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JAEA Reports

Biosphere assessment methodology commonly applicable to various disposal concepts

Kato, Tomoko; Fukaya, Yukiko*; Sugiyama, Takeshi*; Nakai, Kunihiro*; Oda, Chie; Oi, Takao

JAEA-Data/Code 2019-002, 162 Pages, 2019/03

JAEA-Data-Code-2019-002.pdf:2.78MB

The radioactive waste generated from Fukushima Daiichi nuclear power station (FDNPS) accident have features such as wide range of radioactivity level (from low to high) and huge amount etc. It would be necessary for the waste from the FDNPS accident to develop suitable disposal concept and to be disposed safely and reasonably. When considering such appropriate disposal concepts in site-generic phase, it is necessary to appropriately develop models and parameters depending on the disposal concepts, such as disposal depth and specification of engineered barrier. In addition, it is desirable to evaluate the safety of repository with common models and parameters independent on the disposal concepts. In the safety assessment of disposal, it is useful to show the difference in performance of repository with "dose" as an indicator of safety assessment. Biosphere model and parameter set and flux-to-dose conversion factors calculated using them are originally dependent on the disposal concepts. However, the biosphere models and the parameter set in safety assessment of near-surface disposal, sub-surface disposal and geological disposal are prepared in each case, and are different according to the age and purpose of the discussion. In this study, an example of biosphere model and parameter-set of groundwater sceinario commonly applicable to various disposal concepts were shown, to calculate flux-to-dose conversion factors, as common indicators independent to disposal concept. And, a set of flux-to-dose conversion factors was also calculated by using the commonly available biosphere model and parameter set. By applying the flux-to-dose conversion factors, it is possible to compare the performance of disposal concepts to the waste generated from FDNPS accident, focusing on the parts depending on the disposal concepts.

Journal Articles

Status of studies on HLW glass performance for confirming its validity in assessment

Inagaki, Yaohiro*; Mitsui, Seiichiro*; Makino, Hitoshi*; Ishiguro, Katsuhiko*; Kamei, Gento*; Kawamura, Kazuhiro*; Maeda, Toshikatsu; Ueno, Kenichi*; Bamba, Tsunetaka*; Yui, Mikazu*

Genshiryoku Bakkuendo Kenkyu, 10(1-2), p.69 - 84, 2004/03

https://doi.org/10.3327/jnuce.10.69

no abstracts in English

Journal Articles

Validation and verification of accident consequence assessment models

Homma, Toshimitsu; Takahashi, Tomoyuki*; Togawa, Orihiko

Hoken Butsuri, 36(4), p.308 - 313, 2001/12

https://doi.org/10.5453/jhps.36.308

no abstracts in English

JAEA Reports

Reliability evaluation of simulation models for nearfield groundwater flow and radionuclide transport computation

*; *; *; *

JNC TJ8400 2000-005, 71 Pages, 2000/05

JNC-TJ8400-2000-005.pdf:4.0MB

In this research, simulations with some parameters which characterize ground water flow and the reliability evaluation for the expansion of the calculation method of groundwater flow were carried out by using the radionuclide transport computations in nearfield heterogeneous porous media. Concretely contents are follows: (1)With the series of calculation method for three-dimensional saturated/unsaturated groundwater flow and one-dimensional radionuclide transport, the computational analyses with the parameters used in JNC report in 2000 was carried out and the influence of the different input flux was evaluated. (2)The examination of the application for the different ways of inverse laplace transformation which is used in one-dimensional radionuclide transport analysis code "MATRICS" was carried out. (3)The examination of the application of multi-element "MATRICS" (m-MATRICS) for radionuclide transport computations in nearfield heterogeneous porous media was carried out. (4)The series of calculation methods from three-dimensional saturated/unsaturated ground water flow simulation code to one-dimensional radionuclide transport simulation code was integrated.

