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菅原 隆徳; 渡辺 奈央; 小野 綾子; 西原 健司; 市原 京子*; 半澤 光平*
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 10 Pages, 2022/03
JAEAでは、LBE冷却型加速器駆動システム(ADS)の研究開発を行っている。ADSは既存の臨界炉に比べて安全性が高いと考えられているが、除熱源喪失事象が生じた場合、炉心の破損の可能性が考えられる。これを回避するためDRACS(直接炉心補助冷却システム)を備えているが、その機器信頼性と流量確保が重要となる。本研究では、CFD解析によりADS炉容器内の自然循環時の流れを解析し、DRACS熱交換器位置において十分な流量が確保されるかを検討した。解析の結果、定格運転時の約4.3%の流量が確保され、崩壊熱が適切に除去されることを確認した。また、炉容器内のLBE流れを整えるために設けられている内筒の有無が、自然循環時にどのような影響を与えるかについても検討を行った。
青柳 光裕; 内堀 昭寛; 高田 孝
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 13 Pages, 2022/03
The SPECTRA code has been developed as an integrated safety analysis tool for sodium-cooled fast reactors (SFRs). In this study, the capability of SPECTRA is enhanced by establishing a sodium pool and floor concrete module commonly used in individual physical modules. This paper describes the framework of modelling for the mass and heat transfer in the sodium pool and the floor concrete. Considering concrete ablation due to sodium and debris, bottom of the sodium pool changes during event progression. This change in material composition in certain position is modeled by volume and mass fraction of each component. A simple convection model for the pool is implemented to ensure the conservation of heat and mass. This model is tested through the verification analysis in comparison with the existing model. As the result, it is confirmed the behavior of pool spreading and concrete ablation can be simulated by this module correctly.
間所 寛; 山下 拓哉; 佐藤 一憲; Gaus-Liu, X.*; Cron, T.*; Fluhrer, B.*; St
ngle, R.*; Wenz, T.*; Vervoortz, M.*; 溝上 伸也
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03
Debris and molten pool behavior in the reactor pressure vessel (RPV) lower plenum is a key factor to determine its failure mode, which affects the initial condition of ex-vessel accident progression and the debris characteristics. These are necessary information to accomplish safe decommissioning of the Fukushima Daiichi Nuclear Power Station. After dryout of the solidified debris in the lower plenum, metallic debris is expected to melt prior to the oxide debris due to its lower melting temperature. The lower head failure is likely be originated by the local thermal load attack of a melting debris bed. Numerous experiments have been conducted in the past decades to investigate the homogeneous molten pool behavior with external cooling. However, few experiments address the transient heat transfer of solid-liquid mixture without external cooling. In order to enrich the experimental database of melting and heat transfer process of debris bed consisted of materials with different melting temperatures, LIVE-J1 experiment was conducted using ceramic and nitrate particles as high melting and low melting temperature simulant materials, respectively. The test results showed that debris height decreased gradually as the nitrate particles melt, and molten zone and thermal load on vessel wall were shifted from bottom upwards. Both conductive and convective heat transfer could take place in a solid-liquid mixture pool. These results can support the information from the internal investigations of the primary containment vessel and deepen the understanding of the accident progression.
吉村 一夫; 堂田 哲広; 田中 正暁; 藤崎 竜也*; 村上 諭*; Vilim, R. B.*
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03
原子力機構では、ナトリウム冷却高速炉の設計最適化や安全強化のため、プラント全体応答から局所現象までの一貫した評価を可能とするマルチレベルシミュレーション(MLS)システムを構築している。MLSシステムによる連成手法の妥当性確認のため、1次元プラント動特性解析コードとしてSuper-COPDを、多次元熱流動解析コードとしてFluentを使用した1D-CFD連成解析手法をEBR-IIの自然循環試験に適用した。その結果、プラント全体応答を押さえつつ、上部プレナム,Z型配管やコールドプールの温度成層化現象(多次元熱流動現象)を予測可能であることを確認した。また、実測データとの比較から本手法の自然循環試験への適用性を確認した。
Bentaib, A.*; Chaumeix, N.*; Nyrenstedt, G.*; Bleyer, A.*; Maas, L.*; Gastaldo, L.*; Kljenak, I.*; Dovizio, D.*; Kudriakov, S.*; Schramm, B.*; et al.
