Uncertainty Analysis for a Depressurised Loss of Forced Cooling Event of the PBMR Reactor

2006 ◽  
Author(s):  
Pieter A. Jansen van Rensburg ◽  
Martin G. Sage
Keyword(s):  
Monte Carlo ◽  
Uncertainty Analysis ◽  
Power Distribution ◽  
Sensitivity Study ◽  
Fuel Temperature ◽  
Distribution Profile ◽  
Pebble Bed ◽  
Decay Heat ◽  
Input Parameters ◽  
Forced Cooling

This paper presents an uncertainty analysis for a Depressurised Loss of Forced Cooling (DLOFC) event that was performed with the systems CFD (Computational Fluid Dynamics) code Flownex for the PBMR reactor. An uncertainty analysis was performed to determine the variation in maximum fuel, core barrel and reactor pressure vessel (RPV) temperature due to variations in model input parameters. Some of the input parameters that were varied are: thermo-physical properties of helium and the various solid materials, decay heat, neutron and gamma heating, pebble bed pressure loss, pebble bed Nusselt number and pebble bed bypass flows. The Flownex model of the PBMR reactor is a 2-dimensional axi-symmetrical model. It is simplified in terms of geometry and some other input values. However, it is believed that the model adequately indicates the effect of changes in certain input parameters on the fuel temperature and other components during a DLOFC event. Firstly, a sensitivity study was performed where input variables were varied individually according to predefined uncertainty ranges and the results were sorted according to the effect on maximum fuel temperature. In the sensitivity study, only seven variables had a significant effect on the maximum fuel temperature (greater that 5°C). The most significant are power distribution profile, decay heat, reflector properties and effective pebble bed conductivity. Secondly, Monte Carlo analyses were performed in which twenty variables were varied simultaneously within predefined uncertainty ranges. For a one-tailed 95% confidence level, the conservatism that should be added to the best estimate calculation of the maximum fuel temperature for a DLOFC was determined as 53°C. This value will probably increase after some model refinements in the future. Flownex was found to be a valuable tool for uncertainly analyses, facilitating both sensitivity studies and Monte Carlo analyses.

scholarly journals Prediction calculations for the first criticality of the HTR-PM using the PANGU code

2021 ◽  
Vol 32 (9) ◽  
Author(s):  
Ding She ◽  
Bing Xia ◽  
Jiong Guo ◽  
Chun-Lin Wei ◽  
Jian Zhang ◽  
...  
Keyword(s):  
Monte Carlo ◽  
High Temperature ◽  
Power Plant ◽  
Uncertainty Analysis ◽  
Nuclear Data ◽  
High Fidelity ◽  
Pebble Bed ◽  
Input Model ◽  
High Temperature Reactor ◽  
High Temperature Gas

AbstractThe high-temperature reactor pebble-bed module (HTR-PM) is a modular high-temperature gas-cooled reactor demonstration power plant. Its first criticality experiment is scheduled for the latter half of 2021. Before performing the first criticality experiment, a prediction calculation was performed using PANGU code. This paper presents the calculation details for predicting the HTR-PM first criticality using PANGU, including the input model and parameters, numerical results, and uncertainty analysis. The accuracy of the PANGU code was demonstrated by comparing it with the high-fidelity Monte Carlo solution, using the same input configurations. It should be noted that keff can be significantly affected by uncertainties in nuclear data and certain input parameters, making the criticality calculation challenge. Finally, the PANGU is used to predict the critical loading height of the HTR-PM first criticality under design conditions, which will be evaluated in the upcoming experiment later this year.

Study on the Core Cooling Scheme After Accident Shutdown of the Pebble-Bed Modular High Temperature Gas-Cooled Reactor

2012 ◽  
Author(s):  
Yanhua Zheng ◽  
Lei Shi ◽  
Fubing Chen
Keyword(s):  
High Temperature ◽  
Temperature Distribution ◽  
Normal Operation ◽  
Helium Temperature ◽  
Fuel Temperature ◽  
Pebble Bed ◽  
Decay Heat ◽  
The Core ◽  
Core Cooling ◽  
High Temperature Gas

One of the most important properties of the modular high temperature gas-cooled reactor is that the decay heat in the core can be carried out solely by means of passive physical mechanism after shutdown due to accidents. The maximum fuel temperature is guaranteed not to exceed the design limitation, so as to the integrity of the fuel particles and the ability of retaining fission product will keep well. Nonetheless, the auxiliary active core cooling should be design to help removing the decay heat and keeping the reactor in an appropriate condition effectively and quickly in case of reactor scram due to any transient and the main helium blower or steam generator unusable. Based on the preliminary design of the 250 MW pebble-bed modular high temperature gas-cooled reactor, assuming that the core cooling will be started up 1 hour after the scram, different core cooling schemes are studied in this paper. After the reactor shutdown, a certain degree of natural convection will come into being in the core due to the non-uniform temperature distribution, which will accordingly change the core temperature distribution and in turn influence the outlet hot helium temperature. Different cooling flow rates are also analyzed, and the important parameters, such as the fuel temperature, outlet hot helium temperature and the pressure vessel temperature, are studied in detail. A feasible core cooling scheme, as well as the reasonable design parameters could be determined based on the analysis. It is suggested that, considering the temperature limitation of the structure material, the coolant flow direction should be same as that of the normal operation, and the flow rate could not be too large.

