Yuichi Yamane

Review of studies on criticality safety evaluation and criticality experiment methods

Since the early 1960s, many studies on criticality safety evaluation have been conducted in Japan. Computer code systems were developed initially by employing finite difference methods, and more recently by using Monte Carlo methods. Criticality experiments have also been carried out in many laboratories in Japan as well as overseas. By effectively using these study results, the Japanese Criticality Safety Handbook was published in 1988, almost the intermediate point of the last 50 years. An increased interest has been shown in criticality safety studies, and a Working Party on Nuclear Criticality Safety (WPNCS) was set up by the Nuclear Science Committee of Organisation Economic Co-operation and Development in 1997. WPNCS has several task forces in charge of each of the International Criticality Safety Benchmark Evaluation Program (ICSBEP), Subcritical Measurement, Experimental Needs, Burn-up Credit Studies and Minimum Critical Values. Criticality safety studies in Japan have been carried out in cooperation with WPNCS. This paper describes criticality safety study activities in Japan along with the contents of the Japanese Criticality Safety Handbook and the tasks of WPNCS.

Power Profile Evaluation of the JCO Precipitation Vessel Based on the Record of the Gamma-Ray Monitor

Abstract The gamma-ray monitor installed at Processing Facility 1 in the JCO Tokai-works recorded the gamma dose rate change over time in proportion to power (nuclear fission rate) in the precipitation vessel from the beginning of the criticality accident to the end of the critical condition. The shape of the gamma dose rate record from 25 min after occurrence to the point of termination (“plateau” part) was normalized using the fission number 2.2 × 1018, and the absolute value of the power was evaluated.

Release of radioactive materials from high active liquid waste in small-scale hot test for boiling accident in reprocessing plant

The release behavior of radioactive materials from high active liquid waste (HALW) has been experimentally investigated under boiling accident conditions. In the experiments using HALW obtained through laboratory-scale reprocessing, the release ratio was measured for fission product (FP) nuclides such as Ru, Tc-99, Cs, Sr, Nd, Y, Mo, Rh and actinides such as Cm-242 and Am-241. As a result, the release ratio was 0.20 for Ru and was around 1×10−4 for the FP and actinide nuclides. Ru was released into the gas phase in the form of both mist and gas. For its released amount, weak dependency was found to its initial concentration in the test solution. The release ratio decreased with the increase in the initial concentration. For other FP nuclides and actinides as non-volatile, released into the gas phase in the form of mist, the released amount increased with the increase in the initial concentration. The release ratio of Ru and NOx concentration increased with the increase in the temperature of the test solutions. They were released together almost at the same temperature between 200 and 300 °C. Size distribution of particles like mist was measured. The data show that there was a difference between distributions at the temperatures below 150 °C and over 200 °C.

Release of Radioactive Materials from Simulated High-Level Liquid Waste at Boiling Accident in Reprocessing Plant

Abstract The release behavior of radioactive materials from high active liquid waste (HALW) has been investigated under boiling accident conditions. Results of the experiment using a nonradioactive simulated HALW found Ru to be a volatile element under the accident conditions and to be released into the gas phase in the form of both mist and gas. The Ru release rate and the apparent Ru volatilization rate constant were obtained under the boiling conditions of simulated HALW. The other fission product elements such as Cs were found to be nonvolatile and to be released into the gas phase in the form of mist. The mist size distribution near the surface of the simulated HALW in the reactor vessel was found to range from 0.05 to 20 μm with a peak diameter of ∼2 μm.

Options of Principles of Fuel Debris Criticality Control in Fukushima Daiichi Reactors

In the Three Mile Island Unit 2 reactor accident, a large amount of fuel debris was formed whose criticality condition is unknown, except the possible highest 235U/U enrichment. The fuel debris had to be cooled and shielded by water in which the minimum critical mass is much smaller than the total mass of fuel debris. To overcome this uncertain situation, the coolant water was borated with sufficient concentration to secure the subcritical condition. The situation is more severe in the damaged reactors of Fukushima Daiichi Nuclear Power Station, where the coolant water flow is practically “once through.” Boron must be endlessly added to the water to secure the subcritical condition of the fuel debris, which is not feasible. The water is not borated relying on the circumstantial evidence that the xenon gas monitoring in the containment vessels does not show a sign of criticality. The criticality condition of fuel debris may worsen with the gradual drop of its temperature, or the change of its geometry by aftershocks or the retrieval work, that may lead to criticality. To avoid criticality and its severe consequences, a certain principle of criticality control must be established. There may be options, such as prevention of criticality by coolant water boration or neutronic monitoring, prevention of the severe consequences by intervention measures against criticality, etc. Every option has merits and demerits that must be adequately evaluated toward selection of the best principle.

Improvement in estimation of first peak power based on non-linear temperature feedback reactivity in criticality accident with instantaneous reactivity insertion

A simple equation for the first peak power in a criticality accident due to instantaneous reactivity insertion into nuclear fuel solution system has been developed to improve the accuracy in the estimation of the first peak power keeping the easiness of calculation. The equation is based on the assumption that temperature feedback reactivity is a second-order function of an increase in fuel temperature. Peak power estimated using the equation was in a range between about a half and twice of experimental value. Its applicability to a wide range of initial reactivity and accuracy of estimation have been confirmed in the comparison to one-point kinetics numerical calculation. The expression suggests the first peak power increases with the square of small initial reactivity and three-halves power of large initial reactivity.