Daiichi Nuclear Power Station Unit Research Articles

Early-phase core degradation interpretation by interaction between fuel cladding and fuel components in Fukushima Daiichi nuclear power plant Unit 2

ABSTRACT The early phase of the core degradation at the Fukushima Daiichi Nuclear Power Station Unit 2 was investigated in terms of a fuel-component degradation by materials interaction by carrying out the accident analysis, evaluation of chemical conditions and experimental works. First, the accident scenario was evaluated to specify a target period (from 18:40 to 22:40 on NaN Invalid Date NaN), and then the temperature and H2/H2O pressure ratio in the core region during this period were calculated based on plant parameters. Subsequently, semi-integral experiments were performed using the mocked-up fuel components of a boiling water reactor (BWR). Furthermore, separate-effects tests using Ni-alloy/Zircaloy4 were also performed to understand the phenomenon observed in the semi-integral experiments, where the Ni-Zr reaction occurred even with the oxide layer on the Zr-based alloy. Consequently, it was estimated that the Zr-Ni eutectic melting reaction above 1100°C under steam-starved conditions resulted in a fuel failure at 19:08, melting of spacer locations at 19:37, and upper-tie-plate at 21:26. The first and second degradations could have caused the release of radioactive elements or particles leading to the neutron detection at 21:00 in front of the main gate, and the third degradation could have resulted in the separation of the upper-tie-plate.

MAAP code analysis for the in-vessel phase of Fukushima-Daiichi Nuclear Power Station Unit 1 and comparison of the results among Units 1 to 3

The accident progression of the in-vessel phase of Fukushima Daiichi Nuclear Power Station Unit 1 was analyzed using the MAAP code. Although there is a large uncertainty in the initial stage of accident progression behavior in Unit 1 with little measurement data, it is presumed to have similarities to that of Unit 3. As a result, in Unit 1, since there was almost no alternative water injection during the in-vessel phase, cooling of the debris transferred to the lower plenum was small. It was likely that a large molten pool of metals had formed, and that the steam supply to the high-temperature core materials was suppressed and metal oxidation was relatively small. The analysis results for Unit 1 were compared with those for Units 2 and 3, and differences between units such as the thermal conditions of the debris that relocated to the pedestal and the degree of metal oxidation were shown.

Estimated 137Cs Radioactivity in the Gap Between the Top and Middle Covers at a Shield Plug in the Fukushima Daiichi Nuclear Power Station Unit 2

The contamination density of 137Cs deposited in the gap between the top and middle covers of the shield plug in the Fukushima Daiichi Nuclear Power Station Unit 2 was estimated using three types of measurement results. Owing to the extremely high dose rate on the entire operation floor, including the top of the shield plug, only remotely measured dose rates, such as ambient dose equivalent rates, were obtained using robots. Based on three types of measurements, significantly higher concentrations of 137Cs were observed than previously estimated. An estimation based on the measurements of the ambient dose equivalent rate inside the hole also demonstrated that the contamination density in the gap between the top and middle covers varied significantly at different positions of the cover. The results obtained will significantly aid in future decommissioning scenarios, and will be important for examining the progress of an accident.

Comprehensive Analysis and Evaluation of Fukushima Daiichi Nuclear Power Station Unit 3

The estimation and understanding of the state of fuel debris and fission products inside the plant is an essential step in the decommissioning of the Tokyo Electric Power Company Holdings Fukushima Daiichi Nuclear Power Station (1F). However, the direct observation of the plant interior, which is under a high radiation environment, is difficult and limited. Therefore, in order to understand the plant interior conditions, a comprehensive analysis and evaluation are necessary, based on various measurement data from the plant, analysis of plant data during the accident progression phase, and information obtained from computer simulations for this phase. These evaluations can be used to estimate the conditions of the interior of the reactor pressure vessel (RPV) and the primary containment vessel (PCV). Herein, 1F Unit 3 is addressed as the subject to produce an estimated diagram of the fuel debris distribution from data obtained about the RPV and PCV based on the comprehensive evaluation of various measurement data and information obtained from the accident progression analysis, which were released to the public in November 2022.

Radiation imaging of a highly contaminated filter train inside Fukushima Daiichi Nuclear Power Station unit 2 using an integrated Radiation Imaging System based on a Compton camera

ABSTRACT The Fukushima Daiichi Nuclear Power Station (FDNPS) suffered a meltdown in the aftermath of the large tsunami caused by the Great East Japan Earthquake that occurred on 11 March 2011. A massive amount of radioactive substance was spread over a wide area both inside and outside the FDNPS site. In this study, we present an approach for visualizing a radioactive hotspot on a standby gas-treatment system filter train, a highly contaminated piece of equipment in the air-conditioning room of the Unit 2 reactor building of FDNPS, using radiation imaging based on a Compton camera. In addition to fixed-point measurements using only the Compton camera, data acquisition while moving using an integrated Radiation Imaging System (iRIS), which combines a Compton camera with a simultaneous localization and mapping device and a survey meter, enabled the three-dimensional visualization of the hotspot location on the filter train. In addition, we visualized the hotspot and quantitatively evaluated its radioactivity. Notably, the visualized hotspot location and estimated radioactivity value are consistent with the accident investigation report of the FDNPS. Finally, the extent to which the radioactivity increased the ambient dose equivalent rate in the surrounding environment was explored.

