scholarly journals Advanced Presentation of BETHSY 6.2TC Test Results Calculated by RELAP5 and TRACE

2012 ◽  
Vol 2012 ◽  
pp. 1-15 ◽  
Author(s):  
Andrej Prošek ◽  
Ovidiu-Adrian Berar
Keyword(s):  
User Interface ◽  
Graphical User Interface ◽  
Computer Code ◽  
Nuclear Regulatory Commission ◽  
Equivalent Diameter ◽  
Test Results ◽  
Software Applications ◽  
Analysis Package ◽  
Regulatory Commission ◽  
Physical Phenomena

Today most software applications come with a graphical user interface, including U.S. Nuclear Regulatory Commission TRAC/RELAP Advanced Computational Engine (TRACE) best-estimate reactor system code. The graphical user interface is called Symbolic Nuclear Analysis Package (SNAP). The purpose of the present study was to assess the TRACE computer code and to assess the SNAP capabilities for input deck preparation and advanced presentation of the results. BETHSY 6.2 TC test was selected, which is 15.24 cm equivalent diameter horizontal cold leg break. For calculations the TRACE V5.0 Patch 1 and RELAP5/MOD3.3 Patch 4 were used. The RELAP5 legacy input deck was converted to TRACE input deck using SNAP. The RELAP5 and TRACE comparison to experimental data showed that TRACE results are as good as or better than the RELAP5 calculated results. The developed animation masks were of great help in comparison of results and investigating the calculated physical phenomena and processes.

scholarly journals RELAP5 Calculations of Bethsy 9.1b Test

2012 ◽  
Vol 2012 ◽  
pp. 1-11 ◽  
Author(s):  
Andrej Prošek
Keyword(s):  
Computer Code ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Reactor System ◽  
Equivalent Diameter ◽  
Standard Problem ◽  
Similar Range ◽  
Regulatory Commission ◽  
Physical Phenomena ◽  
Relap5 Code

Recently, several advanced computational tools for simulating reactor system behavior during real and hypothetical transient scenarios were developed. The TRAC/RELAP Advanced Computational Engine (TRACE) is the latest in a series of advanced, best-estimate reactor system codes developed by the United States Nuclear Regulatory Commission (US NRC). Nevertheless, the RELAP5/MOD3.3 computer code will be maintained in the next years. The purpose of the present study was to assess how the accuracy of Bethsy 9.1b test calculation depends on the US NRC RELAP5 code version used. Bethsy 9.1b test (International Standard Problem no. 27) was 5.08 cm equivalent diameter cold leg break without high-pressure safety injection and with delayed ultimate procedure. Seven different RELAP5 code versions were used and as much as possible the same input model. The obtained results indicate that the results obtained by the oldest and latest RELAP5 versions are in general comparable for Bethsy 9.1b test. This is very important for the validity of the results, obtained in the past with older RELAP5 versions. Due to the fact that observation was restricted to Bethsy 9.1b posttest, with its own physical phenomena, this conclusion could be generalized only for scenarios having similar range of the considered Bethsy transient conditions.

Computer Code Applicability Assessment for the Advanced CANDU Reactor

2004 ◽  
Author(s):  
D. J. Wren ◽  
N. Popov ◽  
V. J. Langman ◽  
V. G. Snell
Keyword(s):  
Computer Code ◽  
Nuclear Regulatory Commission ◽  
Safety Analysis ◽  
The United States ◽  
Quality Assurance Program ◽  
Software Quality Assurance ◽  
Formal Validation ◽  
Current Program ◽  
Computer Codes ◽  
Regulatory Commission

AECL Technologies (AECLT), the 100%-owned US subsidiary of Atomic Energy of Canada Ltd. (AECL), is currently the proponents of a pre-licensing review of the Advanced CANDU® Reactor (ACR™)* with the United States Nuclear Regulatory Commission (USNRC). A key focus topic for this pre-application review is the NRC acceptance of the computer codes used in the safety analysis of the ACR. These codes have been developed and their predictions compared against experimental results over extended periods of time in Canada. These codes have also undergone formal validation in the 1990’s. In support of this formal validation effort AECL has developed, implemented and currently maintains a Software Quality Assurance program (SQA) to ensure that its analytical, scientific and design computer codes meet the required standards for software used in safety analyses. This paper discusses the SQA program used to develop, qualify and maintain the computer codes used in ACR safety analysis, including the current program underway to confirm the applicability of these computer codes for use in ACR safety analyses.

scholarly journals Good Or Poor Design: Comprehending Ux From an Interactive Designer Perspective, the Case of Handheld Devices

