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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.

Journal Articles

Preliminary analysis of sodium experimental apparatus PLANDTL-2 for development of evaluation method for thermal-hydraulics in reactor vessel of sodium fast reactor under decay heat removal system operation condition

Ono, Ayako; Tanaka, Masaaki; Miyake, Yasuhiro*; Hamase, Erina; Ezure, Toshiki

Mechanical Engineering Journal (Internet), 7(3), p.19-00546_1 - 19-00546_11, 2020/06

Fully natural circulation decay heat removal systems (DHRSs) are to be adopted for sodium fast reactors, which is a passive safety feature without any electrical pumps. It is required to grasp the thermal-hydraulic phenomena in the reactor vessel and evaluate the coolability of the core under the natural circulation not only for the normal operating condition but also for severe accident conditions. In this paper, the numerical results of the preliminary analysis for the sodium experimental condition with the PLANDTL-2 are discussed to establish an appropriate numerical models for the reactor core including the gap region among the subassemblies and the DHX. From these preliminary analyses, the characteristics of the thermal-hydraulics behavior in the PLANDTL-2 to be focused are extracted.

Journal Articles

Preliminary analysis of sodium experimental apparatus PLANDTL-2 for development of evaluation method for thermal hydraulics in reactor vessel of sodium fast reactor under decay heat removal system operation condition

Ono, Ayako; Tanaka, Masaaki; Miyake, Yasuhiro*; Hamase, Erina; Ezure, Toshiki

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 7 Pages, 2019/05

Decay heat removal system (DHRS) by using the natural circulation without depending on the pump as the mechanical equipment is recognized as one of the most effective methodologies for the sodium-cooled fast reactor from the viewpoint of the safety enhancement. In this paper, the numerical simulation results of the preliminary analysis for the sodium experiment with the apparatus of PLANDTL-2, in which the core and the upper plenum with a dipped-type direct heat exchanger (DHX) were modeled, were discussed, in order to establish appropriate numerical models for the reactor core including the gap region among the subassemblies and the DHX.

Journal Articles

Development of probabilistic risk assessment methodology of decay heat removal function against combination hazard of low temperature and snow for sodium-cooled fast reactors

Nishino, Hiroyuki; Yamano, Hidemasa; Kurisaka, Kenichi

Mechanical Engineering Journal (Internet), 5(4), p.18-00079_1 - 18-00079_17, 2018/08

Journal Articles

ASTRID nuclear island design; Update in French-Japanese joint team development of decay heat removal systems

Hourcade, E.*; Mihara, Takatsugu; Dauphin, A.*; Dirat, J.-F.*; Ide, Akihiro*

Proceedings of 2018 International Congress on Advances in Nuclear Power Plants (ICAPP 2018) (CD-ROM), p.556 - 561, 2018/04

In the framework of the French-Japanese agreement signed in 2014, CEA, AREVA NP, JAEA, and MHI/MFBR is jointly performing components design of ASTRID such as Decay Heat Removal Systems (DHRS). This paper is giving an update concerning ASTRID DHR strategy with description of reference architecture evolution and project objectives. In particular, new developments were made for DHR during normal shutdown and role of Ex-Vessel system. A special focus is made on design process of automatic shutter to hydraulically connect Hot Plenum and cold plenum to enhance primary vessel natural convection.

JAEA Reports

Applicability confirmation test of optimum decay heat evaluation method for HTGR with HTTR (Non-nuclear heating test); Validation of residual heat evaluation model

Honda, Yuki; Inaba, Yoshitomo; Nakagawa, Shigeaki; Yamazaki, Kazunori; Kobayashi, Shoichi; Aono, Tetsuya; Shibata, Taiju; Ishitsuka, Etsuo

JAEA-Technology 2017-013, 20 Pages, 2017/06

JAEA-Technology-2017-013.pdf:2.52MB

Decay heat is one of an important factor for a safety evaluation of depressurized loss-of-forced cooling accident, a representative high consequence accident, in high temperature gas-cooled reactor (HTGR). Traditionally, a conservative decay heat curve is used for safety analysis according to the regulatory standards. On the other hand, there is growing interest in obtaining test data related to decay heat for the use of uncertainty analysis. However, such data has not been obtained for prismatic-type HTGR. Therefore, we have launched a test program to obtain the decay heat data from the HTTR. As an initial step, an applicability confirmation test of decay heat evaluation method for HTGR was conducted in February 2017 without non-nuclear heating condition. This report introduces an estimation method for the decay heat based on test data using HTTR and shows the results of validation of the reactor residual heat evaluation method which will be used to obtain the decay heat data based on test data.

