Development of virtual plant model for design rationalization of fast reactors by multi-level simulation system; Confirmation of functionality in application to U.S. experimental fast reactor EBR-II

Yoshimura, Kazuo; Doda, Norihiro; Nakamine, Yoshiaki*; Fujisaki, Tatsuya*; Igawa, Kenichi*; Iida, Masaki*; Tanaka, Masaaki

Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2023/09

In Japan Atomic Energy Agency, a virtual plant model of the sodium-cooled fast reactor plant composed in a computer is being developed to reduce the development cost, by replacing the experiments to the numerical simulations with coupled analyses of the physical phenomena accounting for the interaction between components under various plant conditions. Through the numerical analysis of the ULOHS test in the U.S. experimental fast reactor named EBR-II, applicability of the virtual plant model was confirmed in comparison with the measured data including the core inlet temperature and the reactor power.

Improvement of reactivity model of core deformation in plant dynamics analysis code during unprotected loss of heat sink event in EBR-II

Yoshimura, Kazuo; Doda, Norihiro; Fujisaki, Tatsuya*; Igawa, Kenichi*; Tanaka, Masaaki; Yamano, Hidemasa

Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 10 Pages, 2023/05

The benchmark analyses for the unprotected loss of heat sink (ULOHS) tests in the pool-type experimental SFR in the United States, EBR-II (BOP-301 and BOP-302R) have been conducted in order to validate the evaluation method of the reactivity feedback equipped in the plant dynamics analysis code named Super-COPD. In this study, 1D-CFD coupled analyses adding the core bowing reactivity model were conducted. Through the analysis, the applicability of the modified reactivity model was confirmed for the BOP-301 test. For the BOP-302R test, consideration of the core restraint system in the core and modeling the control rod driveline expansion reactivity was indicated.

Application of 1D-CFD coupling method to unprotected loss of heat sink event in EBR-II focusing on thermal stratification in cold pool

Yoshimura, Kazuo; Doda, Norihiro; Fujisaki, Tatsuya*; Igawa, Kenichi*; Tanaka, Masaaki; Yamano, Hidemasa

Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 10 Pages, 2022/08

To confirm the applicability of the reactivity model, the authors have been conducting the benchmark exercises of the unprotected loss of heat sink event tests in a pool-type experimental fast reactor EBR-II. In the blind phase in the benchmark analyses using the plant dynamics analysis (1D) code in which the cold pool was modeled by means of the perfect mixing volume, it was found the increase of the core inlet temperature was evaluated lower than that of the measured data and the feedback reactivity was underestimated, because the thermal stratification in the cold pool was ignored. Then, the detailed model of the cold pool for the computational fluid dynamics (CFD) code was introduced and the 1D-CFD codes coupling method was applied to the benchmark analyses. It was confirmed that both the thermal stratification in the cold pool and the increase of the core inlet temperature were successfully reproduced.

Validation of evaluation method of feedback reactivity for plant dynamics analysis code during unprotected loss of heat sink event in sodium-cooled fast reactors

Yoshimura, Kazuo; Doda, Norihiro; Tanaka, Masaaki; Yamano, Hidemasa; Igawa, Kenichi*

Proceedings of 28th International Conference on Nuclear Engineering (ICONE 28) (Internet), 8 Pages, 2021/08

The numerical results of the benchmark analyses for the unprotected loss of heat sink (ULOHS) tests in the pool-type experimental SFR in the United States, EBR-II (BOP-302R and BOP-301) are discussed in order to validate the evaluation method of the reactivity feedback equipped in the in-house plant dynamics analysis code named Super-COPD. By comparing the numerical results and the experimental data, the profiles of the increase of the core inlet temperature and the decrease of the reactor power calculated by Super-COPD were comparable with those of the experimental data and the applicability of the evaluation method for the reactivity feedback was indicated during the ULOHS event.

