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Journal Articles

Study on plutonium burner high temperature gas-cooled reactor in Japan; Introduction scenario, reactor safety and fabrication tests of the 3S-TRISO fuel

Ueta, Shohei; Mizuta, Naoki; Fukaya, Yuji; Goto, Minoru; Tachibana, Yukio; Honda, Masaki*; Saiki, Yohei*; Takahashi, Masashi*; Ohira, Koichi*; Nakano, Masaaki*; et al.

Nuclear Engineering and Design, 357, p.110419_1 - 110419_10, 2020/02

 Times Cited Count:1 Percentile:24.17(Nuclear Science & Technology)

The concept of a plutonium (Pu) burner HTGR is proposed to incarnate highly-effective Pu utilization by its inherent safety features. The security and safety fuel (3S-TRISO fuel) employs the coated fuel particle with a fuel kernel made of plutonium dioxide (PuO$$_{2}$$) and yttria stabilized zirconia (YSZ) as an inert matrix. This paper presents feasibility study of Pu burner HTGR and R&D on the 3S-TRISO fuel.

Journal Articles

Study on Pu-burner high temperature gas-cooled reactor in Japan; Test and characterization for ZrC coating

Ueta, Shohei; Aihara, Jun; Mizuta, Naoki; Goto, Minoru; Fukaya, Yuji; Tachibana, Yukio; Okamoto, Koji*

Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 7 Pages, 2018/10

The security and safety fuel (3S-TRISO fuel) employs the coated fuel particle with a fuel kernel made of plutonium dioxide (PuO$$_{2}$$) and yttria stabilized zirconia (YSZ) as an inert matrix. Especially, a zirconium carbide (ZrC) coating is one of key technologies of the 3S-TRISO, which performs as an oxygen getter to reduce the fuel failure due to internal pressure during the irradiation. R&Ds on ZrC coating directly on the dummy CeO$$_{2}$$-YSZ kernel have been carried in the Japanese fiscal year 2017. As results of ZrC coating tests by the bromide chemical vapor deposition process, stoichiometric ZrC coatings with 3 - 18 microns of thicknesses were obtained with 0.1 kg of particle loading weight.

Journal Articles

Development plan of high burnup fuel for high temperature gas-cooled reactors in future

Aihara, Jun; Ueta, Shohei; Honda, Masaki*; Blynskiy, P.*; Gizatulin, S.*; Sakaba, Nariaki; Tachibana, Yukio

Journal of Nuclear Science and Technology, 51(11-12), p.1355 - 1363, 2014/11

 Times Cited Count:8 Percentile:34.41(Nuclear Science & Technology)

Plan and status of research and development (R&D) were described on coated fuel particle (CFP) and fuel compacts for core of small sized high temperature gas-cooled reactor (HTGR) HTR50S at 2nd step of phase I (second core of HTR50S). Specifications of existing CFPs for high burnup (HTR50S2-type-CFPs) were adopted as specifications of CFPs, to reduce the R&D. HTR50S2-type-CFPs were fabricated based on technology developed in High Temperature Engineering Test Reactor (HTTR) project. First irradiation test of HTR50S2-type-CFPs is now being carried out. In addition, R&D for fuel compact with high packing fraction is needed, because volume fraction of fuel kernel to whole of HTR50S2-type-CFP is rather smaller than that of the HTTR-type-CFP. In addition, we describe outline of R&D plans for core of HTR50S in phase II and naturally safe HTGR.

JAEA Reports

Development of a framework for the neutronics analysis system for next generation, 3

Yokoyama, Kenji; Hirai, Yasushi*; Hyodo, Hideaki*; Tatsumi, Masahiro*

JAEA-Data/Code 2009-012, 208 Pages, 2010/02

JAEA-Data-Code-2009-012.pdf:11.28MB

Japan Atomic Energy Agency promotes development of innovative analysis methods and models in fundamental studies for next-generation nuclear reactor systems. In order to efficiently and effectively reflect the latest analysis methods and models to primary design of prototype reactor and/or in-core fuel management for power reactors, a next-generation analysis system MARBLE has been developed. In the present study, we examined in detail the existing design and implementation of ZPPR critical experiment analysis database followed by unification of models within the framework of the next-generation analysis system by extending to various critical experiment analysis. Furthermore, we examined requirements for functions of analysis results correction which is indispensable for critical analysis system, and designed and implemented an analysis system for various critical experiments including ZPPR.

