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Journal Articles

Mechanisms responsible for two possible electrochemical reactions in Li$$_{1.2}$$Ni$$_{0.13}$$Mn$$_{0.54}$$Co$$_{0.13}$$O$$_{2}$$ used for lithium ion batteries

Konishi, Hiroaki*; Hirano, Tatsumi*; Takamatsu, Daiko*; Gunji, Akira*; Feng, X.*; Furutsuki, Sho*; Okumura, Takafumi*; Terada, Shohei*; Tamura, Kazuhisa

Journal of Solid State Chemistry, 258, p.225 - 231, 2018/02

 Times Cited Count:8 Percentile:45.21(Chemistry, Inorganic & Nuclear)

Li$$_{1.2}$$Ni$$_{0.13}$$Mn$$_{0.54}$$Co$$_{0.13}$$O$$_{2}$$ is known as one of the cathode electrode material for Li ion batteries and its structure during charge and discharge process was investigated using electrochemical method and X-ray diffraction. It was found that in the charge process the structure changes in the order of Li$$_{2}$$MnO$$_{3}$$, LiNi$$_{0.33}$$Mn$$_{0.33}$$Co$$_{0.33}$$O$$_{2}$$, and Li$$_{2}$$MnO$$_{3}$$. On the other hand, in the discharge process, the structure changes in the order of Li$$_{2}$$MnO$$_{3}$$ and LiNi$$_{0.33}$$Mn$$_{0.33}$$Co$$_{0.33}$$O$$_{2}$$.

Journal Articles

Verification and validation procedures of calculation codes for determining corrosive conditions in the BWR primary cooling system based on water radiolysis and mixed potential models

Uchida, Shunsuke*; Wada, Yoichi*; Yamamoto, Seiji*; Takagi, Junichi*; Hisamune, Kenji*

Journal of Nuclear Science and Technology, 51(1), p.24 - 36, 2014/01

 Times Cited Count:8 Percentile:57.39(Nuclear Science & Technology)

ECP in the BWR primary cooling system can be measured only in the restricted region. In order to determine ECP at any location, ECP should be evaluated by computer simulation codes consisting of water radiolysis models to determine the concentrations of corrosive radiolytic species and mixed potential models to determine ECP based on corrosive species. Mitigation of SCC crack growth rate due to decreasing ECP can be authorized by the JSME Standards, while mitigation of ECP due to hydrogen addition has not been authorized yet. In the paper, standard procedures to authorize the computer simulation codes based on the verification and validation method are proposed. The numerical justification of every code applied as a standard code should be verified and its accuracy and applicability for plant analysis should be validated. Benchmark problems for verification processes are proposed and comparison of the calculated results with the measured ones for the plant of evaluation is required.

Journal Articles

Magnetic structure analysis of magnetic multilayers using neutron reflectometry

Hirano, Tatsumi*; Takeda, Masayasu

Magune, 4(1), p.38 - 42, 2009/01

The characterization of a magnetic interface structure in magnetic multilayer is important for producing good magnetic devices because the device properties strongly depend on the structure. In this report we review a polarized neutron reflectometry to analyze the magnetic interface structure and also mention a future prospect of a neutron experiment.

JAEA Reports

Preparation of computer codes for analyzing sensitivity coefficients of burnup characteristics, 2 (Contract research, Translated Document)

Hanaki, Hiroshi*; Sanda, Toshio*; Ohashi, Masahisa*

JAEA-Review 2008-047, 266 Pages, 2008/10

JAEA-Review-2008-047.pdf:62.06MB

It is thought to improve the prediction accuracy for burnup characteristics using many burnup data of "Joyo" effectively. It is thought the best way to adjust cross-sections using sensitivity coefficients of burnup characteristics to utilize burnup data of "Joyo". Therefore in this work computer codes for analyzing sensitivity coefficients of burnup characteristics had been prepared since 1992. The results are as follows: (1) The computer codes which could analyze sensitivity coefficients of burnup characteristics taking into consideration plural cycles and refueling were prepared, therefore it came to be able to adjust cross-sections using burnup data and to estimate the accuracy for design of large LMFBR cores. (2) The adequacy of prepared codes was made sure by comparing with direct calculations. (NB: This document is a translation of PNC TJ9124 94-007 Vol.2 published in March 1997.)

