Influence of In-Vessel Melt Progression on Uncertainty of Source Term During a Severe Accident

2014 ◽  
Author(s):  
Hiroto Itoh ◽  
Xiaoyu Zheng ◽  
Hitoshi Tamaki ◽  
Yu Maruyama
Keyword(s):  
Uncertainty Analysis ◽  
Nuclear Power ◽  
Source Term ◽  
Sampling Technique ◽  
Latin Hypercube Sampling ◽  
Severe Accident ◽  
Source Terms ◽  
Radioactive Materials ◽  
Uncertainty Analyses ◽  
Reactor Pressure

The influence of the in-vessel melt progression on the uncertainty of source terms was examined in the uncertainty analysis with integral severe accident analysis code MELCOR (Ver. 1.8.5), taking the accident at Unit 2 of the Fukushima Daiichi nuclear power plant as an example. The 32 parameters selected from the rough screening analysis were sampled by Latin hypercube sampling technique in accordance with the uncertainty distributions specified for each parameter. The uncertainty distributions of the outputs, including the source terms of the representative radioactive materials (Cs, CsI, Te and Ba), the total mass of in-vessel H2 generation and the total debris mass released from the reactor pressure vessel to the drywell, were obtained through the uncertainty analysis with an assumption of the failure of drywell. Based on various types of correlation coefficient for each parameter, 9 significant uncertain parameters potentially dominating the source terms were identified. These 9 parameters were transferred to the subsequent sensitivity and uncertainty analyses, in which the influence of the transportation of radioactive materials was taken into account.

Estimation of Source Term Uncertainty in a Severe Accident With Correlated Variables

2014 ◽  
Author(s):  
Xiaoyu Zheng ◽  
Hiroto Itoh ◽  
Hitoshi Tamaki ◽  
Yu Maruyama
Keyword(s):  
Uncertainty Analysis ◽  
Nuclear Power ◽  
Source Term ◽  
Correlation Coefficients ◽  
Sampling Technique ◽  
Latin Hypercube Sampling ◽  
Distribution Functions ◽  
Cumulative Distribution ◽  
Severe Accident ◽  
Source Terms

The quantitative evaluation of the fission product release to the environment during a severe accident is of great importance. In the present analysis, integral severe accident code MELCOR 1.8.5 has been applied to estimating uncertainty of source term for the accident at Unit 2 of the Fukushima Daiichi nuclear power plant (NPP) as an example and to discussing important models or parameters influential to the source term. Forty-two parameters associated with models for the transportation of radioactive materials were chosen and narrowed down to 18 through a set of screening analysis. These 18 parameters in addition to 9 parameters relevant to in-vessel melt progression obtained by the preceding uncertainty study were input to the subsequent sensitivity analysis by Morris method. This one-factor-at-a-time approach can preliminarily identify inputs which have important effects on an output, and 17 important parameters were selected from the total of 27 parameters through this approach. The selected parameters have been integrated into uncertainty analysis by means of Latin Hypercube Sampling technique and Iman-Conover method, taking into account correlation between parameters. Cumulative distribution functions of representative source terms were obtained through the present uncertainty analysis assuming the failure of suppression chamber. Correlation coefficients between the outputs and uncertain input parameters have been calculated to identify parameters of great influences on source terms, which include parameters related to models on core components failure, models of aerosol dynamic process and pool scrubbing.

Uncertainty studies on hydrogen source term with MAAP5 code

Kerntechnik ◽  
2021 ◽  
Vol 86 (2) ◽  
pp. 152-163
Author(s):  
T.-C. Wang ◽  
M. Lee
Keyword(s):  
Hydrogen Production ◽  
Nuclear Power ◽  
Hydrogen Generation ◽  
Pearson Correlation ◽  
Latin Hypercube Sampling ◽  
Severe Accident ◽  
Model Parameters ◽  
Power Company ◽  
Sensitivity Studies ◽  
The Impact

Abstract In the present study, a methodology is developed to quantify the uncertainties of special model parameters of the integral severe accident analysis code MAAP5. Here, the in-vessel hydrogen production during a core melt accident for Lungmen Nuclear Power Station of Taiwan Power Company, an advanced boiling water reactor, is analyzed. Sensitivity studies are performed to identify those parameters with an impact on the output parameter. For this, multiple calculations of MAAP5 are performed with input combinations generated from Latin Hypercube Sampling (LHS). The results are analyzed to determine the 95th percentile with 95% confidence level value of the amount of in-vessel hydrogen production. The calculations show that the default model options for IOXIDE and FGBYPA are recommended. The Pearson Correlation Coefficient (PCC) was used to determine the impact of model parameters on the target output parameters and showed that the three parameters TCLMAX, FCO, FOXBJ are highly influencing the in-vessel hydrogen generation. Suggestions of values of these three parameters are given.

