Since the end of cold war, negotiations on nuclear disarmament have made progress and its verification systems are under consideration. One of them is verification system for disposition of excess nuclear materials including weapon grade plutonium between USA and Russia and another is verification system of Fissile Material Cut-off Treaty (FMCT). Such verification systems, which will be applied also to Nuclear-Weapon-States (NWS), can affect IAEA safeguards applied to peaceful nuclear activities in Non-Nuclear-Weapon-States (NNWS). NNWS have much experience to accept IAEA comprehensive safeguards in accordance with the Treaty on the Non-Proliferation of Nuclear Weapons (NPT). Concept of verification systems related to nuclear material for nuclear disarmament is considered and analyzed from the view point of experience of NNWS so that no additional obligation will be born by the NNWS which accept comprehensive safeguards strengthened already by additional protocol.
The shape and size of a NaI (Tl) scintillator that should maximize response variation with γ-ray incident angle was estimated by analytical model calculation. It proved that, even for gamma rays of energy exceeding 1MeV, a slab detector measuring 50cm×50cm×5cm thick should present a ratio of at least 4 between maximum and minimum responses against incidence at different angles. For a sample case of 60keV gamma rays, estimation of the incident angle dependence by means of Monte Carlo simulation agreed well with experiment using a CZT detector. The counts from photoelectric peak varied with incident angle roughly along a sine curve. The foregoing finding served as basis for proposing a practical direction finder for γ-ray source operating on the principle of determining the source direction from variations in count with incident angle.
Many studies to clarify the mechanism, which hydrated cement maintains high pH condition for a long period have been reported. However, due to the lack of solid phase analysis, they do not related to the dissolution process of cement hydrates. We have performed a permeability test in which centrifugal force is used to percolate pore water and have studied the change of aqueous chemistries and that of the solid phase simultaneously in order to investigate the dissolution phenomena of hydrated cement. The dissolution process of the hydrate phase agreed with that of the results of the other dissolution tests that is dissolving in the order of Na2O and K2O, Ca(OH)2 and C-S-H gel. Accordingly, it was ascertained that this permeability test was carried out under the equilibrium condition between percolated pore water and hydrates. It was confirmed that pore volume increased by the dissolving of hydrates and pore structure clearly changed. The relation between the volume of percolated water and its chemical composition was evaluated by the model coupling transport and chemical equilibrium calculations. The calculated calcium and silicon concentrations of the percolated water were in good agreement with the experimental values, and thus, we were considered that this model was useful for evaluating the behavior of cement hydrates dissolution.
In order to investigate applicability of Ti alloy to large scaled structural material for fusion reactors, irradiation effect on the mechanical properties of Ti-6Al-4V alloy and its TIG welded material was investigated after neutron irradiation (temperature: 746-788K, fluence: 2.8×1023n/m2 (>0.18MeV)). The following results were obtained. (1) Irradiated Ti alloy shows about 20-30% increase of its tensile strength and large degradation of facture elongation, comparing with those of unirradiated Ti alloy. (2) TIG welded material behaves as Ti alloy in its tensile test, however, shows 30% increase of area reduction in 373-473K, whereas 1/2 degradation of area reduction over 600K. (3) Irradiated TIG welded material behaves heavier embrittlement than that of irradiated Ti alloy. (4) Charpy impact properties of un- and irradiated Ti alloys shift to ductile from brittle fracture and transition temperature shift, ΔT was estimated as about 100K. (5) Remarkable increase of hardness was found, especially in HAZ of TIG welded material after irradiation.
In Japan there are many kinds of radiation facilities, and a great number of radiation employees are engaged in plant repairing. It is therefore, very important to strive for employee controls, radiation controls, health examinations and data control. Furthermore, it is necessary to establish a total data management system that processes numerous amounts of data concerning radiation employees. The present paper proposes the establishment of a radiation work market on the web using a total data management system. The system will include radiation employee control information service for members who are planning new employment contracts.
A helium accumulation fluence monitor (HAFM) has been developed at the Japan Nuclear Cycle Development Institute's experimental fast reactor JOYO. This monitor provides high-precision neutron dosimetry and direct measurements of helium production in the fast reactor structural component materials. The measurement accuracy of neutron fluence by the HAFM has been evaluated. Calibration tests using helium ion implanted samples showed that the HAFM measurement system can measure helium atoms with an uncertainty of about 5%. The HAFMs were irradiated in the standard neutron spectrum field of the fast neutron source reactor YAYOI at the University of Tokyo. The HAFM measurement system measured neutron fluence of typical fast reactor spectrum fields and low energy neutron fields with an uncertainty of about 7%. The last stage of this study confirmed the applicability of HAFM to fast reactors. In these tests at JOYO, enriched boron-type HAFMs were irradiated and the neutron fluence was measured. The results obtained are in good agreement with those from the existing foil activation method within the experimental uncertainty and verified the reliability of the HAFM. As a result of this study, it was confirmed that the HAFM could be applicable for fast reactor dosimetry.