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 11 Pages, 2022/03
In case of a core melt-down accident in a light water nuclear reactor, hydrogen is produced during reactor core degradation and released into the reactor building. In case of failure of in-vessel corium retention, a large amount of carbon monoxide (CO) in addition to H
and other gases may be produced during molten core concrete interaction (MCCI). This subsequently creates a combustion hazard. A local ignition of the combustible mixture may generate standing flames or initially slow propagating flames. Depending on geometry, mixture composition and turbulence level, the flame can accelerate or be quenched after a certain distance. The pressure and temperature loads generated by the combustion process may threaten the integrity of the containment building and safety equipment. The evaluation of such loads requires validated codes which can be used with a high level of confidence. Currently, turbulence and steam effect on flame propagation mechanisms are not well reproduced by combustion models usually implemented in safety tools and further model enhancement and validation are still needed. For this purpose and at the initiative of the SAMHYCO-NET project consortium and of the European Technical Safety Organization Network (ETSON), a benchmark on hydrogen combustion was organized with the goal to identify the current level of the computational tools in the area of hydrogen combustion simulation under conditions typical for safety considerations in a Nuclear Power Plant (NPP). This benchmark is composed of four main steps with increasing difficulty starting from flame propagation in homogenous dry atmosphere and finishing with more representative conditions with (H/HO/O/N) stratified mixtures. In this paper, only experiments related to flame propagation in homogenous atmosphere are considered.
安部 諭; 小尾 善男*; 佐藤 聡; 岡垣 百合亜; 柴本 泰照
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 13 Pages, 2022/03
A modeling of heat transfers with boiling transition (BT) and that after the occurrence of BT, called post-BT, is one of the key technical issues to estimate the duration of surface dryout and the peak cladding temperature during DBA (Design Basis Accident) and BDBA (Beyond Design Basis Accident) in light water reactors. Recently, CFD (Computational Fluid Dynamics) has emerged as a powerful tool for representing the heat transfer mechanism. Our main purpose is to obtain in-depth physical insight into the BT and post-BT phenomena by combining experiment and CFD simulation research. This paper introduces our developing activity for an integrated three-field two-fluid CFD methodology based on the Eulerian-Eulerian approach toward the accurate prediction of the dryout behavior from annular-mist to mist flow regimes. We implemented following interaction terms and functions into OpenFOAM ver.7, an opensource code developed by OpenFOAM foundation, as (1) Interaction terms between liquid film and droplets due to deposition and entrainment, (2) Interaction terms on the liquid interface between the liquid film and the gas phase on friction and heat conduction, (3) Heat transfer from the heated wall to the liquid film, (4) Dryout occurrence judgement and the switching function on the boundary condition. The dryout occurrence judgment is based on a correlation on critical film thickness, which is originally applied into the MARS (Multi-dimensional Analysis of Reactor Safety) code. A trial calculation with the developed solver called two Phase Three Field Euler Foam was performed to check the solver operation. The CFD could simulate temperature increase behavior due to the dryout occurrence, whereas here were still challenges in reproducing the transition from mist flow to single-phase vapor flow.
