Refine your search:     
Report No.
 - 
Search Results: Records 1-20 displayed on this page of 82

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

Journal Articles

Estimation method of systematic uncertainties in Monte Carlo particle transport simulation based on analysis of variance

Hashimoto, Shintaro; Sato, Tatsuhiko

Journal of Nuclear Science and Technology, 56(4), p.345 - 354, 2019/04

 Times Cited Count:1 Percentile:17.09(Nuclear Science & Technology)

Particle transport simulations based on the Monte Carlo method have been applied to shielding calculations. Estimation of not only statistical uncertainty related to the number of trials but also systematic one induced by unclear physical quantities is required to confirm the reliability of calculated results. In this study, we applied a method based on analysis of variance to shielding calculations. We proposed random- and three-condition methods. The first one determines randomly the value of the unclear quantity, while the second one uses only three values: the default value, upper and lower limits. The systematic uncertainty can be estimated adequately by the random-condition method, though it needs the large computational cost. The three-condition method can provide almost the same estimate as the random-condition method when the effect of the variation is monotonic. We found criterion to confirm convergence of the systematic uncertainty as the number of trials increases.

Journal Articles

Quasielastic neutron scattering of brucite to analyse hydrogen transport on the atomic scale

Okuchi, Takuo*; Tomioka, Naotaka*; Purevjav, N.*; Shibata, Kaoru

Journal of Applied Crystallography, 51, p.1564 - 1570, 2018/12

AA2018-0399.pdf:1.07MB

 Times Cited Count:0 Percentile:0.01(Chemistry, Multidisciplinary)

It is demonstrated that quasielastic neutron scattering (QENS) is a novel and effective method to analyse atomic scale hydrogen transport processes occurring within a mineral crystal lattice. The method was previously characterized as sensitive for analysing the transport frequency and distance of highly diffusive hydrogen atoms or water molecules in condensed matter. Here are shown the results of its application to analyse the transport of much slower hydrogen atoms which are bonded into a crystal lattice as hydroxyls. Two types of hydrogen transport process were observed in brucite, Mg(OH)$$_{2}$$ : a jump within a single two-dimensional layer of the hydrogen lattice and a jump into the next nearest layer of it. These transport processes observed within the prototypical structure of brucite have direct implications for hydrogen transport phenomena occurring within various types of oxides and minerals having layered structures.

Journal Articles

Neutron transport

Tamura, Itaru

Hamon, 28(4), p.204 - 207, 2018/11

A Neutron guide is one of the devices to transport neutron beam for long distance without sacrificing much neutrons; therefore, it can supply neutrons to many experimental instruments distributed in a large experimental hall. Also, by using a curved guide, only the neutrons in a required energy range can be transported, and $$gamma$$ rays and fast neutrons can be effectively eliminated, therefore the signal to background ratio is improved. In addition, a neutron beam can be branched by applying curved guides. Neutron guides are also used to control the divergence angle and intensity of the neutron beam supplied to the neutron instrument.

Journal Articles

Influence of the neutron transport tube on neutron resonance densitometry

Kitatani, Fumito; Tsuchiya, Harufumi; Koizumi, Mitsuo; Takamine, Jun; Hori, Junichi*; Sano, Tadafumi*

EPJ Web of Conferences, 146, p.09032_1 - 09032_3, 2017/09

 Times Cited Count:0 Percentile:0.08

Journal Articles

Materials and Life Science Experimental Facility at the Japan Proton Accelerator Research Complex, 1; Pulsed spallation neutron source

Takada, Hiroshi; Haga, Katsuhiro; Teshigawara, Makoto; Aso, Tomokazu; Meigo, Shinichiro; Kogawa, Hiroyuki; Naoe, Takashi; Wakui, Takashi; Oi, Motoki; Harada, Masahide; et al.

