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Journal Articles

Systematic measurements and analyses for lead void reactivity worth in a plutonium core and two uranium cores with different enrichments

Fukushima, Masahiro; Goda, J.*; Oizumi, Akito; Bounds, J.*; Cutler, T.*; Grove, T.*; Hayes, D.*; Hutchinson, J.*; McKenzie, G.*; McSpaden, A.*; et al.

Nuclear Science and Engineering, 194(2), p.138 - 153, 2020/02

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

To validate lead (Pb) nuclear cross sections, a series of integral experiments to measure lead void reactivity worth was conducted systematically in three fast spectra with different fuel compositions on the Comet critical assembly of the National Criticality Experiments Research Center. Previous experiments in a high-enriched uranium (HEU)/Pb and a low-enriched uranium (LEU)/Pb systems had been performed in 2016 and 2017, respectively. A follow-on experiment in a plutonium (Pu)/Pb system has been completed. The Pu/Pb system was constructed using lead plates and weapons grade plutonium plates that had been used in the Zero Power Physics Reactor (ZPPR) of Argonne National Laboratory until the 1990s. Furthermore, the HEU/Pb system was re-examined on the Comet critical assembly installed newly with a device that can guarantee the gap reproducibility with a higher accuracy and precision, and then the experimental data was re evaluated. Using the lead void reactivity worth measured in these three cores with different fuel compositions, the latest nuclear data libraries, JENDL 4.0 and ENDF/B VIII.0, were tested with the Monte Carlo calculation code MCNP version 6.1. As a result, the calculations by ENDF/B-VIII.0 were confirmed to agree with lead void reactivity worth measured in all the cores. It was furthermore found that the calculations by JENDL 4.0 overestimate by more than 20% for the Pu/Pb core while being in good agreements for the HEU/Pb and LEU/Pb cores.

Journal Articles

Study on the two-phase flow in simulated LWR fuel bundle by CFD code

Ono, Ayako; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki

Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.666 - 677, 2019/08

An evaluation methodology of critical heat fluxes (CHFs) based on a mechanism for fuel assemblies in light water reactors (LWRs) is needed in order to design and evaluate the safety for the fuel assemblies in LWRs. In our study, the numerical simulation with surface-tracking will be applied for the two-phase flow in fuel assemblies in order to obtain the detail data relating to the size and velocity of bubbles in the subchannel, which is needed to predict the CHF based on the mechanism. In this study, the numerical simulation of two-phase flow in 4$$times$$4 bundle was implemented by using JUPITER in order to establish the evaluation method of the size and velocity of bubbles by the numerical simulation, which is the multi-physics simulation code and enable to track the gas-liquid surface. The simulation results are validated by the curve of flow regime for air-water under the adiabatic condition. The bubble and velocity of bubbles obtained by simulation results are analyzed.

Journal Articles

Thermal-hydraulic analysis of fuel assembly with inner duct structure of an advanced loop-type sodium-cooled fast reactor using ASFRE code

Kikuchi, Norihiro; Imai, Yasutomo*; Yoshikawa, Ryuji; Doda, Norihiro; Tanaka, Masaaki; Ohshima, Hiroyuki

Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 12 Pages, 2017/07

In the design study of an advanced loop-type SFR in JAEA, a specific fuel assembly (FA) named FAIDUS (Fuel Assembly with Inner DUct Structure) has been adopted as one of the measures to enhance safety of the reactor. Thermal-hydraulics evaluations of FAIDUS under various operation conditions are required to confirm its design feasibility. In this study, after the applicability of ASFRE to FAs was confirmed through the numerical analysis using simulated FA tests, thermal-hydraulic analyses of a FA without an inner duct and a FAIDUS were conducted. Through the numerical analyses, it was indicated that asymmetric temperature distribution in a FAIDUS would not be occurred and characteristics of the temperature distribution was almost the same as that in a FA without an inner duct. Under the low flow rate condition, it was expected that the local flow acceleration caused by the buoyancy force in a FAIDUS could bring the flow redistribution and make the temperature distribution flat.

