Improvement of model for cesium chemisorption onto stainless steel in severe accident analysis code SAMPSON (Joint research)

Miwa, Shuhei; Karasawa, Hidetoshi; Nakajima, Kunihisa; Kino, Chiaki*; Suzuki, Eriko; Imoto, Jumpei

JAEA-Data/Code 2021-022, 32 Pages, 2023/01

JAEA-Data-Code-2021-022.pdf:1.41MB

The improved model for cesium (Cs) chemisorption onto stainless steel (SS) in the fission product (FP) chemistry database named ECUME was incorporated into the severe accident (SA) analysis code SAMPSON for the more accurate estimation of Cs distribution within nuclear reactor vessels in the TEPCO's Fukushima Daiichi Nuclear Power Station (1F). The SAMPSON with the improved model was verified based on the analysis results reproducing the experimental results which were subjected to the modeling of Cs chemisorption behavior. Then, the experiment in the facility with the temperature gradient tube to simulate SA conditions such as temperature decrease and aerosol formation was analyzed to confirm availability of the improved model to the analysis of Cs chemisorption onto SS. The SAMPSON with the improved model successfully reproduced the experimental results, which indicates that the improved model and the analytical method such as setting a method of node-junction, models of aerosol formation and the calculation method of saturated CsOH vapor pressure can be applicable to the analysis of Cs chemisorption behavior. As the information on water-solubility of Cs deposits was also prerequisite to estimate the Cs distribution in the 1F because Cs can be transported through aqueous phase after the SA, the water-solubility of chemisorbed Cs compounds was investigated. The chemisorbed compounds on SS304 have been identified to CsFeO at 873 K to 973 K with higher water-solubility, CsFeSiO at 973 K to 1273 K and CsSiO at 1073 K to 1273 K with lower water-solubility. From these results, the water-solubility of chemisorbed Cs compounds can be estimated according to the SA analysis conditions such as temperature in the reactor and the CsOH concentration affecting the amount of chemisorbed Cs.

Comprehensive analysis and evaluation of Fukushima Daiichi Nuclear Power Station Unit 3

Yamashita, Takuya; Honda, Takeshi*; Mizokami, Masato*; Nozaki, Kenichiro*; Suzuki, Hiroyuki*; Pellegrini, M.*; Sakai, Takeshi*; Sato, Ikken; Mizokami, Shinya*

Nuclear Technology, 26 Pages, 2023/00

https://doi.org/10.1080/00295450.2022.2157663

Revolatilization of iodine by bubbly flow in the suppression pool during an accident

Nanjo, Kotaro; Ishikawa, Jun; Sugiyama, Tomoyuki; Pellegrini, M.*; Okamoto, Koji*

Journal of Nuclear Science and Technology, 59(11), p.1407 - 1416, 2022/11

https://doi.org/10.1080/00223131.2022.2062065

BWR lower head penetration failure test focusing on eutectic melting

Yamashita, Takuya; Sato, Takumi; Madokoro, Hiroshi; Nagae, Yuji

Annals of Nuclear Energy, 173, p.109129_1 - 109129_15, 2022/08

https://doi.org/10.1016/j.anucene.2022.109129

Study on initiating phase of core disruptive accident (Validation study of SAS4A code for the unprotected transient overpower accident)

Ishida, Shinya; Fukano, Yoshitaka

Nihon Kikai Gakkai Rombunshu (Internet), 88(911), p.21-00304_1 - 21-00304_11, 2022/07

https://doi.org/10.1299/transjsme.21-00304

In previous studies, the reliability and validity of the SAS4A code was enhanced by applying Phenomena Identification and Ranking Table (PIRT) approach to the Unprotected Loss of Flow (ULOF). SAS4A code has been developed to analyze the early stage of Core Disruptive Accident (CDA), which is named Initiating Phase (IP). In this study, PIRT approach was applied to Unprotected Transient over Power (UTOP), which was one of the most important and typical events in CDA as well as ULOF. The phenomena were identified by the investigation of UTOP event progression and physical phenomena relating to UTOP were ranked. 8 key phenomena were identified and the differences in ranking between UTOP and ULOF were clarified. The code validation matrix was completed and an SAS4A model, which was not validated in ULOF, was identified and validated. SAS4A code became applicable to various scenarios by using PIRT approach to UTOP and the reliability and validity of SAS4A code were significantly enhanced.

