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Luu, V. N.; 谷口 良徳; 宇田川 豊; 田崎 雄大; 勝山 仁哉
Annals of Nuclear Energy, 230, p.112114_1 - 112114_14, 2026/06
被引用回数：1 パーセンタイル：92.65(Nuclear Science & Technology)Fracture behavior of chromium (Cr) coated cladding under loss of coolant accident (LOCA) conditions was investigated utilizing the FEMAXI fuel performance code. Cr coating degradation models were introduced to FEMAXI to calculate oxygen diffusion behavior within the cladding tube. The FEMAXI code reasonably simulated the observed evolution of cladding metallic and oxide layers under the simulated LOCA conditions, accounting for factors such as wall thinning due to cladding high temperature creep, Cr layer thinning by Cr
O
formation and Cr/Zr interdiffusion, weight increase by oxygen absorption, associated oxide growth, and increased oxygen concentration in 
-Zr phase. According to sensitivity analyses of the cladding oxygen concentration, where the effects of wall thickness change and eutectic reactions were taken into account, the fracture condition of the Cr-coated cladding samples can be reasonably modelled by the fracture criteria based on the remaining -Zr thickness with an oxygen concentration of 
0.9 wt%.
谷口 良徳; Luu, V. N.; 田崎 雄大; 宇田川 豊; 勝山 仁哉
Annals of Nuclear Energy, 231, p.112177_1 - 112177_16, 2026/06
被引用回数：0 パーセンタイル：0.00(Nuclear Science & Technology)Advanced technology fuels (ATF) with improved oxidation resistance are under development to enhance the safety of light water reactors. Cr-coated Zr alloy cladding, a promising near-term ATF, exhibits excellent oxidation resistance below the Cr-Zr eutectic temperature. However, its gradual loss of protective effect over time, even without mechanical damage, indicates the need to understand its degradation mechanisms. This article presents a phenomenological model describing degradation due to high-temperature oxidation, focusing on Zr ingress into the Cr coating and the formation of oxygen pathways that accelerate oxygen uptake into the Zr matrix. The model was validated against experimental data at 1200
C and 1300C, reproducing key trends such as oxide growth, weight gain, and oxygen concentration profiles. Applying the same parameters to a different PVD-coated cladding test gave reasonable agreement at 1200C, while discrepancies at 1300C suggest Cr-Zr eutectic reactions from local temperature variations, highlighting the model's sensitivity near the eutectic point.
Luu, V. N.; 谷口 良徳; 宇田川 豊; 勝山 仁哉
Nuclear Engineering and Design, 442, p.114222_1 - 114222_15, 2025/10
被引用回数：2 パーセンタイル：73.73(Nuclear Science & Technology)For near-term application, coated-Zr alloy claddings show potential for enhancing safety by providing better oxidation resistance and minimizing hydrogen absorption under design-basis accidents (DBA). This benefit could extend the burnup and operational cycles of fuel rods. In assessing safety, reactivity-initiated accidents (RIA) are considered as one of the DBA conditions. The current safety criteria for high-temperature oxidation failure, one of the failure modes linked to RIA, are defined by peak fuel enthalpy values that range from 205 to 270 cal/g. This wide variability presents challenges when attempting to generalize criteria for modified-Zr alloy claddings with superior oxidation resistance. Therefore, it may be more relevant to apply failure criteria based on embrittlement mechanisms, such as oxygen concentration in the -Zr phase. This study aimed to assess the failure based on both peak fuel enthalpy and cladding embrittlement by analyzing previous NSRR experiments conducted with conventional materials using the RANNS fuel performance code. The findings suggest that the failure criteria associated with cladding embrittlement can provide a rational evaluation of failure behavior compared to the existing criterion based on peak fuel enthalpy. The local failure criterion leading to the formation of through-wall cracks during quenching is consistent with Chung's proposal (NUREG/CR-1344): -Zr thickness of 0.9 wt% oxygen is less than 0.1 mm, and this corresponds to approximately 35% BJ-ECR.
