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Water cooled ceramic breeder (WCCB) blankets have been regarded as a primary concept in Japan. We in-vestigated the tritium breeding capability of a WCCB blanket with a cylindrical structure, an advantageous shapefor withstanding high coolant pressure, in this study. The breeding area in the blanket was modeled as ahomogeneous mixture or heterogeneous structures. Nuclear analyses with Monte Carlo method were conductedfor evaluating the tritium breeding ratio (TBR) of the blanket. The neutron balance of the blanket was analyzedto elucidate the mechanism related to the increase in TBR when an additional thin breeding layer (envelope) wasapplied to the blanket. The neutron multiplication in (n, 2n) reaction increased concomitantly with increase ofthe beryllium volume ratio. However, numerous neutrons were not used efficiently for tritium production, butwere captured by the container, which was made of the reduced activation ferritic martensitic steel, F82H.Capture by the container can be reduced by introducing an envelope. The effect of the envelope was considerablewhen modeling the internal structure such as U-shaped pipes in the breeding area. The envelope was also appliedto a blanket with different container designs. An increase in TBR was shown irrespective of blanket design. Theenvelope effect was remarkable when it was difficult to achieve the internal configuration, which was equivalentto the homogeneous mixture in the breeding area

A Water-Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) System is being developed as one of the most important steps toward DEMO blanket in Japan. As for the ITER-TBM program, the conceptual design of WCCB-TBM system has been approved by the ITER organization (IO). And two years have already passed after the start of the preliminary design activity. WCCB TBM Team had a concern about TBM box structure withstanding over 15 MPa of the coolant pressure, and decided to change the design as the result of study on structural integrity. Recently, the design change of the TBM structure has been approved domestically. So, WCCB TBM Team has applied the design change to the IO, and agreed with the IO to receive the conceptual design review which is limited the scope to the TBM structure. This paper provides an overview of the recent achievements of the development of the WCCB Test Blanket Module System in Japan.

In test blanket module (TBM) program blanket major functions, which are tritium production, heat extraction for electric power production, and neutron shielding, will be demonstrated under fusion reactor environments in ITER. National Institutes for Quantum and Radiological Science and Technology (QST) has performed developments of water cooled ceramic breeder (WCCB) TBM as a primary concept in Japan. The container design of WCCB TBM was changed from the Box-shaped to a cylindrical structure to increase tritium breeding capability while maintaining pressure resistance under In-Box LOCA caused by water ingress into the container. The nuclear responses related to the blanket major functions would be changed with the container design change. The nuclear responses of WCCB TBM with the container design change were evaluated and the effects of the changed nuclear responses were described in this study. Tritium production rate increased by about 2 times with the container design changes. In contrast neutron shielding performance decreased due to the container design change. Degradation of neutron shielding in the cylindrical TBM led to an increase in nuclear heating of the TBM frame as well as an average of neuron flux behind TBM-set.

Automatic conversion systems from CAD data to MCNP geometry input data have been developed to convert the CAD data of the fusion reactor with very complicated structure. So far, two conversion systems (GEOMIT-1 and ARCMCP) have been developed and the third system (GEOMIT-2) is under developing. The void data can be created in these systems. GEOMIT-1 was developed in 2007, but a lot of manual shape splitting work for the CAD data was required to convert the complicated geometry. ARCMCP was developed in 2008. The algorithm has been drastically improved on automatic creation of ambiguous surface in ARCMCP, but it still required a little manual shape splitting work. The latest system, GEOMIT-2, does not require additional commercial software packages, though the previous systems require them. It also has functions of the CAD data healing and the automatic shape splitting. Geometrical errors of CAD data can be automatically revised by the healing function, and complicated geometries can be automatically split into simple geometries by the shape splitting function. Any manual works for CAD data are not required in GEOMIT-2. GEOMIT-2 is very useful for nuclear analyses of fusion reactors.

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 int...