羽賀 勝洋; 粉川 広行; 直江 崇; 涌井 隆; 若井 栄一; 二川 正敏
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 13 Pages, 2022/03
J-PARCではMWクラスの核破砕中性子源を実現するために水銀ターゲットの流路構造としてクロスフロー型ターゲットを開発し、流路構造の改良を継続してきた。高出力の短パルス陽子ビームが水銀ターゲットに入射すると、急激な水銀の発熱と体積膨張により最大40MPaにも達する圧力波が誘起され、ターゲット容器にキャビテーション損傷を生じさせる。このため、水銀流路にマイクロバブル生成器を配置し、バブルの収縮により水銀の体積膨張をクッションのように吸収することで圧力波を低減する技術や、陽子ビームが入射する容器壁に内壁を設けて、狭隘流路に形成される速い水銀流れの大きな速度勾配を利用してキャビテーション損傷を低減する技術などを水銀ターゲット容器の流路構造に導入した。これらの技術開発により、2020年には36.5時間の1MW連続運転を成功させ、2021年4月から最大740kWの高出力で長期の安定な利用運転を達成した。本報告は、主に水銀ターゲット容器の熱流動設計に関して1MW運転を実現するまでの技術開発をまとめたものである。
田中 正暁; 堂田 哲広; 森 健郎; 横山 賢治; 上羽 智之; 岡島 智史; 松下 健太郎; 橋立 竜太; 矢田 浩基
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03
日本原子力研究開発機構では、ARKADIAと呼ぶ原子炉の革新的な設計システムの開発を進めている。ARKADIAは、安全性かつ経済性を高め、脱炭素エネルギー源となる革新的原子炉の設計を実現するものである。最初の開発段階として、設計検討のためのARKADIA-Designと、安全性評価のためのARKADIA-Safetyを開発する。本報告では、ARKADIA-Designに焦点を当て、システムの概念と、マルチレベル解析及びマルチフィジックス解析を実施するための数値解析コードについて説明する。また、解析コードを組み合わせて構築する機能及び妥当性確認としての対象問題についても言及する。
Lebel, L. S.*; Morreale, A. C.*; Freitag, M.*; Gupta, S.*; Allelein, H.-J.*; Klauck, M.*; 孫 昊旻; Herranz, L. E.*
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03
Properly assessing pool scrubbing decontamination factors or radionuclide re-entrainment rates in a reactor safety analysis needs to be supported by a sufficiently robust experimental dataset, based on well qualified aerosol measurement techniques. A review of different pool scrubbing-related source term experiments has been conducted, along with a comparison of the measurement techniques that have been employed. In most areas, a fairly robust dataset exists to assess decontamination factors, but there is still a need to better understand some of the underlying aerosol mechanisms. The available dataset of re-entrainment experiments is smaller, and has gaps, for example, in pools with high velocity gas injections, or with re-flooded corium applications where the pool is undergoing film boiling. There are also many measurement techniques (e.g., cascade impactors, light scattering techniques, phase Doppler anemometry, etc.) that have different capabilities and are suitable for studying different aspects of the experiments. Linking the results that the techniques give, and how their results can ultimately be employed in safety analysis (including uncertainty quantification), is an important consideration in applying the results. This work was performed as a collaborative activity within the framework of the NUGENIA IPRESCA (Integration of Pool scrubbing Research to Enhance Source-term Calculations) project.
廣瀬 意育; 久木田 豊; 柴本 泰照; 佐川 淳*
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 12 Pages, 2022/03
Electric impedance tomography (EIT) is a non-invasive and radiation-free imaging method applicable to gas/liquid two-phase flow measurements. It determines the electrical resistivity distribution of an object from measurements of boundary potentials in response to current injection. Due to the severely ill posed nature of the problem, the quality of reconstructed image depends much on the quality and amount of information available from potential measurements. We have proposed a DC pulse-driven EIT system design equipped with countermeasures for the influences of electrode polarization on potential measurements (Hirose et al., in preparation). The usefulness of EIT in two-phase flow measurement is however restricted by the intrinsically limited spatial resolution. Due to the diffusive nature of electricity, the spatial resolution degrades quickly with the distance from the boundary. In this study, we attempt to improve the spatial resolution by adding thin electrodes inserted into the flow field away from the boundary. Although this means that non-invasiveness is traded off, the influence of invasive electrodes on flow field could be estimated and limited on the basis of experiences gained with other intrusive methods, e.g., needle probes for measurement of interfacial area. The benefit taken by the addition of invasive electrodes, on the other hand, would depend on two-phase flow regime and other flow parameters. In the present paper we consider dispersed bubbly flow and simulate the bubbles with thin cylindrical insulators. The results obtained with and without invasive electrodes are compared to discuss the effectiveness and limitations in measurement of two-phase flow.