-CsOHRizaal, M.; Luu, V. N.; 中島 邦久; 三輪 周平
Proceedings of International Topical Workshop on Fukushima Decommissioning Research 2024 (FDR2024) (Internet), 4 Pages, 2024/10
Thermochemistry prevailing between gaseous CsOH and concrete main chemical phase CaCO at temperatures up to 570C was investigated with various scenarios using the thermogravimetric method. The aim was to elucidate the decreasing behavior of cesium (Cs) trapping on CaCO observed in the transpiration method. A quasi-two-compartment platinum crucible was developed to realize co-measurements of both CsOH and CaCO during thermal treatment. Post-test X-ray diffraction was conducted to identify the chemical compound formed on the CaCO precursor. The early presence (timely sensitivity) of CsOH near the heated surface of CaCO was found to play a key role in the trapping (in the form of CsCO). Such a factor is crucial because, otherwise, the Ca(OH) would predominate the surface upon CaCO decomposition where leading to no reaction with CsOH.
Luu, V. N.; 中島 邦久; Rizaal, M.; 三輪 周平
Proceedings of International Topical Workshop on Fukushima Decommissioning Research 2024 (FDR2024) (Internet), 4 Pages, 2024/10
The trapping behavior of cesium (Cs) on concrete, such as pedestal and shield plugs, was assessed through multiple deposition tests of CsOH aerosols on concrete phase-CaCO within a temperature range of 170C to 570C under humid conditions. The deposition rate of CsOH aerosols on CaCO as a function of temperature was examined. X-ray diffraction analysis revealed the presence of water-soluble CsCO on CaCO samples up to 420C, but negligible detection above this temperature. This finding is supported by quantitative analyses of Cs weight gain and chemical analysis, demonstrating an increase in deposition rate up to 420C, followed by a decline thereafter. The increased deposition rate is attributed to the chemical interaction between CsOH and CaCO, while the diminished rate beyond 420C may result from the decomposition of CaCO under steam, thereby reducing the available reactant for the chemical reaction with CsOH.
at temperature range 170 - 290CLuu, V. N.; 中島 邦久
Nuclear Engineering and Design, 426, p.113402_1 - 113402_7, 2024/09
被引用回数：3 パーセンタイル：40.43(Nuclear Science & Technology)A field assessment at the Fukushima-Daiichi Nuclear Power Station revealed high radioactivity on the concrete shield plugs, which is estimated above 20 PBq for Cs-137 at units 2 and 3. This leads to significant interest in the retention of Cs on concrete during severe accidents (SA). However, the interaction of CsOH, as one of the main Cs forms released in SA, with concrete surfaces at elevated temperatures remains poorly researched. In this study, we have experimentally investigated the deposition behavior of CsOH on CaCO, which is the primary phase existing on the surface of concrete, under humid atmosphere. As a result, the chemical reaction enhanced deposition rate (N), and increased linearly with CsOH concentration (C
), as following expression: N(
g/cm
s) = v
C, where v is temperature-dependent deposition velocity as given by ln v (cm/s) = -3785.8/T + 3.766, for T in the range of 170 and 290 C. This empirical model can be integrated into severe accident codes to quantify the chemical trapping of cesium on concrete surfaces during ex-vessel release. Moreover, it can contribute to understanding the reasons behind the high dose rate on concrete shield plugs at the Fukushima Daiichi Nuclear power stations and aid in developing effective decommissioning practices for concrete structures.
CLuu, V. N.; 中島 邦久
Mechanical Engineering Journal (Internet), 11(2), p.23-00446_1 - 23-00446_11, 2024/01
Cesium distribution is crucial for decommissioning Fukushima Daiichi Nuclear Power Station (1F). Several experimental studies confirmed Cs retention on stainless steels by performing chemical reactions at high temperatures (typically above 800C)), but the Cs retention on concrete, used in large quantities in light water reactors, is not fully understood. This study demonstrated that Cs might have been deposited and retained on the concrete structures where the temperature was not so high during the 1F accident. Results showed that the CsOH/concrete interaction at around 200C occurred in water-insoluble Cs-(Al,Fe)-Si-O deposits and water-soluble phases, i.e., cesium carbonate hydrate and possibly cesium silicate if Al and Fe are not present. CsOH might be trapped on concrete by chemical reaction with CaCO to form CsCO hydrate, and with aluminosilicate and SiO(quartz) to form Cs-Al-Si-O and Cs-Si-O deposits, respectively. This output could help elucidate the trapping mechanism that caused extremely high radioactivity on concrete shield plugs at 1F and develop an effective decommissioning practice for concrete structures.