JAEA Reports

Investigation of Conceptua1 Facility Structure on the vibration packing fuel manufacturing system Corresponding to wet fuel recycle plant

Kawamura, Fumio*; Aoi, Masakatsu*; Hoshino, Kuniyoshi*

JNC TJ9420 2003-006, 436 Pages, 2004/02

JNC-TJ9420-2003-006.pdf:3.67MB

The stratagem feasibility studies for aim of commercialization of FBR cycle system have been researched in JNC. In these studies economy, load suppression for environmentality, non-proliferation and et. al. have been evaluated for several kinds of systems. In series of these researches, we had researched the vibration packing fuel manufacturing system of wet fuel recycle plant for FBR's fuel, and investigated the concepts of main equipments for the fuel pin fabrication system. In 2002 we had achieved the conceptual design of main equipments and assignment of those in cells from reception process of fuel particles material to storage process of fuel assemblies concerning the production scale about 50t-HM/y manufacturing system, and estimated the applicability of fabrication system in cell. In this study, we have researched the conceptual design and assignment in cells of main equipments of the vibration packing fuel manufacturing system, quality control, and maintenance of equipments concerning the production scale about 200t-HM/y manufacturing system. The building is constructed independently near here the recycle facility building and fuel particle fabrication building. And then the technical matching and economical evaluation of manufacturing system have been investigated. Resulting these investigations, we have got the basic conceptual design of main equipments for the fuel pin fabrication system in the scale of 200t-HM/y, assignment of main procedure cell and auxiliary equipments, and also the ability of 200 days operation per year based on the quality control and maintenance of equipments. And then we have got the aspect of technical matching to the plant and the reasonable economical evaluation of manufacturing system. It is cleared that this manufacturing method is the suitable one of systems concerning of wet fuel recycle plant for FBR's fuel manufacturing system.

JAEA Reports

None

Nakasato*; Yamaji*; Matsuoka*; Kishikawa, Takao*

JNC TJ7410 2005-013, 147 Pages, 2004/02

JNC-TJ7410-2005-013.PDF:6.98MB

no abstracts in English

JAEA Reports

None

Nakasato*; Takahashi*; Yamaji*

JNC TJ7410 2005-012, 91 Pages, 2003/03

JNC-TJ7410-2005-012.PDF:6.55MB

no abstracts in English

JAEA Reports

The Evaluation of thermal striping damages in temperature fluctuations

Sukekawa, Masayuki*; Imo, Kazumichi*; Kawakami, Mitsuo*

JNC TJ9430 2001-002, 98 Pages, 2002/03

JNC-TJ9430-2001-002.pdf:2.68MB

None

JAEA Reports

Improvement for simulation system for emergency release of radio active substance

Miyauchi, Kazuya*; Yamamoto, Asao*; Yatake, Yoichi*; Kasuya, Kazunori*

JNC TJ4440 2002-001, 197 Pages, 2002/03

JNC-TJ4440-2002-001.pdf:8.54MB

None

JAEA Reports

JAEA Reports

Evaluation of mechanical characteristic design models for vibro-packed fuel (2)

*; Onuki, Norihiko*; *; *

JNC TJ8410 2002-001, 114 Pages, 2002/02

JNC-TJ8410-2002-001.pdf:3.03MB

This study has been carried out as a link in the research and development of FBR vibro-packed fuel that fuel particles are compacted directly in the fuel cladding. The calculation program code for vibro-packed fuel design has been developed to analyze of the fuel thermal/mechanical performances. In last year, we investigated material design property equations for vibro-packed fuel by the distinct element method (PFC program code). In this year, considering the vibro-packed fuel pin has variable fuel particle packing conditions, we researched fuel mechanical properties using variable particle packing density and particle size distributions. Mechanical property analyses were carried out at the condition of load cycle and simulated fuel pin conditions, too. In the analysis, the particle packing density was set up approximately 50%$$sim$$70%, and particle size distribution was set up 0$$mu$$m$$sim$$200$$mu$$m. The hysteresis loop of stress - strain was investigated in the analysis of load cycle, and simulated fuel pin with center hole was investigated, too. Vibro-packed fuel's material design property equations (Effect Young's modulus, Effect Poisson's ratio) were proposed based on the above mentioned analysis results.

