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Journal Articles

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.

Journal Articles

Investigation of the directions of special drawers loaded in ZPPR-9, -10A, -10B, and -10C of JUPITER experiments

Yokoyama, Kenji; Ishikawa, Makoto*

International Handbook of Evaluated Reactor Physics Benchmark Experiments (CD-ROM), p.ZPPR-LMFR-EXP-001, 002, 005, 006 - Appendix M, D, G, G, 2023/00

In the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP) edited by OECD/NEA, nine ZPPR experimental cores named ZPPR-9, 10A, 10B, 10C, 13A, 17A, 18A, 18C, and 19A, which are mock-up critical experiments for sodium-cooled MOX-fueled fast reactors in the JUPITER cooperative program, are included as reactor physics benchmarks. In order to build the as-built modeling of a ZPPR experimental core, the "all master model (AMM)" map and the drawer masters are used which are attached in the ZPPR benchmark as EXCEL files. Although these materials are usually enough to reproduce a complete ZPPR as-build modeling, there is one exception for several ZPPR cores with respect to special narrow drawers called "poison safety rods (PSR)" or "shim rods." These narrow drawers have a void region in the half part of the drawer to insert absorber materials in case of emergent shutdown or criticality adjustment. To get the exact direction of the narrow drawers, the direction of the void region (Right or Left, hereafter, R or L) must be designated in the AMM maps and the drawer masters. Unfortunately, those data of JUPITER-I series (ZPPR-9, 10A, 10B and 10C) did not distinguish the direction of the narrow drawers. JAEA has surveyed the narrow drawer directions of these four ZPPR cores from the original experimental core maps and the loading records which were made by ANL experimenters. New appendices for the four ZPPR core benchmarks are prepared as EXCEL files to summarize the narrow drawer directions obtained by JAEA survey, to make it possible to build the as-built modeling of the ZPPR experimental cores.

Journal Articles

Validation and verification for the melting and eutectic models in JUPITER code

Chai, P.; Yamashita, Susumu; Nagae, Yuji; Kurata, Masaki

Proceedings of 9th Conference on Severe Accident Research (ERMSAR 2019) (Internet), 14 Pages, 2019/03

In order to obtain a precise understanding of molten material behavior inside RPV and to improve the accuracy of the SA code, a new computational fluid dynamics (CFD) code with multi-phase, multi-physics models, which is called JUPITER, was developed. It optimized the algorithms of the multi-phase calculation. Besides, the chemical reactions are also modeled carefully in the code so that the melting process could be treated precisely. A series of verification and validation studies are conducted, which show good agreement with analytical solutions and previous experiments. The capabilities of the multi-physics models in JUPITER code provide us another useful tool to investigate the molten material behaviors in the relevant severe accident scenario.

Journal Articles

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

JAEA Reports

None

; Numata, Kazuyuki*; ; ; Oigawa, Hiroyuki*

JNC TY9400 2000-006, 162 Pages, 2000/04

JNC-TY9400-2000-006.pdf:4.57MB

no abstracts in English

JAEA Reports

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.

JAEA Reports

Development of a standard data base for FBR core nuclear design (VI): JUPITER-II experimental data book

PNC TN9450 96-052, 694 Pages, 1996/10

PNC-TN9450-96-052.pdf:45.48MB

The present report compiles the experimental data of JUPITER Phase-II, which was a joint research program between U.S. DOE and PNC of Japan, using the ZPPR facility, which stands for Zero Power Physics Reactor at ANL-Idaho in l982 to l984. The JUPITER-II experiment was a series of critical experiments for conventional radial heterogeneous cores of 650 MWe class LMFBR, including six experimental cores. The nuclear characteristics recorded here include criticality, control rod reactivity, reaction rate distribution, sodium void reactivity, sample reactivity, Doppler reactivity and gamma heating. The present work is a part of efforts to develop a standard data base for LMFBR core nuclear design at PNC. The detail of experimental data is thoroughly recorded here so as to re-analyze these experiments in future. In addition, these experimental data are installed in the computer system at OEC for convenience of analytical code input.

JAEA Reports

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, ...

JAEA Reports

Development of a standard data base for FBR core nuclear design(II); JUPITER-I experlmental data book

PNC TN9410 93-010, 502 Pages, 1992/12

PNC-TN9410-93-010.pdf:17.39MB

The present report compiles the experimental data of JUPITER phase-I, which was a joint research program between U.S.DOE and PNC of Japan, using the ZPPR facility at ANL-Idaho in 1978 to 1979. The JUPITER-I experiment was a series of critical experiments for conventional two-zone homogeneous cores of 600 to 800 MWe-class LMFBR, including seven experimental cores The nuclear characteristics recorded here include criticality, control rod reactivity, reaction rate distribution, sodium void reactivity, sample reactivity, Doppler reactivity, gamma heating and neutron spectrum. (1)ZPPR-9 : two-region cylindrical clean core with volume of app. 4,600 liters, (2)ZPPR-10A : hexagonal engineering-mockup core with 19 cotrol-rod positions(CRPs), (3)ZPPR-10B : changes seven CRPs to control rods(CRs) from ZPPR-10A, (4)ZPPR-10C : volume of app. 6,200 liters with similar core arrangement to ZPPR-10A, (5)ZPPR-10D : 31 CRPs with the same volume as ZPPR-10C, (6)ZPPR-10D/1 : changes the central CRP to a CR from ZPPR-10D, and, (7)ZPPR-10D/2 : changes seven CRPs to CRs from ZPPR-10D. The present work is a part of efforts to develop a standard data base for LMFBR core nuclear design at PNC. The detail of experimental data is thoroughly recorded here so as to re-analyze these experiments in future. In addition, these experimental data are installed in the computer system at OEC for convenience of analytical code input.