CLuu, V. N.; 中島 邦久
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 9 Pages, 2023/05
Information of Cs distribution is important for decommissioning of the Fukushima Daiichi Nuclear Power Station (1F). Several experimental studies confirmed the Cs retention on stainless steels by chemical reaction at very high temperatures (commonly above 800C), but the Cs retention on non-metallic materials, such as concrete and thermal insulators, was not fully understood though they are used with large quantity in light water reactors. This study demonstrated that Cs might be deposited and retained on the concrete structure where the temperature was not so high during the 1F accident. It was revealed that the CsOH/concrete interaction at around 200C resulted in the formation of water-insoluble Cs-(Al,Fe)-Si-O deposits and water-soluble phases, i.e., cesium carbonate hydrate and possibly cesium silicate, if Al and Fe are not present. CsOH might be trapped on concrete by chemical reaction with CaCO to form CsCO hydrate, and with aluminosilicate and SiO(quartz) to form Cs-Al-Si-O and Cs-Si-O deposits, respectively. This output will be useful for elucidating the trapping mechanism that caused an extremely high radioactivity on concrete shield plugs at 1F, and for developing an effective decommissioning practice for concrete structure.
Luu, V. N.; 中島 邦久
Journal of Nuclear Science and Technology, 60(2), p.153 - 164, 2023/02
被引用回数：8 パーセンタイル：61.11(Nuclear Science & Technology)Recently, extremely high dose rates were detected in the three-layer concrete plugs of Units 2 and 3 at the Fukushima Daiichi Nuclear Power Plant. The high dose rates suggest that there are some trapping effects of radioactive materials on shield plugs when gas species and aerosols (e.g., CsOH, CsI) are released from reactor through the plug layers. To determine the trapping mechanism, concrete and commonly used aggregate and minerals are pulverized and mixed with CsOH, followed by heating at different temperatures to clarify the chemical interaction. The results showed that interactions of CsOH and CaCO in concrete occurred even at room temperature to form CsCO(HO). The interaction with aggregates occurred above 100C and resulted in the formation of CsAlSiO
. Additionally, amorphous and crystalline SiO interacted with CsOH, forming a glass-like product above 200C. These results suggest that formation of CsCO(HO) would be one of the main trapping mechanism at shield plugs because CaCO is commonly formed on concrete surface and reacts with CsOH at room temperature.
at temperature range 170 - 290 CLuu, V. N.; 中島 邦久
no journal, ,
A recent field survey revealed extremely high radioactivity on the concrete shield plugs, exceeding 20 PBq for Cs-137 at units 2, 3 of 1F. This leads to significant interest in the retention of Cs on concrete during severe accident. Despite this, the interaction of Cs vapors and/or aerosols in the gas phase with concrete surfaces at high temperatures remains inadequately explored. In this work, we investigated the deposition behavior of CsOH on CaCO, as a primary phase on the concrete surface, under humid atmosphere. As a result, the chemical reaction enhanced deposition rate, and increased linearly with CsOH concentration.
Luu, V. N.; 谷口 良徳; Nguyen, T. H.; 宇田川 豊; 田崎 雄大; 三原 武; 勝山 仁哉; 垣内 一雄
no journal, ,
To obtain the knowledge and data necessary for the regulatory assessment of Cr coated Zr-based alloy claddings, which is expected to be installed to Japanese plants in the near future, integrated LOCA tests were conducted using 190mm-long uncoated and Cr-coated Zr-based alloy claddings filled with zirconia pellets to simulate the thermal behavior of UO pellets. The test rod was initially pressurized to 5 MPa and heated in 100% steam atmosphere at a heating rate of 5 K/s until burst. This was followed by isothermal oxidation at temperatures between 1200 and 1350C and quenching at 700C under an axial constraint of 
500N. The test temperature covered conditions both below and above the Zr/Cr eutectic temperature (1320C) to establish fracture limits under LOCA conditions. Post-test observations showed that fractures into two segments occurred only at the burst locations due to circumferential cracking. Below the eutectic temperature, the fracture limit exceeded 28% BJ ECR (Equivalent Cladding Reacted calculated using the Baker-Just correlation) for both uncoated and Cr-coated claddings. Above the eutectic temperature at 1350C, the fracture limit remained unchanged for Cr-coated claddings but decreased to 20% BJ ECR for uncoated claddings. This reduction is attributed to increased hydrogen uptake (up to 500 wtppm), as confirmed by hydrogen analysis at the burst locations. The similar fracture limits of Cr-coated claddings below and above the eutectic temperature, further supported by metallographic observations and Vickers hardness measurements, suggest that the eutectic reaction did not appear to impact the fracture limit of Cr-coated claddings.
