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Sumita, Junya; Shibata, Taiju; Kikuchi, Takayuki; Ishihara, Masahiro; Iyoku, Tatsuo; Sawa, Kazuhiro; Fujimoto, Nozomu
JAEA-Data/Code 2007-001, 57 Pages, 2007/02
Visual inspection by a TV camera and material properties measurement by surveillance test on core support graphite structures are planned for the High Temperature Engineering Test Reactor (HTTR) to confirm their structural integrity and characteristics. The surveillance test is aimed to investigate the change of material and mechanical properties by aging effects such as fast neutron irradiation and oxidation. The obtained data will be used not only for evaluating the structural integrity of the core support graphite structure of the HTTR but also for design data to advanced Very High Temperature Reactor (VHTR) discussed at generation IV international forum. This report describes the material properties and installed position of surveillance specimens in the HTTR in order to carry out the surveillance test.
Nagai, Haruyasu; Kobayashi, Takuya; Tsuzuki, Katsunori; Kim, K.
JAEA-Data/Code 2007-002, 65 Pages, 2007/02
As a numerical simulation tool of the numerical simulation system SPEEDI-MP, which is applicable for various environmental studies, a model coupling program (model coupler) has been developed. It controls parallel calculations of several models and data exchanges among them to realize the dynamical coupling of the models. It is applicable for any models with three-dimensional structured grid system, which is used by most environmental and hydrodynamic models. A coupled model system for water circulation has been constructed with atmosphere, ocean, wave, hydrology, and land-surface models using the model coupler. Performance tests of the coupled model system for water circulation were also carried out for the flood event at Saudi Arabia in January 2005 and the storm surge case by the hurricane KATRINA in August 2005.
Okumura, Keisuke
JAEA-Data/Code 2007-003, 120 Pages, 2007/02
COREBN is an auxiliary code of the SRAC system for multi-dimensional core burn-up calculation based on the diffusion theory and interpolation of macroscopic cross-sections tabulated to local parameters such as burn-up degree, moderator temperature and so on. The macroscopic cross-sections are prepared by cell burn-up calculations with the collision probability method of SRAC. SRAC and COREBN have wide applicability for various types of cell and core geometries. They have been used mainly for the purpose of core burn-up management of research reactors in Japan Atomic Energy Agency. The report is a revision of the users manual for the latest version of COREBN served with the SRAC released in 2006.
Okumura, Keisuke; Kugo, Teruhiko; Kaneko, Kunio*; Tsuchihashi, Keiichiro*
JAEA-Data/Code 2007-004, 313 Pages, 2007/02
The SRAC is a code system applicable to neutronics analysis of a variety of reactor types. Since the publication of the second version of the users manual (JAERI-1302) in 1986 for the SRAC system, a number of additions and modifications to the functions and the library data have been made to establish a comprehensive neutronics code system. The current system includes major neutron data libraries (JENDL-3.3, JENDL-3.2, ENDF/B-VII, ENDF/B-VI.8, JEFF-3.1, JEF-2.2, etc.), and integrates five elementary codes for neutron transport and diffusion calculation; PIJ based on the collision probability method applicable to 16 kind of lattice models, SN transport codes ANISN(1D) and TWOTRAN(2D), diffusion codes TUD(1D) and CITATION(multi-D). The system also includes an auxiliary code COREBN for multi-dimensional core burn-up calculation.
Makino, Hitoshi; Kawamura, Makoto; Wakasugi, Keiichiro; Okubo, Hiroo*; Takase, Hiroyasu*
JAEA-Data/Code 2007-005, 67 Pages, 2007/02
In this study, based on an improvement method for treatment of FEP interaction a computer tool to support scenario analysis by specialists of performance assessment has been developed. Anticipated effects of this tool are to improve efficiency of complex and time consuming scenario analysis work and to reduce possibility of human errors in this work. This tool also enables to describe interactions among a vast number of FEP's and the related information as interaction matrix, and analysis those interactions from a variety of perspectives.
Asano, Yoshihiro*; Sakamoto, Yukio
JAEA-Data/Code 2007-006, 44 Pages, 2007/02
Buildup factors up to 40 mean free path in ranging from 0.015 MeV to 15 MeV photon energy were evaluated by using the Monte Carlo simulation code, EGS4 for two typical heavy concretes. One is iron-contained and the other is barium-contained heavy concretes. The parameters of Geometrical Progression approximation for buildup factors were also presented for simplified calculations such as using the point kernel method.
Chiba, Go
JAEA-Data/Code 2007-007, 19 Pages, 2007/03
When we evaluate uncertainties in nuclear parameters which are induced by uncertainties in neutron-nuclide reaction cross sections, covariance data for energy-averaged cross sections are necessary. ERRORJ is a processing code to transform cross section covariance given in the ENDF format into energy-averaged cross section covariance. This document describes the revision in the version 2.3 and how to use it.
