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.

scholarly journals On the effect of model parameters on forecast objects

2018 ◽  
Vol 11 (4) ◽  
pp. 1577-1590 ◽  
Author(s):  
Caren Marzban ◽  
Corinne Jones ◽  
Ning Li ◽  
Scott Sandgathe
Keyword(s):  
Forest Fires ◽  
Eddy Currents ◽  
Numerical Models ◽  
Weather Prediction ◽  
Clustering Algorithms ◽  
Latin Hypercube Sampling ◽  
Numerical Weather Prediction Model ◽  
Model Parameters ◽  
Multiple Model ◽  
The Impact

Abstract. Many physics-based numerical models produce a gridded, spatial field of forecasts, e.g., a temperature map. The field for some quantities generally consists of spatially coherent and disconnected objects. Such objects arise in many problems, including precipitation forecasts in atmospheric models, eddy currents in ocean models, and models of forest fires. Certain features of these objects (e.g., location, size, intensity, and shape) are generally of interest. Here, a methodology is developed for assessing the impact of model parameters on the features of forecast objects. The main ingredients of the methodology include the use of (1) Latin hypercube sampling for varying the values of the model parameters, (2) statistical clustering algorithms for identifying objects, (3) multivariate multiple regression for assessing the impact of multiple model parameters on the distribution (across the forecast domain) of object features, and (4) methods for reducing the number of hypothesis tests and controlling the resulting errors. The final output of the methodology is a series of box plots and confidence intervals that visually display the sensitivities. The methodology is demonstrated on precipitation forecasts from a mesoscale numerical weather prediction model.

Proposal for Evaluation With Strain-Based Criterion of Impact Phenomena Between Steel Structure and Tornado Missile

2016 ◽  
Author(s):  
Yuko Sakamoto ◽  
Koji Shirai ◽  
Toshiko Udagawa ◽  
Shunsuke Kondo
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Impact Analysis ◽  
Structural Integrity ◽  
Steel Structures ◽  
Severe Accident ◽  
Numerical Code ◽  
Pressurized Water ◽  
Impact Phenomena ◽  
The Impact

In Japan, nuclear power plants must be protected from tornado missiles that are prescribed by Nuclear Regular Authority (NRA). When evaluating the structural integrity of steel structures in the plant with impact analysis by numerical code, strain-based criteria are appropriate because the tornado missiles have huge impact energy and may cause large deformation of the structures. As one of the strain-based criteria, the Japan Society of Mechanical Engineers (JSME) prescribes limiting triaxial strain for severe accident of Pressurized Water Reactor (PWR) steel containment. To confirm whether or not this criterion is appropriate to the evaluation of the impact phenomena between the steel structures and the tornado missiles, a free drop impact experiment to steel plates (carbon steel and austenitic stainless steel) was carried out with heavy weights imitated on one of the tornado missiles, followed by an impact analysis of the experiment with AUTODYN code and the JSME strain-based criterion. Consequently, it was confirmed that the strain-based criterion of JSME standard was for evaluating the fracture of steel structures caused by tornado missiles.

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.

scholarly journals Uncertainty and Sensitivity Analysis of the In-Vessel Hydrogen Generation for Gen-III PWR and Phebus FPT-1 with MELCOR 2.2

Energies ◽  
2021 ◽  
Vol 14 (16) ◽  
pp. 4884
Author(s):  
Piotr Darnowski ◽  
Piotr Mazgaj ◽  
Mateusz Włostowski
Keyword(s):  
Sensitivity Analysis ◽  
Hydrogen Generation ◽  
Rank Correlation ◽  
Correlation Coefficients ◽  
Latin Hypercube Sampling ◽  
Severe Accident ◽  
Sensitivity Analyses ◽  
Pressurized Water ◽  
Water Reactor ◽  
Best Estimate