JAEA Reports

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-9$$sim$$10D/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 (5$$sim$$11%). Those of reaction rate distributions also indicate similar spatial variations (2$$sim$$5%), 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 $$sim$$4% on the radial positions were found in sample ...

Oral presentation

Development of advanced neutronics/thermal-hydraulics coupling simulation system, 11; MVP/JUPITER coupling simulation using JAMPAN for fuel bundle

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

no journal, , 

JAEA has developed a platform JAMPAN for multi-physics simulations, has improved a neutronics analysis code, and has improved and validated thermal-hydraulics analysis codes to improve the design and the safety of light water reactors. The objective is implementing the coupling modules between the neutronics code MVP and the thermal-hydraulics code JUPITER, and verifying the modules. A fuel bundle geometry under a normal operation condition of a BWR was used for the neutronics and thermal-hydraulics coupling simulation to verify the modules. In this presentation, we will explain how to send and receive data between MVP and JUPITER through JAMPAN and show the results of the neutronics/thermal-hydraulics coupling simulations using MVP and JUPITER.

Oral presentation

Development of numerical simulation method to evaluate heat transfer of fuel debris in air cooling, 7; Effect of porous internal structure of fuel debris on thermal behavior in PCV

Uesawa, Shinichiro; Ono, Ayako; Yamashita, Susumu; Sano, Yoshihiko*; Yoshida, Hiroyuki

no journal, , 

To evaluate the thermal behavior of fuel debris of porous media in primary containment vessels (PCVs) of TEPCO's Fukushima Daiichi nuclear power station for air cooling, JAEA has developed a numerical simulation method using JUPITER code with porous medium models. In this presentation, we report the numerical simulation results of the thermal behavior in the PCV considering three porous structures of fuel debris, such as bubble dispersion, truss structure, and particle packed bed types. The results showed the temperature of the fuel debris was different for each model. It is important to understand the porous structure of the fuel debris for the thermal behavior because the structure has a strong effect on it.

Oral presentation

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

Oral presentation

Development of advanced neutronics/thermal-hydraulics coupling simulation system, 1; Outline of advanced neutronics/thermal-hydraulics coupling simulation system

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

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. This simulation system uses a continuous energy Monte Carlo calculation code MVP and CFD calculation code TPFIT and JUPITER. The advanced system can treat rigorous bubble shape and it does not adopt the approximate expression and empirical formula. The advanced system will be a reference system for the current neutronics/thermal-hydraulics coupling simulation system. In this presentation, we show the outline of the advanced system.

Oral presentation

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.

Oral presentation

Research and development of multi-physics coupling simulation based on CFD

Yoshida, Hiroyuki; Kamiya, Tomohiro; Tada, Kenichi

no journal, , 

no abstracts in English

Oral presentation

Oral presentation

Development of advanced neutronics/thermal-hydraulics coupling simulation system, 2; Impact of the fine-scale void distribution in sub-channels on neutronics calculations

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

no journal, , 

Japan Atomic Energy Agency is developing an advanced neutronics/thermal-hydraulics coupling simulation system for the design advancement and the safety improvement of light water reactors. In this presentation, an impact of the fine-scale void distribution in sub-channels on neutronics calculations is studied.

Oral presentation

Boiling simulation in 8$$times$$8 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 8$$times$$8 STEP-II single fuel assembly system was performed considering boiling using the temperature recovery method.

Oral presentation

Numerical simulation of breakup characteristics of sodium liquid jet

Sun, G.*; Yamashita, Susumu; Okawa, Tomio*; Yoshida, Hiroyuki

no journal, , 

When coolant leak occurs in sodium-cooled fast reactors, liquid sodium jet ejected from damaged piping can impact solid surfaces such as floors, generating splashing droplets. These droplets significantly increase the contact area between sodium and atmosphere, thereby intensifying combustion reactions. The quantity and size of splashing droplets, which influence the contact area, depend on the state of the jet during impact. Therefore, understanding the falling behavior of the jet is crucial. In this study, the breakup characteristics were analyzed to investigate the applicability of the numerical fluid dynamics simulation code JUPITER to the falling behavior of liquid sodium jets. The simulation results were similar to the extrapolated values from the empirical equations obtained from the experimental results with alternative fluids. Prospects for applying JUPITER to the analysis of sodium jet falling behavior were obtained.

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