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Journal Articles

Analysis on the effect of forest decontamination on reducing the air dose rate using the three-dimensional transport code MCNP

Hemmi, Ko; Yamaguchi, Tetsuji; Takeda, Seiji; Kimura, Hideo

Genshiryoku Bakkuendo Kenkyu (CD-ROM), 24(1), p.3 - 14, 2017/06

Conditions of contaminated sources and ranges of forest decontamination that significantly reduce the air dose rate in residential areas were investigated by means of a sensitivity analysis related to the decontamination of the forest contaminated by radiocesium deriving from the accident at Fukushima Daiichi Nuclear Power Station. The contaminated sources including $$^{134}$$Cs and $$^{137}$$Cs were assumed to be a layer of sedimented organic matter (the A$$_{0}$$ layer) and surface soils (the A$$_{1}$$ layer). The air dose rates were calculated using the three-dimensional Monte Carlo transport code MCNP. A slope number of the forest, angles, state of contaminant distribution, radiocesium content in the forest soils, decontamination ranges, distance from the forest boundary to an evaluation point, and height at the evaluation point were adopted as the parameters. The decontamination of a litter (A$$_{0}$$) layer within the distance of 20 m from the forest boundary was revealed to be effective in reducing the air dose rate when the source distribution was homogeneous. The air dose rates were significantly reduced by the decontamination of the A$$_{0}$$ layer within a distance of 40 m from the forest boundary on condition that the radiocesium content of the A$$_{0}$$ layer was larger than that of the A$$_{1}$$ layer and the source distribution was non-homogeneous, such as the forest areas beyond 20 m from the forest boundary, which were more heavily contaminated than those within 20 m.

Journal Articles

A Sensitivity analysis for construction of the seismic response analysis model of a nuclear reactor building by using a three-dimensional finite element model

Choi, B.; Nishida, Akemi; Nakajima, Norihiro

Kozo Kogaku Rombunshu, B, 63B, p.325 - 333, 2017/03

The Japan Atomic Energy Agency promotes research and development of three-dimensional vibration simulation technologies for nuclear facilities. In this paper, we report a seismic response analysis of the Tohoku Pacific Coast Earthquake using three-dimensional models of the High-Temperature Engineering Test Reactor (HTTR) building. We conducted a sensitivity study using input parameters with uncertainty. Furthermore, we examined the variation of the seismic response results against the input parameters.

Journal Articles

Sensitivity analysis on safety functions of engineered and natural barriers for fuel debris disposal

Shimada, Taro; Nishimura, Yuki; Takeda, Seiji

MRS Advances (Internet), 2(12), p.687 - 692, 2017/01

A disposal measure for fuel debris generated at the accident in the Fukushima Daiichi Nuclear Power Station has been studied so far. However, physical and chemical properties of the fuel debris have not yet investigated in reactor containment vessels. In order to investigate the safety function of barriers required for disposal of fuel debris, sensitivity analyses for radionuclide migration were carried out, considering with uncertainty of the properties. As a result, it is indicated that it was important for evaluation of fuel debris disposal to obtain the physical and chemical properties of $$^{14}$$C and $$^{129}$$I during release to groundwater, in addition to $$^{238}$$U.

JAEA Reports

Discussion of uncertainties associated with parameters of biosphere model for safety assessment of geologieal disposal through sensitivity analysis

Kato, Tomoko; ; *;

JNC-TN8400 2001-014, 212 Pages, 2001/03

JNC-TN8400-2001-014.pdf:8.21MB

Reference Biospheres are regarded as tools which can be used for making reasonable estimates of radiological impacts for the purposes of safety assessment of geological disposal. Moreover, those are available for reducing the uncertainties based on future human environments and lifestyles. On the other hand, it is recognised that the parameter values have some uncertainties derived from experimental or sampling errors. It is possible to estimate the impacts of these uncertainties throughout the model by sensitivity analysis. Thus for this study, to evaluate the impact of the variation of migration conditions and exposure pathways, we changed some of migration and exposure parameters in turn, which were used in the compartment model where the geosphere-biosphere interface is a river in a plain.

