Development of a handy criticality analysis tool for fuel debris

Tada, Kenichi

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 4 Pages, 2019/05

The decommissioning of Fukushima Daiichi Nuclear Power Plant accident is one of the most important issues in Japan. The criticality safety of fuel debris is imperative to prevent exposure of workers. The investigating criticality monitoring system cannot detect the criticality of fuel debris quickly. The estimation of criticality of fuel debris is required for the fuel debris retrieval. Though the expert knowledge of reactor physics is necessary to estimate the criticality of fuel debris, many people who make a plan of fuel debris retrieval may not know well about criticality analysis. We developed a handy criticality analysis tool HAND to quickly estimate the criticality of fuel debris without expert knowledge of reactor physics. Since the input data of HAND is so simple and users can intuitively understand the calculation results, this tool is expected to be the effective tool to estimate the criticality of fuel debris.

Development of fission product chemistry database ECUME for the LWR severe accident

Miwa, Shuhei; Miyahara, Naoya; Nakajima, Kunihisa; Nishioka, Shunichiro; Suzuki, Eriko; Horiguchi, Naoki; Liu, J.; Miradji, F.; Imoto, Jumpei; Mohamad, B. A.; et al.

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05

We constructed the first version of fission product (FP) chemistry database named ECUME for LWR severe accident. The first version of ECUME is equipped with dataset of the chemical reactions and their kinetics constants for the reactions of cesium(Cs)-iodine(I)-boron(B)-molybdenum(Mo)-oxygen(O)-hydrogen(H) system in gas phase, the elemental model for the high temperature chemical reaction of Cs with stainless steel, and thermodynamic data for CsBO vapor species and solids of CsSiO and CsFeSiO. The ECUME will provide more accurate estimation of Cs distribution due to the evaluation of effects of interaction with BWR control material B and stainless steel on Cs behavior in the Fukushima Daiichi Nuclear Power Station.

Activities of the GIF safety and operation project of sodium-cooled fast reactor systems

Yamano, Hidemasa; Vasile, A.*; Kang, S.-H.*; Summer, T.*; Tsige-Tamirat, H.*; Wang, J.*; Ashurko, I.*

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 7 Pages, 2019/05

The Generation IV (GEN-IV) international forum is a framework for international co-operation in research and development for the next generation of nuclear energy systems. Within the GEN-IV sodium-cooled fast reactor (SFR) system arrangement, the SFR Safety and Operation (SO) project addresses the areas of safety technology and reactor operation technology developments. The aims of the SO project include (1) analyses and experiments that support establishing safety approaches and validating performance of specific safety features, (2) development and verification of computational tools and validation of models employed in safety assessment and facility licensing, and (3) acquisition of reactor operation technology, as determined largely from experience and testing in operating SFR plants. The tasks in the SO topics are categorized into the following three work packages (WP): WP-SO-1 "Methods, Models and Codes", WP-SO-2 "Experimental Programs and Operational Experience", and WP-SO-3 "Studies of Innovative Design and Safety Systems". This paper reports recent activities within the SO project.

Study on loss-of-cooling and loss-of-coolant accidents in spent fuel pool, 4; Investigation of fuel loading effects in BWR spent fuel rack

Tojo, Masayuki*; Kanazawa, Toru*; Nakashima, Kazuo*; Iwamoto, Tatsuya*; Kobayashi, Kensuke*; Goto, Daisuke*; Nemoto, Yoshiyuki; Kaji, Yoshiyuki

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 13 Pages, 2019/05

In this study, fuel loading effects in BWR spent fuel rack accidents are widely investigated using three-dimensional analysis methods from both nuclear and thermal hydraulics viewpoints, including: (a) Decay heat of spent fuel after discharge, (b) The maximum temperature of spent fuel cladding in the spent fuel rack depending on heat transfer phenomena, and (c) Criticality of the spent fuel rack after collapsing of the fuel due to a severe accidents in the BWR spent fuel pool (SFP).

Effect of experimental setting and surface roughness on oxidation behavior of Zry-4 in steam at 1273 K

Negyesi, M.; Amaya, Masaki

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05

Simulation analysis on local damage to reinforced concrete panels subjected to oblique impact by different projectiles, 1; Comparison of impact behavior for rigid projectiles with flat and hemispherical nose shape

Kang, Z.; Nagai, Minoru*; Nishida, Akemi; Tsubota, Haruji; Li, Y.

