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Doda, Norihiro; Kato, Shinya; Hamase, Erina; Kuwagaki, Kazuki; Kikuchi, Norihiro; Ohgama, Kazuya; Yoshimura, Kazuo; Yoshikawa, Ryuji; Yokoyama, Kenji; Uwaba, Tomoyuki; et al.
Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.946 - 959, 2023/08
An innovative design system named ARKADIA is being developed to realize the design of advanced nuclear reactors as safe, economical, and sustainable carbon-free energy sources. This paper focuses on ARKADIA-Design for design studies as a part of ARKADIA and introduces representative verification methods for numerical analysis methods of the core design. ARKADIA-Design performs core performance analysis of sodium-cooled fast reactors using a multiphysics approach that combines neutronics, thermal-hydraulics, core mechanics, and fuel pin behavior analysis codes. To confirm the validity of these analysis codes, validation matrices are identified with reference to experimental data and reliable numerical analysis results. The analysis models in these codes and the representative practices for the validation matrices are described.
Tanaka, Masaaki; Miyake, Yasuhiro*; Ezure, Toshiki; Hamase, Erina
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 9 Pages, 2023/05
The numerical analysis model for the computational fluid dynamics (CFD) code for the design study is developed to evaluate the thermal-hydraulics in the core under the core-plenum interaction (CPI) during the decay heat removal using the dipped type direct heat exchanger (D-DHX). To judge the adequacy of the numerical results for a validation study with the sodium experiment results conducted at PLANDTL-2 facility, the degree of difference (DoD) between the numerical and experimental results must be measured by using the area validation metrics (AVM). Through the examinations, the applicability of the AVM and MAVM based on the p-box method was confirmed.
Tanaka, Masaaki; Kudo, Yoshiro*; Nakada, Kotaro*; Koshizuka, Seiichi*
Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.1473 - 1484, 2019/08
Verification and validation (V&V) including uncertainty quantification on modeling and simulation activities has been very much focused on. Due to increase of requirement for standardization of the procedures on the V&V and prediction process to enhance the simulation credibility, "Guideline for Credibility Assessment of Nuclear Simulations (AESJ-SC-A008: 2015)" was published on July 2016 from the AESJ through ten-year discussion. The paper describes brief history of discussion in the AESJ to the publication and introductory explanation of the procedures in the five major elements and one scheme described in the Guideline. And also, a practical experience of the V&V activity according to the fundamental concept indicated in the Guideline is introduced.
Nakamura, Hideo
Nihon Genshiryoku Gakkai-Shi ATOMO
, 61(4), p.270 - 272, 2019/04
no abstracts in English
Tanaka, Masaaki
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 14 Pages, 2018/10
A numerical simulation code MUGTHES has been developed to estimate high cycle thermal fatigue in SFRs. In development of numerical simulation code, verification, validation, and uncertainty quantification (VVUQ) are indispensable. In this study, numerical simulation at impinging jet condition in the WATLON experiment which was the water experiment of a T-junction piping system was performed for the fundamental validation. Based on the previous studies, the simplified least square version GCI method and the area validation metrics were employed as reference methods to quantify uncertainty and to measure the degree of difference between the numerical and the experimental results, respectively. Through the examinations, the potential applicability of the MUGTHES to the thermal striping phenomena was indicated and requirements of modification in the simulation was suggested in accordance with the uncertainty values.
Kikuchi, Norihiro; Doda, Norihiro; Hashimoto, Akihiko*; Yoshikawa, Ryuji; Tanaka, Masaaki; Ohshima, Hiroyuki
Dai-23-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 5 Pages, 2018/06
For the thermal-hydraulic design regarding a fuel assembly of sodium-cooled fast reactors, a subchannel analysis code ASFRE has been developed by JAEA. ASFRE was applied to numerical simulations of several kinds of water and sodium experiments as its validation studies and it was confirmed that pressure drops and temperature distributions measured in the experiments can be well reproduced. To enhance safety of sodium-cooled fast reactor, it is required to evaluate thermal-hydraulics in a core during decay heat removal by natural circulation. It is necessary to estimate radial heat transfer phenomena between fuel assemblies. In this study, a numerical simulation of a 37-pin bundle sodium experiment with radial heat flux was carried out and it was confirmed that ASFRE can be qualitatively reproduced temperature distributions in a fuel assembly affected by radial heat transfer.
