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SIMMER-IV application to safety assessment of severe accident in a small SFR
Tagami H.,Tobita Y. 한국원자력학회 2024 Nuclear Engineering and Technology Vol.56 No.3
A sodium-cooled fast reactor (SFR) core has a potential of prompt criticality due to a change of core material distribution during a severe accident, and the resultant energy release has been one of the safety issues of SFRs. In this study, the safety assessment of an unprotected loss-of-flow (ULOF) in a small SFR (SSFR) has been performed using the SIMMER-IV computer code, which couples the models of space- and time-dependent neutronics and multi-component, multi-field thermal hydraulics in three dimensions. The code, therefore, is applicable to the simulations of transient behaviors of extended disrupted core material motion and its reactivity effects during the transition phase (TP) of ULOF, including a potential of prompt-criticality power excursions driven by fuel compaction. Several conservative assumptions are used in the TP analysis by SIMMER-IV. It was found out that one of the important mechanisms that drives the reactivity-inserting fuel motion was sodium vapor pressure resulted from a fuel-coolant interaction (FCI), which itself was non-energetic local phenomenon. The uncertainties relating to FCI is also evaluated in much conservative way in the sensitivity analysis. From this study, the ULOF characteristics in an SSFR have been understood. Occurrence of recriticality events under conservative assumptions are plausible, but their energy releases are limited.
Proton-Coupled Electron Shuttling in a Covalently Linked Ruthenium–Copper Heterodinuclear Complex
Ishizuka, Tomoya,Tobita, Kengo,Yano, Yuichi,Shiota, Yoshihito,Yoshizawa, Kazunari,Fukuzumi, Shunichi,Kojima, Takahiko American Chemical Society 2011 JOURNAL OF THE AMERICAN CHEMICAL SOCIETY - Vol.133 No.46
<P>A heterodinuclear complex based on a Ru<SUP>II</SUP>–TPA [TPA = tris(2-pyridylmethyl)amine] complex having a peripheral Cu<SUP>II</SUP>(bpy)<SUB>2</SUB> (bpy = 2,2′-bipyridine) group bonded through an amide linkage displayed reversible intramolecular electron transfer between the Ru and Cu complex units that can be controlled by protonation and deprotonation of the bridging amide moiety.</P><P><B>Graphic Abstract</B> <IMG SRC='http://pubs.acs.org/appl/literatum/publisher/achs/journals/content/jacsat/2011/jacsat.2011.133.issue-46/ja208141b/production/images/medium/ja-2011-08141b_0004.gif'></P><P><A href='http://pubs.acs.org/doi/suppl/10.1021/ja208141b'>ACS Electronic Supporting Info</A></P><P><A href='http://pubs.acs.org/doi/suppl/10.1021/ja208141b'>ACS Electronic Supporting Info</A></P>
A PRELIMINARY EVALUATION OF UNPROTECTED LOSS-OF-FLOW ACCIDENT FOR A PROTOTYPE FAST-BREEDER REACTOR
TOHRU SUZUKI,Yoshiharu Tobita,KENICHI KAWADA,HIROTAKA TAGAMI,JOJI SOGABE,KENICHI MATSUBA,KEI ITO,HIROYUKI OHSHIMA 한국원자력학회 2015 Nuclear Engineering and Technology Vol.47 No.3
In the original licensing application for the prototype fast-breeder reactor, MONJU, the event progression during an unprotected loss of flow (ULOF), which is one of the technically inconceivable events postulated beyond design basis, was evaluated. Through this evaluation, it was confirmed that radiological consequences could be suitably limited even if mechanical energy was released. Following the Fukushima-Daiichi accident, a new nuclear safety regulation has become effective in Japan. The conformity of MONJU to this new regulation should hence be investigated. The objectives of the present study are to conduct a preliminary evaluation of ULOF for MONJU, reflecting the knowledge obtained after the original licensing application through CABRI experiments and EAGLE projects, and to gain the prospect of in-vessel retention for the conformity of MONJU to the new regulation. The preliminary evaluation in the present study showed that no significant mechanical energy release would take place, and that thermal failure of the reactor vessel could be avoided by the stable cooling of disrupted-core materials. This result suggests that the prospect of in-vessel retention against ULOF, which lies within the bounds of the original licensing evaluation and conforms to the new nuclear safety regulation, will be gained.
CHARACTERISTICS OF SELF-LEVELING BEHAVIOR OF DEBRIS BEDS IN A SERIES OF EXPERIMENTS
Cheng, Songbai,Yamano, Hidemasa,Suzuki, TYohru,Tobita, Yoshiharu,Nakamura, Yuya,Zhang, Bin,Matsumoto, Tatsuya,Morita, Koji Korean Nuclear Society 2013 Nuclear Engineering and Technology Vol.45 No.3
During a hypothetical core-disruptive accident (CDA) in a sodium-cooled fast reactor (SFR), degraded core materials can form roughly conically-shaped debris beds over the core-support structure and/or in the lower inlet plenum of the reactor vessel from rapid quenching and fragmentation of the core material pool. However, coolant boiling may ultimately lead to leveling of the debris bed, which is crucial to the relocation of the molten core and heat-removal capability of the debris bed. To clarify the mechanisms underlying this self-leveling behavior, a large number of experiments were performed within a variety of conditions in recent years, under the constructive collaboration between the Japan Atomic Energy Agency (JAEA) and Kyushu University (Japan). The present contribution synthesizes and gives detailed comparative analyses of those experiments. Effects of various experimental parameters that may have potential influence on the leveling process, such as boiling mode, particle size, particle density, particle shape, bubbling rate, water depth and column geometry, were investigated, thus giving a large palette of favorable data for the better understanding of CDAs, and improved verifications of computer models developed in advanced fast reactor safety analysis codes.