Under the ITER/ITA task, we have conducted the neutron streaming experiment simulating narrow and deep gaps at boundaries between ITER vacuum vessel and equatorial port plugs. Micro-fission chambers and some activation foils were used to measure fission rates and reaction rates to evaluate the relative fast and slow neutron fluences along the gap in the experimental assembly. The MCNP4C, TORT and Attila codes were used for the experimental analysis. From comparing our measurements and calculations, the following facts were found: (1) in case of a such narrow and deep gap structure, the calculation with MCNP, TORT and Attila codes and FENDL-2.1 is sufficient to predict fast neutron field inside the gap; (2) by scattering neutrons in the experimental room, experimental error considerably increased at the deeper region than 100 cm; (3) angular quadrature set of upward biased U315 and last collided source calculation on TORT and Attila were very important technique for accurate estimation of neutron transport.

We developed a CAD/MCNP interface program prototype to easily perform Monte Carlo fusion nuclear design calculation with high accuracy using detailed calculation geometry. Usually, there is only CAD data on the solid region. However, CAD void region data must be input to MCNP. Our newly developed CAD/MCNP interface program creates the void region data automatically. The void region data is created by subtracting the solid region data using Boolean operations. The program has also the function of uniformly dividing the overall data into small cubes to simplify the void region data. It has been applied to create 3D CAD data in the ITER benchmark model. It was demonstrated that data of the void region in the ITER model were successfully created using our program.

The design of the ITER machine was presented in 2001 [1] , [2] . A nuclear analysis was performed at this time, using fairly detailed models and the best assessed nuclear data and codes that were available. As the construction phase of ITER is approaching, the design of the main components has been optimized/finalized and several minor design changes/optimizations have been made, some with the object to mitigate critical radiation shielding problems. These have required refined calculations to confirm that the nuclear design requirements are met. This paper reviews some of the most recent neutronic work with emphasis on critical nuclear responses in the TF coil inboard legs and vacuum vessel related to design modifications made to the blanket modules and vacuum vessel.

For the evaluation of gamma-ray dose rates around the duct penetrations after shutdown of nuclear fusion reactor, the calculation method is proposed with an application of the Monte Carlo neutron and decay gamma-ray transport calculation. For the radioisotope production rates during operation, the Monte Carlo calculation is conducted by the modification of the nuclear data library replacing a prompt gamma-ray spectrum with a decay gamma-ray spectrum. By multiplying each correction factor, which is ratio of the actual activation level after shutdown to the production rate during operation, with each decay gamma-ray flux due to each radioisotope, the decay gamma-ray dose rate is evaluated. In order to improve the statistical error, a variance reduction method is proposed by the application of the weight window importance technique and the specification of the decay gamma-ray generation location. We identify the cell producing the decay gamma-ray which can contribute the decay gamma-ray flux in evaluation lo...

In the ITER project, the estimation of the dose equivalent rate levels after reactor shutdown for hands-on maintenance around the torus is a key point. The Sn transport and activation codes, because of their poor ability for modelling complex geometry (as in the ITER machine), yield a large uncertainty in radiation transport calculations where the geometry is not simple. In this paper, we propose a method that solves the above problem using a Monte Carlo code that allows a detailed geometry description. This new method requires modification of nuclear data library replacing a prompt gamma spectrum with a decay gamma spectrum and also a modest change in the computer program (MCNP). A simple geometry benchmark problem was conducted, comparing the new method and an existing method (THIDA-2). The agreement of the two methods is fairly good, suggesting that the new method is useful for very complex geometry devices. The slight difference observed in the results from both methods likely comes from the difference in the nuclear data library used in both methods.

The design has progressed on the vacuum vessel and First Wall (FW)/blanket for the Reduced Technical Objective/Reduced Cost (RTO/RC) ITER. The basic functions and structures are the same as for the 1998 ITER design. The design has been improved to achieve, along with the size reduction, ∼50% target reduction of the fabrication cost. The number of blanket modules has been minimized according to smaller dimensions of the machine and a higher payload capacity of the blanket Remote Handling tool. A concept without the back plate has been designed and assessed. The blanket module concept with flat separable FW panels has been developed to reduce the fabrication cost and future radioactive waste.

In the shielding design of the ITER machine, which is a tokamak fusion experimental reactor and has a very complex geometry, it is important to have a reliable estimation of the dose rate levels after reactor shutdown for realising hands-on maintenance around the torus. The ITER project position is that dose rates inside the cryostat be kept low enough to allow for human access shortly after shutdown for limited periods to provide rescue and/or maintenance activities.The methodology of estimation for dose rates after shutdown in such a complex geometry machine is discussed. The Monte Carlo method is preferable to conduct neutron transport calculations in the ITER geometry. The Conversion Factor method, which was used for dose rate estimation in the 1998 ITER shielding design, is described with an example of dose rate estimation around penetrations in the vacuum vessel. A Full Monte Carlo method is proposed showing the possibility of eliminating uncertainty accompanied with conversion factors from neutron ...

