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.

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

https://doi.org/10.1299/jsmeicone.2019.27.2318

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.

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.

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.

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.

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.

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)(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.

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

https://doi.org/10.1088/0029-5515/41/3/311

no abstracts in English

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

https://doi.org/10.1016/S0920-3796(00)00193-9

no abstracts in English

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

https://doi.org/10.1016/S0920-3796(99)00079-4

no abstracts in English