Journal Articles

Development of thermal hydraulics analysis code ASFRE for fuel assembly of sodium-cooled fast reactor; Modification of distributed resistance model and validation analysis

Kikuchi, Norihiro; Ohshima, Hiroyuki; Tanaka, Masaaki; Hashimoto, Akihiko*

Dai-21-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 4 Pages, 2016/06

For the thermal-hydraulic design and safety assessment regarding a fuel assembly of sodium-cooled fast reactors, a subchannel analysis code ASFRE has been and is continuously developed in JAEA. In the numerical simulation of ASFRE confirmed that the tendency to overestimate the maximum coolant temperature in a fuel assembly still remains. In this study, Distributed Resistance Model (DRM), which deals with wire-spacer wrap volumetric effect in subchannels on peripheral and axial directions, was modified and its calibration factor was optimized in order to improve the prediction accuracy of the maximum coolant temperature. A numerical simulation of a 37-pin bundle sodium experiment was also carried out and the result showed the validity of the modified DRM.

Journal Articles

Validation of MOSRA-SRAC for burnup of a BWR fuel assembly

Kojima, Kensuke

Proceedings of International Conference on the Physics of Reactors; Unifying Theory and Experiments in the 21st Century (PHYSOR 2016) (USB Flash Drive), p.3283 - 3292, 2016/05

The MOSRA system has been developing to improve the applicability of the neutronic characteristic analyses. The cell calculation module MOSRA-SRAC is a core module of MOSRA, and applicability tests for realistic problems are required. As a test, MOSRA-SRAC is validated by comparison with measured values. As the measurement, the post irradiation examination SFCOMPO 99-5 is chosen. In the examination, the compositions of major heavy metal and fission product nuclides in a UO$$_{2}$$-Gd$$_{2}$$O$$_{3}$$ fuel rod pulled from the 8$$times$$8 BWR fuel assembly used in TEPCO's Fukushima-Daini-2 were measured. The result shows good agreement between calculated and measured value. For uranium and plutonium nuclides, calculated values agree within 5% except for $$^{238}$$Pu. $$^{238}$$Pu composition is overestimated by 30%, and the overestimation is caused by the unclearness of the void faction history of the fuel rod. For fission products, calculated values agree within approximately 10%.

JAEA Reports

Development of BDI behavior evaluation method in the fast reactor fuel assembly; Improvement of out-of-pile bundle compression test technology

Higashiuchi, Atsushi; Ishimi, Akihiro; Katsuyama, Kozo; Uwaba, Tomoyuki; Ichikawa, Shoichi

JAEA-Technology 2015-057, 72 Pages, 2016/03


Bundle-duct interaction (BDI) in fast reactors (FRs) is one of the limiting factors for burnup. To study the high performance fuel for FR fuel, it is important to establish the method to predict accurately the BDI behavior for the fuel assembly of large-diameter fuel pins. Therefore, it was adopted a new method that the bundle compression test apparatus is placed outside the cell, the bundle specimen is put in the airtight container for contamination prevention, and the bundle specimen is carried in the cell for internal observation by X-ray CT examination apparatus. From the result of this test, it was confirmed that the new method of out-of-pile bundle compression test is carried out as it was before. The results of this test are available to study integrity assessment of fast reactor fuel, validation of the BDI analysis code and substantiation of the safety design guidelines of fast reactor. In addition, it is possible to reflect in the BDI behavior evaluation for "ASTRID".

JAEA Reports

The States of the art of the nondestructive assay of spent nuclear fuel assemblies; A Critical review of the Spent Fuel NDA Project of the U.S. Department of Energy's Next Generation Safeguards Initiative

Bolind, A. M.*; Seya, Michio

JAEA-Review 2015-027, 233 Pages, 2015/12


This report surveys the 14 advanced NDA techniques that were examined by the Spent Fuel NDA Project of the Next Generation Safeguards Initiative (NGSI) of the U.S. DOE-NNSA. It discusses and critique NDA techniques from a view point of obtaining higher accuracies. The report shows the main problem, large uncertainties in the assay results are caused primarily by using too few independent NDAs. In this report authors shows that at least three independent NDA techniques are required for obtaining better accuracies, since the physics of the NDA of SFAs is three dimensional.

JAEA Reports

Development of thermal-hydraulic design code for transmutation system with lead-bismuth cooled accelerator driven reactor

Akimoto, Hajime

JAEA-Data/Code 2014-031, 75 Pages, 2015/03


A thermal-hydraulic analysis code for transmutation system with lead-bismuth cooled accelerator-driven system (ADS) has been developed using the Japanese-version of Transient Reactor Analysis Code (J-TRAC) as the framework to apply the design studies of ADS. To identify the required capabilities of the thermal-hydraulic analysis code for ADS, previous thermal-hydraulic analyses of light water reactors, sodium-cooled fast reactor and ADS have been surveyed. To make up for insufficient capabilities of the J-TRAC code as a thermal-hydraulic analysis code of ADS, physical properties of lead-bismuth eutectic (LBE), argon gas and nitride nuclear fuel were implemented to the J-TRAC code. It was confirmed that the implemented capabilities worked as expected through verification calculations on (1) single-phase LBE flow, (2) heat transfer in a fuel assembly, and (3) heat transfer in a steam generator.

