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Journal Articles

Model updates and performance evaluations on fuel performance code FEMAXI-8 for light water reactor fuel analysis

Udagawa, Yutaka; Amaya, Masaki

Journal of Nuclear Science and Technology, 56(6), p.461 - 470, 2019/06

no abstracts in English

JAEA Reports

Development of fuel performance code FEMAXI-8; Model improvements for light water reactor fuel analysis and systematic validation

Udagawa, Yutaka; Yamauchi, Akihiro*; Kitano, Koji*; Amaya, Masaki

JAEA-Data/Code 2018-016, 79 Pages, 2019/01


FEMAXI-8 is the latest version of the fuel performance code FEMAXI developed by JAEA. A systematic validation work has been achieved against 144 irradiation test cases, after many efforts have been made, in development of new models, improvements in existing models and the code structure, bug-fixes, construction of irradiation-tests database and other infrastructures.

Journal Articles

Thermal-hydraulics technological strategy roadmap 2017; An Approach for continuous safety improvement of LWRs

Itoi, Tatsuya*; Iwaki, Chikako*; Onuki, Akira*; Kito, Kazuaki*; Nakamura, Hideo; Nishida, Akemi; Nishi, Yoshihisa*

Nippon Genshiryoku Gakkai-Shi, 60(4), p.221 - 225, 2018/04

no abstracts in English

Journal Articles

Technical basis of accident tolerant fuel updated under a Japanese R&D project

Yamashita, Shinichiro; Nagase, Fumihisa; Kurata, Masaki; Nozawa, Takashi; Watanabe, Seiichi*; Kirimura, Kazuki*; Kakiuchi, Kazuo*; Kondo, Takao*; Sakamoto, Kan*; Kusagaya, Kazuyuki*; et al.

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09

In Japan, the research and development (R&D) project on accident tolerant fuel and other components (ATFs) of light water reactors (LWRs) has been initiated in 2015 for establishing technical basis of ATFs. The Japan Atomic Energy Agency (JAEA) has coordinated and carried out this ATF R&D project in cooperation with power plant providers, fuel venders and universities for making the best use of the experiences, knowledges in commercial uses of zirconium-base alloys (Zircaloy) in LWRs. ATF candidate materials under consideration in the project are FeCrAl steel strengthened by dispersion of fine oxide particles(FeCrAl-ODS) and silicon carbide (SiC) composite, and are expecting to endure severe accident conditions in the reactor core for a longer period of time than the Zircaloy while maintaining or improving fuel performance during normal operations. In this paper, the progresses of the R&D project are reported.

JAEA Reports

Compilation of the data book on light water reactor benchmark to develop the next version of JENDL; Utilization of criticality data in ICSBEP and IRPhEP open databases

JENDL Committee, Reactor Integral Test Working Group

JAEA-Data/Code 2017-006, 152 Pages, 2017/05


A benchmark database which is devoted to the evaluation of the future JENDL against the criticality of light water reactors was prepared, where the ICSBEP and IRPhEP handbooks by OECD/NEA were utilized effectively. Specific features of this report can be described as follows: (1) The recommendation for benchmarking is based on careful reviewing for the document and related information. Validity of the original benchmark evaluation is carefully checked, and numerical results obtained with JENDL-4.0 are considered. (2) Heterogeneity effect of PuO$$_{2}$$ particles dispersed in fuel medium is consistently quantified for the MOX fuel-loaded experimental data. This precise evaluation is realized by the newly developed finite fuel pin bundle model with the Monte Carlo neutron transport code. (3) Sensitivity analysis is conducted in order to specify nuclear data whose difference between recent nuclear data libraries significantly affects the critical parameter calculation.

Journal Articles

Thermal-hydraulics technological strategy roadmap that improves safety of LWRs

Arai, Kenji*; Umezawa, Shigemitsu*; Oikawa, Hirohide*; Onuki, Akira*; Nakamura, Hideo; Nishi, Yoshihisa*; Fujii, Tadashi*

Nippon Genshiryoku Gakkai-Shi, 58(3), p.161 - 166, 2016/03

no abstracts in English

JAEA Reports

Influence of fuel assembly loading pattern and fuel burnups upon leakage neutron flux spectra from light water reactor core (Joint research)

