An Assessment of Reactor Vessel Integrity Under In-Vessel Vapor Explosion Loads

Bang, Kwang-Hyun,Cho, Jong-Rae,Park, Soo-Yong Korean Nuclear Society 2000 Nuclear Engineering and Technology Vol.32 No.4

A safety assessment of reactor vessel lower head integrity under in-vessel vapor explosion loads has been performed. The core melt relocation parameters were chosen within the ranges of physically realizable bounds. The premixing and explosion calculations were performed using TRACER-II code. Using the calculated explosion pressures imposed on the lower head inner wall, strain calculations were peformed using ANSYS code. Then, the calculated strain results and the established failure criteria were used in determining the failure probability of the lower head, In the explosion analyses, it is shown that the explosion impulses are not altered significantly by the uncertain parameters of triggering location and time, fuel and vapor volume fractions in uniform premixture bounding calculations. Strain analyses show that the vapor explosion-induced lower head failure is not possible under the present framework of assessment. The result of static analysis using the conservative explosion-end pressure of 50 MPa also supports the conclusion. It is recommended, however, that an assessment of fracture mechanics for preexisting cracks be also considered to obtain a more concrete conclusion.

SEINA: A two-dimensional steam explosion integrated analysis code

Wu Liangpeng,Sun Ruiyu,Chen Ronghua,Tian Wenxi,Qiu Suizheng,Su G.H. 한국원자력학회 2022 Nuclear Engineering and Technology Vol.54 No.10

In the event of a severe accident, the reactor core may melt due to insufficient cooling. the hightemperature core melt will have a strong interaction (FCI) with the coolant, which may lead to steam explosion. Steam explosion would pose a serious threat to the safety of the reactors. Therefore, the study of steam explosion is of great significance to the assessment of severe accidents in nuclear reactors. This research focuses on the development of a two-dimensional steam explosion integrated analysis code called SEINA. Based on the semi-implicit Euler scheme, the three-phase field was considered in this code. Besides, the influence of evaporation drag of melt and the influence of solidified shell during the process of melt droplet fragmentation were also considered. The code was simulated and validated by FARO L-14 and KROTOS KS-2 experiments. The calculation results of SEINA code are in good agreement with the experimental results, and the results show that if the effects of evaporation drag and melt solidification shell are considered, the FCI process can be described more accurately. Therefore, it is proved that SEINA has the potential to be a powerful and effective tool for the analysis of steam explosions in nuclear reactors.

다중벽 탄소나노튜브의 분진폭발 특성

한인수 ( In Soo Han ),이근원 ( Keun Won Lee ),최이락 ( Yi Rac Choi ) 한국화학공학회 2017 Korean Chemical Engineering Research(HWAHAK KONGHA Vol.55 No.1

가연성 분진이 제조·취급되는 공정에서의 분진폭발 위험성은 항상 존재한다. 그러나 산업현장에서 취급되는 분진에 대한 분진폭발 특성 정보는 아주 미흡한 실정으로 사업장에서는 화학사고 예방대책 수립에 어려움을 겪고 있다. 본 연구에서는 입도분포가 다른 두 종류의 다중벽 탄소나노튜브(MWCNT)에 대한 분진폭발 특성을 실험적으로 조사하였으며, NFPA 499 Code를 적용하여 MWCNT 제조·취급 공정의 분진폭발 위험장소 구분을 검토하였다. 그 결과 평균입도가124.2 μm인 MWCNT 1의 P_max, K_st, LEL, MIE, 및 MIT는 각각 6.3 bar, 56 bar·m/s, 125 g/m³, 1000 mJ 초과 및 650 ℃ 초과로 나타났다. 평균입도가 293.5 μm인 MWCNT 2의 P_max, K_st, LEL, MIE, MIT는 각각 6.2 bar, 42 bar·m/s, 100 g/m³, 1000 mJ 초과 및 650 ℃ 초과로 나타났다. NFPA 499 Code에 따른 MWCNT 1, 2의 폭발강도와 점화감도는 각각 0.35와 0.01 미만으로 나타났기 때문에 MWCNT는 NFPA 499 Code에서 제시된 분진폭발 위험장소로 구분하여야 하는 가연성 분진으로 분류되지 않았다. Dust explosion hazards are always present when combustible dusts are manufactured or handled in the process. However, industries is experiencing difficulty in establishing chemical accident prevention measures because of insufficiency of information on dust explosion characteristics of combustible dust handled in industry. In this study, we investigated experimentally dust explosion characteristics of two kinds of multi-walled carbon nano tubes (MWCNT) different in particle size distribution and examined classification of dust explosion hazardous area for MWCNT manufacturing or handling process by applying the NFPA 499 code. As a result, P_max, K_st, LEL, MIE and MIT of MWCNT 1 having 124.2 μm median diameter are obtained 6.3 bar, 56 bar·m/s, 125 g/m³, over 1000 mJ, and over 650 ℃. P_max, K_st, LEL, MIE and MIT of MWCNT 2 having 293.5 μm median diameter are 6.2 bar, 42 bar·m/s, 100 g/m³, over 1000 mJ, and over 650 ℃, respectively. MWCNT 1, 2 are not categorized as combustible dust listed in the NFPA 499 Code for classification of dust explosion hazardous area because explosion severity and ignition sensitivity of MWCNT 1, 2 are below 0.35 and 0.01, respectively.