JAEA Reports

Comparison between TRU burning reactors and commercial fast reactor

Fujimura, Koji*; Sanda, Toshio*; Ogawa, Takashi*

JNC TJ9400 2001-015, 85 Pages, 2001/03

JNC-TJ9400-2001-015.pdf:2.4MB

Research and development for stabilizing or shortening the radioactive wastes included in spent nuclear fuel are widely conducted in view point of reducing the environmental impact. Especially it is effective way to irradiate and transmute long-lived TRU by fast reactors. Two types of loading way were previously proposed. The former is loading relatively small amount of TRU in all commercial fast reactors and the latter is loading large amount of TRU in a few TRU burning reactors. This study has been intended to contribute to the feasibility studies on commercialized fast reactor cycle system. The transmutation and nuclear characteristics of TRU burning reactors were evaluated and compared with those of conventional transmutation system using commercial type fast reactor based upon the investigation of technical information about TRU burning reactors. Major results are summarized as follows. (1) Investigation of technical information about TRU burning reactors. Based on published reports and papers, technical information about TRU burning reactor concepts and transmutation system using conventional commercial type fast reactors were investigated. Transmutation and nuclear characteristics or R&D issue were investigated based on these results. (a) Homogeneously loading of about 5wt% MAs on core fuels in the conventional commercial type fast reactor may not cause significant impact on the nuclear core characteristics. Transmutation of MAs being produced in about five fast reactors generating the same output is feasible. (b) The helium cooled MA burning fast reactor core concept propose by JAERI attains criticality using particle type nitride fuels which contain more than 60 wt% MA. This reactor could transmute MAs being produced in more than ten 1000 MWe-LWRs. (c)Ultra-long life core concepts attaining more than 30 years operation without refueling by utilizing MA's nuclear characteristics as burnable absorber and fertile nuclides were proposed. (d)Those ...

JAEA Reports

Summary of requirements on SASS for a large FBR Core

*; Sawada, Shusaku*; *; *; *

JNC TJ9400 2001-011, 159 Pages, 2001/03

JNC-TJ9400-2001-011.pdf:5.21MB

In order to improve the reliability of safety system of FBR employing passive safety functions, requirements on a self-actuated shutdown system (SASS) have been summarized, which employs Curie-point magnets in its safety rods, for a large homogeneous reactor core (1500MWe), which was designed by JNC in JFY1999 as one of candidates in the Feasibility Studies on Commercialized FBR System. The requirements were based on the sensitivity analyses, conducted by using the safety analysis code SAS4A, of uncertainty factors that affect the maximum coolant temperature in unprotected loss-of-flow events. The study has given the following requirements: (1)The detachment temperature of the SASS magnet is to be set above 638$$^{circ}$$C (911K) to avoid the possibility of unintended rod drop under normal operations, and to be set below 666$$^{circ}$$C (939K) to prevent coolant boiling under ULOF conditions conservatively assuming a boiling temperature of 960$$^{circ}$$C in the plant. (2)The parameter that has the largest sensitivity on the maximum coolant temperature under ULOF conditions is the exit coolant temperatures of the adjacent fuel assemblies to the SASS assembly, with a sensitivity coefficient of 20.5$$^{circ}$$C/$$sigma$$ (1$$sigma$$= 6.7$$^{circ}$$C for the exit coolant temperature). It has been concluded based on the parametric analyses that a SASS design even with a nominal detachment temperature of 666$$^{circ}$$C, the SASS has a safety margin of about 3.4$$sigma$$ for the exit coolant temperature of the most sensitive adjacent fuel assembly in preventing coolant boiling under ULOF conditions, if the incoherency of SASS detachments are appropriately taken into account together with a more realistic criterion on the coolant boiling temperature (1018$$^{circ}$$C) under the ULOF conditions of this plant.

JAEA Reports

Mercury flow tests, 2; Characteristic of gear pump for mercury circulation

Kaminaga, Masanori; Kinoshita, Hidetaka; Haga, Katsuhiro; Hino, Ryutaro; Nakamura, Fumihito*; Ohashi, Masahisa*

JAERI-Tech 2000-044, 25 Pages, 2000/06

JAERI-Tech-2000-044.pdf:4.64MB

no abstracts in English

JAEA Reports

Improvement on reaction model for sodium-water reaction; Jet code and application analysis

