Development of numerical simulation method of natural convection around heated porous medium by using JUPITER

Uesawa, Shinichiro; Yamashita, Susumu; Shibata, Mitsuhiko; Yoshida, Hiroyuki

Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 9 Pages, 2023/05

For contaminated water management in decommissioning Fukushima Daiichi Nuclear Power Stations, reduction in water injection, intermittent injection water and air cooling are considered. However, since there are uncertainties of fuel debris in the PCV, it is necessary to examine and evaluate optimal cooling methods according to the distribution state of the fuel debris and the progress of the fuel debris retrieval work in advance. We have developed a method for estimating the thermal behavior in the air cooling, including the influence of the position, heat generation and the porosity of fuel debris. Since a large-scale thermal-hydraulics analysis of natural convection is necessary for the method, JUPITER developed independently by JAEA is used. It is however difficult to perform the large-scale thermal-hydraulics analysis with JUPITER by modeling the internal structure of the debris which may consist of a porous medium. Therefore, it is possible to analyze the heat transfer of the porous medium by adding porous models to JUPITER. In this study, we report the validation of JUPITER applied the porous model and discuss which heat transfer models are most effective in porous models such as series, parallel and geometric mean models. To obtain validation data of JUPITER for the natural convective heat transfer analysis around the porous medium, we performed the heat transfer and the flow visualization experiments of the natural convection in the experimental system including the porous medium. In the comparison between the experiment and the numerical analysis with each model, the numerical result with the geometric mean model was the closest of the models to the experimental results. However, the numerical results of the temperature and the velocity were overestimated for those experimental results. In particular, the temperature near the interface between the porous medium and air was more overestimated.

Validation of free-convective heat transfer analysis with JUPITER to evaluate air-cooling performance of fuel debris in dry method

Uesawa, Shinichiro; Yamashita, Susumu; Shibata, Mitsuhiko; Yoshida, Hiroyuki

Mechanical Engineering Journal (Internet), 5(4), p.18-00115_1 - 18-00115_13, 2018/08

https://doi.org/10.1299/mej.18-00115

Development of a standard database for FBR core nuclear design (XI); Analysis of the experimental fast reactor "JOYO" MK-I start up test and oparation data

; Numata, Kazuyuki*

JNC TN9400 2000-036, 138 Pages, 2000/03

JNC-TN9400-2000-036.pdf:10.16MB

Japan Nuclear Cycle Development lnstitute (JNC) had developed the adjusted nuclear cross-section library in which the results of the JUPITER experiments were renected. Using this adjusted library, the distinct improvement of the accuracy in nuclear design of FBR cores had been achieved. As a recent research, JNC develops a database of other integral data in addition to the JUPITER experiments, aiming at further improvement for accuracy and reliability. ln this report, the authors describe the evaluation of the C/E values and the sensitivity analysis for the Experimental Fast Reactor "JOYO" MK-l core. The minimal criticality, sodium void reactivity worth, fuel assembly worth and burn-up coefficient were analyzed. The results of both the minimal criticality and the fuel assembly worth, which were calculated by the standard analytical method for JUPITER experiments, agreed well with the measured values. 0n the other hand, the results of the sodium void reactivity worth have a tendency to overestimate. As for the burn-up coefficient, it was seen that the C/E values had a dispersion among the operation cycles. The authors judged that further investigation for the estimation of the experimental error will increase the applicability of the integral data to the adjusted library. Furthermore, sensitivity analyses for the minimal criticality, sodium void reactivity worth and fuel assembly worth showed the characteristics of "JOYO" MK-l core in comparison with ZPPR-9 core of JUPITER experiments.

Development of a standard data base for FBR core nuclear design (V); Consistency evaluation of JUPITER experimental analysis

