Innovative System Software Research Articles

Uncertainty and Sensitivity Evaluation of the QUENCH-02 Experiment Simulation Using the ASYST Code

Abstract Uncertainty and sensitivity methods are increasingly used in safety analyzes of nuclear power plants to address the unreliability of input data, numerical models, and, in general, the lack of knowledge regarding certain physical phenomena, in determining safety margins and acceptance criteria. The ASYST code, developed as part of an international nuclear technology ASYST Development and Training Program (ADTP) managed by Innovative Systems Software (ISS), is used to perform an uncertainty analysis of the QUENCH-02 experiment conducted at the Karlsruhe Institute of Technology. The code uses a probabilistic methodology based on the propagation of input uncertainties. The QUENCH facility contains electrically heated PWR fuel rod simulators and the aim of the experiment is to examine hydrogen generation and the behavior of the fuel rod cladding during core reflood. For selected input parameters, such as steam/water flow, electrical power, and other relevant boundary conditions, it is necessary to define their probability density functions. Input databases are then prepared for individual calculations based on the selected confidence level and confidence interval. The number of performed calculations is 60, large enough to ensure at least 95% coverage of expected output results and uncertainty limits. The results of the calculations are compared with the experimental measurements. The Pearson correlation coefficient is used to obtain correlation between the input uncertain parameters and the output data. Sensitivity analyses cover the influence of variations in the heater's electrical power and the steam flow rate on the hydrogen production and the maximum cladding temperature.

Advanced data integration in banking, financial, and insurance software in the age of COVID-19.

This study contributes to our understanding of how the emergence of the COVID‐19 pandemic changes the global Banking Financial Services and Insurance (BFSI) landscape. Before the COVID‐19 pandemic, BFSIs corporate strategy was solely aligned to the quest for operational efficiency. However, during the ongoing COVID‐19 pandemic, global BFSIs are forced to adopt digital transformation in their operations due to a rise in transaction volumes. The ongoing COVID‐19 pandemic already triggers holistic innovations concerning the global BFSI's product, process, concept, trend, or idea. Thus, the BFSI cannot survive without efficient and innovative system software for global operations. The study plots the hype cycle to identify relevant technologies to deal with real‐world business problems. The hype cycle indicates that the need for advanced data integration is growing and COVID‐19 pandemic has already triggered it. The study argues that the incorporation of data integration might be challenging initially for BFSIs but eventually it may result in an efficient model to handle these types of pandemic or unexpected circumstances.

Development of flow regime maps for lead lithium eutectic–helium flows

Abstract Institute for Plasma Research (IPR), Gandhinagar (India) is currently involved in the design and development of its Lead-Lithium Ceramic Breeder (LLCB) module for testing in the International Thermo-nuclear Experimental Reactor (ITER). In order to fulfill the ITER safety requirements, some postulated events need to be analyzed. Among them, the internal loss of coolant accident is being studied using RELAP/SCDAPSIM/MOD4.0 code. To this aim, RELAP/SCDAPSIM/MOD4.0 capabilities of modeling lead lithium eutectic (LLE) and helium flows are being extended nowadays by a collaboration of three institutions, namely IPR, Innovative System Software (ISS) and UPC-BarcelonaTech. The current study is part of this effort and is focused on the adjustments of existing RELAP/SCAPSIM/MOD4.0 flow regime maps in order to deal with the LLE-helium pair. Due to the lack of experimental data, the current study is based on a numerical assessment of LLE-helium flows. The volume of fluid method implemented in OpenFOAM toolkit has been chosen to characterize the flow regime at different mixture velocities and helium fractions. The paper includes a description of the followed methodology and its validation. Results of the systematic simulations of both horizontal and vertical upward flows are shown and the proposed flow diagrams presented. It is worthwhile to mention that the numerically obtained flow regime maps must be validated by an experimental campaign. However, the current approach allows the use of RELAP/SCDAPSIM/MOD4.0 code to properly initiate the study of breeding blanket off-normal events postulated for Lead-Lithium Ceramic Breeder, Helium Cooled Lithium-Lead and Dual Coolant Lithium-Lead designs.