Severe accident simulation for VVER-1000 reactor using ASTEC-V2.1.1.3

Kerntechnik ◽  
2021 ◽  
Vol 86 (6) ◽  
pp. 454-469
Author(s):  
S. H. Abdel-Latif
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Fission Products ◽  
Reactor Pressure Vessel ◽  
Severe Accident ◽  
Potential Core ◽  
Safety Requirements ◽  
The Core ◽  
Active Safety Systems ◽  
Reactor Pressure

Abstract The station black-out (SBO) is one of the main accident sequences to be considered in the field of severe accident research. To evaluate a nuclear power plant’s behavior in the context of this accident, the integral ASTEC-V2.1.1.3 code “Accident Source Term Evaluation Code” covers sequences of SBO accidents that may lead to a severe accident. The aim of this work is to discuss the modelling principles for the core melting and in-vessel melt relocation phenomena of the VVER-1000 reactor. The scenario of SBO is simulated by ASTEC code using its basic modules. Then, the simulation is performed again by the same code after adding and activating the modules; ISODOP, DOSE, CORIUM, and RCSMESH to simulate the ex-vessel melt. The results of the two simulations are compared. As a result of SBO, the active safety systems are not available and have not been able to perform their safety functions that maintain the safety requirements to ensure a secure operation of the nuclear power plant. As a result, the safety requirements will be violated causing the core to heat-up. Moreover potential core degradation will occur. The present study focuses on the reactor pressure vessel failure and relocation of corium into the containment. It also discusses the transfer of Fission Products (FPs) from the reactor to the containment, the time for core heat-up, hydrogen production and the amount of corium at the lower plenum reactor pressure vessel is determined.

IRSN R&D Actions on FP Behaviour for RCS, Containment and FCVS in Severe Accident Conditions

2016 ◽  
Author(s):  
Laurent Cantrel ◽  
Thierry Albiol ◽  
Loïc Bosland ◽  
Juliette Colombani ◽  
Frédéric Cousin ◽  
...  
Keyword(s):  
Porous Materials ◽  
Nuclear Power ◽  
Source Term ◽  
Cooling System ◽  
Fission Products ◽  
Simulation Software ◽  
Severe Accident ◽  
Status Report ◽  
Diagnosis And Prognosis ◽  
Accident Conditions

This paper deals with near past, ongoing and planned R&D works on fission products (FPs) behaviour in Reactor Cooling System (RCS), containment building and in Filtered Containment Venting Systems (FCVS) for severe accident (SA) conditions. For the last topic, in link with the Fukushima post-accident management and possible improvement of mitigation actions for such SA, the FCVS topic is again on the agenda (see Status Report on Filtered Containment Venting, OECD/NEA/CSNI, Report NEA/CSNI/R(2014)7, 2014.) with a large interest at the international scale. All the researches are collaborative works; the overall objective is to develop confident models to be implemented in ASTEC SA simulation software. After being initiated in the International Source Term Program (ISTP), researches devoted to the understanding of iodine transport through the RCS are still ongoing in the frame of a bilateral agreement between IRSN and EDF with promising results. In 2017, a synthesis report of the last 10 years of researches, which have combined experimental and fundamental works based on the use of theoretical chemistry tools, will be issued. For containment, the last advances are linked to the Source Term Evaluation and Mitigation (STEM) OECD/NEA project operated by IRSN. The objective of the STEM project was to improve the evaluation of Source Term (ST) for a SA on a nuclear power plant and to reduce uncertainties on specific phenomena dealing with the chemistry of two major fission products: iodine and ruthenium. More precisely, the STEM project provided additional knowledge and improvements for calculation tools in order to allow a more robust diagnosis and prognosis of radioactive releases in a SA. STEM data will be completed by a follow-up, called STEM2, to further the knowledge concerning some remaining issues and be closer to reactor conditions. Two additional programmes deal with FCVS issues: the MItigation of outside Releases in the Environment (MIRE) (2013–2019) French National Research Agency (NRA) programme and the Passive and Active Systems on Severe Accident source term Mitigation (PASSAM) (2013–2016) European project. For FCVS works, the efficiencies for trapping iodine with various FCVS, covering scrubbers and dry filters, are examined to get a clear view of their abilities in SA conditions. Another part, performed in collaboration with French universities (Lorraine and Lille 1), is focused on the enhancement of the performance of these filters with specific porous materials able to trap volatile iodine. For that, influence of zeolites materials parameters (nature of the counter-ions, structure, Si/Al ratio …) will be tested. New kind of porous materials constituted by Metal organic Frameworks (MOF) will also be looked at because they can constitute a promising way of trapping.

