[en] CAMS (Computerised Accident Management Support) is a system that will provide assistance in case of the accidents in a nuclear power plant. The PSA module was developed in order to give useful information in this situation applying the PSA method, which is a comprehensive source of safety knowledge. This module contains plant-specific PSA data, comprising event trees, failure probabilities etc. It has several event trees categorised according to the initiating events. Each event tree has an initiating event frequency and branching probabilities. The various support systems for branches are considered and their dependencies are calculated logically. This module can be activated by data from the state identification (SI) module of CAMS. If an initiating event occurs, the event tree is re-calculated and the PSA module shows which systems of the plant should be activated to bring the plant to a safe state. If the plant responds to the event in the normal way, the plant will be shut down and come to a safe state. However, if some functions do not work, the PSA module generates another path and gives information about the critical systems. If the state of the plant is changed, either by the operators or automatically by the control system, the PSA module follows the new path. Because the estimation of the core damage frequency should be very quick in the accident situation, a simplified model of the event tree and fault trees was adopted. It enabled the PSA module to calculates the CDF within 5 seconds on a standard type work station. The development of the module has been successful. However, further development of the functionality of the module is suggested like real connection to a plant and to the strategy generator module of CAMS, applications for operational support, low power operation optimisation, etc. (author)

[en] The device of the present invention can conveniently measure γ-rays and neutron fluxes under water in a reactor without drawing out coolants in a reactor upon reactor shutdown. Namely, the device comprises an aluminum rod disposed in a vessel of a sealed-structure having substantially the same size as a fuel material and disposed in a pressure tube. A dosimeter for high radiation dose, or a neutron flux radiation foil is attached along the aluminum rod. With such a constitution, the device provides such advantageous effects that (1) the vessel of the sealed-structure as a measuring vessel, can be disposed to any position instead of a fuel material, (2) measurement for the reactor is possible for entire region in a vertical direction, (3) the measuring device can be attached and removed by remote control of a fuel exchanger, and (4) there is no need to draw coolants out of the reactor, although the inside of the reactor has an extremely narrow structure. (I.S.)

[en] The Fugen Nuclear Power Station (NPS) was shut down permanently in March 2003, and preparatory activities are underway to decommission the Fugen NPS. It is necessary to accomplish the decommissioning economically and rationally by optimizing the workload, exposure dose and waste mass. This is important at the planning stage of the decommissioning. Virtual reality (VR) technology may prove beneficial to this process with regard to minimizing the workers' radiation exposure as well as contributing towards achieving efficient use of manpower. It could also be a valuable tool in the actual dismantling phase. In addition to this, VR provides an effective medium in presentations for public acceptance as well as for communication with relevant engineers. The VRdose project conducted by Japan Nuclear Cycle Development Institute (JNC) and Halden Virtual Reality Centre is doing research and development of VR technology for use in the decommissioning process at the Fugen NPS. This is technically an extensive project, touching on many of the present challenges in the VR area such as visual simulation and animation, interaction with objects in a virtual environment and scenario generation and optimisation. This paper describes the present status and future of the system. (Author)

[en] Radioactive aerosol disperses slightly via contamination prevention systems such as control enclosures and filters when the nuclear installation is dismantled, and it might impact the environment. Therefore, when decommissioning is planned, it is necessary to assess the safety such as exposure dose evaluation to the public. For the radioactive aerosol, it is possible that the dispersion ratio is different according to the contamination condition, the dismantlement method of the material, nuclides (elements), etc. The radiation exposure evaluation for the decommissioning plan has been executed by operators in Japan based on a number of experiments (mostly cold tests) and overseas results. The decommissioning is now being carried out at the Tokai Power Station (GCR) and Fugen Decommissioning Engineering Center in Japan. In this study, the results data is acquired at the decommissioning sites, and the methodology and data for the exposure dose evaluation are verified and confirmed. These examination results will lead to the upgrading and improvement of the exposure evaluation methodology. In particular, the dismantlement work of connected piping of the heat exchanger (steam generator) was executed in the Tokai Power Station in 2008. In this study, we paid attention to the radionuclides of Co-60 and Cs-137 that adhered to piping, and the dispersion behavior of aerosol was measured and contamination prevention effect was assured. As a result, the data show that the cesium concentrates about four times higher than cobalt. Moreover, the effects of the prevention measures of contamination were confirmed and the behavior of the radioactive aerosol became clear and the effective findings about the dose evaluation of the dismantling were collected. (authors)