Characteristics of the 250MW Pebble-Bed Modular High Temperature Gas-Cooled Reactor in Depressurized Loss of Coolant Accidents

2008 ◽  
Author(s):  
Yanhua Zhengy ◽  
Lei Shi
Keyword(s):  
High Temperature ◽  
Power Distribution ◽  
Safety Margin ◽  
Reactor Core ◽  
Reactor Power ◽  
Fuel Temperature ◽  
Pebble Bed ◽  
Loss Of Coolant ◽  
High Temperature Gas ◽  
Residual Heat

Depressurized loss of coolant accident (DLOCA) is one of the most important design basis accidents for high temperature gas-cooled reactors. Analysis of the reactor characteristic behavior during DLOCA can provide useful reference to the physics, thermo-hydraulic and structure designs of the reactor core. In this paper, according to the preliminary design of the 250MW Pebble-bed Modular High Temperature Gas-cooled Reactor (HTR-PM), three cases of DLOCA: a instantaneous depressurization along with a flow coastdown and scram at zero time, a main pipe with a diameter of 65mm rupture, and a instrument pipe with a diameter of 10mm broken, are studied by the help of two different kinds of software THERMIX and TINTE. The key parameters of different cases including reactor power, temperature distribution of the core and pressure vessel, and the decay power removal by the passive residual heat remove system (RHRS) are compared in detail. Some uncertainties, such as residual heat calculation, power distribution, heat conductivity of fuel element, etc., are analyzed in order to evaluate the safety margin of the maximum fuel temperature during DLOCA. The calculating results show that, the decay heat in the DLOCA can be removed from the reactor core solely by means of physical processes in a passive way, so that the temperature limits of fuel and components are still obeyed. It also illustrates that the HTR-PM can reach 250MW reactor power per unit and still can keep the inherent safety.

A Neutronic Evaluation of the HTR-10 Using Scale, MCNPX and MCNP5 Nuclear Codes

2014 ◽  
Author(s):  
Rômulo V. Sousa ◽  
Clarysson A. M. Silva ◽  
Ângela Fortini ◽  
Cláubia Pereira ◽  
Maria Auxiliadora F. Veloso ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Heat Removal ◽  
Fission Products ◽  
Negative Temperature ◽  
Electricity Production ◽  
Three Dimension ◽  
Fuel Temperature ◽  
Pebble Bed ◽  
Coated Particles ◽  
The Core

The HTR-10 (High Temperature Gas-cooled Test Reactor) is a 10 MW modular pebble bed type reactor, built by the Institute of Nuclear Energy Technology (INET), Tsinghua University, China. As an advanced reactor, it has good passive safety characteristics: capacity of retaining all fission products inside the coated particles (up to 1,600° C), passive decay heat removal, large heat capacity of the core to mitigate temperature transition, large fuel temperature margin and negative temperature reactivity coefficient sufficient to accommodate reactivity insertion and small amount of excess reactivity in the core. This reactor, which core is filled with 27,000 spherical fuel elements, e.g. TRISO coated particles, is used to test and develop fuel, verify PBR safety features, demonstrate combined electricity production and cogeneration of heat, and provide experience in PBR design, operation and construction. Using the SCALE 6.0 (Standardized Computer Analysis for Licensing Evaluation), the MCNPX 2.6.0 (Monte Carlo N-Particle eXtended) and the MCNP 5 (Monte Carlo N-Particle) nuclear codes, the HTR-10 first critical core described in the Evaluation of The Initial Critical Configuration of The HTR-10 Pebble-Bed Reactor was modeled and analyzed. A three-dimension model was simulated and the keff was obtained and compared with the reference. The result presents good agreement with experimental value. The goal is to validate the DEN/UFMG model to be applied in transmutation studies changing the fuel.

scholarly journals Sensitivity and Uncertainty Analysis of the Maximum Fuel Temperature under Accident Condition of HTR-PM

2020 ◽  
Vol 2020 ◽  
pp. 1-21
Author(s):  
Chen Hao ◽  
Peijun Li ◽  
Ding She ◽  
Xiaoyu Zhou ◽  
Rongrui Yang
Keyword(s):  
Standard Deviation ◽  
Uncertainty Analysis ◽  
Effective Conductivity ◽  
Thermal Analyses ◽  
Mean Value ◽  
Reactor Power ◽  
Fuel Temperature ◽  
Decay Heat ◽  
Uncertainty Sources ◽  
Accident Condition