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.

The shutdown dose rates around the equatorial and divertor maintenance ports of ITER were evaluated with the 2-D/3-D combined approach, using the three-dimensional continuous-energy Monte Carlo code, MCNP-4B and the two-dimensional discrete ordinate code, DOT3.5. The neutron flux-to-shutdown dose rate conversion factor is derived with the two-dimensional geometry using the THIDA code system and the assumed operation scenario, i.e. the neutron fluence of 0.3 MWa/m2 in ten years of operation. The dose rate around the equatorial and divertor ports after 106 seconds (11.6 days) after reactor shutdown ranges from 100 to 200 μ Sv/h. Attempts to further reduce the dose rate by improving the shield design were made to follow the principle of ALARA.

Since the main features of the design of the neutral beam (NB) system for the International Thermonuclear Experimental Reactor (ITER) were first reported, integration with the tokamak and with the rest of the plant has been the main priority. Moreover, operational requirements and maintainability have been considered in the evolution of the design. Each of the three NB injectors is connected to the tokamak vacuum vessel with the NB duct on an equatorial port. The article describes the integration of the NB port/duct with the blanket, the vacuum vessel, the toroidal field and poloidal field coils, the cryostat and the bioshield. Two main design modifications are reported. The insulation of the source, originally done with compressed gas, is now achieved with vacuum to limit the power losses caused by the radiation induced conductivity. Large cylindrical insulators are still required but their inner diameter has been reduced from 2.7 to 1.8 m. The improvements on the compensation system needed to reduce the magnetic field in the NB volume are also described. Finally, the progress in R&D for the ITER NB system is reported, including an overview of the achievements in the critical areas of negative ion production at high current density (tests of a large size, low pressure, steady state caesiated ion source), acceleration up to 1 MV (tests of two alternative accelerator concepts) and neutralization (tests of an experimental plasma neutralizer to investigate it as an alternative to the gas target neutralizer).

Shielding analyses for the International Thermonuclear Experimental Reactor (ITER) neutral beam injector (NBI) ports have been performed using two-dimensional discrete ordinates S n method with activation analyses and three-dimensional Monte Carlo method. From the two-dimensional S n /activation analyses, it was found that a conversion ratio relating fast neutron flux to the biological dose rates at 10 6 s after reactor shutdown is 1.5–2.0×10 −5 μSv/h per (cm −2 s −1 ) for the fast neutron flux with energies above 0.1 MeV. From the results of three-dimensional Monte Carlo analyses, the fast neutron flux above 0.1 MeV at the cryostat is about 3.6×10 6 cm −2 s −1 for the case of a 50–60 cm thick NBI port wall (thin case), and 5.2×10 5 cm −2 s −1 for a 60–65 cm thick port wall (thick case). The corresponding biological dose-rates at the cryostat are about 7 and 2×10 1 μSv/h for the thin and thick cases, respectively. These dose rate levels satisfy the tentative design target of 100 μSv/h specified for the ITER Engineering Design Activity (EDA).

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.

In a D-T fusion tokamak, neutron activation of structural material will result in a source of heat from radioactive decay after shutdown of the plasma. Although relatively small, in certain postulated loss-of-cooling accidents this decay heat could drive a temperature transient of modest proportions. In safety studies of the International Thermonuclear Experimental Reactor (ITER), such accident scenarios are analysed for their possible consequences, requiring an accurate knowledge of the level and time-dependence of the decay heat in components in which it is significant. This is calculated by activation modelling using computer codes with libraries of nuclear data. This paper describes an international benchmark activity to validate several of these codes and data, using the results of direct measurements of decay heat in ITER-relevant materials after exposure to a flux of 14 MeV neutrons. The results show consistency between the codes and, when using data from the recently-compiled FENDL:A-2 activation data library, good agreement with the experiments over the time-scales of interest in ITER safety studies. © 1999 Elsevier Science S.A. All rights reserved.

