Toshimitsu Homma

A new computational method of a moment-independent uncertainty importance measure

Abstract For a risk assessment model, the uncertainty in input parameters is propagated through the model and leads to the uncertainty in the model output. The study of how the uncertainty in the output of a model can be apportioned to the uncertainty in the model inputs is the job of sensitivity analysis. Saltelli [Sensitivity analysis for importance assessment. Risk Analysis 2002;22(3):579–90] pointed out that a good sensitivity indicator should be global, quantitative and model free. Borgonovo [A new uncertainty importance measure. Reliability Engineering and System Safety 2007;92(6):771–84] further extended these three requirements by adding the fourth feature, moment-independence, and proposed a new sensitivity measure, δi. It evaluates the influence of the input uncertainty on the entire output distribution without reference to any specific moment of the model output. In this paper, a new computational method of δi is proposed. It is conceptually simple and easier to implement. The feasibility of this new method is proved by applying it to two examples.

A New Importance Measure for Sensitivity Analysis

Uncertainty is an integral part of risk assessment of complex engineering systems, such as nuclear power plants and space crafts. The aim of sensitivity analysis is toidentify the contribution of the uncertainty in model inputs to the uncertainty in the model output. In this study, a new importance measure that characterizes the influence of the entire input distribution on the entire output distribution was proposed. It represents the expected deviation of the cumulative distribution function (CDF) of the model output that would be obtained when one input parameter of interest were known. The applicability of this importance measure was tested with two models, a nonlinear nonmonotonic mathematical model and a risk model. In addition, a comparison of this new importance measure with several other importance measures was carried out and the differences between these measures were explained.

Probabilistic Assessment of Doses to the Public Living in Areas Contaminated by the Fukushima Daiichi Nuclear Power Plant Accident

Many residents are exposed to radiation in their daily lives in the areas contaminated by radioactive materials by the Fukushima Daiichi Nuclear Power Plant accident. To protect the people from radiation exposures adequately, dose assessment is necessary. The aim of this study is to provide the scientifically based quantitative information about a range of received doses to the people from the evacuation areas and the deliberate evacuation areas. To achieve this aim, we adopted a probabilistic approach that can provide the information about a range of doses and their likelihood of occurrence taking into account uncertainty and variability of input data. The dose assessment was performed based on the measurement data of the surface activity concentrations of 137Cs and the results of actual survey on behavioral patterns of the population groups living in Fukushima Prefecture. As the result of assessment, the 95th percentile of the annual effective dose received by the inhabitants evacuated was mainly in the 1–10 mSv dose band in the first year after the contamination. However, the 95th percentile of the dose received by some outdoor workers and inhabitants evacuated from highly contaminated areas was in the 10–50 mSv dose band.

Radiation protection issues on preparedness and response for a severe nuclear accident: experiences of the Fukushima accident

Radiation protection issues on preparedness and response for a severe nuclear accident are discussed in this paper based on the experiences following the accident at Fukushima Daiichi nuclear power plant. The criteria for use in nuclear emergencies in the Japanese emergency preparedness guide were based on the recommendations of International Commission of Radiological Protection (ICRP) Publications 60 and 63. Although the decision-making process for implementing protective actions relied heavily on computer-based predictive models prior to the accident, urgent protective actions, such as evacuation and sheltering, were implemented effectively based on the plant conditions. As there were no recommendations and criteria for long-term protective actions in the emergency preparedness guide, the recommendations of ICRP Publications 103, 109, and 111 were taken into consideration in determining the temporary relocation of inhabitants of heavily contaminated areas. These recommendations were very useful in deciding the emergency protective actions to take in the early stages of the Fukushima accident. However, some suggestions have been made for improving emergency preparedness and response in the early stages of a severe nuclear accident.

Evaluation of the precautionary action zone using a probabilistic consequence analysis

To develop an effective emergency plan for a nuclear accident, the size of the precautionary action zone (PAZ) was evaluated using a Level 3 probabilistic safety analysis (PSA) code, the OSCAAR. For accident scenarios identified through a Level 2 PSA analysis for representative boiling water reactor (BWR) and pressurized water reactor (PWR) plants, the absorbed doses to red marrow were calculated as a function of the distance from the accidental site for possible weather sequences in a year, and then the evaluation was made for the distance which satisfies the dose criterion of 1 Gy for acute exposure, as established in the Basic Safety Standards (BSS) of the International Atomic Energy Agency (IAEA). Consequently, the present study indicated that the size of the PAZ should be approximately a 3-km radius under average weather conditions.

Sensitivity analysis of a passive decay heat removal system under a post-loss of coolant accident condition

Passive safety features are now of interest to the design of future generation reactors because of their peculiar characteristics of simplicity, reduction of human interaction, and avoidance of failures of active components. However, the large uncertainty associated with the responses of passive systems might not be ignored. Therefore, it is necessary to identify the uncertain inputs that have the important impact on the uncertainty of the system performance. In this study, two global sensitivity measures, the first-order sensitivity index and the total-order sensitivity index, are applied to a natural circulation decay heat removal system of a gas-cooled fast reactor for identifying the important system inputs. It is found that the uncertainty in the system pressure contributes the most to the uncertainty in the system outputs. In addition, the cooler (the heat exchanger of the emergency cooling system) wall temperature, the Nusselt number in the mixed convection regime, and the friction factor in the mixed convection flow regime also have small impact on the uncertainty of the system outputs.

Study on the Intervention and Return Criteria for Relocation Using PSA method

Relocation is one of the long-term protective actions for a nuclear and radiological emergency. The objective of relocation should be taken into account the concepts of “safe living condition” and “return to normality”, and prepare the strategy adopted these concepts. In order to prepare the strategy, we have examined the model using Return Level which is basis on these concepts. We have investigated the dependence of number of people to be relocated, relocation time, avertable collective dose and avertable per caput dose per year to criteria, i.e. Intervention Level and Return Level. From these results, it would be impossible to justify the strategy in terms of cost-benefit analysis. Therefore, to justify the strategies which are taken into account “safe living condition” and “return to normality”, we have to consider the other factor such as social physiological aspects.