Tanaka, Masaaki; Kobayashi, Jun; Nagasawa, Kazuyoshi*
Dai-22-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 4 Pages, 2017/06
In JAEA, a numerical simulation code named MUGTHES which can deal with conjugate heat transfer between the fluid and the structure parts has been developed for estimation of the thermal fatigue issue. In fundamental validation, the benchmark analysis was considered using the experiment of planar triple parallel jet sodium test (PLAJEST). Three specific experimental conditions at Vr=1, 1.56, and 5.56 were employed for the benchmark analyses according to the knowledge in the literatures. Through the benchmarks, applicability of the large eddy simulation (LES) approach with the standard Smagorinsky model in MUGTHES to simulate thermal striping phenomena was potentially confirmed and issues to be modified in the future works were indicated.
Tanaka, Masaaki
Keisan Kogaku Koenkai Rombunshu (CD-ROM), 22, 4 Pages, 2017/05
In the development of the simulation code and the numerical estimation for high cycle thermal fatigue on a structure caused by thermal striping phenomena in sodium cooled fast reactors, implementation of verification and validation (V&V) process is indispensable. A procedure named V2UP (Verification and Validation plus Uncertainty quantification and Prediction) has been made by referring to the existing guidelines regarding the V&V and the methodologies of the safety assessment. In this paper, a challenging installation of quality management procedures into the V2UP procedure is attempted based on the JSCES Standard for "A Model Procedure for Engineering Simulation".
Doda, Norihiro; Ohira, Hiroaki; Kamide, Hideki
Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.1618 - 1625, 2016/04
Sodium-cooled fast reactors have been developed aiming at introducing natural circulation decay heat removal systems by utilizing the characteristic of having a large coolant temperature difference between at the inlet and at the outlet of reactor vessel. In this study, as part of validation for core hot spot evaluation method which is required for adoption of natural circulation decay heat removal systems, an analysis of EBR-II (Experimental Breeder Reactor II) shutdown heat removal test using the method was performed. The results demonstrated that the evaluation method sufficiently predicts the whole plant thermal hydraulic behaviors and the maximum coolant temperature in a fuel subassembly during natural circulation decay heat removal operations.
Munakata, Masahiro; Kimura, Hideo; Tanaka, Tadao; Mukai, Masayuki; Maeda, Toshikatsu; Ogawa, Hiromichi
Nihon Genshiryoku Gakkai Wabun Rombunshi, 2(3), p.361 - 367, 2003/09
no abstracts in English
Shibata, Mitsuhiko; Takase, Kazuyuki; Watanabe, Hironori; Akimoto, Hajime
Fusion Engineering and Design, 63-64, p.217 - 222, 2002/12
Times Cited Count：5 Percentile：34.58(Nuclear Science & Technology)no abstracts in English
Mori, Koji*; Neyama, Atsushi*; Nakagawa, Koichi*
JNC TJ8400 2000-064, 175 Pages, 2000/03
JNC-TJ8400-2000-064.pdf:5.23MB
In this study, the following tasks have been performed in order to evaluate the stability of earthquake resistance for the engineered barrier system(EBS) of High Level Waste (HLW) geological isolation system. (1)validation studies for the liquefaction model. The function of single-phase analysis without interaction between soil and pore water in three-dimensional effective stress analysis code, which had been developed in this study, have been verified using by actual vibration test data. This fiscal year, some validation studies for the function of liquefaction analysis was conducted usig by actual measured data through the laboratory liquefaction test. (2)Supplemental Studies for JNC Second Progress Report. Through the JNC second progress report, it was considered that the stability of earthquake resistance of the engineered barrier system would be maintained under the major seismic event. At the same time we have recognized that several model parameters for joint-crack element, which takes into account for the response behavior of material discontinuous surface such as between overpack and buffer material, will become important in the response behavior of the whole EBS. This year, we have studied about several topics, which arise from technical discussion on JNC second progress report and we have discussed about total seismic stability of EBS. (3)Supplemental Studies for joint study with NRIDP. At this fiscal year, the joint study with National Research Institute for Disaster Prevention (NRIDP) will be final stage. UP to this day, incremental validation studies had been continued using by mesuared data obtained from vibration test. In this final stage, validation analysis has been conducted again using by current version new analysis code and maintained the validation data which will be contribute to the joint study mentioned above.
