scholarly journals Demonstration of Emulator-Based Bayesian Calibration of Safety Analysis Codes: Theory and Formulation

2015 ◽  
Vol 2015 ◽  
pp. 1-17 ◽  
Author(s):  
Joseph P. Yurko ◽  
Jacopo Buongiorno ◽  
Robert Youngblood

System codes for simulation of safety performance of nuclear plants may contain parameters whose values are not known very accurately. New information from tests or operating experience is incorporated into safety codes by a process known as calibration, which reduces uncertainty in the output of the code and thereby improves its support for decision-making. The work reported here implements several improvements on classic calibration techniques afforded by modern analysis techniques. The key innovation has come from development of code surrogate model (or code emulator) construction and prediction algorithms. Use of a fast emulator makes the calibration processes used here with Markov Chain Monte Carlo (MCMC) sampling feasible. This work uses Gaussian Process (GP) based emulators, which have been used previously to emulate computer codes in the nuclear field. The present work describes the formulation of an emulator that incorporates GPs into a factor analysis-type or pattern recognition-type model. This “function factorization” Gaussian Process (FFGP) model allows overcoming limitations present in standard GP emulators, thereby improving both accuracy and speed of the emulator-based calibration process. Calibration of a friction-factor example using a Method of Manufactured Solution is performed to illustrate key properties of the FFGP based process.

Author(s):  
J. Birchley

Calculations of PHEBUS FPT-1 are performed in the frame of CSNI International Standard Problem ISP-46. The objective of ISP-46 is to assess the capability of computer codes to provide an integral simulation of a severe accident in a Pressurised Water Reactor (PWR), from the initial stages of core heat-up to the behaviour of released fission products in the containment. The present calculations are performed using MELCOR, chosen as the main tool for assessment of Swiss nuclear plants by virtue of its whole-plant simulation capability, using modelling practices as similar as possible to those used in plant analyses. The calculations cover the bundle heat-up, degradation, the release, transport and retention of fission products and other materials, and the thermal-hydraulic and aerosol behaviour in the containment. Comparison between a best-estimate case and experiment demonstrates the code’s ability to capture most aspects of the sequence with fair to good accuracy. Uncertainties remain, particularly in regard to core degradation, and the chemistry and transport of fission products. Weaknesses of code models in these areas largely reflect limitations in current knowledge.


Author(s):  
Alan D. Chockie ◽  
M. Robin Graybeal ◽  
Scott D. Kulat

The risk-informed inservice inspection (RI-ISI) process provides a structured and systematic framework for allocating inspection resources in a cost-effective manner while improving plant safety. It helps focus inspections where failure mechanisms are likely to be and where enhanced inspections are warranted. To date, over eighty-five percent of US nuclear plants and a number of non-US plants have implemented, or are in the process of implementing, RI-ISI programs. Many are already involved in the periodic update of their RI-ISI program. The development of RI-ISI methodologies in the US has been a long and involved process. The risk-informed procedures and rules were developed to take full advantage of PRA data, industry and plant experiences, information on specific damage mechanisms, and other available information. An important feature of the risk-informed methodologies is the requirement to make modifications and improvements to the plant’s RI-ISI application as new information and insights become available. The nuclear industry, ASME Section XI, and the Nuclear Regulatory Commission have all worked together to take advantage of the lessons learned over the years to refine and expand the use of risk-informed methodologies. This paper examines the lessons learned and the benefits received from the application and refinement of risk-informed inservice inspection programs. Also included in the paper is a review of how the information and insights have been used to improve the risk-informed methodologies.


Author(s):  
Richard Tilley ◽  
Robin Dyle

United States (US) and International utilities are actively engaged in assessing the economic and societal benefits of operating nuclear plants beyond their initial license periods. Nuclear plant generated electricity is still the largest contributor to non-carbon dioxide emitting generation. In the US, a majority of operating plants has already received approval for an additional 20 years of operation, and soon it is expected that utilities will begin the process to seek a second 20 year renewal. The keys to successful renewal are to maintain safe and reliable operations by building a sound technical case through the following activities: • Develop comprehensive understanding of aging degradation issues for systems, structures and components (SSCs) • Implement specific plant aging management programs to address aging degradation • Confirm behavior of degradation mechanisms for the entire period of operation This paper will step through the above elements to illustrate how a strong technical case may be created for safe and reliable long-term operation. Examples or case studies will be provided to clearly link the fundamental science of materials degradation to the inspection, testing and evaluation efforts implemented at a plant and to the confirmatory data that is provided by both actual operating experience and the extensive research and development projects pursued by industry, governments, and the academic community.