Estimation of the core degradation and relocation at the Fukushima Daiichi Nuclear Power Station Unit 2 based on RELAP/SCDAPSIM analysis

Estimation of the final debris distribution at the Fukushima Daiichi Nuclear Power Station (1F) is inevitable for a safe and effective decommissioning. It is necessary to clarify possible failure modes of the reactor pressure vessel (RPV), which is influenced by the thermal status of slumped debris that highly depends on the in-vessel accident progression. In this study, the accident analysis of 1F Unit 2 (1F2) was conducted using the RELAP/SCDAPSIM code, in order to understand better the in-vessel accident progression. One of the unsolved issues of 1F2 is the mechanism of three pressure peaks measured through late Mar. 14 to early Mar. 15, 2011. Despite various analyses, its mechanism is not clearly understood. Comparing the results of previous boiling water reactor (BWR) core degradation experiments and that of 1F2 numerical analysis, it can be estimated that most relocated metallic materials had solidified at the core bottom at the onset of first pressure peak. It is likely that the pressure increase occurred due to the evaporation of injected water reaching the heated core plate structures. Between the first and second pressure peaks, the water is assumed to have been injected continuously and the water level was likely to have recovered to BAF at the initiation of the second pressure peak. Probable slumping of a certain amount of molten materials initiated the second pressure peak and the subsequent gradual pressure increase continued possibly due to massive reaction between coolant and remaining Zircaloy in the core. Assuming the closure of the safety relief valve (SRV) at 0:00 on Mar. 15, the third pressure peak was well reproduced in the analysis. Although the total amount of the slumped material was evaluated, large deviation exists among the cases and uncertainty is still large.

Evaluation of core material energy change during the in-vessel phase of Fukushima Daiichi Unit 3 based on observed pressure data utilizing GOTHIC code analysis

ABSTRACT The vapor formation within the reactor pressure vessel (RPV) is regarded to represent heat removal from core materials to the coolant, while the hydrogen generation within the RPV is regarded to represent heat generation by metal oxidation. Based on this understanding, the history of the vapor/hydrogen generation in the in-vessel phase of Fukushima Daiichi Nuclear Power Station Unit 3 was evaluated based on the comparison of the observed pressure data and the GOTHIC code analysis results. The resultant vapor/hydrogen generation histories were then converted to heat removal by coolant and heat generation by oxidation. The effects of the decay power and the heat transfer to the structures on the core material energy were also evaluated. The core materials are suggested to be significantly cooled by water within the RPV, especially when the core materials are relocated to the lower plenum. The results of this study combined with the existing severe accident (SA) code analyses suggested that the debris relocated to the lower plenum was effectively cooled down near the liquid water temperature within a few hours after the relocation.

Comprehensive Analysis and Evaluation of Fukushima Daiichi Nuclear Power Station Unit 2

Estimation and understanding of the state of the fuel debris and fission products inside the plant comprise an essential step in the decommissioning of Tokyo Electric Power Company Holdings’ Fukushima Daiichi nuclear power station (1F). However, because of the plant’s high-radiation environment, direct observation of the plant interior is difficult. Therefore, in order to understand the plant’s interior conditions, comprehensive analysis and evaluation based on various measurement data from the plant, analysis of plant data during the accident progression phase, and information obtained from computer simulations for this phase are necessary. These evaluations can be used to estimate the conditions of the interior of the reactor pressure vessel (RPV) and the primary containment vessel (PCV). This paper addresses 1F Unit 2 as the subject to produce an estimated map of the fuel debris distribution from data obtained about the RPV and PCV based on comprehensive evaluation of various measurement data and information obtained from the accident progression analysis, which were released to the public in June 2018.

Verifications for Large-Scale Parallel Simulations of 3D Fluid–Structure Interaction Using Moving Particle Simulation (MPS) and Finite Element Method (FEM)

Fluid–Structure Interaction (FSI) is an important phenomenon in many fields. In this study, the Moving Particle Simulation (MPS) along with Finite Element Method (FEM) are coupled together. MPS-FEM coupled method is used as the numerical simulators to estimate the effects for structures by the fluid pressure of the tsunami. This simulator consists of two solvers: fluid analysis and structure analysis. The fluid analysis is based on MPS and the structural part is analyzed by using FEM. In this paper, for verification of the MPS-FEM coupled method, the problem of hydrostatic pressure on a structure is calculated. The results are compared with the theoretical values and the results obtained by the original FEM analysis. We confirmed that the MPS-FEM coupling method can achieve adequate accuracy. Furthermore, the MPS-FEM coupled simulation for the turbine building of Fukushima Dai-ichi Nuclear Power Station Unit 1 has been done.

