Development of analytical approach of source term for accident of evaporation to dryness by boiling of reprocessed high level liquid waste

Yoshida, Kazuo; Tamaki, Hitoshi; Hiyama, Mina*

JAEA-Research 2023-001, 26 Pages, 2023/05

JAEA-Research-2023-001.pdf:1.61MB

An accident of evaporation to dryness by boiling of high-level radioactive liquid waste (HLLW) is postulated as one of the severe accidents caused by the loss of cooling function at a fuel reprocessing plant. In this case, volatile radioactive materials, such as ruthenium (Ru) are released from the tanks with water and nitric-acid mixed vapor into the atmosphere. Accurate quantitative estimation of released Ru is one of the important issues for risk assessment of those facilities. To resolve this issue, an analytical approach has been developed using computer simulation programs to assess the radioactive source term from those facilities. The proposed approach consists analyses with three computer programs. At first, the simulation of boiling behavior in the HLLW tank is conducted with SHAWED code. Next step, the thermal-hydraulic behavior in the facility building is simulated with MELCOR code based on the results at the first step simulation such as flowed out mixed steam flow rate, temperature and volatilized Ru from the tank. The final analysis step is carried out for estimating amount of released radioactive materials with SCHERN computer code which simulates chemical behaviors of nitric acid, nitrogen oxide and Ru based on the condition also simulated MELCOR. Series of sample simulations of the accident at a hypothetical typical facility are presented with the data transfer between those codes in this report.

Uncertainty analysis of dynamic PRA using nested Monte Carlo simulations and multi-fidelity models

Zheng, X.; Tamaki, Hitoshi; Takahara, Shogo; Sugiyama, Tomoyuki; Maruyama, Yu

Proceedings of Probabilistic Safety Assessment and Management (PSAM16) (Internet), 10 Pages, 2022/09

https://www.iapsam.org/PSAM16/

MELCOR validation study on sodium pool fire model with comparison to SPHINCS

Louie, D. L. Y.*; Aoyagi, Mitsuhiro

Proceedings of International Topical Meetings on Advances in Thermal Hydraulics (ATH 2022) (Internet), p.316 - 329, 2022/06

Development of evaluation framework for ex-vessel core coolability

Matsumoto, Toshinori; Iwasawa, Yuzuru; Sugiyama, Tomoyuki

Proceedings of Reactor core and Containment Cooling Systems, Long-term management and reliability (RCCS 2021) (Internet), 8 Pages, 2021/10

A methodological framework is being developed in JAEA for evaluating debris coolability at ex-vessel during the severe accident (SA) of BWR under the wet cavity strategy. The probability of ex-vessel debris coolability under the wet cavity strategy is analyzed to demonstrate the evaluation approach. Probabilistic distribution of the melt conditions ejected from the RPV was obtained as the result of the iterative analyses with MELCOR code. Five uncertainty parameters relating with the core degradation and transfer process were chosen. Parameter sets were generated by Latin hypercube sampling (LHS). JASMINE code plays the physical model to predict the mass fraction of agglomerated debris and melt pool spreading on the floor. Fifty-nine input parameter set for JASMINE code were generated by LHS again using the probabilistic distribution of melt condition determined from the results of MELCOR analyses. The depth of the water pool was set as 0.5, 1.0 and 2.0 m. The accumulated debris height was compared with the criterion to judge the debris coolability. As the result, the success probability of debris cooling was obtained through the sequence of calculations.

Sodium fire collaborative study progress; CNWG fiscal year 2019

Louie, D. L. Y.*; Uchibori, Akihiro

SAND2019-15043 (Internet), 35 Pages, 2019/12

https://doi.org/10.2172/1592872

This report describes the progress on the sodium fire research in fiscal year 2019 in the Civil Nuclear Energy Research and Development Working Group (CNWG). In this study, the validation study of the sodium pool fire model incorporated into the MELCOR code, which was originally developed for accident analysis in light water reactors, was carried out through the numerical analysis on the sodium pool fire experiment named F7-1. In this preliminary analysis, pool and atmosphere temperature went up to the same level with the measured results, while the unnatural behavior appeared in the latter half of the analysis. Based on this result, recommendations for improvement were made for a new analysis in next fiscal year, 2020.