Characteristics of Self-Leveling Behavior of Debris Beds in a Series of Experiments
Songbai Cheng,Hidemasa Yamano,Tohru Suzuki,Yoshiharu Tobita,Yuya Nakamura,Bin Zhang,Tatsuya Matsumoto,Koju Morita 한국원자력학회 2013 Nuclear Engineering and Technology Vol.45 No.3
During a hypothetical core-disruptive accident (CDA) in a sodium-cooled fast reactor (SFR), degraded core materials can form roughly conically-shaped debris beds over the core-support structure and/or in the lower inlet plenum of the reactor vessel from rapid quenching and fragmentation of the core material pool. However, coolant boiling may ultimately lead to leveling of the debris bed, which is crucial to the relocation of the molten core and heat-removal capability of the debris bed. To clarify the mechanisms underlying this self-leveling behavior, a large number of experiments were performed within a variety of conditions in recent years, under the constructive collaboration between the Japan Atomic Energy Agency (JAEA)and Kyushu University (Japan). The present contribution synthesizes and gives detailed comparative analyses of those experiments. Effects of various experimental parameters that may have potential influence on the leveling process, such as boiling mode, particle size, particle density, particle shape, bubbling rate, water depth and column geometry, were investigated, thus giving a large palette of favorable data for the better understanding of CDAs, and improved verifications of computer models developed in advanced fast reactor safety analysis codes.
지진이후의 수압 거동이 소규모 흙제방에 미치는 영향성 검토
이영학 ( Younghak Lee ),이달원 ( Dalwon Lee ),한지상 ( Jisang Han ),( Testuo Tobita ),( Soichiro Yamakawa ) 한국농공학회 2022 한국농공학회 학술대회초록집 Vol.2022 No.-
일본에서 지진에 의해 피해를 입은 제방댐의 대부분은 소규모 흙댐이었으며, 진도 M5.4의 지진에서도 제방의 파손 및 붕괴가 발생된 바 있다. 이러한 결과들은 일시적인 침투 또는 지진문제로 발생되었다는 시각보다 오랫동안 지진의 영향을 받아오면서 제체 약화에 의해 발생된 문제로 보는 시각이 적합하다. 또한, 소규모 흙댐은 유지관리 측면에서 내진성능이 현저히 부족한 상태로, 침투와 지진의 연계적 거동에 대한 제체의 안전성을 재평가할 필요가 있다. 본 연구에서는 제방댐에서 침투와 지진의 연속성이 제체에 미치는 영향을 시각화하고 지진 이후의 문제점을 상세히 검토하기 위하여 원심모형실험에 의해 1차 침투 - 지진 - 2차 침투로 이어지는 연계적 수압-변형 거동을 모사하였다. 실험모형은 원심가속도 50G 에 맞추어 1/50 로 축소하였고, 수압의 지속적 작용을 모사하기 위해 하류측에서 수중펌프에 의한 순환방식을 채택하였다. 상류측 수위는 홍수위 상태로 유지한 상태에서 점성유체를 사용하여 50로 조정·적용하였고 지진조건은 1.0m/s2의 지진 가속도를 적용하였다. 지진 이후 하류사면 천단부와 댐마루 경계부는 균열부위가 미세함에도 불구하고 2차 침투에 의해 크게 확장되었으며, 사면 활동의 임계원이 시작되는 지점도 중첩되어 붕괴 취약부로 검토되었다. 또한, 지진에 의해 댐마루 표면에서 형성된 균열은 2차 침투가 진행됨에 따라 종방향 균열과 함께 단시간내에 제체높이의 73.3% 지점 부근까지 균열의 깊이가 확장되어 1.0m/s2의 지진 가속도만으로도 제체에 이상변화가 발생되었다. 댐마루에서 균열에 따른 변형률을 비교한 결과, 상류사면은 침투가 중요한 지배요소가 되고, 댐마루는 지진이 지배요소가 되는 것으로 나타났다. 지진 후 공극수압은 상류사면과 중앙에서 1.4-1.5배, 하류측은 2.1배, 비탈끝은 3.0배증가하는 것으로 나타났다. 이와 같은 결과는 상대적으로 상류수위에 가까운 상류사면과 중앙에서 지진으로 인한 제 체 내부의 상태변화를 야기하여 하류측과 비탈끝의 수압상승을 촉진시키는 결과로 나타났다. 본 연구에서 검토된 변위, 변형 형상, 공극수압 결과로부터 각 단계에서의 침투와 지진에 따른 붕괴현상을 발생되지 않았지만, 오랫동안 지진과 침투의 영향을 받게 되는 경우, 제체내부에 균열을 발생시키고 전단강도를 약화시켜 제체 안정성에 심각한 영향을 미칠 수 있음을 알 수 있다. 따라서 중·장기적인 관점에서 재해대응은 댐마루 가까운 하류사면 천단부에서 균열 깊이 억제와 개선방법이 필요하고, 제체 하류존의 수압을 저감시킬 수 있는 방안이 지진재해와 관련된 저수지의 재해예방 및 재해완화조치의 방향이 될 것으로 판단된다.