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論文

Scaling-up capabilities of TRACE integral reactor nodalization against natural circulation phenomena in small modular reactors

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.

論文

The Working group on the analysis and management of accidents (WGAMA); A Historical review of major contributions

Herranz, L. E.*; Jacquemain, D.*; Nitheanandan, T.*; Sandberg, N.*; Barr$'e$, F.*; Bechta, S.*; Choi, K.-Y.*; D'Auria, F.*; Lee, R.*; 中村 秀夫

Progress in Nuclear Energy, 127, p.103432_1 - 103432_14, 2020/09

WGAMA started on Dec. 31st 1999 to assess and strengthen the technical basis needed for the prevention, mitigation and management of potential accidents in NPP and to facilitate international convergence on safety issues and AM analyses and strategies. WGAMA addresses reactor thermal-hydraulics (Thys), in-vessel behavior of degraded cores, containment behavior and protection, and FP release, transport, deposition and retention, for both current and advanced reactors. This paper summarizes such WGAMA contributions in Thys, CFD and severe accidents, which include the Fukushima-Daiichi accident impacts on the WGAMA activities and their substantial outcomes. Around 50 technical reports have become reference in the related fields, which appear in References. Recommendations in these reports include further research, some of which have given rise to the joint projects conducted or underway within the OECD framework. Ongoing WGAMA activities are numerous and a number of them are to be launched in the near future, which are shortly mentioned too.

論文

Scaling issues for the experimental characterization of reactor coolant system in integral test facilities and role of system code as extrapolation tool

Mascari, F.*; 中村 秀夫; Umminger, K.*; De Rosa, F.*; D'Auria, F.*

Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.4921 - 4934, 2015/08

The phenomenological analyses and thermal hydraulic characterization of a nuclear reactor are the basis for its design and safety evaluation. Scaled down tests of Integral Effect Test (IET) and Separate Effect Test (SET) are feasible to develop database. Though several scaling methods such as Power/Volume, Three level scaling and H2TS have been developed and applied to the IET and SET design, direct extrapolation of the data to prototype is in general difficult due to unavoidable scaling distortions. Constraints in construction and funding for test facility demand that a scaling compromise is inevitable further. Scaling approaches such as preservation of time, pressure and power etc. have to be adopted in the facility design. This paper analyzes some IET scaling approaches, starting from a brief analysis of the main characteristics of IETs and SETFs. Scaling approaches and their constraints in ROSA-III, FIST and PIPER-ONE facility are used to analyze their impact to the experimental prediction in Small Break LOCA counterpart tests. The liquid level behavior in the core are discussed for facility scaling-up limits.

論文

Effectiveness of Core Exit Thermocouple (CET) indication in accident management of light water reactors

中村 秀夫; T$'o$th, I.*; Sandervag, O.*; Umminger, K.*; Dreier, J.*; Prior, R.*; Alonso, J. R.*; Muellner, N.*; D'Auria, F.*; M$"u$hleisen, A.*; et al.

Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14) (CD-ROM), 15 Pages, 2011/09

OECD原子力機関(NEA)の原子力施設安全委員会(CSNI)の事故の分析と管理ワーキンググループ(WGAMA)は、軽水炉(LWR)のアクシデントマネジメント(AM)における炉心出口温度計(CET)の有効性に関するタスクを実施した。同タスクはNEA加盟国に対してCETの利用に関する調査を行い、軽水炉事故時のAM策へのCETの利用に際するCETの設計基準を調べるとともに、特に事故時の炉心温度上昇に対するCET温度表示の時間遅れや温度表示の乖離に焦点を当て、LOFT, ROSA/LSTF, PKL, PSB-VVERなどのシステム効果実験装置で行われてきた実験結果を調べた。また、それらの実験結果の実機LWRへの外挿適用について、スケーリング上の課題を議論した。本論文は、同タスクによって得られた成果をまとめ、今後の課題を示すものである。

口頭

Scaling rationale design and extrapolation problem for ITF and SETF

Mascari, F.*; 中村 秀夫; De Rosa, F.*; D'Auria, F.*

no journal, , 

In the development and safety evaluation of LWRs, thermal hydraulic analysis of the reactor, containment and their coupling is essential to understand the accident phenomena. To reproduce the behavior in a scaled model, it is necessary to properly characterize thermal hydraulics both in the local and integral responses. The facility geometry and test conditions should then be correctly derived according to scaling laws to avoid scaling distortions that could compromise the target phenomena identified by PIRT process. Many scaling approach and methods have thus been developed. Together with the scaling analysis, computer codes may be used in supporting the design of test facilities, assessing the scale distortions, and verifying the selected scaling method. However, since the closure laws in the computer code are mainly based on scaled test data, the extrapolation of code results remains a challenging and open issue. This paper provides some insights about the methods used in the scaling.

口頭

Scaling approaches and system code as extrapolation tool

Mascari, F.*; 中村 秀夫; Umminger, K.*; Moon, S.-K.*; Lien, P.*; Bestion, D.*; D'Auria, F.*

no journal, , 

軽水炉事故時の安全評価には解析コードが用いられるが、採用される多数のモデルや相関式はほとんど全て、実機から縮小された実験により得られており、解析結果への現象のスケーリングの影響を考慮する必要がある。本発表では、2016年にOECD NEAによって取りまとめられたScalingに関する最新情報レポート(State-of-Art Report)などを振り返り、現象のスケーリングに際する体積比や高さ比などパラメータの影響、事故模擬を行うシステム効果試験装置の特徴やデータの範囲、カウンターパート試験の例としてのBWR事故模擬試験などを基に、スケーリング(外挿)を行うツールとしての解析コードの有効性や限界を概説する。

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