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Journal Articles

Study on mechanism and threshold conditions for fuel fragmentation during loss-of-coolant accident conditions

Narukawa, Takafumi; Udagawa, Yutaka

Proceedings of TopFuel 2021 (Internet), 10 Pages, 2021/10

JAEA Reports

Report of summer holiday practical training 2020; Feasibility study on nuclear battery using HTTR core; Feasibility study for nuclear design, 3

Ishitsuka, Etsuo; Mitsui, Wataru*; Yamamoto, Yudai*; Nakagawa, Kyoichi*; Ho, H. Q.; Ishii, Toshiaki; Hamamoto, Shimpei; Nagasumi, Satoru; Takamatsu, Kuniyoshi; Kenzhina, I.*; et al.

JAEA-Technology 2021-016, 16 Pages, 2021/09

JAEA-Technology-2021-016.pdf:1.8MB

As a summer holiday practical training 2020, the feasibility study for nuclear design of a nuclear battery using HTTR core was carried out, and the downsizing of reactor core were studied by the MVP-BURN. As a result, it is clear that a 1.6 m radius reactor core, containing 54 (18$$times$$3 layers) fuel blocks with 20% enrichment of $$^{235}$$U, and BeO neutron reflector, could operate continuously for 30 years with thermal power of 5 MW. Number of fuel blocks of this compact core is 36% of the HTTR core. As a next step, the further downsizing of core by changing materials of the fuel block will be studied.

JAEA Reports

HTTR burnup characteristic analysis with detailed axial burning region using MVP-BURN

Ikeda, Reiji*; Ho, H. Q.; Nagasumi, Satoru; Ishii, Toshiaki; Hamamoto, Shimpei; Nakano, Yumi*; Ishitsuka, Etsuo; Fujimoto, Nozomu*

JAEA-Technology 2021-015, 32 Pages, 2021/09

JAEA-Technology-2021-015.pdf:2.74MB

Burnup calculation of the HTTR considering temperature distribution and detailed burning regions was carried out using MVP-BURN code. The results show that the difference in k$$_{rm eff}$$, as well as the difference in average density of some main isotopes, is insignificant between the cases of uniform temperature and detailed temperature distribution. However, the difference in local density is noticeable, being 6% and 8% for $$^{235}$$U and $$^{239}$$Pu, respectively, and even 30% for the burnable poison $$^{10}$$B. Regarding the division of burning regions to more detail, the change of k$$_{rm eff}$$ is also small of 0.6%$$Delta$$k/k or less. The small burning region gives a detailed distribution of isotopes such as $$^{235}$$U, $$^{239}$$Pu, and $$^{10}$$B. As a result, the effect of graphite reflector and the burnup behavior could be evaluated more clearly compared with the previous study.

Journal Articles

Concepts and basic designs of various nuclear fuels, 5; Fuels for high temperature gas-cooled reactor and molten salt reactor

Ueta, Shohei; Sasaki, Koei; Arita, Yuji*

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 63(8), p.615 - 620, 2021/08

no abstracts in English

JAEA Reports

Improvement of intragranular fission gas behavior model for fuel performance code FEMAXI-8

Udagawa, Yutaka; Tasaki, Yudai

JAEA-Data/Code 2021-007, 56 Pages, 2021/07

JAEA-Data-Code-2021-007.pdf:5.05MB

Japan Atomic Energy Agency (JAEA) has released FEMAXI-8 in 2019 as the latest version of the fuel performance code FEMAXI, which has been developed to analyze thermal and mechanical behaviors of a single fuel rod in mainly normal operation conditions and anticipated transient conditions. This report summarizes a newly developed model to analyze intragranular fission gas behaviors considering size distribution of gas bubbles and their dynamics. While the intragranular fission gas behavior models implemented in the previous FEMAXI versions have supported only treating single bubble size for a given fuel element, the new model supports multiple gas groups according to their size and treats their dynamic behaviors, making the code more versatile. The model was first implemented as a general module that takes arbitrary number of bubble groups and spatial mesh division for a given fuel grain system. An interface module to connect the model to FEMAXI-8 was then developed so that it works as a 2 bubble groups model, which is the minimum configuration of the multi-grouped model to be operated with FEMAXI-8 at the minimum calculation cost. FEMAXI-8 with the new intragranular model was subjected to a systematic validation calculation against 144 irradiation test cases and recommended values for model parameters were determined so that the code makes reasonable predictions in terms of fuel center temperature, fission gas release, etc. under steady-state and power ramp conditions.