Dehbi, A.*; Cheng, X.*; Liao, Y.*; 岡垣 百合亜; Pellegrini, M.*
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 15 Pages, 2022/03
Nuclear degraded cores produce fission product aerosols that may reach the environment if not removed by natural processes and/or filtering equipment. The transport paths of aerosols usually include transits through stagnant water pools. It is therefore essential to develop computational tools to predict the aerosols retention by water pools. Currently, this is mostly done with 1-D lumped-parameter codes that are too simplistic to capture the physics. It is hence worthwhile to attempt the CFD approach, which has recently become reasonably mature to address bubble hydrodynamics in low momentum two-phase flows. In this first comparative exercise, we restrict the investigation to a hypothetical parallelepiped water pool (2
28 cm
) into which air is injected through a circular 4 mm ID orifice at low velocity of 0.2 m/s. We present predictions of the gas phase dynamics (void and velocity profiles) for both Euler-Euler and Interface Tracking (IT, Volume-of-Fluid (VOF)) methodologies. In addition, we compare bubble shape, volume and detachment frequency from various IT simulation codes (CFX, Fluent, Star-CCM+, OpenFOAM). Reasonable agreement is found between IT simulations near the injector, but discrepancies increase as one moves towards the free surface. The disagreement between the Euler-Euler and IT results is substantial throughout the domain. Future studies will consist of validation exercises against experimental data to highlight potential model deficiencies and point to ways of remedying them.
Marchetto, C.*; Ha, K. S*; Herranz, L. E.*; 廣瀬 意育; Jankowski, T.*; Lee, Y.*; Nowack, H.*; Pellegrini, M.*; Sun, X.*
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 17 Pages, 2022/03
After the Fukushima Daiichi accident of March 2011, one of the main concerns of the nuclear industry has been the research works for improving atmospheric radioactive release mitigation systems. Pool scrubbing is an important process in reactors that mitigates radioactive release. It is based on the injection of gases containing fission products through a water pool. Bubble hydrodynamics, as a result of gas injection and the associated water pool thermal-hydraulics, is an important aspect of the process since the bubble size, shape, velocity, etc. influence the fission product trapping at the bubble interface with the water. Computer codes dedicated to the pool scrubbing have been mainly developed in the 90's last century and modelling drawbacks have been identified in particular for bubble hydrodynamics. One of IPRESCA project objectives is to improve the pool scrubbing modelling. In order to highlight the main modelling issues, a benchmark exercise has been performed focusing on the bubble hydrodynamics. This benchmark, performed by nine organisations coming from six countries, aims at simulating a basic configuration, a single upward injector in ambient conditions, experimentally characterized in the RSE tests carried out in the European PASSAM project. In this paper, a short description of the code modelling and a comparison between the code results and the experimental data are presented and discussed. Then, outcomes from the benchmark result analysis and proposals of improvements are emphasized.