Journal Articles

An Experimental study on heat transfer from a mixture of solid-fuel and liquid-steel during core disruptive accidents in Sodium-Cooled Fast Reactors

Kamiyama, Kenji; Konishi, Kensuke; Sato, Ikken; Toyooka, Junichi; Matsuba, Kenichi; Suzuki, Toru; Tobita, Yoshiharu; Pakhnits, A. V.*; Vityuk, V. A.*; Vurim, A. D.*; et al.

Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 8 Pages, 2014/12

Journal Articles

Research for thermal-hydraulic performance in tight-lattice fuel assembly, 1; Outline of research program

Akimoto, Hajime; Tamai, Hidesada; Onuki, Akira; Takase, Kazuyuki

Nippon Konsoryu Gakkai Nenkai Koenkai 2005 Koen Rombunshu, p.229 - 230, 2005/08

A thermal-hydraulic research program for Reduced-Moderation Water Reactor (RMWR) has been performed since 2002. The RMWR has a tight-lattice core to attain the breeding of nuclear fuel for the effective use of Plutonium in a light-water reactor system. In this R&D program, large-scale thermal-hydraulic tests, several model experiments and development of advanced numerical analysis codes are being carried out to confirm the cooling performance in tight-lattice fuel assembly of the RMWR. In this paper, outline of the research program is described as well as the latest results of critical power measurement in the large-scale thermal-hydraulic tests and model experiments, which simulates the tight-lattice core of the RMWR.

Journal Articles

Numerical analysis of three-dimensional two-phase flow behavior in a fuel assembly

Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*; Akimoto, Hajime

WIT Transactions on Engineering Sciences, Vol.50, p.183 - 192, 2005/00

no abstracts in English

JAEA Reports

Measurement of doppler effect on resonance materials for ROX fuel using FCA

Ando, Masaki; Nakano, Yoshihiro; Okajima, Shigeaki; Kawasaki, Kenji

JAERI-Research 2003-029, 72 Pages, 2003/12


The objectives of this study is to clarify calculation accuracy for the Doppler effect of the resonance materials; erbium (Er), tungsten (W) and thorium (ThO$$_{2}$$). Doppler effect measurements were carried out in a fast neutron spectrum (XX-2 core) and in an intermediate neutron spectrum (XXI-1D2 core) by the sample-heated and reactivity worth measurement method up to 800$$^{circ}$$C using FCA. The experiment was analyzed with the standard analysis method for fast reactor cores at FCA with the use of the JENDL-3.2. The SRAC system was also used to investigate the calculation accuracy of the system and to compare it with that of the FCA standard analysis method. The standard analysis method underestimated for the XX-2 core and agreed the experiments within the experimental errors for the XXI-1D2 core. The analysis with the SRAC system gave smaller values by 3%$$sim$$10% for the Er sample and bigger values by 2%$$sim$$5% for the W sample than the standard analysis method.

Journal Articles

Classification of criticality calculations with correlation coefficient method and its application to OECD/NEA burnup credit benchmarks phase III-A and II-A

Okuno, Hiroshi

Journal of Nuclear Science and Technology, 40(7), p.544 - 551, 2003/07

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

A method for classifying benchmark results of criticality calculations according to similarity was proposed in this paper. After formulation of the method utilizing correlation coefficients, it was applied to burnup credit criticality benchmarks Phase III-A and II-A, which were conducted by the Expert Group on Burnup Credit Criticality Safety under auspices of the OECD/NEA. Phase III-A benchmark was a series of criticality calculations for irradiated BWR fuel assemblies, whereas Phase II-A benchmark was a suite of criticality calculations for irradiated PWR fuel pins. These benchmark problems and their results were summarized. The correlation coefficients were calculated and sets of benchmark results were classified according to the criterion that the correlation coefficients were no less than 0.15 for Phase III-A and 0.10 for Phase II-A benchmarks. When a couple of results were in a same group, one result was found predictable from the other. An example was shown for each of the Benchmarks. The evaluated nuclear data seemed the main factor of errors.