Quantum Beam Science (Internet), 1(2), p.8_1 - 8_26, 2017/09

At the Japan Proton Accelerator Research Complex (J-PARC), a pulsed spallation neutron source provides neutrons with high intensity and narrow pulse width to promote researches on a variety of science in the Materials and life science experimental facility. It was designed to be driven by the proton beam with an energy of 3 GeV, a power of 1 MW at a repetition rate of 25 Hz, that is world's highest power level. A mercury target and three types of liquid para-hydrogen moderators are core components of the spallation neutron source. It is still on the way towards the goal to accomplish the operation with a 1 MW proton beam. In this paper, distinctive features of the target-moderator-reflector system of the pulsed spallation neutron source are reviewed.

JAEA Reports

MVP/GMVP version 3; General purpose Monte Carlo codes for neutron and photon transport calculations based on continuous energy and multigroup methods

Nagaya, Yasunobu; Okumura, Keisuke; Sakurai, Takeshi; Mori, Takamasa

JAEA-Data/Code 2016-018, 421 Pages, 2017/03

JAEA-Data-Code-2016-018.pdf:3.89MB
JAEA-Data-Code-2016-018-appendix(CD-ROM).zip:4.02MB
JAEA-Data-Code-2016-018-hyperlink.zip:1.94MB

In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two Monte Carlo codes MVP (continuous-energy method) and GMVP (multigroup method) have been developed at Japan Atomic Energy Agency. The codes have adopted a vectorized algorithm and have been developed for vector-type supercomputers. They also support parallel processing with a standard parallelization library MPI and thus a speed-up of Monte Carlo calculations can be achieved on general computing platforms. The first and second versions of the codes were released in 1994 and 2005, respectively. They have been extensively improved and new capabilities have been implemented. The major improvements and new capabilities are as follows: (1) perturbation calculation for effective multiplication factor, (2) exact resonant elastic scattering model, (3) calculation of reactor kinetics parameters, (4) photo-nuclear model, (5) simulation of delayed neutrons, (6) generation of group constants, etc. This report describes the physical model, geometry description method used in the codes, new capabilities and input instructions.

Journal Articles

Improvement of neutron startup source handling work by developing new transportation container for High-Temperature engineering Test Reactor (HTTR)

Shimazaki, Yosuke; Sawahata, Hiroaki; Shinohara, Masanori; Yanagida, Yoshinori; Kawamoto, Taiki; Takada, Shoji

Journal of Nuclear Science and Technology, 54(2), p.260 - 266, 2017/02

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

The High-Temperature engineering Test Reactor (HTTR) has three neutron startup sources (NSs) in the reactor core, each of which consists of $$^{252}$$Cf with 3.7 GBq and is contained in a small capsule, installed in NS holder and subsequently in a control guide block (CR block). The NSs are exchanged at the interval of approximately 7 years. The NS holders are transported from the dealer's hot cell to the reactor facility of HTTR using a transportation container. The loading work of NS holders to the CR blocks is subsequently carried out in the fuel handling machine maintenance pit of HTTR. Technical issues, which are the reduction and prevention of radiation exposure of workers and the exclusion of falling of NS holder, were extracted from the experiences in past two exchange works of NSs to develop a safety handling procedure. Then, a new transportation container special to the NSs of HTTR was developed to solve the technical issues while keeping the cost as low as that for overhaul of conventional container. As the results, the NS handling work using the new transportation container was safely accomplished by developing the new transportation container which can reduce the risks of radiation exposure dose of workers and exclude the falling of NS holder.

JAEA Reports

Shielding calculation by PHITS code during replacement works of startup neutron sources for HTTR operation

Shinohara, Masanori; Ishitsuka, Etsuo; Shimazaki, Yosuke; Sawahata, Hiroaki

JAEA-Technology 2016-033, 65 Pages, 2017/01

JAEA-Technology-2016-033.pdf:11.14MB

To reduce the neutron exposure dose for workers during the replacement works of the startup neutron sources of the High Temperature Engineering Test Reactor, calculations of the exposure dose in case of temporary neutron shielding at the bottom of fuels handling machine were carried out by the PHITS code. As a result, it is clear that the dose equivalent rate due to neutron radiation can be reduced to about an order of magnitude by setting a temporary neutron shielding at the bottom of shielding cask for the fuel handling machine. In the actual replacement works, by setting temporary neutron shielding, it was achieved that the cumulative equivalent dose of the workers was reduced to 0.3 man mSv which is less than half of cumulative equivalent dose for the previous replacement works; 0.7 man mSv.