In this study, uncertainty and sensitivity analyses were performed with MELCOR 2.2.18 to study the hydrogen generation (figure-of-merit (FoM)) during the in-vessel phase of a severe accident in a light water reactor. The focus of this work was laid on a large generation-III pressurized water reactor (PWR) and a double-ended hot leg (HL) large break loss of coolant accident (LB-LOCA) without a safety injection (SI). The FPT-1 Phebus integral experiment emulating LOCA was studied, where the experiment outcomes were applied for the plant scale modelling. The best estimate calculations were supplemented with an uncertainty analysis (UA) based on 400 input-decks and Latin hypercube sampling (LHS). Additionally, the sensitivity analysis (SA) utilizing the linear regression and linear and rank correlation coefficients was performed. The study was prepared with a new open-source MELCOR sensitivity and uncertainty tool (MelSUA), which was supplemented with this work. The FPT-1 best-estimate model results were within the 10% experimental uncertainty band for the final FoM. It was shown that the hydrogen generation uncertainties in PWR were similar to the FPT-1, with the 95% percentile being covered inside a ~50% band and the 50% percentile inside a ~25% band around the FoM median. Two different power profiles for PWR were compared, indicating its impact on the uncertainty but also on the sensitivity results. Despite a similar setup, different uncertainty parameters impacted FoM, showing the difference between scales but also a significant impact of boundary conditions on the sensitivity analysis.

Analysis of RCS Depressurization Effects on Containment Temperature for Qinshan Phase II Nuclear Power Plant

2014 ◽  
Author(s):  
Liu Lili ◽  
Zhang Ming ◽  
Deng Jian
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Hydrogen Generation ◽  
Hydrogen Combustion ◽  
Severe Accident ◽  
Generation Rate ◽  
Pressurized Water Reactor ◽  
Gas Temperature ◽  
Pressurized Water

A severe accident code was applied for modeling of a typical pressurized water reactor (PWR) nuclear power plant, and the effects of RCS depressurization on the gas temperature of the relief tank cell in the containment during a station blackout (SBO) induced accident was analyzed. The sensitivity calculation indicated that the hydrogen generation rate obviously increased due to RCS depressurization in a critical stage. The results show that RCS depressurization can play an important role in hydrogen generation rate and total accumulation, and the temperature of the containment atmosphere is highly influenced by hydrogen combustion. High temperature induced by hydrogen combustion may degrade the equipment and instruments capabilities. Based on this analysis, a feasible strategy of RCS depressurization for mitigating the accident consequence is provided for developing the capacity of the SBO treatment of Qinshan Phase Nuclear Power Plant (QSP-II NPP).

The Impact of Transient Behavior of Corium in the Lower Head of a Reactor Vessel for In-Vessel Melt Retention Strategies

2016 ◽  
Author(s):  
L. Carénini ◽  
F. Fichot
Keyword(s):  
Nuclear Power ◽  
Metal Layer ◽  
Heat Fluxes ◽  
Severe Accident ◽  
Total Power ◽  
Light Water Reactors ◽  
Light Water ◽  
Accident Management ◽  
Water Reactors ◽  
The Impact

One of the main goals of severe accident management strategies is to mitigate radiological releases to people and environment. To choose the most appropriate strategy, one needs to know the probability of its success taking into account the associated uncertainties. In the field of corium and debris behavior and coolability, research programs are still on going and the possibilities to efficiently cool and retain corium and debris inside the Reactor Pressure Vessel (RPV) then inside the containment are difficult to evaluate. This leads to uncertainties in safety assessments particularly when margins to RPV or containment failure are too weak. In Vessel Melt Retention (IVMR) strategies for Light Water Reactors (PWR, BWR, VVER) intend to stabilize and retain the core melt in the RPV (as it happened during the TMI-2 accident). This would reduce significantly the threats to the last barrier (the containment) and therefore reduce the risk of release of radioactive elements to the environment. This type of Severe Accident Management (SAM) strategy has already been incorporated recently in the SAM guidance (SAMG) of several operating medium size Light Water Reactors (reactor below 500MWe (like VVER440)) and is part of the SAMG strategies for some Gen III+ PWRs of higher power like the AP1000. A European project coordinated by IRSN and gathering 23 organizations (Utilities, Technical Support Organizations, Nuclear Power Plant vendors, Research Institutes…) has been launched in 2015 with as main objective the evaluation of feasibility of IVMR strategies for Light Water Reactors (PWR, VVER, BWR) of total power around 1000MWe (which represent a significant part of the European Nuclear Power Plants fleet). This paper intends to show how it is possible to introduce transient evolutions of the stratified corium pool in the evaluation of the heat flux profile along the vessel wall. Indeed, due to chemical reactions in the U–Zr–O–Fe molten pool, separation between non-miscible metallic and oxide phases may occur, modifying the thermal load applied to the RPV. If stabilized stratified corium configurations are well defined and modeled, transient evolutions of material layers in the corium pool are still difficult to predict. The evaluations presented are based on calculations performed with the severe accident integral code ASTEC (Accident Source Term Evaluation Code) for a typical PWR reactor. The modeling of transient evolution of corium layers leads to configurations with a thin light metal layer on top of the oxidic one, increasing the so called “focusing effect” (intense heat fluxes on the RPV walls adjacent to the top metal layer). A sensitivity study on some uncertain parameters is proposed to evaluate the impact on the kinetics of layers inversion. Depending on the duration of these transient heat fluxes, the mechanical strength of the RPV could be challenged.