A ‘handy method’ has been developed for performing quick estimates of radiation streaming through cylindrical and rectangular holes in shield assemblies. Quick estimates are often needed during conceptual studies or detailed design of nuclear plants. For example, ITER (International Thermonuclear Engineering Reactor) has a large number of penetrations in the blanket and vacuum vessel and biological shield. This method estimates the levels of radiation that leaks through these holes as well as the magnitude of the fluxes along the hole wall. Results obtained using the ‘handy method’ are in good agreement with 3-D Monte Carlo calculations suggesting that the method provides a practical tool for streaming analysis.

Two dimensional radiation transport calculations have been carried out to determine radiation streaming through the ITER equatorial ports. The NBI port has been identified as the most critical. Shielding requirements were estimated to minimize the nuclear heating rates in the toroidal field (TF) coils and the activation of the cryostat, where hands-on maintenance is anticipated. The shielding efficiency of steel/water port walls was investigated as a function of port and wall dimensions and steel volume fraction. For open ports, i.e. the neutral beam injector ducts, 40–50 cm thick port walls composed of 40% SS-60% H 2 O–75% SS-25% H 2 O will reduce the TF coil heating to acceptable levels, while 60–65 cm thick walls are necessary for reducing the dose rates at ∼2 weeks after shutdown to levels of 750 μSv h −1 at the cryostat.

Two-dimensional shielding analyses have been performed on the maintenance and test module mid-plane ports in the International Thermonuclear Experimental Reactor (ITER). Nuclear responses in the toroidal field (TF) and poloidal field (PF) coils around these ports have been calculated. Shield plugs with 20-mm-wide gaps between the blanket modules exist in both ports. In the case of the test module port, they are installed at the back of a 500-mm-thick test module. In order to satisfy design limits, 540- and 150-mm-thick shield plugs are required for the maintenance and the test module port, respectively. In addition, nuclear responses as a function of the gap width have also been estimated. In the case of the 50-mm-wide gap, it is found that 750 and 390-mm-thick shield plugs are required for the maintenance and test module ports, respectively.

Gamma-ray exposure dose rates at the ITER site boundary were estimated for the cases of removal of a failed activated Toroidal Field (TF) coil from the torus and removal of a failed activated TF coil together with a sector of the activated Vacuum Vessel (VV). Skyshine analyses were performed using the two-dimensional SN radiation transport code, DOT3.5. The exposure gamma-ray dose rates on the ground at the site boundary (presently assumed to be 1 km from the ITER building), were calculated to be 1.1 and 84 μSv/year for removal of the TF coil without and with a VV sector, respectively. The dose rate level for the latter case is close to the tentative radiation limit of 100 μSv/year so an additional ∼14 cm of concrete is required in the ITER building roof to satisfy the criterion for a safety factor often for the site boundary dose rate.

A three dimensional (3-D) shielding analysis of the International Thermonuclear Experimental Reactor (ITER) has been performed with the aim of calculating the nuclear heating on the magnet system, correlating it to the existing vertical and horizontal ports. When these openings are left unshielded, more than 50 kW are calculated for the upper half of Toroidal Field Coil system and two of the Poloidal Field Coils. A simple plug, with same thickness as of the vacuum vessel can lower the heating to meet the imposed requirements. 5 refs., 6 figs., 4 tabs.

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 require oxygen for power production. An application conceivable in the near future is that for a submersible. The Japan Atomic Energy Research Institut is conducting a design study of a 150-kW (electric) DRX plat for a deep-sea research vessel. It has a so-called «integrated pressurized water reactor,» having a steam generator inside the reactor vessel. A pressure chell includes a turbine and a generator as well as a reactor vessel, resulting in a very compact electricity producing plant

