Hiromasa Iida

Design study of the deep-sea reactor X

The deep-sea reactor X (DRX) is a small nuclear plant designed to provide undersea power sources. It has the full advantages of nuclear reactors and can provide large power capacity and does not re...

Application of On-Line Digital Noise Analysis to Reactor Diagnosis in JMTR

The reactor noise analysis technique is particularly useful in reactor diagnosis for on-line monitoring if the raw noise signals can be processed in almost real time. An on-line reactor noise analysis system has been developed with use made of the mini-computer HITAC-10. This system utilizes functions for calculating the power spectral density in almost real time, plots the output by digital incremental plotter, and displays the results by means of color graphic display equipment, in order to detect anomalous reactor conditions with the statistical technique. Using this system, reactor noise signals have been measured and analyzed under various operational conditions in the JMTR. The variance of the power spectral density is found to fit a logarithmic probability density function. This function is independent of the frequency, but is dependent on the number of sampling functions. A logical procedure for anomaly detection based on statistical characteristics has been developed. It is applied to a case wher...

Development of CAD-to-MCNP Model Conversion System and Its Application to ITER

Abstract We developed a conversion system from three-dimensional computer-aided design (CAD) drawing data to MCNP geometry input data. This system consists of programs of “void creation” and “conversion into MCNP input data.” By using this system, it is possible to convert large and complicated CAD drawing data such as a fusion reactor into MCNP geometry input data. We applied this system to ITER CAD drawing data and created MCNP input data for ITER nuclear analysis. We calculated the neutron flux and nuclear heating using this input data. The calculation results agreed well with those by the other parties participating in this ITER research and development.

Monte Carlo analysis of helium production in the ITER shielding blanket module

Abstract In order to examine the shielding performances of the inboard blanket module in the International Thermonuclear Experimental Reactor (ITER), shielding calculations have been carried out using a three-dimensional Monte Carlo method. The impact of radiation streaming through the front access holes and gaps between adjacent blanket modules on the helium gas production in the branch pipe weld locations and back plate have been estimated. The three-dimensional model represents an 18° sector of the overall torus region and includes the vacuum vessel, inboard blanket and back plate, plasma region, and outboard reflecting medium. And it includes the 1 m high inboard mid-plane module and the 20 mm wide gaps between adjacent modules. From the calculated results for the reference design, it has been found that the helium production at the plug of the branch pipe is four to five times higher than the design goal of 1 appm for a neutron fluence of 0.9 MW a m−2 at the inboard mid-plane first wall. Also, it has been found that the helium production at the back plate behind the horizontal gap is about three times higher than the design goal. In the reference design, the stainless steel (SS):H2O composition in the blanket module is 80:20%. Shielding calculations also have been carried out for the SS:H2O composition of 70:30, 60:40, 50:50 and 40:60%. From the evaluated results for their design, it has been found that the dependence of helium production on the SS:H2O composition in the blanket module is small at the branch pipe. Altering the steel–water ratio to reduce the amount of steel and increasing the thickness by >170 mm will reduce helium production to satisfy the design goal and not have a significant impact on weight limitations imposed by remote maintenance handling limitations. Also based on the calculated results, about 200 mm thick shields such as a key structure in the vertical gap are suggested to be installed in the horizontal gap as well to reduce the helium production at the back plate and to satisfy the design goal.

Monte Carlo Analyses for ITER NBI Duct by 1/4 Tokamak Model

In the ITER shielding design, the biological dose rates after shutdown in the region around the NBI ducts are critical. We have performed shielding calculations for the ITER/NBI ducts by 3-D Monte Carlo and 2-D SN codes with activation calculations. From comparison between calculated results by 3-D Monte Carlo and 2-D SN calculations, it has been found that the calculated results by the 2-D SN calculation overestimate by a factor of about eight at the cryostat in the case of the 91.5 cm high duct opening. From the 2-D SN calculation with activation calculations, we have deduced the conversion ratio relating fast neutron flux to the biological dose rates of ~1.5 − 2.0 × 10−5 μSv/hour/(cm−2sec.−1). The biological dose rates are about 7 × 10 μSv/hour in 50–60 cm thick duct wall from the fast neutron flux by the 3-D Monte Carlo calculation and the conversion ratio, and they can satisfy the design criteria.

Ceramic turbomolecular pumping system in reactor structure of FER

Abstract A new pumping system consisting of ceramic turbomolecular pumps and their ceramic fore-pumps has been studied as the main pumping system of FER since these pumps have been developed at JAERI recently. The turbomolecular pump has a rotor of silicon nitride, which is driven by a gas turbine, oil free gas bearings and a non-contact spiral groove seal. The effects of neutron radiation, heat loads and magnetic fields on the ceramic pumps are evaluated based on the FER design conditions.

Radiation Streaming Calculations for Intor-J

The effects of radiation streaming through the neutral beam injector (NBI) port and divertor throat of a tokamak fusion reactor, the INTOR-J, was evaluated using Monte Carlo and discrete ordinates methods. Radiation streaming through the NBI port is found to be tolerable when a thick drift tube support acts as an effective shield. Neutron streaming through the divertor throat, however, makes the shutdown dose too high for personnel access into the reactor room. The radiation levels in the reactor room resulting from leakage through the NBI room walls are far smaller than that from leakage through the bulk shield, except behind the NBI room. The Monte Carlo-Monte Carlo and discrete ordinatesMonte Carlo coupling techniques used in the present study are shown to be very effective for the radiation streaming calculations.

Safety analysis and radioactivity confinement for ITER

Abstract The paper summarizes the main results of an collaborative effort by the ITER safety group (within the Systems Project Unit). The work was done in support of ITER design and evolved in collaboration with the other ITER groups. A safety approach has been defined and radioactivity dose limits have been quantified both in terms of target numbers for the design and of estimated limits expected to be requested by the regulatory bodies. Doses due to normal operation effluents of tritium and activation products have been estimated. Accidents of significance have been identified. Emphasis of accident analysis was put on events related to plasma (vacuum) chamber and cooling system. A confinement concept has been specified and assessed in terms of effective release fractions for accidental situations involving mobilization of both tritium and activation products. The masses of decommissioning and operational radioactive waste have been estimated and broadly characterized by the potential disposal options. Overall, the safety work resulted in safety features implemented in the design, recommendations pending for the EDA, and identification of potential problems to be resolved by EDA safety work as well as by safety related R&D.

Optimization of OH coil recharging scenario of quasi-steady operation in tokamak fusion reactor by lower hybrid wave current drive

Abstract Using simple physical model equations optimum plasma and rf parameters for an OH coil recharging scenario of quasi-steady operation in tokamak fusion reactors by lower hybrid wave current drive are studied. In this operation scenario, the minimization of the recharge time of OH coils or stored energy for it will be essential and can be realized by driving sufficient current without increasing the plasma temperature too much. Low density and broad spectrum are shown to be favorable for the minimization. In the case of FER (Fusion Experimental Reactor under design study in JAERI) baseline parameters, the minimum recharge time is 3–5 s/V s.

A Point Detector Scoring Method Compatible with Monte Carlo Transport Calculations of Specularly Reflected Particles

In using a Monte Carlo transport code, particle fluxes are underestimated with a calculational model using specular reflection boundaries when a point detector estimator that scores only the direct contribution to the detector from a collision point within the model is used. This underestimation occurs because the contributions from collision points outside the calculational model are neglected. An additional scoring scheme is developed to compensate for the discrepancy; the new scheme is implemented to threedimensional Monte Carlo transport code MORSE-GG. Validity of the method is shown by test calculations with ANISN and revised MORSE-GG.