Occupational exposure during BWR refueling and annual maintenance periods is determined by radioactive corrosion products, such as ⁶⁰Co and ⁵⁸Co deposited on major components and piping of the primary cooling system.
The authors have investigated behavior of radioactive corrosion products, e.g. generation, activation and deposition processes, in the primary cooling system, and then expressed the findings as mathematical equations. These provide corrosion product simulation models to predict the amounts of radioactive corrosion products depositing on components and piping.
The effects of corrosion product reduction procedures on shutdown dose rate were evaluated using the model. The procedures were incorporated into the Japanese Improvement and Standardization Program. Improvements of operational procedures to control water chemistry, such as Ni/Fe ratio, as well as application of hardware improvements resulted in an extremely low occupational exposure of less than 90 man•rem/yr for currently constructed BWR plants.

抄録全体を表示

A streaming experiment using a D-T neutron source was carried out to verify the calculational technique for neutron transport in a shield assembly with multi-layered slits. Reaction rate distributions of a small spherical NE213 scintillation detector to fast neutrons were measured in the slits made of 304SS and in the mortar surrounding the slits. The energy spectrum of fast neutrons in the slit was also measured with the same detector. These measurement were compared with calculations using the continuous energy Monte Carlo transport code MCNP. The calculated reaction rates in the slits agreed with the measured ones within experimental and calculational errors. Besides, it is suggested that the attenuation of fast neutrons in the mortar is significantly different from that in the slits and the behavior is nearly traced by the calculation with the MCNP code. The measured and calculated spectra at a position close to the exit inside the lower slit agreed within the both errors.

抄録全体を表示

The sensitivity coefficients of neutronic performance parameters in high-conversion LWR cells have been calculated by means of the SAINT code. In order to show the specific features of the sensitivity coefficients in the HCLWR cells, the differences between sensitivities were investigated for cells with different moderator to fuel volume ratios and different Pu enrichments. The burnup dependence of the sensitivities was also discussed with an emphasis on the effect of fission products on the cell parameters.
We have performed the sensitivity analysis for the PROTEUS cores. Group constants of main heavy nuclides were compared for the different cell calculational methods; SRAC and VIM, and the different cross section libraries; JENDL-2 and ENDF/B-IV. The effect of the differences in group constants was estimated for the cell parameters k_∞, reaction rate ratio and coolant void worth. The differences in the ²³⁸U capture and ²³⁹Pu fission group constants in the resolved and unresolved resonance range produced 0.31.0% change in k_∞ and 16% change in the coolant void worth. These effect was largely dependent on the coolant void fraction.

抄録全体を表示

The low and high power core uncovery patterns were observed in the high-pressure quasi-steady core uncovery experiments in a 25-rod bundle. The boundary between the two patterns was obtained in the experiments. The difference of two patterns was considered to be due to the slug-annular transition below the dryout points. The Osakabe's slug-annular transition model was the good boundary between the two patterns.
The small break loss-of-coolant accident (LOCA) experiments were conducted by using the integral experimental facility with the 1, 168-rod core. The transient core uncovery pattern was expected as the low power core uncovery pattern based on the quasisteady experiments mentioned above. The transient core uncovery patterns were classified into the boiloff and hydraulic core uncovery. In the boiloff core uncovery, the dryout points were controlled with the mixture level like the quasi-steady state. In the hydraulic core uncovery, the dryout points were not controlled with the mixture level alone, and the multi-dimensional dryout process in the core and the relatively high heat transfer above the dryout points were observed. It was considered that a part of water was remained above the dryout points due to the rapid depression of core liquid level.

抄録全体を表示

A method to determine the electric conductivity of water continuously and directly at elevated temperature up to 300°C was developed which can be applied as a sensor for corrosion behavior foreknowledge and diagnosis systems using water chemistry data of BWR primary coolant. Complex impedance was measured between a couple of parallel platinum electrodes installed with a constant distance and dipped in the water. By analyzing frequency dependence of the impedance, the resistivity of the water between the platinum electrodes was estimated separately from the impedance caused by surface reactions on the platinum electrodes, which was the source of error in the measurement of electricl conductivity at elevated temperature.
Increase of necessary frequency to obtain the surface impedance at elevated temperature was evaded by the extrapolation of the frequence dependence of the impedance with calculations by using the data up to 100 kHz in which electric conductivity shows negligibly small dependence on the frequency of applied voltage.
The measured conductivity of pure water up to 300°C showed a good agreement with the calculation based on the dissociation data of water, which shows the applicability for the in-line monitor of electrical conductivity at elevated temperature.

抄録全体を表示

The SAFER03 computer code has a newly developed evaluation model for the analysis of various boiling water reactor (BWR) loss-of-coolant accidents (LOCAs). Analyses of the ROSA-IIIbreak area spectrum tests in a recirculation line were performed using the SAFER03 to assess the predictive capability of the code for a BWR LOCA. The ROSA-rn test facility at the Japan Atomic Energy Research Institute (JAERI) was constructed to simulate a LOCA in a BWR/6-251 plant with 848 fuel bundles and 24 jet pumps. This paper summarizes the assessment results of SAFER03 which predicted the system responses and key phenomena well and the conservative peak cladding temperature (PCT) for recirculation line break tests with different break areas.

抄録全体を表示

Both experimental and analytical studies have been performed on the temperature coefficient of reactivity in a light-water moderated and reflected core loaded with highly-enriched-uranium fuel, which was constructed in the Kyoto University Critical Assembly (KUCA). The temperature effect on reactivity was measured for the range of 2070°C to investigate separately the effects of (1) the fuel pitch (H/²³⁵U atomic ratio) and (2) the core shape on this physical quantity. The experimental data were analyzed with use of the SRAC code system. The calculated eigenvalue k_eff agreed with the measured one within 0.5% in the C/E ratio for both the 2- and 3-dimensional diffusion calculations. The experimental data were approximately reproduced by both the eigenvalue and perturbation calculations. It was found that the contribution of the core region was negative to the temperature coefficient of reactivity, whereas that of the reflector region was positive. The synthesis of these contributions made the temperature coefficient negative in total. The degradation of moderation was the main contributor in the core region, whereas the decrease in the neutron absorption in the reflector region. The positive contribution of the reflector region became larger as the H/²³⁵U atomic ratio became smaller and the core shape became more slender.

抄録全体を表示