小野 綾子; 山下 晋; 鈴木 貴行*; 吉田 啓之
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03
原子力機構では、軽水炉燃料の安全性評価において必須である限界熱流束の評価において、新型燃料設計にかかわる大型モックアップ試験によるコストの削減や、想定外事象に対応するためにモックアップ試験の試験範囲よりも幅広い範囲において、限界熱流束を機構論に基づき評価する研究に着手している。本研究では、機構論的流動解析手法であるJUPITERを用いて、スワールベーンおよびスプリットベーン付きの燃料集合体内の国際ベンチマーク問題を対象とした単相流解析および同体系における二相流解析を実施し、各種ベーンによる流動場および気泡挙動に与える効果、解析における課題の抽出を行った結果を報告する。
Lind, T.*; Herranz, L. E.*; Sonnenkalb, M.*; 丸山 結
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 15 Pages, 2022/03
The accident progression and fission product release from the three damaged units of the Fukushima Daiichi Nuclear Power Plant were systematically investigated in the OECD/NEA BSAF project phases 1 and 2. As a result of those investigations, a good progress was achieved in establishing defendable accident scenarios and the corresponding fission product releases to the environment. Nonetheless, there are some areas requiring further work, particularly concerning fission product behavior. They are addressed in the OECD/NEA project "Analysis of Information from Reactor Buildings and Containment Vessels of Fukushima Daiichi NPS" (ARC-F). Based on the outcome of the BSAF project, several focus areas were selected for further investigations in the ARC-F project, one of them being the behavior of fission products and source term. In this paper, five topics which were ranked with a high significance as open issues based on the BSAF project regarding fission product behavior are discussed: i) fission product speciation, ii) iodine chemistry, iii) pool scrubbing, iv) fission product transport and behavior in the buildings, and v) uncertainty analysis and variant calculations. Significant progress has been made in these five topics in the ARC-F project. In this paper, background is given for choosing these topics for specific investigations based on the outcome of the BSAF project. The topics are described and the approach to study them in the ARC-F given along with some exemplary, preliminary results. Finally, the readers' attention is drawn to open issues which are not included in the ARC-F work scope and could need further attention.
Gupta, S.*; Herranz, L. E.*; Lebel, L. S.*; Sonnenkalb, M.*; Pellegrini, M.*; Marchetto, C.*; 丸山 結; Dehbi, A.*; Suckow, D.*; Krkel, T.*
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03
Pool scrubbing is a major topic in water cooled nuclear reactor technology as it is one of the means for mitigating the source-term to the environment during a severe accident. Pool scrubbing phenomena include coupled interactions between bubble hydrodynamics, aerosols and gaseous radionuclides retention mechanisms under a broad range of thermal-hydraulic conditions as per accident scenarios. Modeling pool scrubbing in some relevant accident scenarios has shown to be affected by substantial uncertainties. In this context, IPRESCA (Integration of Pool scrubbing Research to Enhance Source-term CAlculations) project aims to promote a better integration of international research activities related to pool scrubbing by providing support in experimental research to broaden the current knowledge and database, and by supporting analytical research to facilitate systematic validation and model enhancement of the existing pool scrubbing codes. The project consortium includes more than 30 organisations from 15 countries involving research institutes, universities, TSOs, and industry. For IPRESCA activities, partners join the project with in-kind contributions. IPRESCA operates under NUGENIA Technical Area 2/SARNET (Severe Accident) - Sub Technical Area 2.4 (Source-term). The present paper provides an introduction and overview of the IPRESCA project, including its objectives, organizational structure and the main outcomes of completed activities. Furthermore, key activities currently ongoing or planned in the project framework are also discussed.