Post-test analyses of the CMMR-4 test

Yamashita, Takuya; Madokoro, Hiroshi; Sato, Ikken

Journal of Nuclear Engineering and Radiation Science, 8(2), p.021701_1 - 021701_13, 2022/04

https://doi.org/10.1115/1.4051443

France-Japan collaboration on thermodynamic and kinetic studies of core material mixture in severe accidents of sodium-cooled fast reactors

Quaini, A.*; Goss, S.*; Payot, F.*; Suteau, C.*; Delacroix, J.*; Saas, L.*; Gubernatis, P.*; Martin-Lopez, E.*; Yamano, Hidemasa; Takai, Toshihide; et al.

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 10 Pages, 2022/04

CEA and JAEA defined new sub-tasks under the current implementing arrangement: Kinetics of interaction in core material mixtures- Physical properties of core material mixtures, High temperature thermodynamic data for the UO-Fe-BC system, Experimental studies on BC-SS kinetics and BC-SS eutectic material relocation (freezing), BC/SS eutectic and kinetics models for SIMMER code systems, Methodology for the modelling of mixtures liquefaction kinetics. The paper describes major R&D results obtained in the France-Japan collaboration under the previous implementing arrangement as well as experimental and analytical roadmaps under the current arrangement.

Modelling and simulation of source term for sodium-cooled fast reactor under hypothetical severe accident; Primary system/containment system interface source term estimation

Onoda, Yuichi; John Arul, A.*; Klimonov, I.*; Danting, S.*

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 13 Pages, 2022/04

Release behaviors of elements from an Ag-In-Cd control rod alloy at temperatures up to 1673 K

Nagase, Fumihisa; Otomo, Takashi; Uetsuka, Hiroshi*

Nuclear Technology, 208(3), p.484 - 493, 2022/03

https://doi.org/10.1080/00295450.2021.1905472

An Ag-In-Cd control rod alloy was heated in argon or oxygen at 1073-1673 K for 60-3600 s and the release behavior of the elements was examined. Complete liquefaction of the alloy occurred between 1123 and 1173 K, and elemental release was quite limited below the liquefaction temperature. In argon, almost all of the Cd content was released within 3600 s at 1173 K and within 60 s at 1573 K, while the released fractions of Ag and In were 3% and 8%, respectively. In oxygen, the release of Cd, which was quite small at temperatures up to 1573 K, drastically increased to 30-50% at 1673 K for short periods. Releases of Ag and In were also small in oxygen under the examined conditions. Comparison with the experimental data suggests that conventional empirical release models may underestimate the Cd release at lower temperatures just after control rod failure in severe accidents.

Effect of nitrous acid on migration behavior of gaseous ruthenium tetroxide into liquid phase

Yoshida, Naoki; Ono, Takuya; Yoshida, Ryoichiro; Amano, Yuki; Abe, Hitoshi

JAEA-Research 2021-011, 12 Pages, 2022/01

JAEA-Research-2021-011.pdf:1.49MB

In boiling and drying accidents involving high-level liquid waste in fuel reprocessing plants, emphasis is placed on the behavior of ruthenium (Ru). Ru would form volatile species, such as ruthenium tetroxide (RuO), and could be released to the environment with coexisting gases, including nitric acid, water, or nitrogen oxides. In this study, to contribute toward safety evaluations of these types of accidents, the migration behavior of gaseous Ru into the liquid phase has been experimentally measured by simulating the condensate during an accident. The gas absorption of RuO was enhanced by increasing the nitrous acid (HNO) concentration in the liquid phase, indicating the occurrence of chemical absorption. In control experiments without HNO, the lower the temperature, the greater was the Ru recovery ratio in the liquid phase. Conversely, in experiments with HNO, the higher the temperature, the higher the recovery ratio, suggesting that the reaction involved in chemical absorption was activated at higher temperatures.