JAEA Reports

lrradiation behavior and performance model of nitride fuel

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JNC TN9400 2000-041, 29 Pages, 2000/03

JNC-TN9400-2000-041.pdf:1.18MB

Irradiation behavior and performance models were investigated in order to apply for nitride fuel options in feasibility study on fast breeder reactor and related recycle systems. (1)MechanicaI design of nitride fuel pin: The behaviors of fission gas release (increase of internal Pressure) and fuel-to-cladding chemical interaction (decrease of cladding thickness) are needed to evaluate cumulative damage fraction in case of fuel pin mechanical design. The behaviors of fission gas release and fuel-to-cladding chemical interaction were investigated from the past studies up to high burnuP, since the lower fission gas release in nitride fuel than in oxide fuel could contribute to reduce the plenum volume and result in the shortening of fuel Pin length. (2)Fuel pin smear density: The higher fuel smear density is preferred for the higher fissile density to improve the core characteristic. The behaviors of fuel pellet swelling were investigated from the past studies up to higher burnup, since the larger fuel pellet swelling in nitride fuel than in oxide fuel would restrict high burunp capability due to fuel-cladding mechanical interaction. (3)Compatibility of nitride fuel with high Temperature water: Compatibility of nitride fuel with high temperature water were investigated from the past studies to contribute water cooled fast breeder reactor options.

JAEA Reports

Investigations on the evaluation methods of the irradiation performance of FBR metallic fuel for the design study

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JNC TN9400 2000-031, 15 Pages, 2000/03

JNC-TN9400-2000-031.pdf:0.53MB

For the irradiation performance of metallic fuel, many of the analyses were conducted in USA using EBR-l and EBR-II. ln this study, based on the published data and papers on the above results, the appropriate methods to the evaluation of the irradiation performance of FBR metallic fuel for the design study were considered, as the fbasibility study for FBR. The followings are the targets in this work; (1)deformation of cladding (2)deformation of fuel slug (3)FP gas release (4)fluctuation of the bonding Na level in the fuel pin (5)FCCI

JAEA Reports

Distribution of actinide elements among various environmental factors (II)

Kudo, Akira*; Fujikawa, Yoko*

JNC TJ8400 2000-010, 67 Pages, 2000/02

JNC-TJ8400-2000-010.pdf:2.17MB

The first and second environmental releases of man-made $^{239+240}$ Pu came from nuclear explosions at Alamagordo and Nagasaki in 1945. The release at Nagasaki was more serious than at Alamagordo, because it happened in an area with a high population density. Unfissioned $^{239+240}$ Pu and various fission products have been interacting here with various environmental materials (soils, sediments, and plants) under wet and temperature conditions for more than 45 years. To assess the environmental mobility of $^{239+240}$ Pu, the distributions of radionuclides from this release were investigated at Nishiyama where heavy black rain containing unfissioned plutonium and fission products fell 30 minutes after the nuclear explosion. The vertical distributions of $^{137}$ Cs and $^{239+240}$ Pu were determined in unsaturated soil cores up to 450cm deep. Most radionuclides were found in the soil column 30cm from the ground surface. However, $^{239+240}$ Pu were detected in the groundwater as well below a depth of 200cm. No $^{137}$ Cs was found below 40cm from the ground surface or in groundwater. These observations reveal that about 3% of the total $^{239+240}$ Pu have been migrating in the soil at a faster rate than the remaining $^{239+240}$ Pu. Sharp peak of $^{137}$ Cs and $^{239+240}$ Pu, indicating heavy deposition from the Nagasaki local fallout of 1945, were found in sediment cores collected from the Nishiyama reservoir. $^{239+240}$ Pu peaks were unexpectedly discovered in pre-1945 sediment core sections. NO $^{137}$ Cs was found. By contrast to the distribution in sediment cores, $^{137}$ Cs in tree rings had spread by diffusion from the bark to the center of the tree without holding a fallout deposition record. Most of the $^{239+240}$ Pu was distributed in the tree rings following a similar deposition record to that found in sediment cores. Furthermore, a very small amount of $^{239+240}$ Pu (about 1%) was found unexpectedly in pre-1945 tree rings. The only reasonable ...