Investigation inside primary containment vessel (PCV) of Fukushima Daiichi Nuclear Power Station Unit 2 and 3

Investigation inside primary containment vessel (PCV) of Fukushima Daiichi Nuclear Power Station Unit 2 and 3

Investigation on Main Radiation Source at Operation Floor of Fukushima Daiichi Nuclear Power Station Unit 4

Pulse height distributions were measured using a LaBr3 detector set in a 1 cm lead collimator to investigate main radiation source at the operation floor of Fukushima Daiichi Nuclear Power Station Unit 4. It was confirmed that main radiation source above the reactor well was Co-60 from the activated steam dryer in the DS pool (Dryer-Separator pool) and that at the standby area was Cs-134 and Cs-137 from contaminated buildings and debris at the lower floor. Full energy peak count rate of Co-60 was reduced about 1/3 by 12mm lead sheet placed on the floor of the fuel handling machine.

Dose evaluation of working area on the Reactor Building Refueling Floor of Fukushima Daiichi Nuclear Power Station Unit 3

Dose evaluation of working area on the Reactor Building Refueling Floor of Fukushima Daiichi Nuclear Power Station Unit 3

Design and Construction plan of Spent Fuel Removal System for Fukushima Daiichi Nuclear Power Station Unit 3

Design and Construction plan of Spent Fuel Removal System for Fukushima Daiichi Nuclear Power Station Unit 3

Design and Construction plan of Spent Fuel Removal Structure for Fukushima Daiichi Nuclear Power Station Unit 3

Design and Construction plan of Spent Fuel Removal Structure for Fukushima Daiichi Nuclear Power Station Unit 3

Current situation of preparation for fuel removal from the Spent Fuel Pool of Fukushima Daiichi Nuclear Power Station Unit 3

Current situation of preparation for fuel removal from the Spent Fuel Pool of Fukushima Daiichi Nuclear Power Station Unit 3

Rubble removal work, Decontamination and Shield installation on the Reactor Building Refueling Floor of Fukushima Daiichi Nuclear Power Station Unit 3

3号機におけるプール燃料取り出し計画の最初の取り組みは，崩落したトラス鉄骨やコンクリート片などの汚染瓦礫の解体撤去と，著しく汚染した床表面の除染および床上への遮へい体設置などの準備工事であった。解体撤去に当たっては遠隔操作機器の開発を進める一方で，詳細な調査結果に基づいた瓦礫取出の計画と挙動シミュレーションを繰り返し，切断・つかみ位置の詳細な手順および監視方法を検討した。除染・遮へいの計画は崩落瓦礫下のオペフロ状況を推測し，適切な除染方法の選定及び遮へい体の許容荷重の評価を行った上で，いずれも遠隔操作による除染計画，遮へい体設置計画を立案し機器開発を行った。現在，燃料交換機を含むオペフロ上瓦礫撤去と除染作業を終え，遮へい体の設置工事を実施中である。本稿では計画の概要と高線量下で実施したこれまでの取り組みを紹介する。

Development of Remote Guiding System, Rubble Suction Device, etc, for Dismantling Building Cover of Fukushima Daiichi Nuclear Power Station Unit 1

Development of Remote Guiding System, Rubble Suction Device, etc, for Dismantling Building Cover of Fukushima Daiichi Nuclear Power Station Unit 1

Gamma-ray Spectra and Dose Measurement Results on the Reactor Building Refueling Floor of Fukushima Daiichi Nuclear Power Station Unit 3

Gamma-ray Spectra and Dose Measurement Results on the Reactor Building Refueling Floor of Fukushima Daiichi Nuclear Power Station Unit 3

福島第一原子力発電所2号機の原子炉格納容器漏えいを踏まえた格納容器の事故時耐性強化と格納容器ベントの運用について

福島第一原子力発電所2号機の原子炉格納容器漏えいを踏まえた格納容器の事故時耐性強化と格納容器ベントの運用について

Analysis of the accident in the Fukushima Daiichi nuclear power station Unit 3 with MELCOR_2.1

During the major accident occurred at the Fukushima Daiichi nuclear power station in March 2011, three units of the nuclear power plants suffered extensive damage to the reactors and buildings. It is widely believed that all three reactor cores experienced some melting.In the present paper, the Fukushima Unit 3 accident has been studied by using Melcor_2.1. An initial calculation was performed by using design conditions or operators reported actions. Several series of sensitivity cases have been performed in order to reproduce the main accident measurements (e.g. pressure histories of the reactor pressure vessel, dry-well and wet-well; downcomer water levels and the observed hydrogen explosion time). The sensitivity studies consisted in variations of Reactor Core Isolation Cooling and High Pressure Coolant Injection operation, Alternative Water Injection, venting, leaking to the dry-well and failure parameters of the lower head. The outcomes of the sensitivity calculations have pointed out the likely-state of the U3 core after 6days of the accident start.