*; *; *; *; *

JNC TJ9440 2000-010, 132 Pages, 2000/03

JNC-TJ9440-2000-010.pdf:14.85MB

In selecting the reasonable DBL on steam generator (SG), it is necessary to improve analytical method for estimating the sodium temperature on failure propagation due to overheating. Improvement on sodium-water reaction (SWR) jet code (LEAP-JET ver.1.30) and application analysis to the water injection tests for confirmation of code propriety were performed. On the improvement of the code, a gas-liquid interface area density model was introduced to develop a chemical reaction model with a little dependence on calculation mesh size. The test calculation using the improved code (LEAP-JET ver.1.40) were carried out with conditions of the SWAT-3$$cdot$$Run-19 test and an actual scale SG. It is confirmed that the SWR jet behavior on the results and the influence to analysis result of a model are reasonable. For the application analysis to the water injection tests, water injection behavior and SWR jet behavior analyses on the new SWAT-1 (SWAT-1R) and SWAT-3(SWAT-3R) tests were performed using the LEAP-BLOW code and the LEAP-JET code. In the application analysis of the LEAP-BLOW code, parameter survey study was performed. As the results, the condition of the injection nozzle diameter needed to simulate the water leak rate was confirmed. In the application analysis of the LEAP-JET code, temperature behavior of the SWR jet was investigated.

JAEA Reports

None

*; *

JNC TJ8420 2000-010, 171 Pages, 2000/03

JNC-TJ8420-2000-010.pdf:5.34MB

no abstracts in English

JAEA Reports

None

*; *; *; *

JNC TJ3410 2000-021, 73 Pages, 2000/03

JNC-TJ3410-2000-021.pdf:52.78MB

no abstracts in English

JAEA Reports

None

*; *; *; *

JNC TJ3410 2000-020, 80 Pages, 2000/03

JNC-TJ3410-2000-020.pdf:41.34MB

no abstracts in English

JAEA Reports

Verification of blow down code for LEAP code; Verification by RELAP5/Mod.2 and BLOOPH code

*; *; *; *

JNC TJ9440 99-024, 142 Pages, 1999/03

JNC-TJ9440-99-024.pdf:7.16MB

Behavior of over heating tube rupture in sodium-water reaction have to be evaluated practically in order to confirm the propriety of DBL(Design Basis Leak) on steam generator of next large LMFBR. Over heating tube rupture is closely concerned with water / steam condition in tubes, sodium-water reaction temperature and high temperature strength of tube wall. Therefore, it is very important to precisely evaluate water / steam conditions in blow down event especially. On the other hand, as work for MONJU safety general inspection, blow down behavior was analyzed by BLOOPH code and RELAP5/Mod.2 code. LEAP-BLOW code (Ver.1.20) has been developed reflecting the acknowledgment of the MONJU blow down analysis in the blow down code for LEAP. In this code heat transfer model of sodium side of the downcommer was improved. And, using LEAP-BLOW code (Ver.1.20) MONJU blow down characteristic on the following cases was analyzed and compared with the analysis results of RELAP5/Mod.2 code and BLOOPH code. Then, it has been confirmed that there are no meaningfull difference in the results of these code, and the propriety of analysis result of LEAP-BLOW code has been confirmed. (1)Blow-down from 100% power in MONJU. (1 channel model and 2 channel model) (2)Blow-down from 100% power in MONJU. (Improvement equipment model) (3)Blow-down from Partial power in MONJU, (40% power and start-up)

JAEA Reports

Improvement and test calculation on basic code for sodium-water reaction jet

*; *; *; *; *

JNC TJ9440 99-023, 218 Pages, 1999/03

JNC-TJ9440-99-023.pdf:33.77MB

In selecting the reasonable DBL on steam generator (SG), it is necessary to improve analytical method for estimating the sodium temperature on failure propagation due to overheating. Improvement on the basic code for sodium-water reaction (SWR) jet was performed for an actual scale SG. The improvement points of the code are as follows; (1)introduction of advanced model such as heat transfer between the jet and structure (tube array), cooling effect of the structure, heat transfer between analytic cells, and (2)model improvement for heat transfer between two-phase flow and porous-media. The test calculation using the improved code (LEAP-JET ver.1.30) were carried out with conditions of the SWAT-3 $$cdot$$ Run-19 test and an actual scale SG. It is confirmed that the SWR jet behavior on the results is reasonable and Influence to analysis result of a model. Code integration with the blow down analytic code (LEAP-BLOW) was also studied. It is suitable that LEAP-JET was improved as one of the LEAP-BLOW's models, and it was integrated into this. In addition to above, the improvement for setting of boundary condition and the development of the interface program to transfer the analytical results of LEAP-BLOW have been performed in order to consider the cooling effect of coolant in the tube simply. However, verification of the code by new SWAT-1 and SWAT-3 test data planned in future is necessary because LEAP-JET is under development. And furthermore advancement needs to be planned.

133 (Records 1-20 displayed on this page)