Amantadine is not thought to be effective for the treatment of swine-origin influenza virus (S-OIV) based on an analysis of genetic sequences of the M2 protein. However, the actual clinical efficacy of amantadine has not been well documented. Here, we were able to compare the efficacies of amantadine and neuraminidase inhibitors. Subjects consisted of 428 patients, including 144 with seasonal influenza (flu) identified between 2008 and 2009, and 284 with S-OIV identified between July 1 and November 30, 2009. Diagnosis of flu was established using a rapid diagnostic kit obtained commercially in Japan. Body temperature sheets were obtained from 95% of the S-OIV patients. Times required to recover normal body temperature were compared among subjects using different antiviral drugs. Genetic abnormalities in the M2 protein were also investigated in 66 randomly selected subjects from within the patient pool. Overall, the average hours required to recover normal body temperature in S-OIV patients treated with amantadine (160 cases), with oseltamivir (59 cases), or with zanamivir (65 cases) were 33.9 ± 20.7, 31.7 ± 16.0, or 36.3 ± 21.6, respectively. These differences were not statistically significant. The N31S abnormality was found in all 14 samples taken from the H3N2 patients and in all of the 23 samples taken from in S-OIV patients. However, this abnormality was not found in any of the 30 samples taken from seasonal H1N1 patients. Amantadine was found to be equally effective in treating S-OIV patients as neuraminidase inhibitors. The genetic abnormality resulting in S31N amino acid conversion identified in some of the H3N2 and S-OIV patients is thought to alter the function of M2 protein only mildly.

GEOMIT is the CAD/MCNP conversion interface code. The old version of GEOMIT had a limited capability from CAD model handling point of view. It is developed to automatically generate Monte Carlo geometrical data from CAD data due to the difference in the representation scheme. GEOMIT is capable of importing different CAD format as well as exporting different CAD format. GEOMIT has a capability to produce solid cells as well as void cells without using complement operator. While loading the CAD shapes (Solids), each shape is assigning material number and density according to its color. Shape fixing process is been applied to cure the errors in the CAD data. Vertices location correctness is evaluated first, then a removal of free edges and removal of small faces processes. Binary Space Portioning (BSP) tree technique is used to automatically split complicated solids into simpler cells to avoid excessive complicated cells for MCNP to run faster. MCNP surfaces are subjected to an automatic reduction before creating the model. CAD data of International Thermonuclear Experimental Reactor (ITER) benchmark model has been converted successfully to MCNP geometrical input. The first wall heat loading calculations agree very well with other countries results.Copyright © 2009 by ASME

Results of safety analyses and plant design conducted in the ITER Conceptual Design Phase are briefly reported. Safety analyses concluded that radiological doses from operational effluents and accidents are consistent with anticipated regulatory requirements. To improve safety further and meet the ambitious goal of “passive safety”, efforts are needed to reduce inventories of tritium and activation products. In the ITER plant design, no crucial problem is identified. The efforts in the Engineering Design Phase should be done for improving cost efficiency of the plant not jeopardizing safety features.

A series of neutronics experiments with mockups relevant to solid breeder water cooled blanket at FNS/JAEA have been analyzed with MCNP-4C, FENDL-2.1 and JENDL-3.3. Divergences of calculation results to experimental ones (C/Es) on tritium production rates (TPRs) from unity increase in the experiment with a neutron reflector made of SS316. These also increase at boundary between the breeder and beryllium layers. This suggests that there are some problems in back-scattering cross section data of nuclei included in SS316 and beryllium. The divergence can be improved by installation of water at this boundary. The C/Es on integrated TPRs are within 7% in experiments without a reflector.

In a complex and huge system as in ITER fusion reactor, the creation of the geometrical input data of Monte Carlo (MC) codes such as MCNP is a highly exhausting task. Accordingly, it is a general approach to shift the geometric modeling into a computer aided design (CAD) system and to use an interface, which performs the exchange of CAD data into a representation appropriate for MC code. We have developed efficient algorithms and computer code, which are used to convert Parasolid format CAD files including solid and void data into MCNP input data. The CAD-MCNP conversion processes include creating surface equations; determining surface senses; constructing cell geometry and creating MCNP input file. This paper describes the basic algorithms used for the CAD/MCNP interface along with some applications for different geometries.Copyright © 2006 by ASME

The Fusion Experimental Reactor (FER) is a D-T-burning tokamak machine currently being designed. It is expected to provide physical information and technical experiences that will be sufficient to proceed towards the DEMO Fusion Reactor which will demonstrate electric power generation by fusion energy. An efficient ash exhaust, a hybrid current drive operation, the use of a 3% ripple field, the technological achievements in R&D of the magnets, and the negative-ion beam system are expected to allow the FER to achieve its cost-effectiveness. >

(1979). Applicability of Two-Dimensional Sensitivity Calculation Code: SENSETWO. Journal of Nuclear Science and Technology: Vol. 16, No. 7, pp. 530-533.