JAEA Reports

Development of burnup analysis system for fast reactor, 3 (Contract research)

Hirai, Yasushi*; Hyodo, Hideaki*; Tatsumi, Masahiro*; Yokoyama, Kenji

JAEA-Data/Code 2008-021, 110 Pages, 2008/10

JAEA-Data-Code-2008-021.pdf:3.47MB

Development of burnup analysis system for fast reactors with modularity and flexibility is being done that would contribute to actual core design work and improvement of prediction accuracy. In the previous study on "Development of Burnup Analysis System for Fast Reactors (2)" in FY2006, design and implementation of models for detailed geometry of assembly, fuel loading pattern and so on, accompanied with specification and implementation of input file handling to construct data models. In this study, a prototype system has been implemented in which functionalities are embedded for calculation of macroscopic cross section, core calculation and burnup calculation applying the fruits of the study "Development of a Framework for the Neutronics Analysis System for Next Generation (2)". It also implements a fuel reloading/shuffling function controlled with simple description in user input for multi-cycle burnup analysis.

JAEA Reports

Development of a framework for the neutronics analysis system for next generation, 2 (Contract research)

Hirai, Yasushi*; Hyodo, Hideaki*; Tatsumi, Masahiro*; Jin, Tomoyuki*; Yokoyama, Kenji

JAEA-Data/Code 2008-020, 188 Pages, 2008/10

JAEA-Data-Code-2008-020.pdf:15.06MB

Japan Atomic Energy Agency promotes development of innovative analysis methods and models in fundamental studies for next-generation nuclear reactor systems. In order to efficiently and effectively reflect the latest analysis methods and models to primary design of prototype reactor and/or in-core fuel management for power reactors, a next-generation analysis system MARBLE has been developed. In this study, detailed design of a framework, its implementation and tests are conducted so that a Python system layer can drive calculation codes written in C++ and/or Fortran. It is confirmed that various type of calculation codes such as diffusion, transport and burnup codes can be treated in the same manner on the platform for unified management system for calculation codes with a data exchange mechanism for abstracted data model between the Python and the calculation code layers.

Journal Articles

MARBLE; A Next generation neutronics analysis code system for fast reactors

Yokoyama, Kenji; Hirai, Yasushi*; Tatsumi, Masahiro*; Hyodo, Hideaki*; Chiba, Go; Hazama, Taira; Nagaya, Yasunobu; Ishikawa, Makoto

Proceedings of International Conference on the Physics of Reactors, Nuclear Power; A Sustainable Resource (PHYSOR 2008) (CD-ROM), 8 Pages, 2008/09

A development project of the next generation neutronics analysis code system, MARBLE, has been launched in JAEA. A software platform and common data models for fast reactor neutronics analysis were developed to realize the new system. At present, a fast reactor burnup calculation system, ORPHEUS, has been implemented in the MARBLE system. The new system reproduced benchmark results by the conventional code system and it reduced input data preparation works with the help of the capabilities supported by common data model packages. The new system was validated in an analysis of a burnup reactivity coefficient measured in the experimental fast reactor JOYO. These results show that MARBLE/ORPHEUS can be adopted as a new standard neutronics analysis system for fast reactors.

JAEA Reports

Systemization of burnup sensitivity analysis code, 2 (Contract research)

Tatsumi, Masahiro*; Hyodo, Hideaki*

JAEA-Review 2008-038, 95 Pages, 2008/08

JAEA-Review-2008-038.pdf:2.09MB

The analysis of burnup characteristics is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, development of an analysis code for burnup sensitivity, SAGEP-BURN, has been done and confirmed its effectiveness. However, there is a problem that analysis sequences become inefficient because of a big burden to user due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functionalities in the existing large system. Therefore, systemization of the burnup sensitivity analysis code has been done with an object-oriented scripting language. (NB: This document is a translation of JNC TJ9410 2004-002 published in February 2005.)

JAEA Reports

Development of a framework for the neutronics analysis system for next generation (Contract research)

Tatsumi, Masahiro*; Yokoyama, Kenji

JAEA-Data/Code 2007-020, 106 Pages, 2007/11

JAEA-Data-Code-2007-020.pdf:17.99MB

JAEA has been promoting the development of innovative analysis methods and models for next-generation nuclear reactor systems; advanced analysis system has been developing to apply the object-oriented approach in order to reflect such the latest methods and models to basic designs and operations of reactors in the efficient and effective way. The developing system adopt the two-layer model which consists of a control layer written in the Python and a solver layer in the C++. The principle on the two-layer model was examined followed by the design and implementation of a library that enabled transparent transfer of data models between the two layers. In each layer, appropriate numerical library was used for better performance. In the present library, a model proxy was implemented to exchange internal data that is represented in different ways in each layer. With this mechanism, it confirmed that data exchange between the layers can be performed easily and effectively.