Journal Articles

Preparation for restarting the high temperature engineering test reactor; Development of utility tool for auto seeking critical control rod position

Ho, H. Q.; Fujimoto, Nozomu*; Hamamoto, Shimpei; Nagasumi, Satoru; Goto, Minoru; Ishitsuka, Etsuo

Nuclear Engineering and Design, 377, p.111161_1 - 111161_9, 2021/06

 Times Cited Count:1 Percentile:81.22(Nuclear Science & Technology)

JAEA Reports

Technical basis of ECCS acceptance criteria for light-water reactors and applicability to high burnup fuel

Nagase, Fumihisa; Narukawa, Takafumi; Amaya, Masaki

JAEA-Review 2020-076, 129 Pages, 2021/03

JAEA-Review-2020-076.pdf:3.9MB

Each light-water reactor (LWR) is equipped with the Emergency Core Cooling System (ECCS) to maintain the coolability of the reactor core and to suppress the release of radioactive fission products to the environment even in a loss-of-coolant accident (LOCA) caused by breaks in the reactor coolant pressure boundary. The acceptance criteria for ECCS have been established in order to evaluate the ECCS performance and confirm the sufficient safety margin in the evaluation. The limits defined in the criteria were determined in 1975 and reviewed based on state-of-the-art knowledge in 1981. Though the fuel burnup extension and necessary improvements of cladding materials and fuel design have been conducted, the criteria have not been reviewed since then. Meanwhile, much technical knowledge has been accumulated regarding the behavior of high-burnup fuel during LOCAs and the applicability of the criteria to the high-burnup fuel. This report provides a comprehensive review of the history and technical bases of the current criteria and summarizes state-of-the-art technical findings regarding the fuel behavior during LOCAs. The applicability of the current criteria to the high-burnup fuel is also discussed.

Journal Articles

Benchmarks of depletion and decay heat calculation between MENDEL and MARBLE

Yokoyama, Kenji; Lahaye, S.*

Proceedings of Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo 2020 (SNA + MC 2020), p.109 - 116, 2020/10

CEA/DEN/DM2S/SERMA and JAEA/NSEC are working on benchmarks for burnup, isotopic concentrations and decay heat calculations in the collaboration framework between both organisms. Both actors of this benchmark are independently developing their own simulation code systems for computing quantities of interest in nuclear fuel cycle domain: MENDEL in CEA and MARBLE in JAEA. The purpose of the benchmark is to verify each system by comparing both calculation results on specific applications. MENDEL uses a several solvers for the resolution of Bateman equation. Runge-Kutta method or Chebyshev Rational Approximation method (CRAM) are used for irradiation computations. An analytical solver can also be used for decay calculations. MARBLE can use Krylov subspace method or CRAM method. As the first phase of the benchmark, we compared the calculated results of decay heat and isotropic concentrations following by a Pu-239 fast fission pulse. We applied nuclear data from three libraries: (1) JEFF-3.1.1, (2) JENDL/DDF-2015 + JENDL/FPY-2011, and (3) ENDF/B-VII.1. Nuclear data and burnup chain were generated from these libraries independently on each system. We confirmed that the results for both systems were in very good agreement with each other. Numerical results were also compared to experimental data. As the second phase of the benchmark, we are proceeding with a burnup calculation benchmark of MENDEL and MARBLE using the nuclear data and burnup chain provided by ORLIBJ33, which is a set of cross-section data based on JENDL-3.3 for ORIGEN-2 code system. We will also compare with calculation results by the ORIGEN-2 code with ORLIBJ33. Since the series of ORLIB, that is, ORLIBJ32, ORLIBJ33, and ORLIBJ40, have been widely used especially in Japan for many years, the comparison with ORLIB is effective for confirming the performance of MENDEL and MARBLE.

JAEA Reports

Report of summer holiday practical training 2019; Feasibility study on nuclear battery using HTTR core; Feasibility study for nuclear design, 2

Ishitsuka, Etsuo; Nakashima, Koki*; Nakagawa, Naoki*; Ho, H. Q.; Ishii, Toshiaki; Hamamoto, Shimpei; Takamatsu, Kuniyoshi; Kenzhina, I.*; Chikhray, Y.*; Matsuura, Hideaki*; et al.

JAEA-Technology 2020-008, 16 Pages, 2020/08

JAEA-Technology-2020-008.pdf:2.98MB

As a summer holiday practical training 2019, the feasibility study for nuclear design of a nuclear battery using HTTR core was carried out, and the $$^{235}$$U enrichment and burnable poison of the fuel, which enables continuous operation for 30 years with thermal power of 5 MW, were studied by the MVP-BURN. As a result, it is clear that a fuel with $$^{235}$$U enrichment of 12%, radius of burnable poison and natural boron concentration of 1.5 cm and 2wt% are required. As a next step, the downsizing of core will be studied.

