• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.707-712
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• 1997

This paper presents the current state of knowledge on molten material relocation into the lower plenum. Consequences of movement of material to the lower head are considered with regardt to the potential for reactor pressure vessel failure from both thermal hydraulic and mechanical standpoints. The models are applied to evaluating various in-vessel retention strategies for the Korean Standard power plant (KSNPP) reactor The results are summarized in terms of thermal response of the reactor vessel from the very relevant severe accident management perspective.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.713-718
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• 1997

In the TMI-2 accident, approximately twenty(20) tons of molten core material drained into the lower plenum. Early advanced light water reactor (LWR) designs assumed a lower head failure and incorporated various measures for ex-vessel accident mitigation. However, one of the major findings from the TMI-2 Vessel Investigation Project was that one part of the reactor lower head wall estimated to have attained a temperature of 1100$^{\circ}C$ for about 30 minutes has seemingly experienced a comparatively rapid cooldown with no major threat to the vessel integrity. In this regard, recent empirical and analytical studies have shifted interests to such in-vessel retention designs or strategies as reactor cavity flooding, in-vessel flooding and engineered gap cooling of the vessel Accurate thermohydrodynamic and creep deformation modeling and rupture prediction are the key to the success in developing practically useful in-vessel accident/risk management strategies. As an advanced in-vessel design concept, this work presents the COrium Attack Syndrome Immunization Structures (COASIS) that are being developed as prospective in-vessel retention devices for a next-generation LWR in concert with existing ex-vessel management measures. Both the engineered gap structures in-vessel (COASISI) and ex-vessel (COASISO) are demonstrated to maintain effective heat transfer geometry during molten core debris attack when applied to the Korean Standard Nuclear Power Plant(KSNPP) reactor. The likelihood of lower head creep rupture during a severe accident is found to be significantly suppressed by the COASIS options.

• Korean Journal of Otorhinolaryngology-Head and Neck Surgery
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• 제61권12호
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• pp.702-704
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• 2018

The most common cause of treatment failure in oral cavity cancer is when it is found to have local recurrence, usually occurring in the ipsilateral cervical lymph node. On the contrary, it is extremely rare to find local recurrence in soft tissue metastasis (STM) in the contralateral neck. Furthermore, lung cancer and malignant lymphoma are most commonly confined to their primary sites. The poor general condition increases the likelihood of STM, which indicates bad prognosis. A 72-year-old man with a hard and fixed mass on the right submandibular space visited our clinic. He had received a wide excision with local flapreconstruction for squamous cell carcinoma in the left corner of lower lip 18 months ago. We performed the wide excision with bilateral selective neck dissection (I-III), and he was finally diagnosed as STM from contralateral lip cancer. We report this unique and rare disease entity with a literature review.

• Korean Journal of Radiology
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• 제20권4호
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• pp.649-661
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• 2019

Objective: To systematically review the evaluation of the diagnostic accuracy of pre-treatment apparent diffusion coefficient (ADC) and change in ADC during the intra- or post-treatment period, for the prediction of locoregional failure in patients with head and neck squamous cell carcinoma (HNSCC). Materials and Methods: Ovid-MEDLINE and Embase databases were searched up to September 8, 2018, for studies on the use of diffusion-weighted magnetic resonance imaging for the prediction of locoregional treatment response in patients with HNSCC treated with chemoradiation or radiation therapy. Risk of bias was assessed by using the Quality Assessment Tool for Diagnostic Accuracy Studies-2. Results: Twelve studies were included in the systematic review, and diagnostic accuracy assessment was performed using seven studies. High pre-treatment ADC showed inconsistent results with the tendency for locoregional failure, whereas all studies evaluating changes in ADC showed consistent results of a lower rise in ADC in patients with locoregional failure compared to those with locoregional control. The sensitivities and specificities of pre-treatment ADC and change in ADC for predicting locoregional failure were relatively high (range: 50-100% and 79-96%, 75-100% and 69-95%, respectively). Meta-analytic pooling was not performed due to the apparent heterogeneity in these values. Conclusion: High pre-treatment ADC and low rise in early intra-treatment or post-treatment ADC with chemoradiation, could be indicators of locoregional failure in patients with HNSCC. However, as the studies are few, heterogeneous, and at high risk for bias, the sensitivity and specificity of these parameters for predicting the treatment response are yet to be determined.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집A
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• pp.782-787
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• 2001

This paper is concerned with the global rupture of a nuclear reactor pressure vessel(RPV) in a severe accident. During the severe reactor accident of molten core, the temperature and the pressure in the nuclear reactor rise to a certain level depending on the initial and subsequent condition of a severe accident. While the rise of the temperature cause the thermal softening of RPV material, the rise of the internal pressure could cause failure of the RPV lower head. The global rupture of an RPV is simulated by finite element limit analysis for the collapse pressure and mode and this analysis results have been compared with a variation of the internal pressure of RPV. The finite element limit method is a systematic tool to secure the safety criteria of a nuclear reactor and to evaluate the in-vessel corium retention.