Ito, Toshimichi; Otosaka, Shigeyoshi; Suzuki, Takashi; Tanaka, Takayuki; Tsuneyama, Teppei; Togawa, Orihiko
JAEA-Data/Code 2007-008, 41 Pages, 2007/03
The database for the Japan Sea parameters on marine environment and radionuclides (JASPER) is established by the Japan Atomic Energy Agency as one of the final products of the Japan Sea Expeditions (phase I) carried out covering the EEZs of Japan and Russian Federation. And now, the part for anthropogenic radionuclides in the JASPER database is opened to the public, prior to the release of succeeding parts including other radionuclides, chemical tracers and oceanographic parameters. In the present, 253 data records are stored in the database including 193 data for Sr and Cs, 163 data for Pu and 236 data for Pu obtained from seawater, seabed sediment and filtered particle with support data. By establishing the database, recent feature of the Japan Sea environment has been recorded using every possible parameter for us. We believe that this database might be a strong tool for the purposes of monitoring for contamination of the Japan Sea by anthropogenic radionuclides, study of material transport in the sea and development and validation of models for numerical simulations. Furthermore, it is being prepared that the database are linked to MARIS of IAEA-MEL in order to contribute the world-wide study and monitoring of anthropogenic radionuclides in marine environment.
Yoshida, Yasushi*; Kitamura, Akira
JAEA-Data/Code 2007-009, 15 Pages, 2007/03
Thermodynamic data base for compounds and complexes of actinides and fission products with auxiliary species specialized in modeling requirements for safety assessments of radioactive waste disposal systems are being developed by the TDB project of OECD/NEA. In the project, thermochemical data bases for compounds and complexes of Ni, Se, Zr and organic ligands have been published in 2005. The data base files of these data available for geochemical calculation codes have been established in the present study. The procedure for establishment and contents of data base files are described in this report. These data base files are prepared as the formats of major geochemical codes of PHREEQE, PHREEQC, EQ3/6 and Geochemist's Workbench. The thermodynamic data base for the evaluation of alteration behavior of engineered barrier in the TRU 2nd progress report has been already published by JNC. The abstract of this data base file is also shown in the appendix of this report.
Tochigi, Yoshikatsu; Shibata, Masahiro; Sato, Haruo; Kitamura, Akira
JAEA-Data/Code 2007-010, 14 Pages, 2007/03
The Diffusivity Database (DDB) System developed on early 2006 was upgraded to apply the data of effective diffusion coefficient of the nuclides in the rock matrix for the "H12: Project to Establish the Scientific and Technical Basis for HLW Disposal in Japan", and the data in the buffer materials from literature survey was newly added. Some functions of data search and selection were reformed to improve the level of convenience. This DDB system (work on MS-AccessTM) is released to the public through Web server managed by JAEA.
Tamai, Hidesada; Kureta, Masatoshi; Liu, W.; Sato, Takashi; Nakatsuka, Toru; Watanabe, Hironori; Onuki, Akira; Akimoto, Hajime
JAEA-Data/Code 2007-011, 126 Pages, 2007/03
Japan Atomic Energy Agency has been performing tight-lattice rod bundle thermal-hydraulic tests to realize essential technologies for the technological and engineering feasibility of super high burn-up water-cooled breeder reactor featured by a high breeding ratio and super high burn-up by reducing the core water volume in water-cooled reactor. The tests were performing to make clear the fundamental subjects related to the boiling transition (BT) (Subjects: BT criteria under a highly tight-lattice rod bundle, effects of gap-width between rods and of rod-bowing) using 37-rod bundles (Base case test section (1.3 mm gap-width), Two parameter effect test sections (Gap-width effect one (1.0 mm) and Rod-bowing one)). In the present report, we summarize the test results from the rod-bowing effect test section. The thermal-hydraulic characteristics were obtained for the critical power under the steady-state and transient conditions, the pressure drop and the wall heat transfer within a wide range of pressure, flow rate, etc. including normal operational conditions of the designed reactor. Then the rod-bowing effects were also investigated based on the comparison between the results using the base case test section and the rod-bowing effect one.
Ichihara, Akira; Iwamoto, Osamu; Chiba, Satoshi; Kunieda, Satoshi; Shibata, Keiichi
JAEA-Data/Code 2007-012, 86 Pages, 2007/03
A computer code, POD, was developed for neutron-induced nuclear data evaluations. This program is based on four theoretical models, (1) the optical model to calculate shape-elastic scattering and reaction cross sections, (2) the distorted wave Born approximation to calculate neutron inelastic scattering cross sections, (3)the preequilibrium model, and (4) the multi-step statistical model. With this program, cross sections can be calculated for reactions , , , , , , He), , , , , and in the neutron energy range above the resonance region to 20 MeV. The computational methods and input parameters are explained in this report, with sample inputs and outputs.