scholarly journals FRACTURE MECHANICS UNCERTAINTY ANALYSIS IN THE RELIABILITY ASSESSMENT OF THE REACTOR PRESSURE VESSEL: (2D) SUBJECTED TO INTERNAL PRESSURE

2016 ◽  
Vol 18 (2) ◽  
pp. 55
Author(s):  
Entin Hartini ◽  
Roziq Himawan ◽  
Mike Susmikanti
Keyword(s):  
Fracture Mechanics ◽  
Uncertainty Analysis ◽  
Internal Pressure ◽  
Pressure Vessel ◽  
Probabilistic Method ◽  
Latin Hypercube Sampling ◽  
Reactor Pressure Vessel ◽  
J Integral ◽  
Deterministic Method ◽  
Reactor Pressure

ABSTRACT FRACTURE MECHANICS UNCERTAINTY ANALYSIS IN THE RELIABILITY ASSESSMENT OF THE REACTOR PRESSURE VESSEL: (2D) SUBJECTED TO INTERNAL PRESSURE. The reactor pressure vessel (RPV) is a pressure boundary in the PWR type reactor which serves to confine radioactive material during chain reaction process. The integrity of the RPV must be guaranteed either  in a normal operation or accident conditions. In analyzing the integrity of RPV, especially related to the crack behavior which can introduce break to the reactor pressure vessel, a fracture mechanic approach should be taken for this assessment. The uncertainty of input used in the assessment, such as mechanical properties and physical environment, becomes a reason that the assessment is not sufficient if it is perfomed only by deterministic approach. Therefore, the uncertainty approach should be applied. The aim of this study is to analize the uncertainty of fracture mechanics calculations in evaluating the reliability of PWR`s reactor pressure vessel. Random character of input quantity was generated using probabilistic principles and theories. Fracture mechanics analysis is solved by Finite Element Method (FEM) with  MSC MARC software, while uncertainty input analysis is done based on probability density function with Latin Hypercube Sampling (LHS) using python script. The output of MSC MARC is a J-integral value, which is converted into stress intensity factor for evaluating the reliability of RPV’s 2D. From the result of the calculation, it can be concluded that the SIF from  probabilistic method, reached the limit value of  fracture toughness earlier than SIF from  deterministic method.  The SIF generated by the probabilistic method is 105.240 MPa m0.5. Meanwhile, the SIF generated by deterministic method is 100.876 MPa m0.5. Keywords: Uncertainty analysis, fracture mechanics, LHS, FEM, reactor pressure vessels   ABSTRAK ANALISIS KETIDAKPASTIAN FRACTURE MECHANIC PADA EVALUASI KEANDALAN BEJANA TEKAN REAKTOR: 2D DENGAN BEBAN INTERNAL PRESSURE. Bejana tekan reaktor (RPV) merupakan pressure boundary dalam reaktor tipe PWR yang berfungsi untuk mengungkung material radioaktif  yang dihasilkan pada proses reaksi berantai. Maka dari itu integritas bejana tekan reaktor harus senantiasa terjamin baik reaktor dalam keadaan operasi normal, maupun kecelakaan. Dalam melakukan analisis integritas RPV, khususnya yang berkaitan dengan pecahnya bejana tekan reaktor akibat adanya retak dilakukan analisis secara fracture mechanics. Adanya ketidakpastian input seperti sifat mekanik bahan, lingkungan fisik, dan input pada data, maka dalam melakukan analisis keandalan tidak hanya dilakukan secara deterministik saja. Tujuan dari penelitian ini adalah melakukan analisis ketidakpastian input pada perhitungan fracture mechanik pada evaluasi keandalan bejana tekan reaktor PWR. Pendekatan untuk karakter random dari kuantitas input menggunakan  teori probabilistik. Analisis fracture mechanics dilakukan berdasarkan metode elemen hingga (FEM) menggunakan perangkat lunak MSC MARC. Analisis ketidakpastian input dilakukan berdasarkan probability density function dengan Latin Hypercube Sampling (LHS) menggunakan python script. Output dari MSC MARC adalah nilai J-integral untuk mendapatkan nilai stress intensity factor pada evaluasi keandalan bejana tekan reactor 2D. Dari hasil perhitungan dapat disimpulkan bahwa SIF probabilistik lebih dulu mencapai nilai batas fracture tougness  dibanding  SIF deterministik. SIF yang dihasilkan dengan metode probabilistik adalah 105,240 MPa m0,5. Sedangkan SIF metode deterministik adalah 100,876 MPa m0,5. Kata kunci: Analisis ketidakpastian, fracture mechanics, LHS, FEM, bejana tekan reaktor