Journal Articles

Development of transportation container for the neutron startup source of High Temperature engineering Test Reactor (HTTR)

Shimazaki, Yosuke; Ono, Masato; Tochio, Daisuke; Takada, Shoji; Sawahata, Hiroaki; Kawamoto, Taiki; Hamamoto, Shimpei; Shinohara, Masanori

Proceedings of International Topical Meeting on Research Reactor Fuel Management and Meeting of the International Group on Reactor Research (RRFM/IGORR 2016) (Internet), p.1034 - 1042, 2016/03

In High Temperature Engineering Test Reactor (HTTR), three neutron holders containing $$^{252}$$Cf with 3.7 GBq for each are loaded in the graphite blocks and inserted into the reactor core as a neutron startup source which is changed at the interval of approximately ten years. These neutron holders containing the neutron sources are transported from the dealer's hot cell to HTTR using the transportation container. The holders loading to the graphite block are carried out in the fuel handling machine maintenance pit of HTTR. There were two technical issues for the safety handling work of the neutron holder. The one is the radiation exposure caused by significant movement of the container due to an earthquake, because the conventional transportation container was so large ($$phi$$1240 mm, h1855 mm) that it can not be fixed on the top floor of maintenance pit by bolts. The other is the falling of the neutron holder caused by the difficult remote handling work, because the neutron holder capsule was also so long ($$phi$$155 mm, h1285 mm) that it can not be pulled into the adequate working space in the maintenance pit. Therefore, a new and low cost transportation container, which can solve the issues, was developed. To avoid the neutron and $$gamma$$ ray exposure, smaller transportation container ($$phi$$820mm, h1150 mm) which can be fixed on the top floor of maintenance pit by bolts was developed. In addition, to avoid the falling of the neutron holder, smaller neutron holder capsule ($$phi$$75 mm, h135 mm) with simple handling mechanism which can be treated easily by manipulator was also developed. As the result of development, the neutron holder handling work was safely accomplished. Moreover, a cost reduction for manufacturing was also achieved by simplifying the mechanism of neutron holder capsule and downsizing.

JAEA Reports

MOSRA-SRAC; Lattice calculation module of the modular code system for nuclear reactor analyses MOSRA

Okumura, Keisuke

JAEA-Data/Code 2015-015, 162 Pages, 2015/10

JAEA-Data-Code-2015-015.pdf:3.99MB
JAEA-Data-Code-2015-015-appendix(CD-ROM).zip:3.38MB

MOSRA-SRAC is a lattice calculation module of the Modular code System for nuclear Reactor Analyses (MOSRA). This module performs the neutron transport calculation for various types of fuel elements including existing light water reactors, research reactors, etc. based on the collision probability method with a set of the 200-group cross-sections generated from the Japanese Evaluated Nuclear Data Library JENDL-4.0. It has also a function of the isotope generation and depletion calculation for up to 234 nuclides in each fuel material in the lattice. In these ways, MOSRA-SRAC prepares the burn-up dependent effective microscopic and macroscopic cross-section data to be used in core calculations.

Journal Articles

Measurement of neutron spectra produced in the forward direction from thick graphite, Al, Fe and Pb targets bombarded by 350 MeV protons

Iwamoto, Yosuke; Taniguchi, Shingo*; Nakao, Noriaki*; Itoga, Toshio*; Nakamura, Takashi*; Nakane, Yoshihiro; Nakashima, Hiroshi; Satoh, Daiki; Yashima, Hiroshi*; Yamakawa, Hiroshi*; et al.

Nuclear Instruments and Methods in Physics Research A, 562(2), p.789 - 792, 2006/06

 Times Cited Count:6 Percentile:44.93(Instruments & Instrumentation)

Neutron energy spectra produced from thick targets play an important role in validation of calculation codes that are employed in the design of spallation neutron sources and the shielding design of accelerator facilities. However, appropriate experimental data were scarce in the forward direction for the incident energy higher than 100 MeV. In this study, neutron spectra at 0 degree from thick targets bombarded with 350 MeV protons were measured by the time-of-flight technique using an NE213. The targets used were graphite, Al, Fe and Pb and their thicknesses were chosen to be a little thicker than the stopping lengths. The experiment was carried out at the TOF course of the RCNP (Research Center of Nuclear Physics) ring cyclotron, Osaka University. The flight path length between center of the target and of an NE213 were 11.4 m for the measurement of low energy neutrons and 95 m for high energy neutrons. The experimental data are compared with the calculated results by using the Monte Carlo transport codes, such as MCNPX and PHITS codes.