• Nuclear Engineering and Technology
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• 제44권3호
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• pp.237-248
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• 2012

Corium behavior in the lower plenum of the reactor vessel during a severe accident is very important, as this affects a failure mechanism of the lower head vessel and a thermal load to the outer reactor vessel under the IVR-ERVC (In-Vessel corium Retention through External Reactor Vessel Cooling) condition. This paper discusses the state of the art and technical issues on corium behavior in the lower plenum, such as initial corium pool formation characteristics and its transient behavior, natural convection heat transfer in various geometries, natural convection heat transfer with a phase change of melting and solidification, and corium interaction with a lower head vessel including penetrations of the ICI (In-Core Instrumentation) nozzle are discussed. It is recommended that more detailed analysis and experiments are necessary to solve the uncertainties of corium behavior in the lower plenum of the reactor vessel.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.473-478
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• 1997

A severe accident management strategy, in-vessel retention corium through external reactor vessel cooling(ERVC) is being studied worldwide as a means to prevent reactor vessel failure following a core melt accident. An evaluation of feasibility of this ERVC for a large Advanced Light Water Reactor (ALWR) is presented. To account for the coolability of corium and metal in the reactor vessel, a thermal analysis is performed using an existing method. Results show that the peak heat flux along the inner surface of the reactor vessel lower head has a relatively smaller margin than a small capacity reactor such as AP600 in regards with the critical heat flux attainable at the outer surface of the reactor vessel lower head.

• Nuclear Engineering and Technology
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• 제55권11호
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• pp.4146-4158
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• 2023

This paper proposes a combined plastic and creep constitutive model of A533B1 pressure vessel steel to simulate progressive deformation of nuclear pressure vessels under severe accident conditions. To develop the model, recent tensile test data covering a wide range of temperatures (from RT to 1,100 ℃) and strain rates (from 0.001%/s to 1.0%/s) was used. Comparison with experimental data confirms that the proposed combined plastic and creep model can well reflect effects of temperature and strain rate on tensile behaviour up to failure. In the companion paper (Part II), the proposed model will be used to simulate OECD lower head failure (OLHF) test data.

• 대한방사선기술학회지:방사선기술과학
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• 제45권5호
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• pp.439-448
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• 2022

The purpose of this study was to analyze the results from statistical process control (SPC) to recommend upper and lower control limits for planning parameters based on delivery quality assurance (DQA) results and establish our institutional guidelines regarding planning parameters for helical tomotherapy (HT). A total of 53 brain, 41 head and neck (H & N), and 51 pelvis cases who had passing or failing DQA measurements were selected. The absolute point dose difference (DD) and the global gamma passing rate (GPR) for all patients were analyzed. Control charts were used to evaluate upper and lower control limits (UCL and LCL) for all assessed treatment planning parameters. Treatment planning parameters were analyzed to provide its range for DQA pass cases. We confirmed that the probability of DQA failure was higher when the proportion of leaf open time (LOT) below 100 ms was greater than 30%. LOT and gantry period (GP) were significant predictor for DQA failure using the SPC method. We investigated the availability of the SPC statistic method to establish the local planning guideline based on DQA results for HT system. The guideline of each planning parameter in HT may assist in the prediction of DQA failure using the SPC statistic method in the future.

• Nuclear Engineering and Technology
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• 제32권4호
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• pp.379-394
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• 2000

As an in-vessel retention (IVR) design concept in coping with a severe accident in the nuclear power plant during which time a considerable amount of core material may melt, external cooling of the reactor vessel has been suggested to protect the lower head from overheating due to relocated material from the core. The efficiency of the ex-vessel management may be estimated by the thermal margin defined as the ratio of the critical heat flux (CHF）to the actual heat flux from the reactor vessel. Principal factors affecting the thermal margin calculation are the amount of heat to be transferred downward from the molten pool, variation of heat flux with the angular position, and the amount of removable heat by external cooling In this paper a thorough literature survey is made and relevant models and correlations are critically reviewed and applied in terms of their capabilities and uncertainties in estimating the thermal margin to potential failure of the vessel on account of the CHF Results of the thermal margin calculation are statistically treated and the associated uncertainties are quantitatively evaluated to shed light on the issues requiring further attention and study in the near term. Our results indicated a higher thermal margin at the bottom than at the top of the vessel accounting for the natural convection within the hemispherical molten debris pool in the lower plenum. The information obtained from this study will serve as the backbone in identifying the maximum heat removal capability and limitations of the IVR technology called the Cerium Attack Syndrome Immunization Structures (COASISO) being developed for next generation reactors.