Differential pressure changes of a high airflow-type HEPA filter during solvent fire in reprocessing facilities

Tashiro, Shinsuke; Uchiyama, Gunzo; Ono, Takuya; Amano, Yuki; Yoshida, Ryoichiro; Abe, Hitoshi

Nuclear Technology, 208(7), p.1205 - 1213, 2022/07

https://doi.org/10.1080/00295450.2021.2018272

A clogging behavior of a high-efficiency particulate air (HEPA) filter at solvent fire accidents for reprocessing facilities has been studied. In this study, the burning rates of 30% tri-butyl phosphate (TBP)/dodecane (DD) mixed solvent and DD solvent and the differential pressure (P) of a high airflow typed HEPA filter applied in the actual facilities in Japan were measured. It was confirmed that the mainly burned was DD at the early stage of the mixed solvent burning and the TBP at the late stage. Furthermore, it was found that the P rapidly rose at the late stage of the mixed solvent burning. The increase of the release ratio of the unburned particulate composition (TBP, its degraded solvent and inorganic phosphorus (PO)) was considered to contribute to the rapid rise. The correlating formulas with the P and the mass of the loading particulates, except for the region of the rapid rise of P, could be induced.

Experimental investigation on fiber-coupled Raman spectrometry in presence of aerosols; Application for reactor containment gas detection in severe accident conditions

Sun, Haomin; Porcheron, E.*; Magne, S.*; Leroy, M.*; Dhote, J.*; Ruffien Ciszak, A.*; Bentaib, A.*

Proceedings of OECD/NEA Specialist Workshop on Advanced Measurement Method and Instrumentation for enhancing Severe Accident Management in an NPP addressing Emergency, Stabilization and Long-term Recovery Phases (SAMMI 2020) (Internet), 10 Pages, 2020/12

Study on loss-of-cooling and loss-of-coolant accidents in spent fuel pool; Confirmation of fuel temperature calculation function with oxidation reaction in the SAMPSON code

Suzuki, Hiroaki*; Morita, Yoshihiro*; Naito, Masanori*; Nemoto, Yoshiyuki; Kaji, Yoshiyuki

Mechanical Engineering Journal (Internet), 7(3), p.19-00450_1 - 19-00450_17, 2020/06

https://doi.org/10.1299/mej.19-00450

In this study, the SAMPSON code was modified to evaluate severe accidents in a spent fuel pool (SFP). Air oxidation models based on oxidation data obtained on the Zircaroy-4 cladding (ANL model) and the Zircaroy-2 cladding (JAEA model) were included in the modified SAMPSON code. Experiments done by Sandia National Laboratory using simulated fuel assemblies equivalent to those of an actual BWR plant were analyzed by the modified SAMPSON code to confirm the functions for analysis of the severe SFP accidents. The rapid fuel rod temperature rise due to the Zr air oxidation reaction could be reasonably evaluated by the SAMPSON analysis. The SFP accident analyses were conducted with different initial water levels which were no water, water level at bottom of active fuel, and water level at half of active fuel. The present analysis showed that the earliest temperature rise of the fuel rod surface occurred when there was no water in the SFP and natural circulation of air became possible.

Creep damage evaluations for BWR lower head in severe accident

Katsuyama, Jinya; Yamaguchi, Yoshihito; Nemoto, Yoshiyuki; Kaji, Yoshiyuki; Osaka, Masahiko

Transactions of the 24th International Conference on Structural Mechanics in Reactor Technology (SMiRT-24) (USB Flash Drive), 11 Pages, 2017/08

It is difficult to assess rupture behavior of the lower head of reactor pressure vessel in boiling-water-type nuclear power plants due to severe accident like Fukushima Daiichi because Boiling Water Reactor (BWR) lower heads have geometrically complicated structure with a lot of penetrations. Therefore, we have been developing an analysis method to predict time and location of RPV lower head rupture of BWRs considering creep damage mechanisms based on coupled analysis of three-dimensional Thermal-Hydraulics (TH) and thermal-elastic-plastic-creep analyses. In this study, we performed creep damage evaluations to investigate the effects of the debris depth and heat generation locations on failure behavior of lower head. From the analysis results, we discussed the outflow paths of the relocated molten core to the containment, and it was concluded that failure regions of BWR lower head are only the control rod guide tubes or stub tubes under simulated conditions.

Oxidation behavior of fuel cladding tube in spent fuel pool accident condition

Nemoto, Yoshiyuki; Kaji, Yoshiyuki; Ogawa, Chihiro; Nakashima, Kazuo*; Tojo, Masayuki*

Zairyo To Kankyo, 66(5), p.180 - 187, 2017/05

https://doi.org/10.3323/jcorr.66.180

In spent fuel pool (SFP) under loss-of-cooling or loss-of-coolant severe accident condition, the spent fuels will be exposed to air and heated by their own residual decay heat. Integrity of fuel cladding is crucial for SFP safety therefore study on cladding oxidation in air at high temperature is important. Zircaloy-2 (Zry2) and zircaloy-4 (Zry4) were applied for thermogravimetric analyses (TGA) in different temperatures and different air flow rates in this work. Oxidation rate increased with temperature. In range of air flow rate predicted in spent fuel lack during SFP accident, influence of flow rate was not clearly observed below 950C for Zry2 and below 1050C for Zry4. Over these temperature, oxidation rates appeared obviously higher in higher air flow rate, and this trend became clearer when temperature increased. Oxide layers were carefully examined after the oxidation tests and compared with the mass gain data in TGA to investigate detail of air oxidation process. The results revealed that mass gain data in the pre breakaway transition stage reflects growth of the dense oxide film on specimen surface, and in the post breakaway transition stage, it reflects growth of porous oxide layer beneath the breakaway cracking of the oxide film.

Accident measures of serious accidents

Tamaki, Hitoshi

Tekisuto "Kakunenryo Saikuru" (Internet), 3 Pages, 2014/12

http://www.aesj.or.jp/~recycle/nfctxt/nfctxt_1-5.pdf

Revised regulatory rule of nuclear facilities in Japan defines "Serious accident", and require accident measures for them to not only power reactors, but also fuel reprocessing and fabrication facilities. This document describes the basis of accident measures for Serious accident at nuclear fuel reprocessing and fabrication facilities, as a part of "Text of Nuclear Fuel Cycle".

Hot atom chemistry and severe accident of water cooled fission reactor

Saeki, Masakatsu

Handbook of Hot Atom Chemistry, p.637 - 646, 1992/00

no abstracts in English

Peer review on the safety requirements for the prototype fast breeder reactor "Monju", 2; Summary of domestic review

Sakai, Takaaki

no journal, , 

no abstracts in English

Peer review on the safety requirements for the prototype fast breeder reactor "Monju", 1; Overview of safety requirement and domestic/international review

Nakai, Ryodai

no journal, , 

An advisory panel on Monju safety requirements issued a report on the safety requirements for the Prototype Fast Breeder Reactor "Monju" in the last July. The issued report was reviewed both by domestic and international experts on FBR safety. The paper presents the outline of the report and its domestic and international review.

Study on improvement of safety for accident conditions in spent fuel pool, 4; Loss of coolant accident analysis by MAAP code

Nishimura, Satoshi*; Satake, Masaaki*; Soga, Shota*; Nishi, Yoshihisa*; Kaji, Yoshiyuki; Nemoto, Yoshiyuki

no journal, , 

We conducted the transient analysis for the phenomenon of loss of coolant in Spent Fuel Pool by Severe Accident code MAAP version 5.03 and evaluated quantitatively the influence of initial water level and decay heat on failure time and hydrogen generation of fuel cladding.