Thermal-hydraulic experiment on beam window for developing the accelerator-driven transmutation system

Hayashi, Kenichi*; Ono, Mikinori*; Kikuchi, Kenji; Tokunaga, Noriya*; Kitano, Teruaki*; Oigawa, Hiroyuki

Nihon Genshiryoku Gakkai Wabun Rombunshi, 7(1), p.44 - 57, 2008/03

https://doi.org/10.3327/taesj.J07.002

Accelerator driven nuclear transmutation system aims at transmuting minor actinides and long-lived fission products to stable or short-lived nuclei. A design study of proton beam window, which is an interface component between accelerator and nuclear reactor. Thermal-hydraulic experiment of the beam window was done. Two experiments were conducted: one was particle image velocimetry measurement around the beam window in flowing water and the other was temperature measurement at the beam window under flowing lead bismuth. Numerical simulation was also done to validate the beam window model for design work. Results show that heat transfer characteristics of the beam widow averaged in space and with time under flowing lead bismuth was formulized by the experimental equation. Numerical simulation model can estimate the mean heat transfer coefficient. However, a local heat transfer coefficient was not stable: it fluctuates with time and even in space, especially around stagnation point.

Heavy Liquid Metal Cooled FBR; Results 2003

Hayafune, Hiroki; Enuma, Yasuhiro; Soman, Yoshindo; Konomura, Mamoru; Mizuno, Tomoyasu; Igawa, Kenichi*

JNC TN9400 2004-054, 339 Pages, 2004/08

JNC-TN9400-2004-054.pdf:24.71MB

Concepts of the reactor, SG and main coolant pump have been studied considering maintainability and aseismic capability, which is a medium size pool type lead-bismuth cooled reactor. The results are following.(1) Reconsideration of reactor design concepts concerning maintainability. In pursuit of good reactor maintainability, the structural concepts of SG, UIS and core support structures have been changed to be drawn up above the upper area of the reactor system. After a few decade of interval, lead-bismuth inventory in the reactor vessel shall be fully drained for easy ISI operation of in-vessel main components such as core support structures. From the viewpoint of the reactor aseismic capability, the axial length of reactor vessel was reduced and the reactor vessel support location was changed from the top hanging to the circumference of the vessel.(2) SG concept selection in conjunction with a compact reactor vessel.The concept of SG consisting of a once through type with helical coil tube is selected. 6 units of a small scale SG are arranged on a reactor roof deck along the peripheral direction, in addition to 3 units of a centrifugal mechanical pump.(3) Aseismic structural integrity of the reactor components. Aseismic structural integrity of the reactor vessel, core support structures, UIS, FHM, SG and the main pumps has been vigorously examined respectively. These components besides FHM could keep the aseismic structural integrity for strong S2 earthquake under the design condition FHM could also keep the integrity for S1 earthquake.(4) Safety evaluation. Thermal translents following loss of flow type accident due to plant total blackout and typical manual reactor trip incident, have been evaluated to assure the plant safety design, by analyzing thermal hydraulic behavior of transients concerning core flow rate and temperatures of the plant cooling system. *Loss of flow accident due to plant total blackout. The reactor coolant pumps shall be tripped and the

Evaluation on the fretting abrasion of heat-transfer tubes of the integrated IHX/primary sodium pump (1); Workrate analyses model

; Igawa, Kenichi*

JNC TN9400 2002-019, 81 Pages, 2002/05

JNC-TN9400-2002-019.pdf:2.21MB

The cost minimization of commercialized FBR plant systems requires the integration of an intermediate-heat-exchanger (IHX) and a primary sodium mechanical pump into one component. The pump is installed in the center of the integrated component and heat transfer tubes surround the pump. Primary sodium flows down inside the heat transfer tubes and secondary sodium flows up outside the tubes in a zigzag. Therefore, the pump rotation and sodium flow induce the vibration of heat transfer tubes and it leads the tubes to fretting wearing against support plates. Then the tube wearing must be evaluated to confirm its integrity during the plant life span (60 years). However, the knowledge of the pump rotation influence on tube wearing is not sufficiently acquired because the integrated component is a new concept in JNC. To evaluate the tube fretting wearing ratio due to the pump rotation, a new calculation model of FINAS was composed. In the first place, the beam vibration analysis model of a pump shaft, shells, tube bundle etc. of the integrated component reveals its properties such as frequency, amplitude and vibration mode. In the second place, based on the above mentioned vibration analysis, the frequency and amplitude of abrasion between the tubes and support plates can be obtained by a contact analysis model of FINAS. Eventuany, this calculation shows that the tube wearing will not affect the tube integrity during the plant life time. However further evaluation by more detailed analysis and abrasion tests are needed to obtain more accurate results.