Numerical Simulation of the Behavior of Fission Products in the Primary Circuit of the VVER During the LOCA Severe Accident

2009 ◽  
Author(s):  
S. V. Tsaun ◽  
V. V. Bezlepkin ◽  
A. E. Kiselev ◽  
I. A. Potapov ◽  
V. F. Strizhov ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Hydrogen Generation ◽  
Fission Products ◽  
Severe Accident ◽  
Primary Circuit ◽  
Decay Heat ◽  
The Core ◽  
Severe Accidents ◽  
Loss Of Coolant ◽  
Mass Activity

The methods and models for the analysis of the radiological consequences of the design basis and severe accidents in a Nuclear Power Plant (NPP) are presented in this paper when using the system code SOCRAT. The system code SOCRAT/V3 was elaborated for a realistic analysis of radiological consequences of severe accidents in a NPP. The following models of the fission products (FP) behavior are included into the code SOCRAT/V3: (i) the condensation and the evaporation of the FP in the gaseous phase and (ii) the sedimentation, the evaporation, and the coagulation of the aerosol-shape FP. The latter processes are governed by gravity, Brownian and turbulent diffusion, thermophoresis, turbophoresis and so forth. The behavior of the FP during the loss-of-coolant accidents (LOCA) is presented to demonstrate the possibilities of the code SOCRAT/V3. The main stages of the accident (the core dryout, the core reflooding, the core degradation, the hydrogen generation, the FP release, etc.) are described. Corresponding estimations of the mass, activity, and decay heat of the suspended, settled and released into containment the FP (Xe, Te, Cs, CsI, Cs2MoO4, and so forth) are represent as well.

scholarly journals Spatial evaluation of volcanic ash forecasts using satellite observations

2015 ◽  
Vol 15 (17) ◽  
pp. 24727-24749 ◽  
Author(s):  
N. J. Harvey ◽  
H. F. Dacre
Keyword(s):  
Volcanic Ash ◽  
Pearson Correlation ◽  
Skill Score ◽  
Near Miss ◽  
Model Parameters ◽  
Time Period ◽  
Long Time ◽  
The Impact ◽  
Success Index ◽  
Ash Transport

Abstract. The decision to close airspace in the event of a volcanic eruption is based on hazard maps of predicted ash extent. These are produced using output from volcanic ash transport and dispersion (VATD) models. In this paper an objective metric to evaluate the spatial accuracy of VATD simulations relative to satellite retrievals of volcanic ash is presented. The metric is based on the fractions skill score (FSS). This measure of skill provides more information than traditional point-by-point metrics, such as success index and Pearson correlation coefficient, as it takes into the account spatial scale over which skill is being assessed. The FSS determines the scale over which a simulation has skill and can differentiate between a "near miss" and a forecast that is badly misplaced. The idealised scenarios presented show that even simulations with considerable displacement errors have useful skill when evaluated over neighbourhood scales of 200–700 km2. This method could be used to compare forecasts produced by different VATDs or using different model parameters, assess the impact of assimilating satellite retrieved ash data and evaluate VATD forecasts over a long time period.