; ; ; Sato, Wakaei*; ; Sanda, Toshio*

PNC TN9410 95-214, 199 Pages, 1995/08

PNC-TN9410-95-214.pdf:9.27MB

In order to improve the design method and accuracy of large fast breeder cores, extensive work has been performed to accumulate and evaluate many kinds of results of fast reactor physics experiments and analyses. As a part of efforts to develop a standard data base for LMFBR core nuclear design, the present report evaluates the physical consistency of JUPITER experimental analysis, especially concentrating on criticality. Here, the judgment of consistency is based on not only the deviation degree of C/E values from unity, but also various viewpoints such as the comparison with other cores or other nuclear characteristics by sensitivity analysis, the effect of changing nuclear data library, the analysis of FCA and JOYO which have completely different source of data from JUPITER, and the use of the Monte Carlo method as an analytical reference. (1)The C/E values of JUPITER criticality are slightly underestimated in the range of 0.993-0.999, using the JFS-3-J2 (1989) group constant set based on JENDL-2 and three-dimensional XYZ transport theory with the most detailed analytical model. There is an obvious dependency of C/Es on reactor core concepts with homogeneous or heterogeneous structure, the main cause of which is considered to be the effect of internal blanket existence and cross-section errors of JFS-3-J2, judged from sensitivity analysis. (2)The latest analytical method and model based on three-dimensional XYZ transport theory has sufficient ability to predict the relative changes of JUPITER criticality caused by the effect of reactor core size, CRP sodium channel, control rod and internal blankets. (3)The analytical error of JUPITER criticality was evaluated as approximately 0.3%dk and this seems reasonable, because the results of Monte Carlo analysis for ZPPR-9 criticality were almost identical with those of our standard analytical method. (4)The analytical results based on the latest JENDL-3.2 library were very close to those of JENDL-2 results, ...

Development of a standard data base for FBR core nuclear design; Analysis of JUPITER-I Experiments by the latest method

PNC TN9410 92-278, 347 Pages, 1992/09

PNC-TN9410-92-278.pdf:7.93MB

A series of critical experiments for conventional two-zone homogeneous cores of 6 to 8 MWe-class LMFBR, JUPITER phase-I, were analyzed and evaluated using the latest analytical method, which had been established from the preceding numerous studies on fast reactor physics. The present work is a part of efforts to develop a standard data base for LMFBR core nuclear design at PNC. The analytical method and results are summarized as follows: (1)Analytical method (a)Nuclear data : 70-group fast reactor constant set JFS-3-J2(1989 edition) based on the Japanese Evaluated Nudear Data Library, version 2 (JENDL-2). (b)Cell calculation : plate stretch model, cell heterogeneity treatment by Tone's method and transport cross-sections weighted with neutron current. (c)Base core calculation : 18-group, three-dimensiona1 XYZ diffusion theory and Benoist's anisotropic diffusion coefficients. (d)Correction calculation : three-dimensional transport effect, mesh size effect, cell asymmetric effect and all master model effect etc. (2)Analytical results (a)The C.E (calculation/experiment) values of criticality agree quite well among seven cores (ZPPR-910D/2) and do not depend on the core volume or the number of control rod positions (CRP). (b)The C/E values of control rod worths increase gradually from the core center to the core edge positions in each core (511%). Those of reaction rate distributions also indicate similar spatial variations (25%), which is considered to be consistent with the C/E tendency of control rod worths. (c)The reaction rate ratios of C28/F49 and F25/49 give quite stable C/E values of 1.06 and 1.03, respectively. (d)The C/E values of sodium void reactivities are overestimated by +25% at core center region. The C/E dependence on void region size, which was pointed out in the past analyses, is found in the ZPPR-9 core, but not in the ZPPR-10 series. (e)The C/E dependence of 4% on the radial positions were found in sample ...

Development of advanced neutronics/thermal-hydraulics coupling simulation system, 3; Development of a prototype simulation system

Tada, Kenichi; Akie, Hiroshi; Ono, Ayako; Nagaya, Yasunobu; Yoshida, Hiroyuki; Kawanishi, Tomohiro

no journal, , 

JAEA has started to develop the advanced neutronics/thermal-hydraulics coupling simulation system for improvement of the light water reactor analysis and safety. We developed a prototype simulation system to find the issues of coupling simulation and to investigate the optimum mesh size for neutronics and thermal-hydraulics analysis. This presentation explains the overview of the prototype simulation system and its calculations results.

Development of numerical simulation method for relocation behavior of molten materials in nuclear reactors, 12; Measurement of the behavior of flow down liquid film

Hihara, Yutaro*; Monji, Hideaki*; Abe, Yutaka*; Yamashita, Susumu; Yoshida, Hiroyuki

no journal, , 

no abstracts in English

Boiling simulation in 88 single bundle assembly of BWR

Kamiya, Tomohiro; Ono, Ayako; Nagatake, Taku; Tada, Kenichi; Kondo, Ryoichi; Nagaya, Yasunobu; Yoshida, Hiroyuki

no journal, , 

JAEA aims to obtain reference solutions for reactor design codes by coupling the Monte Carlo code MVP and the multiphase and multi-component detailed thermal-hydraulic analysis code JUPITER on the multiphysics platform JAMPAN (JAEA Advanced Multi-Physics Analysis platform for Nuclear systems). For BWR, the thermal-hydraulic analysis code is required to consider boiling around fuel rods. Therefore, a thermal-hydraulic simulation of an 88 STEP-II single fuel assembly system was performed considering boiling using the temperature recovery method.