Development of JOYO operation support system Development of plamt state prediction function

*; *; Fukami, Akihiro*; *; Igawa, Kenichi*

PNC TN9440 89-006, 49 Pages, 1989/09

PNC-TN9440-89-006.pdf:1.7MB

In Experimental Fast Reactor "JOYO", "Development of operation Support System" is continued as the enhancement, of opration support by using computer, in order to improve the availability and reliability of "JOYO" and future fast reactors. Plant State Prediction is one of these systems, and its function is to predict important parameters with respect to the security of safety and to figure them as valid data on CRT. This report deals with the examinations to put this function into JOYDAS (JOYO Data Acqisition System). Main results are as follows. (1) Objective phenomena and signals were selected from Emergency Procedure-Operation (EPO) and so on. (2)From these signls, sodium level in R/V and decay heat were selected and these prediction models were produced. (3)AS for sodium level in R/V, its prediction model was compared with the data at Primary Main Pump Trip (1987-9-7) and this reasonability was confirmed.

Progress in procurement of remote handling equipment for ITER Blanket maintenance

Takeda, Nobukazu; Tanigawa, Hisashi; Ueno, Kenichi; Maruyama, Takahito; Noguchi, Yuto; Kakudate, Satoshi

no journal, , 

In the vacuum vessel of the ITER, dose rate is very high (250 Gy/h) even during the plasma shutdown. Therefore, in-vessel components such as blanket and divertor must be maintained by remote handling. Procurement of the remote handling system of the blanket is allocated to Japan and the JAEA is performing the final design of it toward handover to the ITER Organization scheduled in 2020. This report shows progress of the final design activity. In the final design, the major part of the systems are more detailed and the tools for welding and cutting of blanket pipes are newly designed. In the same time, feasibility of operation scenario is confirmed by system analyses such as the reliability, availability, maintainability and inspectability analysis and structural analysis. By these activities, it is confirmed that the main components whose manufacturing contract will be awarded in this year can be fabricated without any concern and will be handed over to the ITER Organization in 2020 as planned.

ITER blanket cooling pipe maintenance tool design

Ueno, Kenichi; Tanigawa, Hisashi; Noguchi, Yuto; Inoue, Ryuichi; Anzai, Katsunori; Kazawa, Minoru; Takeda, Nobukazu; Kakudate, Satoshi

no journal, , 

At the ITER blanket, there are cooling pipes for cooling. They will be cut and welded by dedicated tools. For the cooling pipe cutting, remaining the dust and swarf at the cooling pipe are prohibited and suitable cutting surface for rewelding are required. To implement these requirement, swage cutter cutting system were made as a mockup and tested for simulated cooling pipes. There were no cutting dust and suitable cutting surface by the swage cutter cutting system. For the blanket cooling pipe welding, relaxation for the positioning of welding groove, improvement for optical system durability of LASER welding are required. The mockup of LASER welding tool head were made and tested for simulated cooling pipes. As the result of welding condition improvement, relaxed welding condition with low sputter generation were taken. For the cooling pipe welding groove alignment, alignment tool was made and tested for correction of simulated cooling pipe misalignment. For less than 1.5mm of linear misalignment and 0.5 degree angular misalignment, required alignment accuracy were confirmed. Positioning accuracy and stability improvement at the tool operation were issued for the ITER blanket cooling pipe tool heads development.