[en] CAMS (Computerized Accident Management Support) is a system that will provide assistance to the staff in the control room, in the technical support centre, and in a national safety centre. These three groups of users do not need the same type of support. Support is offered in identification of the plant state, in assessment of the future development of the accident, and in planning of accident mitigation strategies. Last year the predictive part of the system was tested at a safety exercise arranged by the Swedish Nuclear Inspectorate, and found to be a useful tool, with potential for further development. Now, new methods are added in signal validation, state identification, tracking simulation, predictive simulation, risk monitoring, and man-machine interface design. A prototype will be demonstrated at Loen in May 1996. This prototype is still under development. The purpose of this prototype is to test those methods in a simulated environment to verify that the developed functions, using different techniques, can work together producing the desired result in an efficient way. The plan is to test these techniques at power plants. During the CAMS design, a considerable effort has been given to maintain the generality of the CAMS concept; although the referenced process has been so far a BWR nuclear plant, the use of this structure and design can be applied to other processes, including non-nuclear processes. The research programme is carried out in close cooperation with member organizations (author)

[en] The system of the present invention collectively recognizes a state of a nuclear reactor, on a real time, after occurrence of a scram accident and informs this state to an operator by a method capable of recognizing the state at a glance. Namely, a measuring means measures the state for each reactor portions after occurrence of the scram accident. A calculation means calculates a risk index showing a degree of danger of the reactor state based on predetermined measured information among individual information obtained by the measuring means. A display means displays the risk index calculated by the calculation means. In the system thus constituted, since the risk index is calculated/displayed by using only the predetermined measured information obtained by the measuring means, the operator can recognize the degree of danger of the reactor state collectively on real time at a glance. As a result, the operator can easily perform a countermeasure at an early stage, thereby enabling to effectively prevent the scram accident from developing to severe accidents. (I.S.)

[en] After generation of scram signals, the operation state of a reactor scram means is monitored on real time. When a higher rank reactor scram means is not operated or operated incompletely, lower rank reactor scram means are automatically caused to operate successively. The operation state of a reactor cooling means is monitored on real time, and when a higher rank reactor cooling means is not operated or operated incompletely, lower rank reactor cooling means are automatically caused to operate subsequently. This can reliably provide reactor scram and reactor cooling upon occurrence of a scram accident. Then, such an event that the scram accident develops to a severer accident due to poor countermeasure at the initial stage of the accident can effectively be avoided. Further, the cooling state of the reactor is evaluated with reference to a knowledge base which stores the relation of a reactor water level and a ratio between a predetermined measuring information and a minimum limit thermal flux, thereby capable of recognizing the reactor cooling state in an extremely short period of time. (N.H.)

No abstract available

No abstract available

[en] In Japan, 4 nuclear power stations are under decommissioning and some nuclear fuel cycle facilities are expected to be decommissioned in the future. On the other hand, the safety regulation of decommissioning of nuclear facilities was changed by amending act in 2005. An approval system after review process of decommissioning plan was adopted and applied to the power stations above. In this situation, based on the experiences of the new regulatory system, the system should be well established and moreover, it should be improved and enhanced in the future. Nuclear Industry and Safety Agency (NISA) is in charge of regulation of commercial nuclear facilities in Japan and decommissioning of them is included. Japan Nuclear Energy Safety Organization (JNES) is in charge of technical supports for NISA as a TSO (Technical Support Organization) also in this field. As for decommissioning, based on regulatory needs, JNES has been continuing research activities from October 2003, when JNES has been established. Considering the 'Prioritized Nuclear Safety Research Plan (August 2009)' of the Nuclear Safety Commission of Japan and the situation of operators facilities, 'Regulatory Support Research Plan between FY 2010-2014' was established in November 2009, which shows the present regulatory needs and a research program. This program consists of researches for 1. review process of decommissioning plan of power reactors, 2. review process of decommissioning plan of nuclear fuel cycle facilities, 3. termination of license at the end of decommissioning and 4. management of decommissioning waste. For the item 1, JNES studied safety assessment methods of dismantling, e.g. obtaining data and analysis of behavior of dust diffusion and risk assessment during decommissioning, which are useful findings for the review process. For the item 2, safety requirements for the decommissioning of nuclear fuel cycle facilities was compiled, which will be used in the future review. For the item 3, measuring method, release procedure and analysis code for the site release were studied for the establishment of the license termination process in the future. From FY 2010, based on the new plan, we have started the researches for the standardization of review process of decommissioning plan for power reactors and nuclear fuel cycle facilities, establishing the process and criteria of license termination and appropriate method of management of decommissioning waste based on the waste form confirmation process. (author)