Journal Articles

Doppler effect measurement on resonance materials for rock-like oxide fuels in an intermediate neutron spectrum

Ando, Masaki; Nakano, Yoshihiro; Okajima, Shigeaki; Kawasaki, Kenji

Journal of Nuclear Materials, 319, p.126 - 130, 2003/06

 Times Cited Count:0 Percentile:100

Doppler effect experiments on resonance materials for ROX fuels were carried out to examine the calculation accuracy in the intermediate neutron spectrum using Fast Critical Assembly (FCA) at Japan Atomic Energy Research Institute. This study is the second phase of a series of the Doppler effect experiments on the resonance materials, which is following the measurements in the fast neutron spectrum. The Doppler effect was measured as the sample reactivity change between the heated and unheated samples. The cylindrical samples of the resonance materials such as erbium (Er), tungsten (W) and thorium (ThO$$_{2}$$) were used. The sample was heated up to 800$$^{circ}$$C at the center of the FCA core. The Doppler effect measurements were analyzed using the SRAC 95 code system with the use of JENDL 3.2. The calculated values agreed with the experiment within the experimental error for the W and ThO$$_{2}$$ samples, while the calculation overestimated the experiment for the Er sample about 10 %.

JAEA Reports

Super safe small reactor RAPID-L conceptual design and R&D, JAERI's nuclear research promotion program, H11-002 (Contract research)

Kobe, Mitsuru*; Tsunoda, Hirokazu*; Mishima, Kaichiro*; Kawasaki, Akira*; Iwamura, Takamichi

JAERI-Tech 2003-016, 68 Pages, 2003/03


The 200 kWe uranium nitride fueled lithium cooled fast reactor "RAPID-L" combined with thermoelectric power conversion system that can be operated unmanned without refueling for up to ten years has been demonstrated. The RAPID refueling concept enables quick and simplified refueling, and achieves plant design lifetime over 20 years. A significant advantage of the RAPID-L design, which does not require the use of control rods - is the introduction of the innovative reactivity control systems: lithium expansion module (LEM), lithium injection module (LIM) and lithium release module (LRM). LEM is the most promisiong candidate for improving inherent reactivity feedback. LEMs could realize burnup compensation. LIMs assure sufficient negative reactivity feedback in unprotected transients. LRMs enable an automated reactor startup by detecting the hot standby temperature of the primary coolant. All these systems use $$^{6}$$Li as liquid poison and are actuated by highly reliable physical properties (volume expansion of $$^{6}$$Li for LEM, and freeze seal melting for LIM and LRM).

JAEA Reports

OECD/NEA burnup credit criticality benchmarks phase IIIB; Burnup calculations of BWR fuel assemblies for storage and transport

Okuno, Hiroshi; Naito, Yoshitaka*; Suyama, Kenya

JAERI-Research 2002-001, 181 Pages, 2002/02


The report describes the final results of the Phase IIIB Benchmark conducted by the Expert Group on Burnup Credit Criticality Safety under the auspices of the OECD/NEA. The Benchmark was intended to compare the predictability of current computer code and data library combinations for the atomic number densities of an irradiated BWR fuel assembly model, which was irradiated under specific power of 25.6 MW/tHM up to 40 GWd/tHM and cooled for five years. The void fraction was assumed to be uniform and constant, at 0, 40 and 70%, during burnup. In total, 16 results were submitted from 13 institutes of 7 countries. The calculated densities of 12 actinides and 20 fission product nuclides were found mostly within a range of +- 10% relative to the average, although some results, esp. 155Eu and gadolinium isotopes, exceeded the band. Pin-wise burnup results agreed well among the participants. The results in the multiplication factor also accorded well with each other for void fractions of 0 and 40%; however some results deviated from the average noticeably for the void fraction of 70%.

JAEA Reports

Fast computation of the characteristics method on vector computers

Kugo, Teruhiko

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


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

Libraries based on JENDL-3.2 for ORIGEN2 code; ORLIBJ32

Suyama, Kenya; Katakura, Junichi; Okawachi, Yasushi*; Ishikawa, Makoto*

JAERI-Data/Code 99-003, 83 Pages, 1999/02


no abstracts in English

Journal Articles

Nuclear criticality safety of fuel rod arrays taking irregularity into account

Okuno, Hiroshi; *

Criticality Safety Challenges in the Next Decade, 0, p.150 - 155, 1997/00

no abstracts in English

JAEA Reports

Basic experiments of reactor physics using the critical assembly TCA

Obara, Toru*; Nakajima, Ken; *; Sekimoto, Hiroshi*; Suzaki, Takenori

JAERI-M 94-004, 40 Pages, 1994/02


no abstracts in English

43 (Records 1-20 displayed on this page)