2019 ◽  
Vol 8 (12S) ◽  
pp. 734-742
Keyword(s):  
User Interface ◽  
Graphical User Interface ◽  
Computer Code ◽  
Computing System ◽  
Interactive Design ◽  
Handheld Devices ◽  
Hardware System ◽  
Binary Format ◽  
New Challenges ◽  
Ordinary Person

The terms “system interaction”, “system design”, “interface development”, and “GUI design” may sound different, but in reality could well mean the same thing. A computing system is a set of algorithms which are computer-readable only and which control the hardware system and are accessible via a graphical user interface (GUI), which, if poorly designed, confuses and loses users. The user-invisible algorithms are in binary format, which makes it even more difficult for ordinary users to decipher their content, since the ordinary person would rely on decimal, alphabetical, and alphanumerical representation to comprehend meaning, were they to be exposed to the inner circle of computer code. These days, GUIs are more specific than they used to be, thanks to the ingenuity of some developers, but the fight is not over yet, as new challenges are on the rise. Interaction is what stands between humans and a computing system’s algorithm and provides us with information that we need in a format that humans can better understand. The reality is that, as systems are independently developed, there will be good and poor interactive design products developed by good and poor designers. Thus, here we are discussing knowing what goes on in the design room.

Visualization of Real-Time Radar Data by Integration of X-Band Software

2014 ◽  
Vol 6 (4) ◽  
pp. 73-82 ◽  
Author(s):  
G. Javidi ◽  
E. Sheybani ◽  
D. Mason
Keyword(s):  
User Interface ◽  
Real Time ◽  
Graphical User Interface ◽  
Fourier Transforms ◽  
Initial Step ◽  
Computer Code ◽  
Radar Data ◽  
Scientific Development ◽  
Development Tools ◽  
X Band

To have a fully functional FMCW X-band radar for the SMARTLabs ACHIEVE trailer, it is necessary to produce code to retrieve data from an FPGA board linked to the radar, calculate Fourier transforms and display the power spectrum in near-real time using a computer code based on freely available scientific development tools. In order for the communication between the FPGA board and the computer to be reliable and accurate, developing a specific format through the use of C was an initial step. This was followed by the development of a method to visualize data efficiently. In this case, Python, along with its matplotlib, SciPy, and NumPy modules, were used. Both programs were then integrated together within a graphical user interface.

scholarly journals Transit Analysis Package: An IDL Graphical User Interface for Exoplanet Transit Photometry

2012 ◽  
Vol 2012 ◽  
pp. 1-8 ◽  
Author(s):  
J. Zachary Gazak ◽  
John A. Johnson ◽  
John Tonry ◽  
Diana Dragomir ◽  
Jason Eastman ◽  
...  
Keyword(s):  
User Interface ◽  
Graphical User Interface ◽  
Likelihood Function ◽  
Light Curves ◽  
Parameter Uncertainties ◽  
Data Set ◽  
Specific Data ◽  
Software Release ◽  
Analysis Package ◽  
Good Agreement

We present an IDL graphical user-interface-driven software package designed for the analysis of exoplanet transit light curves. The Transit Analysis Package (TAP) software uses Markov Chain Monte Carlo (MCMC) techniques to fit light curves using the analytic model of Mandal and Agol (2002). The package incorporates a wavelet-based likelihood function developed by Carter and Winn (2009), which allows the MCMC to assess parameter uncertainties more robustly than classicχ2methods by parameterizing uncorrelated “white” and correlated “red” noise. The software is able to simultaneously analyze multiple transits observed in different conditions (instrument, filter, weather, etc.). The graphical interface allows for the simple execution and interpretation of Bayesian MCMC analysis tailored to a user’s specific data set and has been thoroughly tested on ground-based andKeplerphotometry. This paper describes the software release and provides applications to new and existing data. Reanalysis of ground-based observations of TrES-1b, WASP-4b, and WASP-10b (Winn et al., 2007, 2009; Johnson et al., 2009; resp.) and space-basedKepler4b–8b (Kipping and Bakos 2010) show good agreement between TAP and those publications. We also present new multi-filter light curves of WASP-10b and we find excellent agreement with previously published values for a smaller radius.