JAEA Reports

A Study of reactor monitoring method with neural network

Nabeshima, Kunihiko

JAERI 1342, 119 Pages, 2001/03

JAERI-1342.pdf:7.52MB

no abstracts in English

Journal Articles

Sensitivity analysis on the deposition of inhaled radioactive iodine and the effectiveness of iodine prophylaxis

Matsunaga, Takeshi; Kobayashi, Kensuke

Hoken Butsuri, 36(1), p.31 - 44, 2001/03

no abstracts in English

Journal Articles

Sensitivity analysis on the effectiveness of iodine prophylaxis to reduce thyroid gland exposure in nuclear emergency

Matsunaga, Takeshi; Kobayashi, Kensuke

Proceedings of 10th International Congress of the International Radiation Protection Association (IRPA-10) (CD-ROM), 10 Pages, 2000/05

no abstracts in English

JAEA Reports

None

; Numata, Kazuyuki*; ; *; Oigawa, Hiroyuki*

JNC-TY9400 2000-006, 162 Pages, 2000/04

JNC-TY9400-2000-006.pdf:4.57MB

no abstracts in English

JAEA Reports

Reliability evaluation for radionuclide transport analysis code MATRICS

*; Ijiri, Yuji*; *; *

JNC-TN8400 2000-021, 66 Pages, 2000/04

JNC-TN8400-2000-021.pdf:1.74MB

A reliability evaluation for radionuclide transport analysis code, MATRICS, used in radionuclide transport analysis in the natural barrier system PA in H12 report has been carried out. Sensitivity analysis to radionuclide transport parameter in MATRICS and analytical solution has been performed, and the results of each analysis have been compared. Additionally sensitivity analysis using Talbot Method, Crump method and Hosono method has been carried out, and the results of each inverse Laplace transform method has been compared. The conclusions obtained from the results of the evaluation are summarized as follows, (1)In case of the infinite matrix diffusion distance, an error among the results of each calculation is maximum about 0.4% in the range of Pe number from 1.0 to 100. And, an error among the results of each calculation is maximum about 5.5% in the range of transmissivity from 1.0$$times$$10$$^{-10}$$ to 1.0$$times$$10$$^{-5}$$(m$$^{2}$$/s). (2)In case of the finite matrix diffusion distance (0.03$$sim$$1.0(m)), an error among the results of each calculation is maximum about 0.7% in the range of Pe number from 1.0 to 100. And, an error among the results of each calculation is maximum about 2.4% in the range of transmissivity from 1.0$$times$$10$$^{-10}$$ to 1.0$$times$$10$$^{-5}$$(m$$^{2}$$/s). 3)By comparing Talbot method with other inverse Laplace transform method, Talbot method is confirmed to give similar results with other inverse Laplace transform method in the range of Pe number from 5.0$$times$$10$$^{-1}$$ to 2.0$$times$$10$$^{3}$$, and that of transmissivity below 1.0$$times$$10$$^{-7}$$(m$$^{2}$$/s). Therefore, it is concluded that the reliability of MATRICS are confirmed by conducting sensitivity analysis in the range of Pe number and transmissivity coefficient used in H12 report.

Journal Articles

Development of a standard data base for FBR core nuclear design, 12; Analysis of FCA X-1 experiments and consistency evaluation using cross-section adjustment

Yokoyama, Kenji*; Numata, Kazuyuki*; Ishikawa, Makoto*; Oigawa, Hiroyuki; Iijima, Susumu

JNC-TY9400 2000-006, 168 Pages, 2000/04

no abstracts in English

JAEA Reports

ComparaUve analyses on nuclear charaderistics of water-cooled breeder cores

; Sato, Wakaei*;

JNC-TN9400 2000-037, 87 Pages, 2000/03

JNC-TN9400-2000-037.pdf:3.48MB

ln order to compare the nuclear characteristics of water-cooled bleeder cores with that of LMFBR, MOX fuel cell models are established for boiling and non-boiling LWR, non-boiling HWR and sodium-cooled reactor. Frst, the comarison is made between the heterogeneous cell calculation results by SRAC and those by SLAROM. The results show some differences as for neutron energy spectrum, one-grouped cross section and conversion ratio due to the different grouped cross section library (both are based on JENDL-3.2, though) used for each code, however, the difference is acceptably small for grasping the basic characteristics of the above-mentioned cores. Second, using the SLAROM code, main core parameters such as mean neutron energy, ratio of fast neutron and $$eta$$-value, are analyzed. The comparison between the cores show that softened neutron spectrum by the scattering effect of hydrogen or heavy hydrogen increase the contribution of nuclear reaction (especially for neutron capture reaction rather than fission reaction) in lower energy region comparing with LMFBR. ln order to overcome the effect, tighter lattice than LMFBR is necessary for water-cooled cores to realize the breeding of fissile nuclides. Third, effects of Pu isotopic composition on the breeding ratio are evaluated using SRAC burnup calculation. From the results, it is confirmed that degraded Pu (larger ratio of Pu-240) show the larger breeding ratio. At last, sensitivity analyses are made for k-effective and main reaction ratios. As for k-effective, using a temporary covariance data of JENDL-3.2, uncertainty resulting from the cross sections' error is analyzed for a boiling LWR and a sodium-cooled reactor. The boiling LWR core shows larger sensitivity in lower energy region than the sodium-cooled reactor (especially for the energy region lower than 1kev), And, 18-group analysis that is considered acceptably good for LMFBR analysis, should not be enough for accurate sensitivity estimation of ...