内田 真緒*; Alzahrani, H.*; 塩野 幹人*; 堺 公明*; 松下 健太郎; 江連 俊樹; 田中 正暁
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03
ナトリウム冷却高速炉の設計において、炉心反応度への予期せぬ影響の観点からカバーガスの巻込み現象が重要な課題の一つとなり、既往研究において高速炉プレナム内の自由液面部における自由表面渦のガスコア成長を評価するための、渦モデルに基づく評価手法が開発されている。本研究では、非定常渦のガスコア成長の予測精度を明確にするために、開水路試験体系を持つ回流水槽による水試験を実施した。また、実験と同じ体系による数値解析に基づいた評価手法によりガスコア長さを予測し、試験結果と比較した。その結果、試験では、下降流速が大きくなる下流領域においてガスコア長さが大きくなることが観測された。一方、数値解析結果を用いたガスコア長さの予測では、試験とは異なる位置でピークが現れ、ピーク値も過大評価となった。
内堀 昭寛; 椎名 祥己*; 渡部 晃*; 高田 孝*
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 12 Pages, 2022/03
ナトリウム-水反応現象数値解析コードSERAPHIMを開発している。最近の研究で、伝熱管群の存在する複雑形状領域に対応するため非構造格子解析手法を本解析コードへ導入した。本研究では、非構造格子化の効果を確認するため、伝熱管群体系でのナトリウム-水反応試験を構造格子及び非構造格子のそれぞれで解析した。構造格子の影響を受けて生じた流動が、非構造格子では改善されることを確認した。また、試験結果との比較から、非構造格子の解析において、化学反応により上昇した温度のピーク値が測定値と同程度であることを確認した。
大林 寛生; 八巻 賢一*; 北 智士*; 有吉 玄; 佐々 敏信
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 11 Pages, 2022/03
原子力機構では、LBE流れに対し円管外に2つのセンサーを設置する非接触式の超音波流量計を開発した。解決すべき課題は計測配管内での信号の多重反射と固液界面での濡れ性である。これらに対し、固体-固体境界での臨界角の採用と配管内面に対する鏡面仕上げの改善による対応を実施した。350
Cの温度条件における適用試験では、LBE流れに対する有用性が既に実証されているプラグ挿入式に相当する性能を示すことを明らかにするとともに、1500時間以上安定的な信号出力が得られることを実証した。
有吉 玄; 大林 寛生; 斎藤 滋; 佐々 敏信
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 10 Pages, 2022/03
液体重金属を核破砕ターゲットおよび冷却材として利用する核変換施設の建設には、液体重金属の流動特性を事前に明らかにしておくことが重要となる。鉛ビスマス共晶合金(LBE)は、加速器駆動システムや高速炉等の革新的原子力システムにおける核破砕ターゲットや冷却材の候補材料とされており、J-PARCではLBEを利用した核破砕ターゲットシステムの構築に向けた技術開発を推進している。通常、LBEの流動特性は汎用数値流体力学(CFD)コードを用いて予測・把握されるが、このようにCFDコードが多用される一つの要因は、高温液体重金属流れに対する計測手法の確立が不十分なことである。そこで原子力機構では、小型電磁石を使用した局所流速計を開発した。開発した流速計は480CのLBEへ適用され、誘導起電力と流速の良好な相関が得られることが実証された。
Mascari, F.*; Bersano, A.*; Woods, B. G.*; Reyes, J. N.*; Welter, K.*; 中村 秀夫; D'Auria, F.*
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03
Safety analyses have a key role for designing the mitigation strategies and for a safety review process, which are carried-out with best-estimate thermal-hydraulic system codes. Small Modular Reactors (SMRs) adopting passive mitigation strategies under development are characterized by some common features with the current reactors and by other features typical of their designs. While many of Natural Circulation (NC) have been studied, further analyses are necessary to confirm the code capability against experimental data representative of SMR phenomenology. Though different scaling methods have been developed, distortions are unavoidable in the experimental facility design. Then, scaled-down facilities are limited in scaling-up capabilities, which may affect the capability of the code to predict full-scale behavior. Therefore, in a V&V process, uncertainty related to the code scaling-up capability is still an open issue. Since the OSU-MASLWR is scaled in volume and height, this paper aims to assess the scaling-up capability of the OSU-MASLWR Reactor Pressure Vessel nodalization against NC phenomenology typical of SMR, having the OSU-MASLWR-002 single phase NC data as a base. This may give some first insights about the TRACE scaling-up capability against single-phase NC in integral type configuration.