Resolution of the Thermal Binding Issue of Safety Related MOVs at NPP Krško

2013 ◽  
Author(s):  
Matija Balić ◽  
Tomislav Bajs ◽  
Bogoljub Sember
Keyword(s):  
Nuclear Regulatory Commission ◽  
Gear Ratio ◽  
Operational Parameters ◽  
Function Type ◽  
Operational Conditions ◽  
Safety Function ◽  
The Us ◽  
Accident Scenarios ◽  
Regulatory Commission ◽  
Physical Phenomena

Pressure Locking and Thermal Binding (PL/TB) are two different but related physical phenomena, which under certain conditions may prevent the opening of some types of valves. According to the US Nuclear Regulatory Commission (NRC) NPP operators should evaluate all safety.related power operated gate valves in all operational configurations for susceptibility to PL/TB. Corrective actions are required for susceptible valves that shall assure performance of safety function within plant licensing bases. NEK (NEK) first addressed this issue through the analysis of the Motor Operated Valves (MOV) that were included in the NEK MOV program. This approach resulted with 31 valves in Krško NPP being found as susceptible to TB, 21 of which to both TB and PL. Subsequently, a more thorough analysis was performed, which took into account realistic operational parameters and detailed deterministic and probabilistic assessment evaluation of accident scenarios. This produced a list of 8 valves susceptible to PL, and another 8 susceptible to TB. Valves were screened according to their safety significance, safety function, type, operational history, and operational conditions. Where applicable, Probabilistic Safety Analysis (PSA) has been used as a tool to screen the risk from PL/TB to the safe plant operation. To reduce the risk from PL/TB occurrence, either physical modification of the valve (alteration of actuator gear ratio to provide more force), relevant part of the entire fluid system (flooding of sump suction lines), or simulation and testing to confirm actual ability to overcome increased forces (initiating a controlled TB condition, and testing to confirm actuator’s ability to perform), was used. Methods used and results produced for TB are the subject of this paper.

Verifying Calculation of the Parameters of AP1000 Core Design Based on MCNP Model

2018 ◽  
Vol 4 (3) ◽  
Author(s):  
Xiong Wenbin ◽  
Cao Jian ◽  
Huang Chaoyun ◽  
Bie Yewang ◽  
Wang Yanqi ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Reactor Core ◽  
Nuclear Regulatory Commission ◽  
Physical Parameters ◽  
Test Results ◽  
Design Values ◽  
Regulatory Commission ◽  
Ap1000 Reactor

This study investigates the reactor core physical properties of the AP1000®, which applies the MCNP4a program to model the AP1000 reactor core with the parameters and data from the design control document (DCD, Rev. 19) of the AP1000 Nuclear Power Plant, which has been submitted to the nuclear regulatory commission (NRC). The model is applied to calculate and verify the physical parameters of AP1000 core design. The results match well with the design values in the DCD of the AP1000 nuclear power plant. The model will be modified according to the actual reactor core arrangement, such as AP1000 reactors at China's Sanmen and Haiyang sites, and then compared with the commissioning test results in the future.

Analysis of Circumferential Welds in BWRs for Life Beyond 60

2015 ◽  
Author(s):  
Patrick Purtscher ◽  
Simon Sheng ◽  
Terry Dickson
Keyword(s):  
National Laboratory ◽  
Computer Code ◽  
Nuclear Regulatory Commission ◽  
Boiling Water Reactors ◽  
Circumferential Welds ◽  
Oak Ridge ◽  
Original Analysis ◽  
Asme Code ◽  
Water Reactors ◽  
Regulatory Commission

This paper describes the probabilistic fracture mechanics (PFM) analyzes of the conditional probability of failure (CPF) due to brittle fracture of circumferential welds (CW) from a cold overpressurize event in boiling water reactors (BWR) operated for 72 EFPY. This analysis used the Fracture Analysis for Vessels, Oak Ridge (FAVOR) computer code, developed at the Oak Ridge National Laboratory (ORNL), under United States Nuclear Regulatory Commission (NRC) funding. Two typical vessel configurations and the associated material properties for the beltline materials, CW, axial welds (AW), and plates (PL) were used. The analyses consider the potential effects of different fabrication options, shop vs field. Shop-fabrication is mainly by submerged arc weld (SAW) process, while field fabrication used the shielded metal arc weld (SMAW) process. In either case, repairs would have required the SMAW process. The calculations show that field-fabricated vessels would have a slight increase in the CPF compared to shop-fabricated vessels, but the assumed fraction of repair welds was more significant than the fabrication option. The details demonstrate the relative importance of surface-breaking flaws vs. embedded flaws for the assumed transient. The results confirm the conclusions from the original analysis from BWRVIP-05 and BWRVIP-74, the CPF for CW is orders of magnitude less than that of PL and AW regions of the vessel; therefore, the ASME Code-required volumetric examinations of the CW every 10 years as part of the in-service inspection (ISI) program does not change the overall CPF for the vessel. In all the cases analyzed, the total CPF values of the BWRs for 72 EFPY are below the goal for safe operation.

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