JAEA Reports

Development of burnup analysis system for fast reactors, 2 (Contract research)

Hirai, Yasushi*; Hyodo, Hideaki*; Tatsumi, Masahiro*

JAEA-Data/Code 2007-019, 133 Pages, 2007/11

JAEA-Data-Code-2007-019.pdf:16.41MB

There is a problem that analysis work tends to become inefficient for lack of functionality suitable for analysis of composition change due to burnup since the conventional analysis system is targeted to critical assembly systems. Therefore development of burnup analysis system for fast reactors with modularity and flexibility can contribute actual core design work and improvement of prediction accuracy. In the previous study on "Development of Burnup Analysis System (for Fast Reactors)" in FY2005, basic design was conducted to define each component in the system(input, solver, edit) and how to drive them. In this study, detailed design of the system and implementation of the I/O component were conducted according to the results in the basic design followed by proto-typing implementation.

JAEA Reports

Development of burnup analysis system for fast reactor (Contract research)

Hyodo, Hideaki*; Tatsumi, Masahiro*

JAEA-Data/Code 2006-018, 120 Pages, 2006/08

JAEA-Data-Code-2006-018.pdf:8.42MB

In order to utilize the measured burnup data for improvement on accuracy in reactor core design, it is important to minimize the methodological errors to retrieve physical meanings from experimental data. The system for neutronics analyses that has been developed as the JUPITER standard analysis method assumes geometry of critical assembly, thus the system has not been maintained in functionalities for analysis of composition change of fuel materials. Therefore, there is a potential restriction for the purpose of detail analysis due to extreme inefficiency that comes from variety of limitation on its functionalities. It is not sufficient to follow a predefined analysis sequence in burnup analysis for reactor core; it is also needed to change analysis sequence to examine modeling error in analysis or to retrieve calculated values in an intermediate computation step for interpretation of physical meanings. Therefore it is not complete with a simple join of each function; it is needed to develop a new system for burnup analysis of reactor cores with flexibility on composition and decomposition of analysis components such as cell and core calculations. In this work, necessary conditions are examined for a new burnup analysis system targeted to actual reactor cores from the results of a research on the current working set in burnup analysis. With results in the research, a set of conceptual and fundamental designs were done.

JAEA Reports

Systemization of burnup sensitivity analysis code (II)

Tatsumi, Masahiro*; Hyodo, Hideaki*

JNC-TJ9410 2004-002, 98 Pages, 2005/02

JNC-TJ9410-2004-002.pdf:2.14MB

A distinct improvement on accuracy in nuclear core design has been accomplished by development of adjusted nuclear library using the cross-section adjustment method, in which the results of critical experiments of JUPITER and so on are reflected. In the design of large LMFBR core, however, it is important to accurately estimate not only neutronics characteristics but also burnup characteristics. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core.The analysis of burnup characteristics is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, development of an analysis code for burnup sensitivity, SAGEP-BURN, has been done and confirmed its effectiveness. However, there is a problem that analysis sequences become inefficient because of a big burden to user due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functionalities in the existing large system It is not sufficient to unify each computational component for some reasons; computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion. For this purpose, systemization of the burnup sensitivity analysis code has been done with an object-oriented scripting language.In this study, an examination was conducted for the two-layer controlling model of the conventional system using Python, the object-oriented scripting language, With the result of the examination, a new analysis system for burnup sensitivity, PSAGEP (Python-wrapped SAGEP-burn), was implemented.

JAEA Reports

Systemization of Burnup Sensitivity Analysis Code

Tatsumi, Masahiro*; Hyodo, Hideaki*

JNC-TJ9400 2003-012, 109 Pages, 2004/02

JNC-TJ9400-2003-012.pdf:3.56MB

In this study, points that should be improved in the current SAGEP-BURN were clarified through analysis of a computational process. Then a prototype of new SAGEP-BURN system was designed and implemented in Python. It is confirmed that the new system gives the identical results with that by the conventional system. For further implementation, analysis and design based on an object-oriented approach have been done.

Journal Articles

Sorption characteristics of actinium and protactinium onto soils

Sakamoto, Yoshiaki; Ishii, Tomoaki*; Inagawa, Satoshi*; Gunji, Yasuyoshi*; Takebe, Shinichi; Ogawa, Hiromichi; Sasaki, Tomozo*

Journal of Nuclear Science and Technology, 39(Suppl.3), p.481 - 484, 2002/11

Sorption behavior of 227Ac and 233Pa onto several kinds of soils has been studied with a sequential extraction technique, for safety assessment of shallow land disposal of uranium bearing waste. After a batch sorption experiment, the sorbed form of 227Ac and 233Pa onto the soils was fractionated into ion exchange form (extraction by KCl+CaCl2), association with Fe+Mn oxides (extraction by NH2OH-HCl and oxalic acid), association with organic materials (extraction by H2O2) and fixation into soil (residual). From the results of the sequential extraction, major part of 227Ac sorbed onto the soils was found in the fraction of the ion exchange form and the fixation into the soils. On the other hand, major part of 233Pa was found in the fraction of the association with Fe+Mn oxides and the fixation into the soils. These results suggest that the sorption behavior of 227Ac and 233Pa is related to the irreversible sorption reaction onto the soils.