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...

The effect of coupling between vertical and radial plasma motions, originating from the asymmetric features of plasma configuration, provides new and interesting problems. A plasma column moves vertically even when it is disturbed radially by some kinds of perturbations. The tokamak with single-null poloidal divertor configuration and up—down asymmetry, has two types of coupling effects: one is due to the asymmetry of external equilibrium field and the other is due to the asymmetry of eddy currents induced in structures surrounding the plasma, such as first wall and blanket. Stability analysis and simulations on plasma position control have been carried out for the tokamak reactor with up—down asymmetric configuration peculiar to the single-null poloidal divertor. The results indicate that the effect of the asymmetry of the equilibrium field is governing; the up—down asymmetry enhances a growth rate of plasma vertical instability compared to a symmetric case; and the stability criteria for the asymmetric system should be modified. Large vertical displacements toward the divertor plates are observed in the simulations on plasma position control when large disturbances are applied to the plasma, e.g. 40% reduction of poloidal beta, βp, and normalized internal plasma inductance, li. It is quite difficult to suppress the vertical movement with a practical control power level, although the radial movement could be controlled. Plasma with the single-null divertor would therefore interact strongly with the divertor plates, or the first wall near the divertor plates, and would be disrupted at some level of radial perturbations, while for such perturbations a plasma with up—down symmetry (e.g. double-null divertor configuration) could attain equilibrium and be shut down in a controlled manner.

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.

Heat flux distribution on the first wall of a fusion reactor due to the thermal radiation from high temperature protection wall placed in front of the first wall was analyzed. With necessary modifications, a three-dimensional Monte Carlo transport code developed for neuronics calculation was successfully applied in the analysis. That is, reasonable results with sufficiently small statistical error were obtained with reasonable computational time. The heat flux distribution was found to be insensitive to the reflection characteristic of the radiation at the first wall i. e. diffusive or specular.

Conceptual design studies of the Fusion Experimental Reactor (FER) have been performed. The FER has an objective of achieving selfignition and demonstrating engineering feasibility as a next generation tokamak to JT-60. Various concepts of the FER have been considered. The reference design is based on a double-null divertor. Optional design studies with some attractive features based on advanced concepts such as pumped limiter and RF current drive have been carried out. Key design parameters are; fusion power of 440 MW, average neutron wall loading of 1MW/m/sup 2/, major radius of 5.5m, plasma minor radius of 1.1m, plasma elongation of 1.5, plasma current of 5.3MA, toroidal beta of 4%, toroidal field on plasma axis of 5.7T and tritium breeding ratio of above unity.

This paper describes the results of an evaluation on the safety concerns of the fuel-gas purification system (FPS), the fuel-gas isotope separation system (ISS), and the fuel-gas storage system (FSS) for the Fusion Experimental Reactor (FER). The results of this evaluation are presented as a probability-consequence plot. The primary function of the FPS is to remove impurities from the vacuum pumping system in the fuel stream. In this system, a palladium diffuser removes all of the impurities, and then the purified hydrogen isotopes are transferred to the ISS. The ISS separates the stream of mixed hydrogen isotopes into three high purity streams of D2, T2, and DT. The ISS makes the required separation by means of four interlinked cryogenic distillation columns. The FSS stores the purified fuel gas by means of metal bed absorbers. This evaluation was performed by using Probabilistic Risk Assessment (PRA). The first step of this assessment is to determine the accident initiators, which was to perform a failure mode and effects analysis (FMEA). The second step is to develop an event tree for each of the important categories of accident initiators identified; event tree analysis (ETA). The third step is fault tree analysis (FTA). Fault trees are required to determine the branching probabilities in the event trees. The fourth step is the determination of the release magnitudes. The fifth step is to assign the consequences to the accident sequences and to derive probability-consequence curves for risk comparisons.

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.

The poloidal distribution of the first wall 14 MeV neutron flux and the tritium breeding ratio in a Tokamak fusion reactor were calculated using Monte Carlo method. The poloidal distribution of the 14 MeV neutron flux in the first wall was found to be quite different from that of the primary incident flux. The tritium breeding ratio calculated by the Monte Carlo method became about 5% larger than the value obtained from SN transport calculations.