Parametric Study of MISTRA M5 Test Using Lumped Parameter COCOSYS Code

2012 ◽  
Author(s):  
Mantas Povilaitis ◽  
Egidijus Urbonavičius ◽  
Sigitas Rimkevičius
Keyword(s):  
Parametric Study ◽  
Nuclear Power ◽  
Power Plants ◽  
Safety Issue ◽  
Severe Accident ◽  
Test Facility ◽  
Lumped Parameter ◽  
Simulation Results ◽  
The Impact ◽  
Parameter Approach

Gas distribution in the containments of nuclear power plants in a case of severe accident currently is a high priority safety issue. One of the topics in this issue is a formation and break up of stratified conditions inside the containment. CEA (France) performed M5 test in their MISTRA test facility, to form a stratified atmosphere, which later was destroyed during MASP tests that followed. During M5 there was formed vertically stratified atmosphere using steam jet release. Application of lumped parameter codes for simulation of jet release is complicated due to limitations inherent in the lumped-parameter approach. However, measures exist which can be used to take these limitations into account when using lumped parameter approach. This paper presents simulations and parametric study of M5 test, i.e. only formation of the stratified atmosphere. Presented simulations of the experiment were performed using lumped parameter code COCOSYS. The aim of the work is to investigate the capability of the code to simulate correctly jet release and formation of the stratified atmosphere in the M5 experiment and the impact of several parameters to the simulation results.

Severe Accident Analysis for Reactor Core Applying SiC to Fuel Claddings and Channel Boxes

2018 ◽  
Author(s):  
Hideki Horie ◽  
Yutaka Takeuchi ◽  
Kenya Takiwaki ◽  
Fumie Sebe ◽  
Kazuo Kakiuchi ◽  
...  
Keyword(s):  
High Temperature ◽  
Oxidation Reaction ◽  
Nuclear Power ◽  
Reaction Rate ◽  
Hydrogen Generation ◽  
Reactor Core ◽  
The Other ◽  
Severe Accident ◽  
Fuel Cladding ◽  
The Difference

Development of a fuel cladding or a channel box applying silicon carbide (SiC), which has high accident tolerance, in place of zircaloy (Zry) or Steel Use Stainless (SUS) composing current light water reactors, has being proceeded with after the accident of Fukushima Daiichi Nuclear Power Plant (1F). When applying SiC to core structures of a nuclear power plant such as fuel cladding, it is expected that the difference of high temperature oxidation characteristics in the severe accident (SA) conditions would mitigate progression of core damage comparing with the current Zry fuel core. This study performed SA analyses considering high temperature chemical reaction characteristics of SiC by using SA analysis code “MAAP”, and thermal hydraulics analysis code “TRAC Toshiba version (TRAC)”, and compared the difference between SiC and Zry. Both codes originally have no model of oxidation reaction for SiC. Hence, a new model for SiC in addition to the current model for Zry was incorporated into “MAAP”. On the other hand, “TRAC” adjusted reaction rate by changing oxidation reaction coefficients in the current Zry oxidation reaction models such as Baker-Just and Cathcart correlations in order to simulate SiC-water/steam reaction. In analysis using “MAAP”, seven accident sequences from representative Probabilistic Risk Assessment ones were selected to evaluate the difference of SA behavior between two materials. As a result, in the case of replacing current Zry of fuel claddings and channel boxes into SiC, an amount of hydrogen generation reduced to about 1/6 than the case of Zry. In addition to that, in the case of replacing SUS structures in the reactor core into SiC, an amount of hydrogen generation moreover reduced to about 1/6 than the above result, which means just about 2% of an amount in the original case. On the other hand, in analysis using “TRAC”, the accident sequence for unit 3 of 1F (1F3) was selected, and reaction rate in the oxidation reaction model was examined as parameter. In the case of 1.0 time of the reaction rate, which means an original reaction rate, maximum fuel cladding temperature exceeded 2000K in 50 hour after reactor scram. However, using the reaction rate below 0.01 to the original one, the fuel cladding temperature didn’t exceed 1,600K.

Sign in / Sign up

Export Citation Format

Share Document