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 9 Pages, 2019/05

Many empirical formulae have been proposed for evaluating the local damage to reinforced concrete (RC) structures caused by rigid projectile impact. The majority of these formulae aim at impact tests perpendicular to target structures, while few impact tests oblique to the target structure have been studied. The final objective of this study is to propose a new formula for evaluating the local damage to RC structures caused by oblique impact based on past experimental and simulation results. The finite element code LS-DYNA R7.1.2 is used to perform the numerical analysis by adopting Lagrangian finite elements and explicit time integration. So far, we validated the analytical method by comparison with the experimental results and conduct the simulation analysis of impact assessment on RC panel by rigid/soft projectile with flat nose shape using the validated approach. Results of reduction coefficient with respect to rigid/soft projectile and impact angle were obtained. Therefore, in this study, we focus on the impact problems caused by rigid projectile with hemispherical nose shape. The same analytical method is used to simulate the local damage to RC panels caused by oblique impact of rigid projectile with hemispherical nose shape. The results associated with penetration depth of RC structure, energy contribution ratio, etc. are presented. According to the comparison analysis of results of local damage to RC structure by rigid projectiles with flat and hemispherical nose shape, the influence of different nose shapes of rigid projectile on the local damage of RC panels caused by oblique impact is investigated.

Simulation analysis on local damage to reinforced concrete panels subjected to oblique impact by different projectiles, 2; Comparison of impact behavior for soft projectiles with flat and hemispherical nose shape

Nagai, Minoru*; Kang, Z.; Nishida, Akemi; Tsubota, Haruji; Li, Y.

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05

In this study, the final purpose is to propose a new formula for evaluating the local damage caused by oblique impact based on past experimental results and previous research achievements. Up to now, we validated the analytical method by comparison with the experimental results and conducted simulation analysis associated with impact assessment on RC panel by soft/rigid projectile with flat nose shape using the validated approach. In the part 1 of this study, the same procedure of our previous work is followed to investigate the local damage to RC panel caused by rigid projectile with flat and hemispherical nose shape. In the part 2, we focus on the comparison analysis of simulation results of local damage to RC panel subjected to oblique impact by soft missile with flat and hemispherical nose shape. The structural damage of RC panel and projectiles, energy contribution ratio, etc. is studied for each case. The results indicate the difference of nose shape of projectile is of great importance to influence the penetration depth generated by oblique impact of soft projectile.

Coolability evaluation of debris bed on core catcher in a sodium-cooled fast reactor

Matsuo, Eiji*; Sasa, Kyohei*; Koyama, Kazuya*; Yamano, Hidemasa; Kubo, Shigenobu; Hourcade, E.*; Bertrand, F.*; Marie, N.*; Bachrata, A.*; Dirat, J. F.*

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 5 Pages, 2019/05

Discharged molten-fuel from the core during Core Disruptive Accident (CDA) could become solidified particle debris by fuel-coolant interaction in the lower sodium plenum, and then the debris could form a bed on a core catcher located at the bottom of the reactor vessel. Coolability evaluations for the debris bed are necessary for the design of the core catcher. The purpose of this study is to evaluate the coolability of the debris bed on the core catcher for the ASTRID design. For this purpose, as a first step, the coolability calculations of the debris beds formed both in short term and later phase have been performed by modeling only the debris bed itself. Thus, details of core catcher design and decay heat removal system are not described in this paper. In all the calculations, coolant temperature around the debris bed is a parameter. The calculation tool is the debris bed module implemented into a one-dimensional plant dynamics code, Super-COPD. The evaluations have shown that the debris beds formed both in short term and later phase are coolable by the design which secures sufficient coolant flow around the core catcher located in the cold pool.

Study on loss-of-cooling and loss-of-coolant accidents in spent fuel pool, 1; Overview

Kaji, Yoshiyuki; Nemoto, Yoshiyuki; Nagatake, Taku; Yoshida, Hiroyuki; Tojo, Masayuki*; Goto, Daisuke*; Nishimura, Satoshi*; Suzuki, Hiroaki*; Yamato, Masaaki*; Watanabe, Satoshi*

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05

In this research program, cladding oxidation model in SFP accident condition, and numerical simulation method to evaluate capability of spray cooling system which was deployed for spent fuel cooling during SFP accident, have been developed. These were introduced into the severe accident codes such as MAAP and SAMPSON, and SFP accident analyses were conducted. Analyses using Computational Fluid Dynamics (CFD) code were conducted as well for the comparison with SA code analyses and investigation of detail in the SFP accident. In addition, three-dimensional criticality analysis method was developed as well, and safer loading pattern of spent fuels in pool was investigated.

Study on loss-of-cooling and loss-of-coolant accidents in spent fuel pool, 5; Investigation of cooling effects of SFP spray and alternate water injection with MAAP code

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

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05

In this study, accident progression analyses in the SFP were performed to investigate cooling effects of the SFP spray and an alternate water injection in the loss-of-pool water accident with MAAP ver. 5.05 beta. Fuel cladding oxidation model which was created by JAEA based on their experimental data was selected and applied in the present calculations. In case of an assessment of SFP spray effects, decay heat, spray fraction going into the fuel assembly, spray droplet diameter, spray start time were selected as analytical parameters. When the SFP spray of 12.5 kg/s (200 GPM) starts 4 hours after the onset of the accident against the spent fuels with 4 months cooling and if the spray fraction going into the fuel assembly is greater than 30%, the maximum cladding temperature can be maintained under 727C (1000 K), resulting in avoiding the cladding failure.