larger than unity. The extraction rate was much reduced by the decrease of DIDPA concentration. The dependence of the rate on nitric acid concentration and on the temperature was also examined.

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An analysis of fast neutron spectra in an integral experiment with a collimated monoenergetic (d, t) neutron beam has been carried out by means of the MCNP Monte Carlo transport code. The calculations were performed on a 95 × 95 × 95 cm³ graphite cube. Results of the transport calculations before smoothing the output spectra with a Gaussian response function of a real neutron detector, reveal a very close resemblance between the calculated neutron spectra on one hand and angular distributions of elastically scattered and first level inelastically scattered neutrons on the other. This conclusion holds especially for neutron spectra obtained on the axis of the collimated neutron beam. Neutron spectra obtained in an integral experiment with a collimated (d, t) neutron beam can therefore be effectively used for an experimental determination of the neutron angular distributions of elastic and first level inelastic scattering at the energy of the collimated neutrons.

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The No. 2 Unit of Fukushima-Daini Nuclear Power Plant (2F-2; 1, 100 MWe) was commercially operated for 10, 320 effective full power hours (EFPH) as its first fuel cycle. The basic design concept of the 2F-2 incorporated the following two features :
(1) Application of procedures for reducing shutdown dose rate based on the Japanese Improvement and Standardization Program
(2) Low crud generation to minimize radioactive waste by careful material selection for the primary system.
Thus, it was possible to keep the average Fe concentration in the condensate water at less than 6 ppb during the first fuel cycle. As a result of this low value, the average life of powdered resin precoated prefilters was extended to about a month, and the average chemical regeneration period of the deep bed demineralizers was extended to more than one year.
The water chemistry of the 2F-2 was characterized by low ⁶⁰Co and high ⁵⁸Co radioactivities in the reactor water, which resulted in a low shutdown dose rate determined mainly by ⁵⁸Co depositing on the primary piping. For example, average dose rate around the primary piping just after reactor shutdown was about 70 mR/h, about 75% of which was from ⁵⁸Co depositing on the pipe inner surfaces. The contribution of ⁶⁰Co was about 25%.

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