Journal Articles

Numerical analysis of EBR-II shutdown heat removal test-17 using 1D plant dynamic analysis code coupled with 3D CFD code

Doda, Norihiro; Hiyama, Tomoyuki; Tanaka, Masaaki; Ohshima, Hiroyuki; Thomas, J.*; Vilim, R. B.*

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 10 Pages, 2017/06

In sodium-cooled fast reactors, a natural circulation is expected to remove the core decay heat when the plant gets into a station blackout. From a perspective of reactor safety, the core hot spot temperature arising in the natural circulation should be evaluated accurately. To this end, Japan Atomic Energy Agency is trying to couple a 1-D plant dynamics analysis code Super-COPD and a 3-D CFD code AQUA to solve the thermal-hydraulic field in the whole plant under natural circulation condition. As a validation study, the coupled code was applied to an analysis of EBR-II shutdown heat removal test. The obtained numerical results reasonably agreed with the measured data, which demonstrated the validity of the coupled code.

Journal Articles

Study on reactor vessel coolability of sodium-cooled fast reactor under severe accident condition; Water experiments using a scale model

Ono, Ayako; Kurihara, Akikazu; Tanaka, Masaaki; Ohshima, Hiroyuki; Kamide, Hideki; Miyake, Yasuhiro*; Ito, Masami*; Nakane, Shigeru*

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04

The water experiment apparatus simulating the thermal hydraulics in a reactor vessel under operating the decay heat removal systems (DHRSs) was fabricated. The theoretical evaluation for similarity and results of basic experiments show applicability for a scale model experiment of a sodium-cooled fast reactor. This paper, moreover, describes the results of flow visualization experiment under operating a dipped-type passive DHX, which is planned to be installed in both a loop type reactor and pool type reactor, and the calculation results using FLUENT comparing with the result of water experiment.

Journal Articles

Validation of decay heat evaluation method based on FPGS cord for fast reactor spent MOX fuels

Usami, Shin; Kishimoto, Yasufumi; Taninaka, Hiroshi; Maeda, Shigetaka

Proceedings of International Conference on the Physics of Reactors; Unifying Theory and Experiments in the 21st Century (PHYSOR 2016) (USB Flash Drive), p.3263 - 3274, 2016/05

The present paper describes the validation of the new decay heat evaluation method using FPGS90 code with both the updated nuclear data library and the rational extent of uncertainty, by comparing the results of the decay heat measurement of the spent fuel subassemblies in Joyo MK-II core and by comparing with the calculation results of ORIGEN2.2 code. The calculated values of decay heat (C) by FPGS90 based on the JENDL-4.0 library were coincident with the measured ones (E) within the calculation uncertainties, and the C/E ranged from 1.01 to 0.93. FPGS90 evaluated the decay heat almost 3% larger than ORIGEN2.2, and it improved the C/E in comparison with the ORIGEN2.2 code. Furthermore, The C/E by FPGS90 based on the JENDL-4.0 library was improved than that based on the JENDL-3.2 library, and the contribution of the revision of reaction cross section library to the improvement was dominant rather than that of the decay data and fission yield data libraries.

JAEA Reports

Calculation of decay heat by new ORIGEN libraries for high temperature engineering test reactor

Simanullang, I. L.*; Honda, Yuki; Fukaya, Yuji; Goto, Minoru; Shimazaki, Yosuke; Fujimoto, Nozomu*; Takada, Shoji

JAEA-Technology 2015-032, 26 Pages, 2016/01

JAEA-Technology-2015-032.pdf:2.07MB

Decay heat of the High Temperature Engineering Test Reactor had been evaluated by the Shure Equation and/or ORIGEN code based on the LWR's data. However, to evaluate more accurately, a suitable method must be considered because of the differences neutron spectrums from the LWRs. Therefore, the decay heat and the generated nuclides for the neutron spectrums of the core with different graphite moderator amount were calculated by the ORIGEN2 code. As a result, it is clear that the calculated decay heats are similar value with LWRs for about one year after the reactor shutdown, and that the significant differences are observed on the longer period affected by the generated nuclides such as $$^{90}$$Y, $$^{134}$$Cs, $$^{144}$$Pr, $$^{106}$$Rh, $$^{241}$$Am etc. It is also clear that the dose is affected by $$^{241}$$Pu on the initial stage after the reactor shutdown.

Journal Articles

Analysis of natural circulation tests in the experimental fast reactor JOYO

Nabeshima, Kunihiko; Doda, Norihiro; Ohshima, Hiroyuki; Mori, Takero; Ohira, Hiroaki; Iwasaki, Takashi*

Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.1041 - 1049, 2015/08

Natural circulation is one of the most important mechanisms to remove decay heat in the sodium cooled fast reactors from the viewpoint of passive safety. On the other hand, it is difficult to evaluate plant dynamics accurately under low flow natural circulation condition. In this study, Super-COPD has been validated through the application to the analysis of natural circulation tests in the experimental fast reactor JOYO. Almost all plant components in JOYO including four air-coolers were modeled in Super COPD. Furthermore, the full scale modeling of fuel subassembly was also adopted in this analysis. The natural circulation test after reactor scram from 100 MW full power at JOYO was selected and simulated by Super-COPD. The transient behaviors predicted by Super-COPD showed good agreement with the experimental data.