Scientists Debate Nuclear Source Terms

1985 ◽  
Vol 28 (6) ◽  
pp. 17-23
Author(s):  
John Graham
Keyword(s):  
Nuclear Power ◽  
Source Term ◽  
Large Body ◽  
Radioactive Material ◽  
Source Terms ◽  
American Nuclear Society ◽  
Life Threatening ◽  
Accident Scenarios ◽  
American Physical ◽  
Term Data

The nuclear source term, defined as the quantity, timing, and characteristic of the release of radioactive material to the environment following a core-melt accident, was thoroughly debated in 1985. This debate, summarized here, turns on the Nuclear Regulatory Commission's (NRC) source term for radioactive iodine, which is postulated as potentially the most life-threatening radionuclide that might escape in a nuclear power-plant accident. A generic radioiodine source term has been used by NRC as the surrogate for all others; thus, it has become one of the bases on which nuclear-safety regulations are founded. Following the Three Mile Island (TMI) accident, from which only traces of radioiodine escaped, scientists began arguing that nuclear regulations based on source-term calculations are erroneous and should be modified. The American Nuclear Society (ANS) and industry researchers have concluded that warranted reductions in the NRC source terms could range from a factor of ten to several factors of ten in most accident scenarios. The American Physical Society (APS), after agreeing with a large body of the conclusions from the other research groups, has told NRC that its source-term data base is still inadequate because of the existence of a number of uncertainties it found therein. Although APS presented no such conclusion, its findings made clear to NRC that an early reduction of all source terms is not warranted. The anti-nuclear lobby agrees with APS. The NRC has taken a cautious, conservative approach to the revision of its regulations based on new source-term data, although it too concedes that its old methodologies and conclusions must be revised and ultimately superceded.

Fracture Probability Assessment for Embrittled Reactor Pressure Vessels Under Ultimate Response Guideline Operation

2016 ◽  
Author(s):  
Hsoung-Wei Chou ◽  
Chin-Cheng Huang
Keyword(s):  
Nuclear Power ◽  
Structural Integrity ◽  
Water Injection ◽  
Fracture Probability ◽  
Pressure Vessels ◽  
Radioactive Material ◽  
Severe Accident ◽  
Core Damage ◽  
Compound Disaster ◽  
Reactor Pressure