Luu, V. N.; 中島 邦久
no journal, ,
To investigate the source of high dose rates at the concrete shield plugs of Unit 2 and 3 at 1F, high temperature tests on the mixture of CsOH and pulverized concrete/main components were conducted. Results showed that both water-soluble and -insoluble phases were formed below 300C. Namely, CsCO(HO) was formed due to chemical reaction with CaCO at room temperature. The authors will discuss the possibility that this might be one of the main trapping mechanisms on shield plugs.
Luu, V. N.; 中島 邦久
no journal, ,
This study investigated the possibility of chemical interaction between concrete/aggregate and cesium hydroxide at high temperatures. CsOH interaction with CaCO in concrete forms water-soluble CsCO(HO) even at room temperature, whereas CsOH interaction with aluminosilicate minerals in aggregate forms water-insoluble CsAlSiO just above 100C, according to TG/DTA and XRD analyses on the mixture of CsOH and pulverized concrete/aggregate. The results suggest that the formation of CsCO(HO) would be one of the main trapping mechanisms at shield plugs because CaCO is commonly formed on concrete surfaces and reacts with CsOH at room temperature.
三輪 周平; 中島 邦久; 唐澤 英年; Rizaal, M.; Luu, V. N.; Mohamad, A. B.
no journal, ,
東京電力福島第一原子力発電所内のセシウム等のFPの分布や性状を把握することが廃炉に向けた重要な課題であり、日本原子力研究開発機構ではそれらに大きな影響を与えるFP化学に着目した基礎研究を実施し、事故時の原子力発電所内のFP化学を評価するためのデータベースECUMEを開発している。ECUMEは福島第一原子力発電所の事故により明らかになった重要な現象、例えば、セシウムの制御材ホウ素との反応や、構造材との反応に関するデータやモデルを収納している。近年の内部調査により明らかとなったシールドプラグでの高線量等の原因を明らかにするため、コンクリートや他の炉内物質との化学反応を調べ、モデル化を行い、ECUMEの更新を進めている。
唐澤 英年; Rizaal, M.; Luu, V. N.; 中島 邦久; 三輪 周平; 木野 千晶*
no journal, ,
福島第一原子力発電所事故で観察されたシールドプラグ間のCs蓄積機構について検討している。従来シールドプラグ間の隙間にガスの流路は無いと考えられていたが、高濃度のCs-137が蓄積していることが確認された。また、原子力規制委員会は、重力によるシールドプラグの変形によりガス流路が形成されることを明らかにした。今回、過酷事故解析コードSAMPSONのFP移行挙動解析モジュールに、CsOHエアロゾルのコンクリートへの化学吸着速度を追加し、2号機のシールドプラグを対象に試解析を行った。そして、Cs蓄積に及ぼすシールドプラグ間の隙間間隔、ガス流速、ガス温度、CsOHエアロゾル濃度などの影響評価を行い、Cs蓄積の主要メカニズムを検討する。
Luu, V. N.
no journal, ,
This study investigates the oxidation behavior and post-quench mechanical properties of MDA cladding, with and without Cr coating. Uncoated MDA exhibits oxidation behavior consistent with the Cathcart-Pawel correlation, while Cr-coated samples show significantly reduced oxidation, confirming the protective role of the coating. Quenching conditions, particularly at 700C, enhance ductility in Cr-coated MDA, likely due to Cr enrichment in -Zr and annealing effects. Differences in critical HV values between direct and 700C quenching were observed, suggesting a need for further analysis to understand ductile-to-brittle transitions.