Journal Articles

Measurement of distribution coefficients of U series radionuclides on soils under shallow land environment, 2; pH dependence of distribution coefficients

Sakamoto, Yoshiaki; Ishii, Tomoaki*; Inagawa, Satoshi*; Gunji, Yasuyoshi*; Takebe, Shinichi; Ogawa, Hiromichi; Sasaki, Tomozo*

Genshiryoku Bakkuendo Kenkyu, 8(1), p.65 - 76, 2001/09

In order to study adsorption behavior of U series radionuclides(Pb, Ra, Th, Ac, Pa and U) in aerated zone environment (loam-rein water system) and aquifer environment(sand-groundwater system) for safety assessment of U bearing waste), pH dependence of distribution coefficients of each element have been obtained. The pH dependence of distribution coefficients of U, Ac, Th, Ra and Pb was analyzed by model calculation of adsorption behavior based on chemical forms of each elements and soil surface characteristics, which are a cation exchange capacity and surface charge. From model calculation of adsorption behavior, the distribution coefficients' values and adsorption behavior of Pb, Ra, Th, Ac and U could be showed by a combination of cation exchange and surface-complexation adsorption model.

Journal Articles

Measurement of distribution coefficients for uranium series radionuclides under shallow land environment condition, 1

Ishii, Tomoaki*; Inagawa, Satoshi*; Gunji, Yasuyoshi*; Sakamoto, Yoshiaki; Takebe, Shinichi; Ogawa, Hiromichi; Sasaki, Tomozo*

Genshiryoku Bakkuendo Kenkyu, 8(1), p.55 - 64, 2001/09

Distribution coefficients of Uranium series nuclide(Pb,Ra,Ac,Th,Pa and U) were obtained under aerated zone environment and aquifer environment, for the safety evaluation of shallow underground disposal of uranium bearing waste. The distribution coefficients of them on 4 kinds of soil such as the loam in the rain water as for aerated zone environment and on 3 kinds of soil and rock such as the sand in groundwater as for aquifer environment have been measured by batch method. The distribution coefficients in aerated zone environment were one or two orders in magnitude higher than that in aquifer environment, except Ac. And, there was approximately the linear correlation on the relationship between cation exchange capacity and specific surface area, which are representative physical property of the soil, and distribution coefficient of lead, radium and protactinium.

JAEA Reports

None

Honda, Masaki*; *; *; *; *

JNC-TJ8440 99-005, 244 Pages, 1999/03

JNC-TJ8440-99-005.pdf:21.82MB

no abstracts in English

JAEA Reports

None

*

PNC-TJ8005 97-001, 122 Pages, 1997/03

PNC-TJ8005-97-001.pdf:32.41MB

no abstracts in English

JAEA Reports

None

PNC-TJ242 77-01VOL2, 43 Pages, 1977/06

PNC-TJ242-77-01VOL2.pdf:0.63MB

no abstracts in English

JAEA Reports

Fast breeder prototype reactor Monju fuel assembly water flow test

PNC-TJ242 73-02T, 79 Pages, 1973/03

PNC-TJ242-73-02T.pdf:1.89MB

This paper describes the FBR fuel assembly flow test which PowerReactor & Nuclear Fuel Development Corporation(PNC)is nowdeveloping for"MONJU".An experiment was performed by loading fuel assembly into the waterloop and the total pressure drop in the fuel assembly and its componentpressure drop were measured.The tested fuel assemblies consist of 7 reactor core fuel assembliesand 1 blanket fuel assembly. These fuel assemblies were manufacturedby 5 companies as the first preproduction for"MONJU". Reactor corefuel assemblies were classified into 5 grid type spacer fuel assembliesand 2 wire wrapped spacer fuel assemblies.The water temperature was 60-C,but one of the total fuel assemblypressure drop tests was done at 30,40,60-C,and in these tests theeffect of water temperature on the pressure drop was ascertained.The Reynolds number of the investigation at the pin bundle ranged be-tween 2.0x104 and 5.0x104 in the reactor core fuel assembly testsand between 5x103 and 2x104 in the blanket fuel

28 (Records 1-20 displayed on this page)