Kojima, Kensuke; Okumura, Keisuke; Kosako, Kazuaki*; Torii, Kazutaka*

JAEA-Research 2015-019, 90 Pages, 2016/01


At the decommissioning of light water reactors (LWRs), it is important to evaluate an amount of radioactivity in the ex-core structures such as a reactor containment vessel, radiation shieldings, and so on. It is thought that the leakage neutron spectra in these radioactivation regions, which strongly affect the induced radioactivity, would be changed by different reactor core configurations such as fuel assembly loading pattern and fuel burnups. This study was intended to evaluate these effects. For the purpose, firstly, partial neutron currents on the core surfaces were calculated for some core configurations. Then, the leakage neutron flux spectra in major radioactivation regions were calculated based on the provided currents. Finally, influence of the core configurations upon the neutron flux spectra was evaluated. As a result, it has been found that the influence is small on the spectrum shapes of neutron fluxes. However, it is necessary to pay attention to the facts that intensities of the leakage neutron fluxes are changed by the configurations and that intensities and spectrum shapes of the leakage neutron fluxes are changed depending on the angular direction around the core.

Journal Articles

Void reactivity evaluation by modified conversion ratio measurements in LWR critical experiments

Yoshioka, Kenichi*; Kikuchi, Tsukasa*; Gunji, Satoshi*; Kumanomido, Hironori*; Mitsuhashi, Ishi*; Umano, Takuya*; Yamaoka, Mitsuaki*; Okajima, Shigeaki; Fukushima, Masahiro; Nagaya, Yasunobu; et al.

Journal of Nuclear Science and Technology, 52(2), p.282 - 293, 2015/02

 Percentile:100(Nuclear Science & Technology)

We have developed a void reactivity evaluation method by using modified conversion ratio measurements in a light water reactor (LWR) critical lattice. Assembly-wise void reactivity is evaluated from the "finite neutron multiplication factor", $$k^ast$$, deduced from the modified conversion ratio of each fuel rod. The distributions of modified conversion ratio and $$k^ast$$ on a reduced-moderation LWR lattice, for which the improvement of negative void reactivity is a serious issue, were measured. Measured values were analyzed with a continuous-energy Monte Carlo method. The measurements and analyses agreed within the measurement uncertainty. The developed method is useful for validating the nuclear design methodology concerning void reactivity.

Journal Articles

Evaluation of Nuclear Knowledge Management; An Outcome in JAERI

Yanagisawa, Kazuaki

International Journal of Nuclear Knowledge Management, 2(2), p.91 - 104, 2006/00

Taking into consideration of national funds invested to individual research divisions in JAERI during the course of safety project studies, one found that a total input was approximately corresponded to $$4b$$. Qualitatively, JAERI contributed to (1) the governmental policies, (2) the scientific technologies and scholarship, (3) the industries (technology transfer, patents, commissioned research etc.), (4) the local society in the lab location area, and (5) the international cooperation. Quantitatively, JAERI made the creation of added value (cost benefit), cost reduction and reduction of cost loss produced either in direct or indirect manner as research outputs. One found that a total output during the course of nuclear safety project study was approximately $$6b$$ taking into consideration of nuclear market creation at electricity and facilities related to nuclear. This ex post evaluation tells that cost benefit made by nuclear safety project of JAERI is 1.5 ($$6b/4b$$) and such outcome contributes to the increase of general domestic products (GDP).

Journal Articles

Results from studies on high burn-up fuel behavior under LOCA conditions

Nagase, Fumihisa; Fuketa, Toyoshi

NUREG/CP-0192, p.197 - 230, 2005/10

The Japanese regulatory criterion for a loss-of-coolant-accident (LOCA) is based on a threshold of fuel rod fracture during quenching, which was experimentally determined under simulated LOCA conditions. In order to evaluate the fracture threshold of high burn-up fuel rods, JAERI performs integral thermal shock tests simulating LOCA conditions. The tests have been performed with pre-hydrided, unirradiated claddings and high burn-up fuel claddings irradiated to 39 and 44 GWd/t at a PWR. It was shown that fracture/no-fracture threshold primarily depends on the oxidation amount and that the threshold decreases with increases in hydrogen concentration and axial restraint during the quench. It was also shown that fracture conditions of the tested high burn-up fuel claddings are consistent with the fracture threshold derived from unirradiated claddings with similar hydrogen concentrations.

Journal Articles

RANNS code analysis on the local mechanical conditions of cladding of high burnup fuel rods under PCMI in RIA-simulated experiments in NSRR

Suzuki, Motoe; Saito, Hiroaki*; Fuketa, Toyoshi

Proceedings of 2005 Water Reactor Fuel Performance Meeting (CD-ROM), p.579 - 601, 2005/10

The RANNS code analyzes behaviors of a single fuel rod in reactivity-initiated accident (RIA) conditions. The code has two types of mechanical model; one-dimensional deformation model for each axial segment length of rod, and newly-developed two-dimensional local deformation model for one pellet length. Analyses were performed on the two RIA-simulated experiments in the NSRR, OI-10 and OI-11 with high burnup PWR rods, and results of cladding deformation were compared between calculations by the two models and PIE data. RANNS calculated the deformation profiles of claddings during the power transient of the experiments on the basis of the pre-pulse conditions of rods predicted by FEMAXI-6 code. In the calculations by the two-dimensional model, the plastic strain increase at the cladding ridges was compared with those in between the ridges and with the PIE data, and effect of stress variance induced by local non-uniformity of strain on the crack growth was discussed.

