Evaluation of Main Control Room Habitability in Japanese LWR (1) Outline of Evaluation Method and Conditions

摘要

It has been recognized that main control room habitability (CRH) is very important safety item. Its evaluation method and conditions was recently updated as a common method both for Boiling Water Reactor (BWR) and Purresurized Water Reactor (PWR) in Japan. This paper describes the common procedure for CRH evaluation. Postulated accidents for CRH evaluation are Loss of Coolant Accident (LOCA) for BWR and PWR, Main Steam Line Break Accident (MSLBA) for BWR, and Steam Generator Tube Rupture (SGTR) for PWR. Evaluation period is thirty days after each accident occurs. Acceptable criterion for each operator is 100mSv as total effective dose equivalent for thirty days, considering the net staying period of an operator in a control room and frequency of alternation of working staff Total effective dose equivalent is calculated not only for an operator in control room, but also for an operator outside a building for alternation, and then total effective dose is compared with the criterion. The routes for dose evaluation are classified into five patterns, covering three patterns of location of radioactive source (i.e. source inside a building, source released into air, and source leaked into a control room) and covering two patterns of location of an operator (i.e. in control room and outside a building for alternation). Total effective dose by a route for source which is released into air and leaked into a control room is typically the largest contributor, in case of PWR LOCA evaluation, as an example showing relatively severe result. Evaluated dose for this dominant route is significantly affected by two calculations. One is a source term for calculation of radioactive gas release and the other is dispersion calculation in the air for concentration of radioactive gas in the vicinity of a control room. For LOCA 100percent Kr,Xe and 50percent iodine of total core radionuclide inventory are assumed to be released immediately into containment vessel. This is the most conservative assumption among current licensing evaluation in Japan. The characteristic of atmospheric dispersion calculation for CRH evaluation is that downdraft effect that should be taken into account if the building exists near the release point of radioactive gas. In order to simply simulate this effect, it is assumed that released gas is dispersed immediately and uniformly behind the downwind side of the building. A wind tunnel experiment with small-scaled containment vessel and reactor building confirms this effect. The regulatory model in Japan, which is current licensing model for evaluation at still further distance, is extendedly applied for evaluation at short distance like CRH and the model relatively well predicts an averaged concentration on the surface of the building around control room for PWR LOCA. Based on this result, regulatory model with coefficient 0.5 was selected from "Meteorological Guideline for Safety Analysis of Light Water Reactor Facilities," (Nuclear Safety Comission of Japan, 2001) for CRH evaluation of PWR LOCA, in couple with other various conservative conditions such as hypothetical source term and calculation of centerline concentration for all wind directions. The method of this paper for CRH involves enough margins as a total evaluation, compared with worldwide standard method. For the future, however, further investigations are to be continued for extended model both for source term and atmospheric dispersion with more accuracy and with moderate margin.