Nojiri, Naoki; Owada, Hiroyuki; Fujimoto, Nozomu
JAEA-Data/Code 2007-013, 93 Pages, 2007/06
For the future HTGR development and the management of the High Temperature engineering Test Reactor(HTTR), the HTTR operation data base is constructed. The data base consists of the sorted or evaluated data based on the measured values from the HTTR's operation, such as excess reactivity of the core, temperature at facilities of the core and the plant, impurities in coolant and so on. The data base also consists of some sub-databases which have objects related to the future HTGR development or the HTTR's operational management in order to manage the important operation data systematically on a long term. This paper describes examples of the HTTR common data base, the HTTR nuclear characteristics data base, the helium purity control data base and the other data base.
Saito, Yoshihiko; Ochs, M.*; Suyama, Tadahiro*; Kitamura, Akira; Shibata, Masahiro; Sasamoto, Hiroshi
JAEA-Data/Code 2007-014, 24 Pages, 2007/07
Japan Nuclear Cycle Development Institute (JNC) has developed the sorption database (JNC-SDB) for bentonite and rocks in order to estimating a retardation capacity of important radioactive elements on natural barrier and engineered barrier in the H12 report. The database includes distribution coefficient (K) of important radionuclides. And JNC collected the sorption data from 1998 to 2003. In this report, Japan Atomic Energy Agency (JAEA) widely collected the sorption data in order to extend and update the sorption database. The updated database includes the published data which are not registered in the sorption database. In this updated JNC-SDB, 3,205 sorption data for 23 elements, which are important for performance assessment were included. The frequency of K for some elements was clearly shown by addition of the sorption data.
Kunimaru, Takanori; Shibano, Kazunori; Kurikami, Hiroshi; Tomura, Goji; Hara, Minoru; Yamamoto, Hajime*
JAEA-Data/Code 2007-015, 113 Pages, 2007/11
In the Horonobe Underground Research Laboratory (URL) Project, ground water from boreholes, river water and precipitation have been preiodically analyzed for the environmental monitoring since the fiscal year 2001. This report shows the data set of water chemistry since the fiscal year 2001 to the fiscal year 2006.
Yamauchi, Michinori; Hori, Junichi*; Sato, Satoshi; Nishitani, Takeo; Konno, Chikara; Kawasaki, Hiromitsu*
JAEA-Data/Code 2007-016, 58 Pages, 2007/09
The ACT-XN is a revised version of the ACT4 code, which was developed in the Japan Atomic Energy Agency (JAEA) to calculate the transmutation, induced activity, decay heat, delayed -ray source etc. for fusion devices. The ACT4 code has not dealt with the sequential reactions of charged particles generated by primary neutron reactions. However, the reactions cannot be disregarded in the design employing low activation material, and the code was newly supplemented with the function to calculate the activation for sequential reactions and renamed the ACT-XN. The FISPACT data were adopted for (x,n) reaction cross sections, charged particles emission spectra and stopping powers. An application of the code to the analysis of FNS experiment for LiF activation confirmed that the function is enough reliable, and a notice was presented through the design calculation of the Demo-reactor with FLiBe that the dose rate may be enhanced by sequential reactions.
Sato, Shohei; Sakai, Tomohiro*; Okuno, Hiroshi
JAEA-Data/Code 2007-017, 40 Pages, 2007/08
OPT-TWO is a calculation code which calculates the optimum concentration distribution, i.e., the most conservative concentration distribution in the aspect of nuclear criticality safety, of MOX (mixed uranium and plutonium oxide) fuels in the two-dimensional system. To achieve the optimum concentration distribution, we apply the principle of flattened fuel importance distribution with which the fuel system has the highest reactivity. Based on this principle, OPT-TWO takes the following 3 calculation steps iteratively to achieve the optimum concentration distribution with flattened fuel importance: (1) the forward and adjoint neutron fluxes, and the neutron multiplication factor, with TWOTRAN code which is a two-dimensional neutron transport code based on the SN method, (2) the fuel importance, and (3) the quantity of the transferring fuel. In OPT-TWO, the components of MOX fuel are MOX powder, uranium dioxide powder and additive. This report describes the content of the calculation, the computational method, and the installation method of the OPT-TWO, and also describes the application method of the criticality calculation of OPT-TWO.