Cs in the surface soil around the Chernobyl Nuclear Power PlantTakahashi, Tomoyuki; Homma, Toshimitsu
Hoken Butsuri, 34(4), p.365 - 374, 1999/12
no abstracts in English
Takachi, Kazuhiko; Taniguchi, Wataru
JNC TN8400 99-042, 68 Pages, 1999/11
The buffer material is expected to maintain its low water permeability, self-sealing properties, radionuclides adsorption and retardation properties, thermal conductivity, chemical buffering properties, overpack supporting properties, stress buffering properties, etc. over a long period of time. Natural clay is mentioned as a material that can relatively satisfy above. Among the kinds of natural clay, bentonite when compacted is superior because (1)it has exceptionally low water permeability and properties to control the movement of water in buffer, (2)it fills void spaces in the buffer and fractures in the host rock as it swells upon water uptake, (3)it has the ability to exchange cations and to adsorb cationic radioelements. In order to confirm these functions for the purpose of safety assessment, it is necessary to evaluate buffer properties through laboratory tests and engineering-scale tests, and to make assessments based on the ranges in the data obtained. This report describes the procedures, test conditions, results and examinations on the buffer material of dynamic triaxial tests, measurement of elastic wave velocity and liquefaction tests that aim at getting hold of dynamic mechanical properties. MWe can get hold of dependency on the shearing strain of the shearing modulus and hysteresis damping constant, the application for the mechanical model etc. by dynamic triaxial tests, the acceptability of maximum shearing modulus obtained from dynamic triaxial tests etc. by measurement of elastic wave velocity and dynamic strength caused by cyclic stress etc. by liquefaction tests.
Suzudo, Tomoaki; Watanabe, Norio
JAERI-Review 99-027, p.23 - 0, 1999/11
JAERI-Review-99-027.pdf:1.35MB
no abstracts in English
Tomari, Haruo*; *; Shimogori, Kazutoshi*; Wada, Ryutaro*; ; Taniguchi, Naoki
JNC TN8400 99-076, 100 Pages, 1999/10
Effects of bentonite clay, applied potential, pH, of solution and cathodic polarization time on hydrogen absorption into titanium, which is one of the candidate materials of overpack for high-level radioactive waste container, have been investigated in artificial underground water. Considering the result at various test time and assuming the hydrogen absorption is ruled by the paraboric law, the amount of hydrogen after 1000 years exposure calculated to about 17ppm, which will be absorbed at the applied potential of -0.51 vs. SHE corresponds to equilibrium potential of hydrogen. It seems the assumption of the parabolic law and the test period are proper, because the linear relations were obtained between the amount of absorbed hydrogen and the logarithm of the averaged cathodic current and between the slopes of the lines and a square root of the test time. Titanium seems to have a life over 1000 years in deep underground repository according to assumption that about 500ppm absorbed hydrogen is critical for hydrogen embrittlement of titanium.