JAEA Reports

MVP/GMVP 2; General purpose Monte Carlo codes for neutron and photon transport calculations based on continuous energy and multigroup methods

Nagaya, Yasunobu; Okumura, Keisuke; Mori, Takamasa; Nakagawa, Masayuki

JAERI 1348, 388 Pages, 2005/06

JAERI-1348.pdf:2.02MB

To realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two vectorized Monte Carlo codes MVP and GMVP have been developed at JAERI. MVP is based on the continuous energy model and GMVP is on the multigroup model. Compared with conventional scalar codes, these codes achieve higher computation speed by a factor of 10 or more on vector supercomputers. Both codes have sufficient functions for production use by adopting accurate physics model, geometry description capability and variance reduction techniques. The first version of the codes was released in 1994. They have been extensively improved and new functions have been implemented. The major improvements and new functions are (1) capability to treat the scattering model expressed with File 6 of the ENDF-6 format, (2) time-dependent tallies, (3) reaction rate calculation with the pointwise response function, (4) flexible source specification, etc. This report describes the physical model, geometry description method used in the codes, new functions and how to use them.

JAEA Reports

Evaluation on activation activity of reactor in JRR-2 applied 3 dimensional model to neutron flux calculation

Kishimoto, Katsumi; Arigane, Kenji*

JAERI-Tech 2005-016, 83 Pages, 2005/03

JAERI-Tech-2005-016.pdf:10.52MB

Revaluation to activation activity of reactor evaluated at the notification of dismantling submitted in 1997 was carried out in JRR-2 where decommissioning was advanced now. In the revaluation, estimation accuracy on neutron streaming at various horizontal experimental tubes was improved by applying 3 dimensional model to neutron transport calculation that had been carried out by 2 dimensional model, and calculating with TORT. As the result, excessive overestimations on horizontal experimental tubes and biological shield that had greatly contributed to total activation activity in evaluation at the notification of dismantling was revised, sum of their activation activities in the revaluation decreased to 1/18(case after 1 year from the permanent shutdown of reactor) of evaluation at the notification of dismantling, and the structural materials that had large activation activity were changed. By the above, it was shown that introducing 3 dimensional model was effective in evaluation on activation activity of the research reactor that had a lot of various experimental tubes.

Journal Articles

Neutron diagnostics for the energetic ion transport analysis

Nishitani, Takeo; Osakabe, Masaki*; Shinohara, Koji; Ishikawa, Masao

Purazuma, Kaku Yugo Gakkai-Shi, 80(10), p.860 - 869, 2004/10

no abstracts in English

JAEA Reports

Proceedings of the 3rd Workshop on Dosimetry for External Radiations; November 28-29, 2002, Japan Atomic Energy Research Institute, Tokai, Ibaraki, Japan

Yoshizawa, Michio; Endo, Akira

JAERI-Conf 2003-002, 166 Pages, 2003/03

JAERI-Conf-2003-002.pdf:9.79MB

The present report is Proceedings of the Third Workshop on Dosimetry for External Radiations, held at the Tokai Research Establishment, Japan Atomic Energy Research Institute (JAERI), in November 28-29, 2002. The proceedings comprises 16 papers and a summary of general discussion. The Third Workshop, subtitled "On an opportunity of the completion of the facility of calibration standards for neutron at JAERI", focused on neutron dosimetry and included presentations on the status of international neutron standards, the development of calibration techniques of neutron dosimeters using accelerator neutron sources, and dosimetry for high-energy neutrons. The workshop identified the directions for the future research and development in this field.