Ex-vessel Steam Explosion Analysis for Pressurized Water Reactor and Boiling Water Reactor

Matjaz Leskovar,Mitja Ursic 한국원자력학회 2016 Nuclear Engineering and Technology Vol.48 No.1

A steam explosion may occur during a severe accident, when the molten core comes intocontact with water. The pressurized water reactor and boiling water reactor ex-vesselsteam explosion study, which was carried out with the multicomponent threedimensionalEulerian fuelecoolant interaction code under the conditions of the Organisationfor Economic Co-operation and Development (OECD) Steam Explosion Resolution forNuclear Applications project reactor exercise, is presented and discussed. In reactor calculations,the largest uncertainties in the prediction of the steam explosion strength areexpected to be caused by the large uncertainties related to the jet breakup. To obtain someinsight into these uncertainties, premixing simulations were performed with both availablejet breakup models, i.e., the global and the local models. The simulations revealed thatweaker explosions are predicted by the local model, compared to the global model, due tothe predicted smaller melt droplet size, resulting in increased melt solidification andincreased void buildup, both reducing the explosion strength. Despite the lower active meltmass predicted for the pressurized water reactor case, pressure loads at the cavity wallsare typically higher than that for the boiling water reactor case. This is because of thesignificantly larger boiling water reactor cavity, where the explosion pressure wave originatingfrom the premixture in the center of the cavity has already been significantlyweakened on reaching the distant cavity wall.

Ex-vessel Steam Explosion Analysis for Pressurized Water Reactor and Boiling Water Reactor

Leskovar, Matjaz,Ursic, Mitja Korean Nuclear Society 2016 Nuclear Engineering and Technology Vol.48 No.1

A steam explosion may occur during a severe accident, when the molten core comes into contact with water. The pressurized water reactor and boiling water reactor ex-vessel steam explosion study, which was carried out with the multicomponent three-dimensional Eulerian fuel-coolant interaction code under the conditions of the Organisation for Economic Co-operation and Development (OECD) Steam Explosion Resolution for Nuclear Applications project reactor exercise, is presented and discussed. In reactor calculations, the largest uncertainties in the prediction of the steam explosion strength are expected to be caused by the large uncertainties related to the jet breakup. To obtain some insight into these uncertainties, premixing simulations were performed with both available jet breakup models, i.e., the global and the local models. The simulations revealed that weaker explosions are predicted by the local model, compared to the global model, due to the predicted smaller melt droplet size, resulting in increased melt solidification and increased void buildup, both reducing the explosion strength. Despite the lower active melt mass predicted for the pressurized water reactor case, pressure loads at the cavity walls are typically higher than that for the boiling water reactor case. This is because of the significantly larger boiling water reactor cavity, where the explosion pressure wave originating from the premixture in the center of the cavity has already been significantly weakened on reaching the distant cavity wall.

INVESTIGATIONS ON THE RESOLUTION OF SEVERE ACCIDENT ISSUES FOR KOREAN NUCLEAR POWER PLANTS

HEE DONG KIM,DONG HA KIM,김종태,SANG BAIK KIM,송진호,홍성완 한국원자력학회 2009 Nuclear Engineering and Technology Vol.41 No.5

Under the government supported long-term nuclear R&D program, the severe accident research program at KAERI is directed to investigate unresolved severe accident issues such as core debris coolability, steam explosions, and hydrogen combustion both experimentally and numerically. Extensive studies have been performed to evaluate the in-vessel retention of core debris through external reactor vessel cooling concept for APR1400 as a severe accident management strategy. Additionally, an improvement of the insulator design outside the vessel was investigated. To address steam explosions, a series of experiments using a prototypic material was performed in the TROI facility. Major parameters such as material composition and void fraction as well as the relevant physics affecting the energetics of steam explosions were investigated. For hydrogen control in Korean nuclear power plants, evaluation of the hydrogen concentration and the possibility of deflagration-to-detonation transition occurrence in the containment using three-dimensional analysis code, GASFLOW, were performed. Finally, the integrated severe accident analysis code, MIDAS, has been developed for domestication based on MELCOR. The data transfer scheme using pointers was restructured with the modules and the derived-type direct variables using FORTRAN90. New models were implemented to extend the capability of MIDAS.