Yui, Mikazu; Sasamoto, Hiroshi; Randolph C Arthu*
JNC TN8400 99-030, 201 Pages, 1999/07
According to the Japanese program for research and development of high level radioactive waste (HLW) disposal defined by Atomic Energy Commission (AEC), the second progress report (i.e., H-12 report) for performance assessment (PA) of HLW disposal is to be published by the Japan Nuclear Cycle Development Institute (JNC) and submitted to the Japanese government before the year 2,000 (AEC, 1997). This report presents the establishment of generic groundwater chemical compositions for the PA supporting the H-12 report. The following five hypothetical groundwaters are categorized for PA based on the results of the first progress report (i.e., H-3 report) and binaly statistical analyses of the screened groundwater dataset: (1)FRHP(Fresh-Reducing-High-pH) groundwater (2)FRLP(Fresh-Reducing-Low-pH) groundwater (3)SRHP(Saline-Reducing-High-pH) groundwater (4)SRLP(Saline-Reducing-Low-pH) groundwater (5)MRNP(Mixing-Reducing-Neutral-pH) groundwater. In order to define representative groundwater compositions for the PA for the H-12 report, JNC has established the representativeness of the above five hypothetical groundwaters by considering the results of multivariate statistical analyses, data reliability, evidence for geochemical controls on groundwater chemistry and exclusion criteria for potential repository sites in Japan. As a result, the following hypothetical reference groundwaters are selected for the performance assessment analysis in H-12 report, respectively: (1)Reference Case groundwater: FRHP groundwater, and (2)Alternative Geological Environment Case groundwater: SRHP groundwater. In addition, JNC has consulted with overseas experts on the concepts used in groundwater evolution modeling. This modeling effort has focussed on simulating equilibrium water-rock interactions to predict groundwater compositions resulting from reactions between initial water compositions and rock mineral assemblages. These discussions have centered on recommendations for developing ...
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JNC TN4400 99-002, 192 Pages, 1999/03
The tritium transport analysis code, TTT, has been validated using data from the low power test of Monju, and then its behaviour at along term full power operation of Monju in future has been estimated, when the estimated transport and distribution of tritium in the reactor system has been also compared with the result in Joyo and Phenix, which had been already experienced long term operations. The TTT code had been develpped using the tiritium and hydrogen transport model proposed by R. Kumar, ANL, and had been applied to the evaluation in Monju design work. After then, futhermore, the code has been improved using the data from long term operation of Joyo with MK-II core, and in this work the code has been validated for the first time for Monju data. The results from this work are as follows; (1)Comparison of the best fitted tritium source rates from cores in Joyo, Phenix and Monju makes an estimation of the major source from control rods, (2)The calculated tritium concentration in each medium for cooling and its change is a reasonable agreement to the measured, C/E=1.1, (3)The cover gas transport model cosidering isotopic exchange of H and H
can reproduce reasonably the measured concentration distirbution of tritium in sodium and cover gas, (4)The tritium concentration in secondary sodium of Monju was about l/50 times as much as the primary one, which shows the acceraration effect on cold tarapping of tritium due to coprecipitation with permeated hydrogen through Evaporater (EV) heat conduction tube walls. The tritium cold trapping efficiency was estimated to be 1 for coprecipitation with hydrogen and 0.3 for isotopic exchange, respectively, (5)Tritium transport and distribution for along term full power operation of Monju in future was estimated, which could involve a excess factor to 4 at the maximum. The tritium concentration in sodium and Steam Generator (SG) water will be substantially saturated after somthing like 10 years full power operation, ...
Brear, D. J.
PNC TN9410 98-005, 53 Pages, 1998/01
When liquid fuel makes contact with steel structure the liquid can freeze as a crust and the structure can melt at the surface. The melting and freezing processes that occur can influence the mode of fuel freezing and hence fuel relocation. Furthermore the temperature gradients established in the fuel and steel phases determine the rate at which heat is transferred from fuel to steel. In this memo the 1-D transient heat conduction equations are applied to the case of initially liquid UO
brought into contact with solid steel using up-to-date materials properties. The solutions predict criteria for fuel crust formation and steel melting and provide a simple algorithm to determine the interface temperature when one or both of the materials is undergoing phase change. The predicted steel melting criterion is compared with available experimental results.