Pham, N. S.*; Katakura, Junichi
JAEA-Data/Code 2007-018, 12 Pages, 2007/10
The precise knowledge of decay heat is one of the most important factors in safety design and operation of nuclear power facilities. Furthermore, decay heat data also play an important role in design of fuel discharges, fuel storage and transport flasks, and in spent fuel management and processing. In this study, a new application program, called DHP (Decay Heat Power program), has been developed for exact decay heat summation calculations, uncertainty analysis, and for determination of the individual contribution of each fission product. The analytical methods were applied in the program without any simplification or approximation, in which all of linear and non-linear decay chains, and 12 decay modes, including ground state and meta-stable states, are automatically identified, and processed by using a decay data library and a fission yield data file, both in ENDF/B-VI format. The window interface of the program is designed with optional properties which is very easy for users to run the code.
Hirai, Yasushi*; Hyodo, Hideaki*; Tatsumi, Masahiro*
JAEA-Data/Code 2007-019, 133 Pages, 2007/11
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.
Tatsumi, Masahiro*; Yokoyama, Kenji
JAEA-Data/Code 2007-020, 106 Pages, 2007/11
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.
Endo, Akira; Eckerman, K. F.*
JAEA-Data/Code 2007-021, 28 Pages, 2007/11
Nuclear decay data used for dose calculations have been compiled for 214 radionuclides with half-lives of less than 10 minutes. The decay data were assembled from decay data sets of the Evaluated Nuclear Structure Data File (ENSDF), the latest version as of May, 2007. Basic nuclear properties in the ENSDF that are particularly important for calculating the energies and intensities of radiations were examined and updated by referring to NUBASE2003/AME2003, the database for nuclear and decay properties of nuclides. In addition, modification of the incomplete ENSDF was done to determine their format errors, level schemes, normalization records, and so on. The energies and intensities of emitted radiations by the nuclear decay and the subsequent atomic process were computed from the ENSDF using the computer code EDISTR04. The compiled data are presented to enhance the nuclear decay database DECDC2, which was previously developed by the authors. The data will be used for dose calculations in the safety analysis for induced radionuclides in accelerator facilities and in treatment planning for the medical use of short-lived nuclides.
Hiraga, Naoto; Ishii, Eiichi
JAEA-Data/Code 2007-022, 100 Pages, 2008/02
Geological Isolation Research and Development Directorate, Horonobe Underground Research Unit conducted following three analysis for the 1st phase of Horonobe Underground Research Laboratory Project. (1) Mineral composition analysis of core sample, (2) Whole rock chemical composition analysis of core sample, (3) Surface gas composition analysis. This document summarizes the results of these analysis. Keywords: Horonobe Underground Research Laboratory Project, Mineral composition, Chemical and isotopic composition, Surface gas composition.
Yokoyama, Kenji; Numata, Kazuyuki*
JAEA-Data/Code 2007-023, 39 Pages, 2008/01
A new cross section adjustment and nuclear design accuracy evaluation solver was developed as a set of modules for MARBLE (multi-purpose advanced reactor physics analysis system based on language of engineering). In order to enhance the system extendibility and flexibility, the object-oriented design and analysis technique was adopted to the development. In the new system, it is easy to add a new design accuracy evaluation method because a new numerical calculation module is independent from other modules. In order to validate the new solver, a test calculation was performed for a realistic calculation case of creating a new unified cross section library. In the test calculation, results calculated by the new solver agreed well with those by the conventional code system. Because the new solver implements all main functions of the conventional code system, MARBLE can be used as a new calculation code system for cross section adjustment and nuclear design accuracy evaluation.
Nakano, Masanao
JAEA-Data/Code 2007-024, 37 Pages, 2008/02
LAMER is a calculation code which calculates the radionuclides distribution in seawater by simulating long-term advection, duffusion, and scavenging processes in worldwide scale. Then, the effective dose to the public is calculated by considering the transfer process from seawater to the human beings via marine products. The oceanic general circulation model was adopted for calculation of three-dimensional velocity field, particle tracking model for advection process, random walk model for diffusion process, and exchangable model for scavenging process. For an example, LAMER calculated the averaged effective dose to the human beings from the main radionuclides which were discharged by the atmospheric nuclear tests conducted between 1945 and 1980. In addition, the programs for output of water profile of radionuclides and the file structure of LAMER are described in the appendices.
Kumagai, Yasuhito; Funaki, Hironori; Yamazaki, Masanao; Yamaguchi, Takehiro; Orukawa, Go*; Sanada, Hiroyuki; Abe, Hironobu
JAEA-Data/Code 2007-025, 106 Pages, 2008/07
The shafts in construction of the underground facilities were started in earnest in 2006. The measurement plan applied to measurements in the underground facilities for validation of the geological environmental model was drew up and was carried out. This data code summarizes the measurements data acquired in 2006 of Phase 2 (investigations during construction of the underground facilities).