JAEA Reports

Development of a standard database for FBR core nuclear design (XI); Analysis of the experimental fast reactor "JOYO" MK-I start up test and oparation data

; Numata, Kazuyuki*

JNC-TN9400 2000-036, 138 Pages, 2000/03

JNC-TN9400-2000-036.pdf:10.16MB

Japan Nuclear Cycle Development lnstitute (JNC) had developed the adjusted nuclear cross-section library in which the results of the JUPITER experiments were renected. Using this adjusted library, the distinct improvement of the accuracy in nuclear design of FBR cores had been achieved. As a recent research, JNC develops a database of other integral data in addition to the JUPITER experiments, aiming at further improvement for accuracy and reliability. ln this report, the authors describe the evaluation of the C/E values and the sensitivity analysis for the Experimental Fast Reactor "JOYO" MK-l core. The minimal criticality, sodium void reactivity worth, fuel assembly worth and burn-up coefficient were analyzed. The results of both the minimal criticality and the fuel assembly worth, which were calculated by the standard analytical method for JUPITER experiments, agreed well with the measured values. 0n the other hand, the results of the sodium void reactivity worth have a tendency to overestimate. As for the burn-up coefficient, it was seen that the C/E values had a dispersion among the operation cycles. The authors judged that further investigation for the estimation of the experimental error will increase the applicability of the integral data to the adjusted library. Furthermore, sensitivity analyses for the minimal criticality, sodium void reactivity worth and fuel assembly worth showed the characteristics of "JOYO" MK-l core in comparison with ZPPR-9 core of JUPITER experiments.

JAEA Reports

None

*; *

JNC-TJ7440 2000-004, 52 Pages, 2000/03

JNC-TJ7440-2000-004.pdf:12.62MB

no abstracts in English

JAEA Reports

An experimental study of sodium aerosol detection sensitivity by laser induced breakdown spectroscopy

;

JNC-TN9400 2000-020, 54 Pages, 1999/11

JNC-TN9400-2000-020.pdf:2.36MB

A Laser-induced breakdown spectroscopy Leak Detection technique (abbreviated LLD) of sodium is accomplished by plasmafying the sodium aerosol, and then selectively detecting the sodium specific optical spectrum. This method is potentially more reliable as a means of detecting of sodium small leakage. This report, describes test results of detection characteristics using sodium aerosol, carried out to verify the principle of LLD in addition to evaluating the response under various conditions. 0ur main objective is to examine the applicability of LLD for small sodium leakage. The main results are as follows; (1)We confirmed the principle of LLD, specifically detecting the sodium optical spectru.m. (2)The relation between LLD fluorescence intensity and sodium aerosol concentration is nearly proportional within a relatively Na concentration ranges 10$$^{-11}$$ $$sim$$ 10$$^{-8}$$ g/cm$$^{3}$$. (3)The LLD signal appeared insensitive to the effect of sampling gas flow rate, oxygen concentration, and humidity in the examined range. ln fact, a high S/N ratio is obtained for small sodium leakage, and the reliability of the leakage detection is high, because LLD showed sensitive to sodium concentration. From these results and others discussed in this report, LLD appears to be an applicable technique in small leakage detection both in terms of response and reliabilily in the leakage phase.

JAEA Reports

Radionuclide migration analysis in porous rock

Ijiri, Yuji; ; *; Watari, Shingo; K.E.Web*; *; *

JNC-TN8400 99-092, 91 Pages, 1999/11

JNC-TN8400-99-092.pdf:6.62MB

JNC has been developed the performance assessment approaches for both fractured rock and porous rock. An equivalent continuum model is incorporated for solving the radionuclide migration in porous rock, while a discrete fracture network model is incorporated for solving the radionuclide migration in fractured rock (see more detail in Sawada et al. [1999]). This report describes the methodology, the data and the results of the performance assessment of porous rock. From the results of radionuclide migration analyses that were based on the hydrogeological properties obtained from the Neogene sedimentaly rock at the Tono mine, it was found that the release rate of selenium-79 and cesium-135 are dominant in porous rock. The sensitivity analyses using one-dimensional porous model revealed that hydraulic conductivity has more influences on the results than porosity does. In addition, it was found that smaller distribution coefficients of sandstone yield higher release rate than mudstone and tuff, and smaller distribution coefficients of saline water conditions yield higher release rate than fresh water conditions. The radionuclide migration in Neogene sedimentaly rock, where flow in rock matrix as well as in fractures are significant, was evaluated by superposing the results of porous model and fracture model. Since fracture model tends to yield more conservative results than porous model, it is obvious that the performance of Neogene sedimentary rock can be conservatively assessed by fracture model alone. The nuclide migration analyses performed in this report were based on the hydrogeological properties obtained at the depth between 20 meters and 200 meters frrom the ground surface. Therefore, it should be noted that the release rate at the depth of a future repository in Neogene sedimentary rock, 500 m, will be smaller than that shown in this report due to peemeability decrease from 200 m to 500 m.