INVESTIGATIONS ON THE RESOLUTION OF SEVERE ACCIDENT ISSUES FOR KOREAN NUCLEAR POWER PLANTS

Kim, Hee-Dong,Kim, Dong-Ha,Kim, Jong-Tae,Kim, Sang-Baik,Song, Jin-Ho,Hong, Seong-Wan Korean Nuclear Society 2009 Nuclear Engineering and Technology Vol.41 No.5

Under the government supported long-term nuclear R&D program, the severe accident research program at KAERI is directed to investigate unresolved severe accident issues such as core debris coolability, steam explosions, and hydrogen combustion both experimentally and numerically. Extensive studies have been performed to evaluate the in-vessel retention of core debris through external reactor vessel cooling concept for APR1400 as a severe accident management strategy. Additionally, an improvement of the insulator design outside the vessel was investigated. To address steam explosions, a series of experiments using a prototypic material was performed in the TROI facility. Major parameters such as material composition and void fraction as well as the relevant physics affecting the energetics of steam explosions were investigated. For hydrogen control in Korean nuclear power plants, evaluation of the hydrogen concentration and the possibility of deflagration-to-detonation transition occurrence in the containment using three-dimensional analysis code, GASFLOW, were performed. Finally, the integrated severe accident analysis code, MIDAS, has been developed for domestication based on MELCOR. The data transfer scheme using pointers was restructured with the modules and the derived-type direct variables using FORTRAN90. New models were implemented to extend the capability of MIDAS.

금속이 함유된 코륨을 이용한 TROI 증가폭발 실험

김종환(Jong-Hwan Kim),민병태(Beong-Tae Min),홍성완(Seong-Wan Hong),박익규(Seong-Ho Hong),송진호(Jin-Ho Song),김희동(Hee-Dong Kim) 대한기계학회 2007 대한기계학회 춘추학술대회 Vol.2007 No.5

Two steam explosion experiments were performed in the TROI facility by using metal-added molten corium (core material) which is produced during a postulated severe accident in the nuclear reactor. A triggered steam explosion occurred in a case, but no triggered steam explosion did in the other case. The dynamic pressure and the dynamic load measured in the former experiment show a stronger explosion that those performed previously with oxidic corium. A steam explosion is prohibited when the melt temperature is low, because the melt is easily solidified to prevent a liquid-liquid interaction.

Uncertainty and Importance Analysis Based on SAUNA System for Severe Accident

Sunhee Park(박선희),Kwang-Il Ahn(안광일),Hee-Dong Kim(김희동),Jae-Min Sohn(손재민) 한국정보기술학회 2011 한국정보기술학회논문지 Vol.9 No.12

The steam explosion in nuclear power plant, which is one of the severe accident, is found to be greatly affected by initial conditions such as the melt temperature, the water temperature in the cavity. the discharged melt radius, and the discharged melt velocity. In this study, the SAUNA system is suggested, which applies the uncertainty analysis technique, one of the adopted AI methods in severe accident research. For these purpose, the TEXAS code which simulates a steam explosion is used, and a numerical analysis methods such as random sampling technique, statistical distribution and importance analysis are also applied. To analyze the results, the execution results of 200 code results were analyzed. From the results of the analysis using the regression analysis model, it is confirmed to understand the major variables that have a big impact. It shows that the use of the SAUNA system will be a good tool in an uncertainty analysis of severe accidents in nuclear power plants.

PARAMETER DEPENDENCE OF STEAM EXPLOSION LOADS AND PROPOSAL OF A SIMPLE EVALUATION METHOD

Kiyofumi Moriyama,박현선 한국원자력학회 2015 Nuclear Engineering and Technology Vol.47 No.7

The energetic steam explosion caused by contact between the high temperature moltencore and water is one of the phenomena that may threaten the integrity of the containmentvessel during severe accidents of light water reactors (LWRs). We examined thedependence of steam explosion loads in a typical reactor cavity geometry on selectedmodel parameters and initial/boundary conditions by using a steam explosion simulationcode, JASMINE, developed at Japan Atomic Energy Agency (JAEA). Among the parameters,we put an emphasis on the water pool depth that has significance in terms of accidentmitigation strategies including cavity flooding. The results showed a strong correlationbetween the load and the premixed mass, defined as the mass of the molten material inlow void zones (void fraction < 0.75). The jet diameter and velocity that comprise the flowrate were the primary factors to determine the premixed mass and the load. The waterpool depth also showed a significant impact. The energy conversion ratio based on theenthalpy in the premixed mass was in a narrow range ~4%. Based on this observation, weproposed a simplified method for evaluation of the steam explosion load. The resultsshowed fair agreement with JASMINE.