Journal Articles

Embrittlement and fracture behavior of pre-hydrided cladding under LOCA conditions

Nagase, Fumihisa; Fuketa, Toyoshi

Proceedings of 2005 Water Reactor Fuel Performance Meeting (CD-ROM), p.668 - 677, 2005/10

A systematic research program on high burnup fuel behavior under LOCA conditions is being conducted at JAERI. As a part of the program, integral thermal shock tests simulating the whole LOCA sequence were conducted with Zircaloy-4 fuel claddings, irradiated to 39 and 44 GWd/t at a PWR, to investigate behavior and condition of cladding fracture during quenching for safety evaluation. Differences were not clearly observed between irradiated and unirradiated claddings at similar hydrogen concentrations in terms of threshold of fracture during quenching, though the threshold is reduced as initial hydrogen concentration increases. Ductility of pre-hydrided, oxidized and quenched claddings was also evaluated by using ring-tensile and ring-compression tests. Embrittlement criteria (zero-ductility limits) from both the tests were lower than the fracture conditions in the integral thermal shock tests. This indicates that loading conditions should be well simulated to evaluate cladding performance under LOCA conditions.

Journal Articles

Investigation on Innovative Water Reactor for Flexible Fuel Cycle (FLWR), 2; Recycle characteristics

Okubo, Tsutomu; Uchikawa, Sadao; Kugo, Teruhiko; Akie, Hiroshi; Takeda, Renzo*

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 6 Pages, 2005/10

In order to ensure sustainable energy supply in the future based on the commercialized LWR technologies, a concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been investigated in JAERI. Results on the FLWR recycling characteristics under possible various reprocessing schemes are presented in the present paper. The results show the recycling is possible a few times at most as long as the fissile Pu content stays over 60%, even in the high conversion type core with the conversion ratio around 0.9, under the simplified PUREX reprocessing, with relatively high average decontamination factor. For breeding core, the results have indicated that even under the reprocessing with relatively low DFs and with whole MA, the recycling is also feasible, suggesting all MAs from the core can be possibly recycled itself, although the core performances are a little degraded depending on MA and FP contents.

Journal Articles

Investigation on Innovative Water Reactor for Flexible Fuel Cycle (FLWR), 1; Conceptual design

Uchikawa, Sadao; Okubo, Tsutomu; Kugo, Teruhiko; Akie, Hiroshi; Nakano, Yoshihiro; Onuki, Akira; Iwamura, Takamichi

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 6 Pages, 2005/10

no abstracts in English

JAEA Reports

Report on the 8th Workshop on the Innovative Water Reactor for Flexible Fuel Cycle; February 10, 2005, Koku-kaikan, Minato-ku, Tokyo

Kobayashi, Noboru; Okubo, Tsutomu; Uchikawa, Sadao

JAERI-Review 2005-029, 119 Pages, 2005/09


The research on Innovative Water Reactor for Flexible fuel cycle (FLWR) has been performed in JAERI for the development of future innovative reactors. The workshop on the FLWRs has been held every year since 1998 aiming at information exchange between JAERI and other organizations. The 8th workshop was held on Feb. 10, 2005 under the joint auspices of JAERI and North Kanto and Kanto-Koetsu branches of Atomic Energy Society of Japan with 75 participants. The workshop began with 3 presentations on FLWRs entitled "Framework and Status of Research and Development on FLWRs", "Long-Term Fuel Cycle Scenarios for Advanced Utilization of Plutonium from LWRs", and "Experiments on Characteristics on Hydrodynamics in Tight-Lattice Core". Then 3 lectures followed: "Development of Evaluation Method for Accuracy in Predicting Neutronics Characteristics of Tight-Lattice Core" by Osaka University, "Development of Cost-Reduced Low-Moderation Spectrum Boiling Water Reactor" by Toshiba Corporation and "Design and Analysis on Super-Critical Water Cooled Power Reactors" by Tokyo University.