JAEA Reports

Scoping calculation of nuclides migration in engineering barrier system for effect of volume expansion due to overpack corrosion and intrusion of the buffer material

; ; Ishiguro, Katsuhiko; Nakajima, Kunihiko*;

JNC-TN8400 99-087, 41 Pages, 1999/11

JNC-TN8400-99-087.pdf:2.78MB

Corrosion of the carbon steel overpack leads to a volume expansion since the specific gravity of corrosion products is smaller than carbon steel. The buffer material is compressed due to the corrosive swelling, reducing its thickness and porosity. On the other hand, Buffer material may be extruded into fractures of the surrounding rock and this may lead to a deterioration of the planned functions of the buffer, including retardation of nuclides migration and colloid filtration. In this study, the sensitivity analyses for the effect of volume expansion and intrusion of the buffer material on nuclide migration in the engineering barrier system are carried out. The sensitivity analyses were performed on the decrease in the thickness of the buffer material in the radial direction caused by the corrosive swelling, and the change in the porosity and dry density of the buffer caused by both compaction due to corrosive swelling and intrusion of buffer material. As results, it was found the maximum release rates of relatively shorter half-life nuclides from the outside of the buffer material decreased for taking into account of a volume expansion due to overpack corrosion. On the other hand, the maximum release rates increased when the intrusion of buffer material was also taking into account. It was, however, the maximum release rates of longer half-life nuclides, such as Cs-137 and Np-237, were insensitive to the change of buffer material thickness, and porosity and dry density of buffer.

JAEA Reports

Super-Phenix Benchmark used for Comparison of PNC and CEA Calculation Methods,and of JENDL-3.2 and CARNAVAL IV Nuclear Data

Hunter

PNC-TN9410 98-015, 81 Pages, 1998/02

PNC-TN9410-98-015.pdf:3.14MB

The study was carried out within the framework of the PNC-CEA collaboration agreement. Data were provided, by CEA, for an experimental loading of a start-up core in Super-Phenix. This data was used at PNC to produce core flux snapshot calculations. CEA undertook a comparison of the PNC results with the equivalent calculations carried out by CEA, and also with experimental measurements from SPX. The resu1ts revealed a systematic radial flux tilt between the calculations and the reactor measurements, with the PNC tilts only $$sim$$30-401 of those from CEA. CEA carried out an analysis of the component causes of the radial tilt. It was concluded that a major cause of radia1 tilt differences between the PNC and CEA calculations lay in the nuclear datasets used: JENDL-3.2 and CARNAVAL IV. For the final stage of the study, PNC undertook a sensitivity analysis, to examine the detailed differences between the two sets of nuclear data. The PNC flux calculations modelled SPX in both 2D (RZ) and 3D (hex-Z) geometries, using the diffusion programs CITATION and MOSES. The sensitivity analysis of the differences between the JENDL-3.2 and CARNAVAL IV nuclear datasets used the SAGEP calculational route. Both datasets were condensed to a single, non-standard, set of energy group boundaries. There were some incompatibilities in the cross-section formats of the two datasets. The sensitivity analysis showed that a relatively small number of nuclear data items contributed the bulk of the radial tilt difference between calculations with JENDL-3.2 and with CARNAVAL IV. A direct comparison between JENDL-3.2 and CARNAVAL IV data revealed the following. The Nu values showed little difference (<5|%). The only large fission cross-section differences were at low energy (<30% otherwise, with <10% typical). Although down-scattering reactions showed some large fractional differences, absolute differences were negligible compared with in-group scattering; for in-group scattering fractional ...

JAEA Reports

None

Oyamada, Kiyoshi*; Ikeda, Takao*

PNC-TJ1281 98-006, 63 Pages, 1998/02

PNC-TJ1281-98-006.pdf:2.08MB

None

JAEA Reports

None

Oyamada, Kiyoshi*; Ikeda, Takao*

PNC-TJ1281 98-005, 400 Pages, 1998/02

PNC-TJ1281-98-005.pdf:10.68MB

None

JAEA Reports

None

Ikeda, Takao*; Yoshida, Hideji*

PNC-TJ1281 98-002, 123 Pages, 1998/02

PNC-TJ1281-98-002.pdf:5.58MB

None

59 (Records 1-20 displayed on this page)