Journal Articles

Radioactivity of the vanadium-alloy induced by D-T neutron irradiation

Sato, Satoshi; Tanaka, Teruya*; Hori, Junichi; Ochiai, Kentaro; Nishitani, Takeo; Muroga, Takeo*

Journal of Nuclear Materials, 329-333(Part2), p.1648 - 1652, 2004/08

 Times Cited Count:2 Percentile:17.98(Materials Science, Multidisciplinary)

no abstracts in English

JAEA Reports

Reliability assessment of high energy particle induced radioactivity calculation code DCHAIN-SP 2001 by analysis of integral activation experiments with 14MeV neutrons

Kai, Tetsuya; Maekawa, Fujio; Kasugai, Yoshimi; Kosako, Kazuaki*; Takada, Hiroshi; Ikeda, Yujiro

JAERI-Research 2002-005, 65 Pages, 2002/03

JAERI-Research-2002-005.pdf:2.75MB

Reliability assessment for the High Energy Particle Induced Radioactivity Calculation Code DCHAIN-SP 2001 was carried out through analysis of integral activation experiments with 14-MeV neutrons. The following three series of experiments conducted at the D-T neutron source facility, FNS, in JAERI were employed: (1) the decay gamma-ray measurement experiment for fusion reactor materials, (2) the decay heat measurement experiment for 32 fusion reactor materials, and (3) the integral activation experiment on mercury.As a result, it was found that the calculations with DCHAIN-SP 2001 predicted the experimental data for (1)$$sim$$(3) approximately within 30%, 20% and 20%, respectively. It was concluded that the activation cross section data below 20 MeV and the associated decay data as well as the calculation algorithm for solving the Beteman equation that was the master equation of DCHAIN-SP were adequate.

Journal Articles

DORT analyses of decay heat experiment on tungsten for ITER

Konno, Chikara; Maekawa, Fujio; Wada, Masayuki*; Ikeda, Yujiro; Takeuchi, Hiroshi

Fusion Engineering and Design, 58-59, p.961 - 965, 2001/11

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

no abstracts in English

Journal Articles

Neutronics experiments for ITER at JAERI/FNS

Konno, Chikara; Maekawa, Fujio; Kasugai, Yoshimi; Uno, Yoshitomo; Kaneko, Junichi; Nishitani, Takeo; Wada, Masayuki*; Ikeda, Yujiro; Takeuchi, Hiroshi

Nuclear Fusion, 41(3), p.333 - 337, 2001/03

 Times Cited Count:3 Percentile:11.53(Physics, Fluids & Plasmas)

no abstracts in English

Journal Articles

Experimental validation of decay heat calculation codes and associated nuclear data libraries for fusion energy

Maekawa, Fujio; Wada, Masayuki*; Ikeda, Yujiro

Journal of Nuclear Science and Technology, 38(1), p.53 - 62, 2001/01

no abstracts in English

Journal Articles

Decay heat experiment featuring low-energy neutron induced tungsten-187 production in ITER baffle plates and its analysis

Maekawa, Fujio; Wada, Masayuki*; Konno, Chikara; Kasugai, Yoshimi; Ikeda, Yujiro

Fusion Engineering and Design, 51-52(Part.B), p.809 - 814, 2000/11

 Times Cited Count:6 Percentile:44.06(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Development of whole energy absorption spectrometer for decay heat measurement

Maekawa, Fujio; Wada, Masayuki*; Ikeda, Yujiro

Nuclear Instruments and Methods in Physics Research A, 450(2-3), p.467 - 478, 2000/08

 Times Cited Count:6 Percentile:45.19(Instruments & Instrumentation)

no abstracts in English

Journal Articles

Decay heat experiment on thirty-two fusion reactor relevant materials irradiated by 14-MeV neutrons

Maekawa, Fujio; Ikeda, Yujiro

Fusion Engineering and Design, 47(4), p.377 - 388, 2000/01

 Times Cited Count:11 Percentile:61.56(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Decay heat experiment and validation of calculation code systems for fusion reactor

Maekawa, Fujio; Wada, Masayuki*; Ikeda, Yujiro

JAERI-Research 99-055, p.188 - 0, 1999/10

JAERI-Research-99-055.pdf:8.59MB

no abstracts in English

66 (Records 1-20 displayed on this page)