Journal Articles

Four-point-bend tests on high-burnup advanced fuel cladding tubes after exposure to simulated LOCA conditions

Narukawa, Takafumi; Amaya, Masaki

Journal of Nuclear Science and Technology, 57(7), p.782 - 791, 2020/07

 Times Cited Count:2 Percentile:62.87(Nuclear Science & Technology)

Journal Articles

Research and development on high burnup HTGR fuels in JAEA

Ueta, Shohei; Mizuta, Naoki; Sasaki, Koei; Sakaba, Nariaki; Ohashi, Hirofumi; Yan, X.

Mechanical Engineering Journal (Internet), 7(3), p.19-00571_1 - 19-00571_12, 2020/06

JAEA has been progressing to design HTGR fuels for not only small-type practical HTGRs but also VHTR proposed in GIF which can be utilized for various purposes with high-temperature heat at 750 to 950 $$^{circ}$$C. To increase economy of these HTGRs, JAEA has been upgrading the design method for the HTGR fuel, which can maintain their integrities at the burnup of three to four times higher than that of the conventional HTTR fuel. Design principles and specifications of various concepts of the high burnup HTGR fuels designed by JAEA are reported. As the latest results on post-irradiation examinations of the high burnup HTGR fuel progressing in a framework of international collaboration with Kazakhstan, irradiation shrinkage rate of the fuel compact as a function of fast neutron fluence was obtained at around 100 GWd/t. Furthermore, the future R&Ds needed for the high burnup HTGR fuel are described based on these experimental results.

Journal Articles

Fracture limit of high-burnup advanced fuel cladding tubes under loss-of-coolant accident conditions

Narukawa, Takafumi; Amaya, Masaki

Journal of Nuclear Science and Technology, 57(1), p.68 - 78, 2020/01

 Times Cited Count:1 Percentile:23.13(Nuclear Science & Technology)

Journal Articles

Thresholds for failure of high-burnup LWR fuels by pellet cladding mechanical interaction under reactivity-initiated accident conditions

Udagawa, Yutaka; Sugiyama, Tomoyuki; Amaya, Masaki

Journal of Nuclear Science and Technology, 56(12), p.1063 - 1072, 2019/12

 Times Cited Count:3 Percentile:57.07(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Behavior of high-burnup advanced LWR fuel cladding tubes under LOCA conditions

Narukawa, Takafumi; Amaya, Masaki

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.912 - 921, 2019/09

Journal Articles

Behavior of high-burnup LWR-MOX fuel under a reactivity-initiated accident condition

Taniguchi, Yoshinori; Udagawa, Yutaka; Mihara, Takeshi; Amaya, Masaki; Kakiuchi, Kazuo

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.551 - 558, 2019/09

Journal Articles

Cross-section-induced uncertainty evaluation of MA sample irradiation test calculations with consideration of dosimeter data

Sugino, Kazuteru; Numata, Kazuyuki*; Ishikawa, Makoto; Takeda, Toshikazu*

Annals of Nuclear Energy, 130, p.118 - 123, 2019/08

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

In MA sample irradiation test data calculations, the neutron fluence during irradiation period is generally scaled by using dosimetry data in order to improve calculation accuracy. In such a case, appropriate correction is required to burnup sensitivity coefficients obtained by the conventional generalized perturbation theory because some cancellations occur in the burnup sensitivity coefficients. Therefore, a new formula for the burnup sensitivity coefficient has been derived with the consideration of the neutron fluence scaling effect (NFS). In addition, the cross-section-induced uncertainty is evaluated by using the obtained burnup sensitivity coefficients and the covariance data based on the JENDL-4.0.

Journal Articles

Oxidation behavior of high-burnup advanced fuel cladding tubes in high-temperature steam

Narukawa, Takafumi; Amaya, Masaki

Journal of Nuclear Science and Technology, 56(7), p.650 - 660, 2019/07

 Times Cited Count:6 Percentile:81.33(Nuclear Science & Technology)

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

 Times Cited Count:3 Percentile:57.07(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Research and development on high burnup HTGR fuels in JAEA

Ueta, Shohei; Mizuta, Naoki; Sasaki, Koei; Sakaba, Nariaki; Ohashi, Hirofumi; Yan, X.

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05

JAEA has been progressing to design HTGR fuels for not only small-type practical HTGRs but also VHTR proposed in GIF which can be utilized for various purposes with high-temperature heat at 750 to 950 $$^{circ}$$C. To increase economy of these HTGRs, JAEA has been upgrading the design method for the HTGR fuel, which can maintain their integrities at the burnup of three to four times higher than that of the conventional HTTR fuel. Design principles and specifications of various concepts of the high burnup HTGR fuels designed by JAEA are reported. As the latest results on post-irradiation examinations of the high burnup HTGR fuel progressing in a framework of international collaboration with Kazakhstan, irradiation shrinkage rate of the fuel compact as a function of fast neutron fluence was obtained at around 100 GWd/thm. Furthermore, the future R&Ds needed for the high burnup HTGR fuel are described based on these experimental results.

209 (Records 1-20 displayed on this page)