The maximum fuel temperature under accident condition is the most important parameter of inherently safe characteristics of HTR-PM, and the DLOFC accident may lead to a peak accident fuel temperature. And there are a variety of uncertainty sources in the maximum fuel temperature calculations, and thus the contributions of these uncertainty sources to the final calculated maximum fuel temperature should be quantified to check whether the peak value exceed the technological limit of 1620°C or not. Eight uncertainty input parameters are selected for inclusion in this uncertainty study, and their associated 2 standard deviation uncertainties and probability density functions are specified. Then, the DLOFC thermal analyses and uncertainty analysis are performed with the home-developed ATHENA and CUSA. The numerical results indicate that the pebble-bed effective conductivity and the decay heat contribute the most of the uncertainty in the DLOFC maximum fuel temperature while this peak fuel temperature is most sensitive to the initial reactor power and the decay heat. In short, uncertainties in these selected eight parameters lead to the two standard deviation (2σ) uncertainty of ±77.6°C (or 5.2%) around the mean value of 1493°C for the maximum fuel temperature under DLOFC accident of HTR-PM. At the same time, the LHS-SVDC method of CUSA is recommended to propagate uncertainties in inputs and 100–200 model simulations seem to be sufficient to get an uncertainty prediction with full confidence.

scholarly journals Uncertainty and Sensitivity Analyses of a Pebble Bed HTGR Loss of Cooling Event

2013 ◽  
Vol 2013 ◽  
pp. 1-16 ◽  
Author(s):  
Gerhard Strydom
Keyword(s):  
National Laboratory ◽  
Input Parameter ◽  
Latin Hypercube Sampling ◽  
Temperature Tolerance ◽  
Sensitivity Analyses ◽  
Data Sets ◽  
Fuel Temperature ◽  
Pebble Bed ◽  
Energy Agency ◽  
Input Parameters

The Very High Temperature Reactor Methods Development group at the Idaho National Laboratory identified the need for a defensible and systematic uncertainty and sensitivity approach in 2009. This paper summarizes the results of an uncertainty and sensitivity quantification investigation performed with the SUSA code, utilizing the International Atomic Energy Agency CRP 5 Pebble Bed Modular Reactor benchmark and the INL code suite PEBBED-THERMIX. Eight model input parameters were selected for inclusion in this study, and after the input parameters variations and probability density functions were specified, a total of 800 steady state and depressurized loss of forced cooling (DLOFC) transient PEBBED-THERMIX calculations were performed. The six data sets were statistically analyzed to determine the 5% and 95% DLOFC peak fuel temperature tolerance intervals with 95% confidence levels. It was found that the uncertainties in the decay heat and graphite thermal conductivities were the most significant contributors to the propagated DLOFC peak fuel temperature uncertainty. No significant differences were observed between the results of Simple Random Sampling (SRS) or Latin Hypercube Sampling (LHS) data sets, and use of uniform or normal input parameter distributions also did not lead to any significant differences between these data sets.

Improvement of sensitivity and uncertainty analysis capabilities of generalized response in Monte Carlo code RMC

2021 ◽  
Vol 154 ◽  
pp. 108099
Author(s):  
Guanlin Shi ◽  
Yuchuan Guo ◽  
Conglong Jia ◽  
Zhiyuan Feng ◽  
Kan Wang ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Uncertainty Analysis ◽  
Monte Carlo Code ◽  
Sensitivity And Uncertainty Analysis

scholarly journals Mathematical Formulation and Analytic Solutions for Uncertainty Analysis in Probabilistic Safety Assessment of Nuclear Power Plants

Energies ◽  
2021 ◽  
Vol 14 (4) ◽  
pp. 929
Author(s):  
Gyun Seob Song ◽  
Man Cheol Kim
Keyword(s):  
Monte Carlo ◽  
Uncertainty Analysis ◽  
Probability Density ◽  
Nuclear Power ◽  
Safety Assessment ◽  
Power Plants ◽  
Mathematical Formulation ◽  
Analytic Solutions ◽  
Probabilistic Safety Assessment ◽  
Probabilistic Safety

Monte Carlo simulations are widely used for uncertainty analysis in the probabilistic safety assessment of nuclear power plants. Despite many advantages, such as its general applicability, a Monte Carlo simulation has inherent limitations as a simulation-based approach. This study provides a mathematical formulation and analytic solutions for the uncertainty analysis in a probabilistic safety assessment (PSA). Starting from the definitions of variables, mathematical equations are derived for synthesizing probability density functions for logical AND, logical OR, and logical OR with rare event approximation of two independent events. The equations can be applied consecutively when there exist more than two events. For fail-to-run failures, the probability density function for the unavailability has the same probability distribution as the probability density function (PDF) for the failure rate under specified conditions. The effectiveness of the analytic solutions is demonstrated by applying them to an example system. The resultant probability density functions are in good agreement with the Monte Carlo simulation results, which are in fact approximations for those from the analytic solutions, with errors less than 12.6%. Important theoretical aspects are examined with the analytic solutions such as the validity of the use of a right-unbounded distribution to describe the uncertainty in the unavailability/probability. The analytic solutions for uncertainty analysis can serve as a basis for all other methods, providing deeper insights into uncertainty analyses in probabilistic safety assessment.

Sign in / Sign up

Export Citation Format

Share Document