Design and modeling of the passive residual heat removal system for VVERs

The results and effects of the Fukushima Daiichi accident show that the passive safety systems, which require no user input in order to operate, are very significant in nuclear industry. Even after stopping the reactor, heat dissipation remains a primary concern. If the reactor is not properly managed after shutdown, residual heat may cause catastrophic failures. The passive residual heat removal system (PRHRS) is designed to increase the inherent safety features of Nuclear Power Plants (NPP). In the present study, thermal hydraulic performance of the PRHRS of VVER type NPP is investigated using RELAP5 MOD3.4 (developed by Innovative Systems Software) system code. The PRHRS is designed to remove decay heat when normal heat removal system is not available and it consists of steam generator secondary side, a heat exchanger cooling by atmospheric air, and corresponding pipes, valves and air gates. In accident conditions, residual heat is transferred to the ambient air by natural circulation in both steam and air cycles of PRHRS. In the part which include analytical calculations of this study, geometric design parameters of PRHRS was investigated for 5.50MW heat capacity. Selected system parameters, which are found by performing analytical modeling, are tested and analyzed with RELAP5 code and compared with the analytical results. Steady-state behavior of the system has been simulated, and cooling capacity of PRHRS has been investigated during station blackout scenario (including loss of coolant accident case). The results of steady-state behavior obtained by using RELAP5 code show that the PRHRS is able to take residual heat away from the primary coolant system.

Insertion of lead lithium eutectic mixture in RELAP/SCDAPSIM Mod 4.0 for Fusion Reactor Systems

RELAP/SCDAPSIM Mod 4.0 code was developed by Innovative System Software (ISS) for the analysis of nuclear power plants (NPPs) cooled by light water and heavy water. Later on the code was expanded to analyze the NPPs cooled by liquid metal, in this sequence: lead bismuth eutectic mixture, liquid sodium and lead lithium eutectic mixture (LLE) are inserted in the code. This paper focuses on the insertion of liquid LLE as a coolant for NPPs in the RELAP/SCDAPSIM Mod 4.0 code. Evaluation of the code was made for a simple pipe problem connected with heat structures having liquid LLE as a coolant in it. The code is predicting well all the thermodynamic and transport properties of LLE.

Preliminary Assessment of the Possible BWR Core/Vessel Damage States for Fukushima Daiichi Station Blackout Scenarios Using RELAP/SCDAPSIM

Immediately after the accident at Fukushima Daiichi, Innovative Systems Software and other members of the international SCDAP Development and Training Program started an assessment of the possible core/vessel damage states of the Fukushima Daiichi Units 1–3. The assessment included a brief review of relevant severe accident experiments and a series of detailed calculations using RELAP/SCDAPSIM. The calculations used a detailed RELAP/SCDAPSIM model of the Laguna Verde BWR vessel and related reactor cooling systems. The Laguna Verde models were provided by the Comision Nacional de Seguridad Nuclear y Salvaguardias, the Mexican nuclear regulatory authority. The initial assessment was originally presented to the International Atomic Energy Agency on March 21 to support their emergency response team and later to our Japanese members to support their Fukushima Daiichi specific analysis and model development.

Structure, Technology, and Development of the AEG-Telefunken TR 440 Computer

The computer AEG-Telefunken TR 440 was first installed in 1970 at the University of Bochum. At that time, it was the most powerful computer that had ever been developed and manufactured in Europe. As a successor of the TR 4, the TR 440 introduced paging, multiple modes of operation, multiprocessing, vastly extended memory, and satellite configurations and included innovative system software.

Role of RELAP/SCDAPSIM in Nuclear Safety

The RELAP/SCDAPSIM code, designed to predict the behaviour of reactor systems during normal and accident conditions, is being developed as part of the international SCDAP Development and Training Program (SDTP). SDTP consists of nearly 60 organizations in 28 countries supporting the development of technology, software, and training materials for the nuclear industry. The program members and licensed software users include universities, research organizations, regulatory organizations, vendors, and utilities located in Europe, Asia, Latin America, and the United States. Innovative Systems Software (ISS) is the administrator for the program. RELAP/SCDAPSIM is used by program members and licensed users to support a variety of activities. The paper provides a brief review of some of the more important activities including the analysis of research reactors and Nuclear Power Plants, design and analysis of experiments, and training.

PARAMETR-M facility capabilities for severe accident code verification

The paper covers the capabilities of PARAMETR-M facility from the point of view of severe accident code verification. A description of PARAMETR-M facility, its main design features including the improvements made to implement tests with rod bundle superheating and melting is presented. The results of computations made by RATEG/SVECHA/HEFEST (RATEG) [RATEG/SVECHA/HEFEST Severe Accident Code, 2002. Description of application. User's manual. Parts 1, 2. SPbAEP] and RELAP/SCDAPSIM MOD3.2 (RELAP) [Allison, C.M., Wagner, R.J., 2001. RELAP/SCDAPSIM/MOD3.2 Supplemental Input Manual, Innovative Systems Software, LLC, April] severe accident codes are discussed and compared with PARAMETR-M test data. On the basis of experimental data of rod bundle heating up to 1600 K and subsequent top flooding two main conclusions were made: 1. A possibility to quench the rod bundle from top at such temperature was corroborated. 2. The capability of RATEG and RELAP codes to simulate the main physical phenomena of the test (zirconium oxidation, temperature behaviour, bundle top flooding) was verified.