To perform Boron Neutron Capture Therapy (BNCT) with epithermal neutron beam, it is essential to make treatment planning based on the calculated irradiation dose and its distribution because epithermal beam penetrate to deeper region which can cause an unwanted radiation damage outside the target. A Computational Dosimetry System (JCDS) has been developed at Japan Atomic Energy Research Institute, which can calculate the irradiation dose and provide dose distribution by using MCNP calculation. Using this system, target dose, normal tissue dose, dose to the critical organ can be calculated. The JCDS treatment planning should be adjusted to the actual irradiation conditions (i.e. accurate patient irradiation position, especially distance and angle relative to the neutron beam collimator) which is important to perform an accurate irradiation previously calculated from to the treatment planning. For this purpose, a Patient Setting System has been developed which can reflect the data obtained from JCDS. Basic experiments using the head phantom and clinical application of this system in cases of BNCT, the accuracy of this system have been evaluated and have been proven to be accurate with the error within 5mm. In conclusion, by combining JCDS and the Patient Setting System, an efficient and safe clinical BNCT trial using epithermal beam could be performed.
A nitrogen gas extinguisher system will be installed at MONJU as one of countermeasures against sodium fire. The basic design specifications of this system were determined by some experiments. Following experiments were conducted with the object of confirming (1) an oxygen concentration to suppress sodium fire, (2) a nitrogen gas mixing efficiency to decrease the oxygen concentration, and (3) a nitrogen gas feed rate to prevent air in-leak from the outside to keep the low oxygen atmosphere. (1) The sodium combustion rate, which was evaluated by a calorimetric heat, depended on oxygen concentration, but it could be rearranged simply by a mole flow rate of oxygen gas. The sodium combustion heat was sharply changed at 1 mole/min, and the oxygen concentration of less than 5% was sufficient to suppress sodium fire under the natural convection condition. Since the sodium combustion heat was approx. 15kW/m2 at 5% of oxygen concentration, while approx. 200kW/m2 at 21%. (2) The mixing efficiency at the initial nitrogen gas feed rate is important to determine the capacity of the nitrogen gas tank. It was verified that over 0.9 of the mixing efficiency could be obtained by a 1/10 scale model and a large scale model test. (3) It was confirmed that 4, 000m3/h of nitrogen gas should be fed for the largest partitioned area of the secondary cooling system in order to keep the low oxygen atmosphere.
In Japanese BWR plants, thermal recombiners have been installed as the flammability control system (FCS) to keep hydrogen and oxygen concentrations below the flammability limit during a loss-of-coolant accident (LOCA). In the meantime, the passive autocatalytic recombiner which needs no electric power supply and heating sources and has an outstanding cost performance and reliability consequently, has been developed recently. This type of recombiner is already introduced to many PWR plants in the United States and Europe. In this situation, Japanese BWR utilities planned to introduce the passive autocatalytic recombiner to new plants to reduce the equipment and maintenance costs and improve system reliability by eliminating dynamic devices and support systems. The performance of catalytic recombination systems, however, has to be evaluated whether they satisfy the regulatory licensing requirements specific to Japan. Accordingly, the Japanese BWR utilities and plant manufacturers carried out a joint study, and experimental tests and analyses were conducted on the catalytic FCS technology. Based on the data from the tests, necessary number of the recombiners and lay-out planning for the ABWR containment vessel was estimated as a reference to ensure their applicability to ABWR plants.
It is anticipated that the modification of frequency of surveillance tests, preventive maintenance or parts replacement of safety related components might affect the component failure probability and result in the change of core damage frequency. In this study, the change of core damage frequency is assessed using a simplified PSA model capable of calculating core damage frequency in a short time period, which was developed by the USNRC to address accident sequence precursors. As a result of the analysis, the following is clarified: (1) Core damage frequency increases sensitive, when failure probability of motor driven pumps increase from 5th percentile lower bound to 95th percentile upper bound. (2) Change of core damage frequency is little affected, when failure probability of motor operated valves, turbine driven auxiliary feed water pump and emergency diesel generators vary between 5th percentile lower bound and 95th percentile upper bound.
This paper describes a model for human error probability (HEP) for seismic probabilistic safety assessments (PSAs) of nuclear power plants and its application. Considering the factors unique to seismic events such as level of seismic motion and stress to the operators, the authors adopted a model called a limited ramp model (LRM), where HEP is assumed to be the same as that for internal event PSAs for zero level of seismic motion, and then increases linearly with seismic level to a constant at a certain seismic level. This model was applied to estimate the effect of human error on core damage frequency (CDF). Here, operator actions in the accident sequences initiated by loss of offsite power were categorized into two groups, short term actions and medium/long term actions, and parameters of the LRM for the HEP for each action were determined from available information on shaking table tests and an existing human reliability analysis method. The results showed that the effect of human error on the CDF was not significant for this case and indicated that, the modeling approach presented here is useful for examining the importance of various operator actions in seismic events.