JAEA Reports

Measurements of $$^{238}$$U doppler effect in the soft neutron spectra using FCA (Joint research)

Ando, Masaki; Kawasaki, Kenji*; Okajima, Shigeaki; Fukushima, Masahiro; Matsuura, Yutaka*; Kaneko, Yuji*

JAERI-Research 2005-026, 39 Pages, 2005/09


$$^{238}$$U Doppler effect measurements in moderated neutron spectra (uranium fuel and MOX simulated fuel) were carried out using FCA for the purpose of contributing to the improvement in prediction accuracy for Doppler coefficient in LWR. In the mockup cores for MOX fuel, the measurements were performed in different neutron spectra, where the voidage of moderator material was varied systematically. The experimental data were obtained using cylindrical uranium samples with different outer diameter up to 800$$^{circ}$$C. Analyses were performed using a standard code system designed to analyze fast reactor mock-up experiments at FCA with the use of the JENDL-3.2 library. The results of the analyses showed that the calculation accuracy did not depend on the types of the core fuel or the Doppler samples. The calculated values agreed with the experimental ones within the experimental error. Any dependency of the prediction accuracy on the neutron spectra was not observed in the MOX simulated fuel cores.

Journal Articles

RIA- and LOCA-simulating experiments on high burnup LWR fuels

Fuketa, Toyoshi; Nagase, Fumihisa; Sugiyama, Tomoyuki

Proceedings of IAEA Technical Meeting on Fuel Behaviour Modelling under Normal, Transient and Accident Conditions, and High Burnups (CD-ROM), 15 Pages, 2005/09

To provide a data base for the regulatory guide of light water reactors, behaviors of reactor fuels during off-normal and postulated accident conditions such as reactivity-initiated accident (RIA) and loss-of-coolant accident (LOCA) are being studied at the Japan Atomic Energy Research Institute (JAERI). A series of RIA-simulating experiments with high burnup fuel rods is being performed by using pulse-irradiation capability of the Nuclear Safety Research Reactor (NSRR). Fuel behaviors during a LOCA are also examined in an extensive program comprising of integral thermal shock tests and separate tests for oxidation rate and mechanical properties of fuel cladding.

Journal Articles

Rationalization of the fuel integrity and transient criteria for the super LWR

Yamaji, Akifumi*; Oka, Yoshiaki*; Ishiwatari, Yuki*; Liu, J.*; Koshizuka, Seiichi*; Suzuki, Motoe

Proceedings of 2005 International Congress on Advances in Nuclear Power Plants (ICAPP '05) (CD-ROM), 7 Pages, 2005/05

Ensuring the fuel integrities is one of the most fundamental parts in the High Temperature Supercritical-Pressure Light Water Reactor. Most abnormal transient events of SCLWR-H last for a short period of time and the fuel rods are replaced after being irradiated in the core. In this study, the fuel integrity criteria are rationalized based on the fact that the fuel rod mechanical failures can be represented by the strain of the fuel rod cladding. A new fuel rod is designed with a Stainless Steel cladding. It is internally pressurized to reduce the stress on the cladding and also to increase the gap conductance between the pellet and the cladding. The fuel integrities both at normal operation and abnormal transient conditions are evaluated using the fuel analysis code FEMAXI-6 of JAERI.

Journal Articles

Behavior of pre-hydrided Zircaloy-4 cladding under simulated LOCA conditions

Nagase, Fumihisa; Fuketa, Toyoshi

Journal of Nuclear Science and Technology, 42(2), p.209 - 218, 2005/02

 Times Cited Count:25 Percentile:12.44(Nuclear Science & Technology)

Regarding high burn-up fuel behavior under LOCA conditions, LOCA-simulated experiments were performed with unirradiated Zircaloy-4 claddings. Claddings containig 100 to 1450 ppm were isothermally oxidized at at 1220 to 1500 K in steam flow, and quenched by flooding water. Axial shrinkage of the rods during the quench was restrained controlling the maximum restraint load at four different levels. Primarily depending on fraction of the cladding thickness oxidized, the claddings fractured into two pieces during the quench, with circumferential cracking. The fracture/non-fracture threshold as for the oxidized fraction decreases as both initial hydrogen concentration and axial restraint load increase. Consequently, it was shown that the threshold is higher than 20% cladding oxidation, e.g. sufficiently higher than the limit in the Japanese ECCS acceptance criteria, irrespective of hydrogen concentration, when the restraint load is below 535 N.

Journal Articles

Introduction to nuclear fuel engineering, 9; LWR fuel behavior

Fuketa, Toyoshi; Nagase, Fumihisa; Sasahara, Akihiro*

Nippon Genshiryoku Gakkai-Shi, 47(2), p.112 - 119, 2005/02

Behavior of LWR fuel during reactivity-initiated accident (RIA) and loss-of-coolant accident (LOCA) is described.

212 (Records 1-20 displayed on this page)