Nguyen, T. H.; Luu, V. N.; 谷口 良徳; 宇田川 豊; 勝山 仁哉; 垣内 一雄
no journal, ,
The proposed use of chromium (Cr)-coated zirconium (Zr)-based alloy cladding as an accident-tolerant fuel (ATF) highlights the importance of understanding its mechanical response under loss-of-coolant-accident (LOCA) conditions. In anticipation of the deployment of Cr-coated cladding, it is necessary to establish appropriate LOCA safety criteria and verify existing fuel performance codes. Under LOCA scenarios, fracture may occur either at the locations with oxidation or secondary hydriding. Nagase and Fuketa evaluated fracture maps at the rupture site as a function of the initial hydrogen content, however, no correlation was reported between fracture limits and post-LOCA hydrogen content. Alakiozidis at al. reported a hydrogen threshold of approximately 2700 ppm for out of burst failure in both coated and uncoated claddings; however, experimental data for coated claddings remains limited. This study aims to evaluate the effects of hydrogen on cladding embrittlement and fracture characteristics of Cr-coated cladding. To this aim, oxidation tests were conducted on Cr-coated claddings with controlled pre-hydrogen absorption to simulate low hydrogen concentrations, and integral LOCA tests were performed on both Cr-coated and uncoated claddings to evaluate secondary hydriding effects. For both tests, cladding tube samples were two sided steam oxidized to achieve equivalent cladding reacted (ECR) values ranging from 10 to 40%, corresponding to oxidation temperatures of 1200 and 1350C for various exposure times in a vertical furnace, followed by water quenching at 700C. Post-test examination includes the hydrogen measurement, microstructure observation by optical microscopy (OM) and scanning electro-microscopy (SEM), Vickers hardness test and nanoindentation. According to the results obtained, hydrogen promotes embrittlement of prior -Zr after exposure at 1200C, while having no significant effect after exposure at 1350C. The eutectic layer forms after exposure of Cr-coated cladding to 1350C. This region consists of two distinct phases: a Cr-rich and a prior -Zr phase. Cr-rich phase exhibits higher nano-hardness than that of prior -Zr phase and hydrogen absorption enhances the hardness of both phases. The fracture threshold of Cr-coated cladding is determined by different controlling mechanisms at burst and out-of-burst positions. At the burst, fracture limits were found to be decreased with increasing hydrogen content (up to approximately 500 ppm), and the Cr-coated cladding fractured at longer oxidation time than uncoated cladding. At out-of-burst positions, no fracture observed up to 2700 ppm hydrogen, in agreement with Alakiozidis et al. and fracture limit was determined based on the remaining prior -Zr thickness at oxygen concentrations 0.9 wt% by Chung [3]. These findings indicate that the Cr-coated cladding exhibits hydrogen-related embrittlement behavior similar to conventional Zr alloys, with the additional advantage of reduced oxidation kinetics, which mitigates oxidation-driven embrittlement.
CLuu, V. N.; 中島 邦久
no journal, ,
Chemisorption tests of CsOH onto concrete and aggregate were carried out at 200C in both dry and humid conditions to investigate the source of high dose rates at the concrete shield plugs of Units 2 and 3 at 1F. Post-test analyses using XRD, Raman, and SEM revealed that water-insoluble Cs-bearing deposits formed different morphologies under dry and humid conditions. Namely, uniform and large Cs deposits were formed under humid conditions, while heterogenous growth and smaller size Cs deposits were observed under dry conditions.
Rizaal, M.; 中島 邦久; 唐澤 英年; Luu, V. N.; 三輪 周平
no journal, ,
Our research focused on, but is not limited to, cesium (Cs) and iodine (I) chemistry due to their high impact on the overall source term. The retention or release of both elements is largely affected by chemical interaction with materials that are present in the reactor. To understand their chemistry during transport in the event of a nuclear severe accident (SA), we studied the interaction phenomena taking place from high- to low-temperature conditions. We have succeeded in elucidating these phenomena (particularly Cs) and summarized them in a fission product (FP) chemistry database ECUME. This database not only could deepen our understanding of the mechanism of Cs and I chemistry in an SA, but could also improve source term analysis. Improvement in the reaction between Cs vapor and stainless steel was shown by the use of the ECUME database in SA analysis code SAMPSON. Better reproducibility of Cs retention at high temperatures of the large-scale experiment was obtained, in contrast to using the MELCOR Cs interaction model (i.e. widely used model in SA code) that was worse in reproducing such phenomenon. Taking into consideration of near-term implementation of Accident Tolerant Fuel (ATF) materials such as chromium (Cr)-coated Zircaloy, further study on the interaction with FP would be important to ensure the material impact on source term because the reaction between Cs and Cr can be thermodynamically expected.