Author(s):  
F. A. Simonen ◽  
S. R. Gosselin ◽  
B. O. Y. Lydell ◽  
D. L. Rudland ◽  
G. M. Wilkowski

This paper describes an application of data on cracking, leak and rupture events from nuclear power plant operating experience to estimate failure frequencies for piping components that had been previously evaluated using the PROLOCA and PRAISE probabilistic fracture mechanics (PFM) computer codes. The calculations had addressed the failure mechanisms of stress corrosion cracking, intergranular stress corrosion cracking and fatigue for materials and operating conditions that were known to have failed components. The first objective was to benchmark the calculations against field experience. A second objective was a review of uncertainties in the treatments of the data from observed failures and in the structural mechanics models. The database PIPExp-2006 was applied to estimate failure frequencies. Because the number of reported failure events was small, there were also statistical uncertainties in the estimates of frequencies. Comparisons of predicted and observed failure frequencies showed that PFM codes correctly predicted relatively high failure probabilities for components that had experienced field failures. However, the predicted frequencies tended to be significantly greater than those estimated from plant operating experience. A review of the PFM models and inputs to the models showed that uncertainties in the calculations were sufficiently large to explain the differences between the predicted and observed failure frequencies.


Author(s):  
Roberto Passalacqua

In case of a High Pressure Melt Ejection (HPME) heated gas and corium may be expelled from the bottom head of a reactor vessel reaching the containment atmosphere, leading to a Direct Containment Heating (DCH). In addition, released gases might burn (e.g. hydrogen) causing a high load of the reactor containment building. Corium dispersal phenomena also strongly affect consequences of Molten Core-Concrete Interaction (MCCI) since the corium mass, which remains within the cavity, may remarkably diminish. Several computer codes are able to simulate the response of nuclear plants during hypothetical severe accidents: MELCOR, MAAP, ESCADRE and ASTEC have the capability to describe corium slump into the reactor cavity, vessel gases blow-down and possible corium entrainment. In this paper the various steps of model development, validation, plant-specific applications, etc., are described in the attempt of establishing a risk-oriented methodology with the target of solving this particular risk-issue. ENEA mature expertise in level-2 PSA analyses shows that the DCH phenomenology can be considered a solved risk issue. The applied methodology gives also hints and/or guidelines for solving similar risk issues in current PSA (level 2) analysis.


2006 ◽  
Vol 91 (10-11) ◽  
pp. 1301-1309 ◽  
Author(s):  
Marc C. Kennedy ◽  
Clive W. Anderson ◽  
Stefano Conti ◽  
Anthony O’Hagan

Biometrika ◽  
2009 ◽  
Vol 96 (3) ◽  
pp. 663-676 ◽  
Author(s):  
S. Conti ◽  
J. P. Gosling ◽  
J. E. Oakley ◽  
A. O'Hagan

Author(s):  
C. Keller ◽  
D. Schmidt

Development reports on closed-cycle gas turbines (CCGT) as proposed by Ackeret and Keller (AK system) and promoted mainly by Escher Wyss Ltd., Zurich, Switzerland, and Gutenhoffnungshütte (GHH), Germany, have been presented since 1945 at ASME meetings about every five years (1). This, the sixth paper, reports on the operating experience with some newer fossil-fuel fired plants made by different manufacturers and gives the study results of European designers for nuclear gas turbines which can be built already with today’s technology for the 600 to 1000-Mw range. The special physical properties of air and helium and their influence on plant design are discussed. The combination of a CCGT and a high-temperature reactor offers many possibilities for simplifications of nuclear plants and lowering capital costs.


1983 ◽  
Vol 6 (3) ◽  
pp. 133-151 ◽  
Author(s):  
J.R. Aupied ◽  
A. Le Coguiec ◽  
H. Procaccia

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