After the Fukushima nuclear accident occurred in Japan on March 11, 2011, the compound disaster beyond design basis which may severely impact the nuclear safety has been noticed and paid much more attention. In addition to the original emergency operating procedures (EOPs) and severe accident management guidance (SAMG) of nuclear power plant, the licensee in Taiwan developed the ultimate response guideline (URG) when EOPs and SAMG cannot be performed effectively due to loss of power and water supply by the Fukushima-like compound disaster. Once the URG procedures are initiated, the operators will conduct reactor depressurization, low pressure water injection and containment venting to strictly prevent the core damage and the release of radioactive material. In the paper, the fracture probabilities of boiling water reactor (BWR) pressure vessels with incremental levels of radiation embrittlement under URG operation are evaluated by probabilistic fracture mechanics (PFM) analysis. First, the models of PFM FAVOR code concerning the beltline shell welds of reactor pressure vessels (RPVs) associated with a very conservative flaw distribution are built. Then, the hypothetical transients of URG operation obtained from the thermal hydraulic analyses for Taiwan domestic BWRs are applied as the loading condition. The analysis results demonstrate that performing URG operation will not cause significant fracture probability for RPV, even at an extremely embrittled condition. The URG procedures can ensure the prevention of core damage as well as maintenance of structural integrity of RPV in the situation of long-term loss of electric power when suffering from the Fukushima-like accidents.

Uncertainty Analysis of Power Distribution for NESTOR Based on the Double Latin Hypercube Sampling Method

2018 ◽  
Author(s):  
Hongkuan Liao ◽  
Qing Li ◽  
Yingrui Yu ◽  
Yuying Hu ◽  
Lei Wu ◽  
...  
Keyword(s):  
Uncertainty Analysis ◽  
Power Distribution ◽  
Nuclear Power ◽  
Power Plants ◽  
Sampling Method ◽  
Latin Hypercube Sampling ◽  
Physical Models ◽  
Latin Hypercube ◽  
Analysis Process ◽  
Design Software

Due to the complexity of the reactor system, many approximations are used in the nuclear design and calculations inevitably. The accuracy of the nuclear design software is closely related to the safety of the reactor design and operation. Besides, improving the accuracy is an effective way to excavating the economy of nuclear power plants. To analyze the uncertainty of power distribution for NESTOR nuclear design software, we suggested an uncertainty analysis method based on the double Latin Hypercube Sampling (LHS) method and the random sampling statistical analysis (RSSA) method, and built an uncertainty analysis process based on LHS. The uncertainty of physical model and the uncertainty of the change of parameters were both taken into consideration with the double samplings, and 3481 core states were generated by the double samplings. Therefore, the uncertainty of power distribution could be directly analyzed through modeling computation, and the uncertainty of radial power distribution was achieved as ±3.856% under the condition of 95% confidence coefficient and 95% probability. Meanwhile, according to deviation transmission idea, we obtained the uncertainty of power distribution from physical models and the change of parameters based on the measured method. The result shows that the accuracy using the double sampling method is nearly the same to which achieved by the deviation transmission idea, and more conservative.

Influence of Boron Carbide on Core Degradation During Severe Accidents in LWRs

2012 ◽  
Author(s):  
Martin Steinbrück
Keyword(s):  
Boron Carbide ◽  
Nuclear Power ◽  
Source Term ◽  
Current Knowledge ◽  
Steel Structures ◽  
Severe Accident ◽  
Control Rod ◽  
Severe Accidents ◽  
Fuel Bundle ◽  
Water Reactors

Boron carbide (B4C) is widely used as neutron absorbing control rod material in light water reactors (LWRs). It was also applied in all units of the Fukushima Dai-ichi nuclear power plant. Although the melting temperature of B4C is 2450 °C, it initiates local, but significant melt formation in the core at temperatures around 1250 °C due to eutectic interactions with the surrounding steel structures. The B4C containing melt relocates and hence transports material and energy to lower parts of the fuel bundle. It is chemically aggressive and may attack other structure materials. Furthermore, the absorber melt is oxidized by steam very rapidly and thus contributes to the hydrogen source term in the early phase of a severe accident. After failure of the control rod cladding B4C reacts with the oxidizing atmosphere. This reaction produces CO, CO2, boron oxide and boric acids, as well as significant amount of hydrogen. It is strongly exothermic, thus causing considerable release of energy. No or only insignificant formation of methane was observed in all experiments with boron carbide. The paper will summarize the current knowledge on boron carbide behavior during severe accidents, and will try, also in the light of the Fukushima accidents, to draw some common conclusions on the behavior of B4C during severe accidents with the main focus on the consequences for core degradation and hydrogen source term.

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