Examination of I and Cs releases during late phase of Fukushima Daiichi NPP accident by using I/Cs ratio of source terms evaluated reversely by WSPEEDI code with environmental monitoring data

Hidaka, Akihide; Yokoyama, Hiroya

Journal of Nuclear Science and Technology, 54(8), p.819 - 829, 2017/08

AA2016-0500.pdf:0.44MB

https://doi.org/10.1080/00223131.2017.1323691

To clarify what happened during the Fukushima accident, the phenomena within RPV and the discussion of ties with the environmental monitoring are very important. However, the previous study has not necessarily advanced until the present that passed almost six years from the accident. The present study investigated I and Cs release behaviors during the late phase of the accident based on I/Cs ratio of the source terms that were recently evaluated backward by WSPEEDI code based on environmental monitoring data. The I release from the contaminated water in the basement of 1F2 and 1F3 reactor buildings was evaluated to be about 10% of I source term. The increase in Cs release from March 21 to 23 and from March 30 to 31 could be explained by the release of CsBO which is formed as a result of chemical reactions of Cs with BC due to re-ascension of the core temperature caused by slight shortage of the core cooling water.

Analysis of pressure- and temperature- induced steam generator tube rupture during PWR severe accident initiated from station blackout

Hidaka, Akihide; Maruyama, Yu; Nakamura, Hideo

Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 15 Pages, 2004/00

Severe accident studies showed that Direct Containment Heating issue was resolved for PWRs because a creep rupture at pressurizer surge line would occur prior to the melt-through of Reactor Pressure Vessel during station blackout (TMLB'). However, it was recently concerned that, if the secondary system is depressurized during TMLB', the creep rupture at SG U-tubes would occur earlier than the surge line. This pressure- and temperature-induced SG U-tube rupture (PTI-SGTR) is not preferable because of the increase in offsite consequences. The SCDAP/RELAP5 analyses by USNRC showed that the surge line would fail earlier than the U-tubes. However, the analyses used a coarse nodilization for steam mixing at the SG inlet plenum that could affect the temperature of U-tubes. To investigate the effect of steam mixing, an analysis was performed with MELCOR1.8.4. The analysis showed that the surge line would fail earliest during TMLB' while the U-tubes could fail earliest during TMLB' with secondary system depressurization. Further investigation is needed for occurrence conditions of PTI-SGTR.

Depressurization analyses of PWR station blackout with MELCOR 1.8.4

Antariksawan, A. R.*; Hidaka, Akihide; Moriyama, Kiyofumi; Hashimoto, Kazuichiro*

JAERI-Tech 2001-011, 116 Pages, 2001/03

JAERI-Tech-2001-011.pdf:4.02MB

no abstracts in English

Analysis of steam generator tube rupture as a severe accident using MELCOR1.8.4

H.Yang*; Hidaka, Akihide; Sugimoto, Jun

JAERI-Tech 99-013, 97 Pages, 1999/03

JAERI-Tech-99-013.pdf:3.24MB

no abstracts in English

Influence of iodine chemistry in accumulated water at reactor buildings on late phase source term at Fukushima NPP accident

Hidaka, Akihide

no journal, , 

During core cooling at Fukushima Daiichi NPP accident, large amount of contaminated water was accumulated in the basements of reactor buildings at Units 1 to 4. The estimated ratios of I and Cs quantities in water to the core inventories are 0.51%, 0.85% at Unit 1, 74%, 38% at Unit 2 and 26%, 18% at Unit 3, respectively. According to the Henry's law, certain fraction of iodine in water could be released to atmosphere. Many evaluations for I-131 release have been performed so far by MELCOR or the reverse estimation with SPEEDI. The SPEEDI reverse predicted significant release until March 26 while no prediction in MELCOR after March 17. The present study showed that iodine release from accumulated water due to radiolytic conversion from I to I and gas-liquid partition of I may explain the release between March 17 and 26. This strongly suggests a need for improvement of current MELCOR approach which treats the release only from containment breaks.