JAEA Reports

Journal Articles

Fast vector computation of the characteristics method

Kugo, Teruhiko

Journal of Nuclear Science and Technology, 39(3), p.256 - 263, 2002/03

 Times Cited Count:3 Percentile:24.2(Nuclear Science & Technology)

Two vector computation algorithms; an odd-even sweep (OES) method and an independent sequential sweep (ISS) method, have been developed for the characteristics method to solve the neutron transport equation in a heterogeneous geometry. They realize long vector lengths without recursive operations for effective use of vector computers. Their efficiency has been investigated to a realistic fuel assembly calculation. For both methods, a vector computation is 15 times faster than a scalar computation. From a viewpoint of a comparison between the OES and ISS methods, the ISS method is superior to the OES method because the ISS method shows a faster convergence and saves a computer memory without reducing a computation speed.

JAEA Reports

Activity report of Working Party on Reactor Physics of Accelerator-driven System; July 1999 to March 2001

Research Committee on Reactor Physics

JAERI-Review 2001-047, 180 Pages, 2002/02

JAERI-Review-2001-047.pdf:10.03MB

Under the Research Committee on Reactor Physics, the Working Party on Reactor Physics of Accelerator-Driven System (ADS-WP) was set in July 1999 to review and investigate special subjects related to reactor physics research for the Accelerator-Driven Subcritical System (ADS).The ADS-WP, at the first meeting, discussed a task guideline of its activity for two years and decided to concentrate upon three subjects: (1) neutron transport calculations in high energy range, (2) static and kinetic (safety-related) characteristics of subcritical system, and (3) system design including ADS concepts and elemental technology developments required.The activity of ADS-WP continued from July 1999 to March 2001. In this duration, the members of ADS-WP met together four times and discussed the above subjects. In addition, the ADS-WP conducted a questionnaire on requests and proposals for the plan of Transmutation Physics Experimental Facility in the High-Intensity Proton Accelerator Project, which is a joint project between JAERI and KEK (High Energy Accelerator Research Organization).This report summarizes the results obtained by the above ADS-WP activity. The report will be useful to overview those results and moreover to set up a new guideline of future research activity in this field.

JAEA Reports

Fast computation of the characteristics method on vector computers

Kugo, Teruhiko

JAERI-Research 2001-051, 39 Pages, 2001/11

JAERI-Research-2001-051.pdf:2.04MB

Fast computation of the characteristics method to solve the neutron transport equation in a heterogeneous geometry has been studied. Two vector computation algorithms; an odd-even sweep (OES) method and an independent sequential sweep (ISS) method have been developed and their efficiency to a typical fuel assembly calculation has been investigated. For both methods, a vector computation is 15 times faster than a scalar computation. From a viewpoint of comparison between the OES and ISS methods, the ISS method is superior to the OES method because the ISS method shows a faster convergence and saves a computer memory without reducing a computation speed. In the vector computation, a table-look-up method to reduce computation time of an exponential function saves only 20% of a whole computation time. Both the coarse mesh rebalance method and the Aitken acceleration method are effective as acceleration methods for the characteristics method, a combination of them saves 70-80% of outer iterations compared with a free iteration.

JAEA Reports

Study on residual radioactive inventory estimation in reactor decommissioning program (Contract research)

Sukegawa, Takenori; Hatakeyama, Mutsuo; Yanagihara, Satoshi

JAERI-Tech 2001-058, 81 Pages, 2001/09

JAERI-Tech-2001-058.pdf:5.98MB

In general, neutron transport and activation calculation codes are used for residual radioactive inventory estimation; however, it is essential to verify calculations by measurement results because of geometrical complexity of the reactor and so on. The comparison between measured and calculated radioactivity in the JPDR core components showed a relatively good agreement (factor of 2), and it was cleared that water content and weight ratio of steel bars to concrete materials significantly influenced the neutron flux distribution in the biological shield (factor of 2-10 error). The measured radioactivity inside of the reactor pressure vessel wall and at the inner part of the biological shield was compared in detail with the calculations to verify the methodology applied to calculations of radioisotope production. Then it was found that the radioactive inventory could be estimated accurately with combination of calculations and measurement of radioactivity in samples and dose rate distribution for planning of dismantling activities.

82 (Records 1-20 displayed on this page)