Innovative Systems Software

Idaho Falls, ID, United States

Innovative Systems Software

Idaho Falls, ID, United States
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Toth B.,European Commission | Bieliauskas A.,European Commission | Bandini G.,Ente per le Nuove Tecnologie | Birchley J.,Paul Scherrer Institute | And 4 more authors.
Nuclear Technology | Year: 2010

This paper presents the results of posttest calculations of thephebus FPT2 experiment. While the exercise concentrates mainly on code-to-code benchmarking, a comparison is also made with selected experimental results. The test scenario with the appropriate initial and boundary conditions was provided by the Institut de Radioprotection et de SÛreté Nucléaire. For the analyses, seven severe accident analysis codes were used: ASTEC, ATHLET-CD, MELCOR, ICARE2, ICARE/CATHARE, SCDAP/RELAP5, and RELAP/SCDAPSIM. The calculations focused on the following phenomena occurring in the FPT2 bundle: thermal behavior; hydrogen production, mainly due to cladding oxidation; severe degradation of irradiated fuel; and the release of fission products, control rod, and structure materials. Using the same postdefined boundary and initial conditions, the code-data differences are typically within 10% for most parameters, and not more than 25%. More importantly, the codes were able to capture the major features of the transient evolution. Given that Phebus FPT2 exhibited almost all of the major low-pressure severe accident phenomena except for core cooling by water injection and late-phase core melt behavior in the lower head, the results engender a degree of confidence in the code predictive capability for sequences similar toFPT2.


Allison C.M.,Innovative Systems Software | Hohorst J.K.,Innovative Systems Software | Allison B.S.,Innovative Systems Software | Konjarek D.,ENCONET | And 4 more authors.
Science and Technology of Nuclear Installations | Year: 2012

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. Copyright © 2012 C. M. Allison et al.


Trivedi A.K.,Indian Institute of Technology Kanpur | Khanna A.,Indian Institute of Technology Kanpur | Munshi P.,Indian Institute of Technology Kanpur | Allison C.,Innovative Systems Software
Canadian Nuclear Society - 33rd Annual Conference of the Canadian Nuclear Society and 36th CNS/CNA Student Conference 2012: Building on Our Past... Building for the Future | Year: 2012

Station Black out (SBO) in Boiling Water Reactor of Laguna Verde Nuclear Power Plant (LVNPP) is analyzed. Each flow channel is modeled as a pipe divided into 14 nodes. Physical and thermodynamic properties of both recirculation loops are identical. There are four steam lines and they are considered separately in the model. SBO lead to loss of cooling in the core, severe damage of the fuel and hydrogen production. The maximum core surface temperature has gone up to 3000 K with total hydrogen accumulation about 430 kg. The maximum debris temperature in the lower plenum is 4233 K.


Nistor-Vlad R.-M.,Polytechnic University of Bucharest | Allison C.M.,Innovative Systems Software | Hohorst J.K.,Innovative Systems Software
International Congress on Advances in Nuclear Power Plants, ICAPP 2016 | Year: 2016

A new experimental version, RELAP/SCDAPSIM/MOD3.6, developed by Innovative Systems Software to support the analysis of Pressurized Heavy Water Reactors (PHWRs) under severe accident conditions allows the user to develop a very detailed model of the CANDU core. The modeling improvements treat the unique features of fuel channel reactor designs such as the radial and axial power distribution in the core, radial heat transfer between the fuel channels, and the behavior of the fuel channels during the heat-up and boildown of the liquid in the calandria vessel. Up until this version of RELAP/SCDAPSIM severe accident computer codes were not able to model a CANDU Reactor in detail. The detailed core model is based on grouping the 380 fuel channels in a CANDU core, taking into account the four core pass channels in the core, to allow for the flow of the coolant in both directions. The new representation of the core also allows a very detailed modeling of the high power and low power feeders connecting the fuel channels and a more accurate configuration of the feeders as they are connected to the headers. This paper will present 1) a brief description of the model improvements in RELAP/SCDAPSIM that allow the user to more accurately model a CANDU reactor 2) a detailed description the input model, 3) preliminary analytical results using the new model and finally 4) some suggestions for additional improvements and testing.


Trivedi A.K.,Indian Institute of Technology Kanpur | Allison C.,Innovative Systems Software | Khanna A.,Indian Institute of Technology Kanpur | Munshi P.,Indian Institute of Technology Kanpur
Nuclear Engineering and Design | Year: 2016

The AP1000 is a Westinghouse 2-loop pressurized water reactor (PWR) with all emergency core cooling systems based on natural circulation. Its core design is very similar to a 3-loop PWR with 157 fuel assemblies. Westinghouse has reported their results of the safety analysis in its design control document (DCD) for a large break loss of coolant accident (LOCA) using WCOBRA/TRAC and for a small break LOCA using NOTRUMP. The current study involves the development of a representative RELAP5/SCDASIM model for AP1000 based on publically available data and its verification for a double ended cold leg (DECL) break in one of the cold legs in the loop containing core makeup tanks (CMT). The calculated RELAP5/SCDAPSIM results have been compared to publically available WCOBRA-TRAC and TRACE results of DECL break in AP1000. The objective of this study is to benchmark thermal hydraulic model for later severe accident analyses using the 2D SCDAP fuel rod component in place of the RELAP5 heat structures which currently represent the fuel rods. Results from this comparison provides sufficient confidence in the model which will be used for further studies such as a station blackout. The primary circuit pumps, pressurizer and steam generators (including the necessary secondary side) are modeled using RELAP5 components following all the necessary recommendations for nodalization. The core has been divided into 6 radial rings and 10 axial nodes. For the RELAP5 thermal hydraulic calculation, the six groups of fuel assemblies have been modeled as pipe components with equivalent flow areas. The fuel including the gap and cladding is modeled as a 1d heat structure. The final input deck achieved all steady state thermal hydraulic conditions as reported in the DCD. The analysis has been performed for the primary safety criteria, the peak clad temperature (PCT) as it is well established for a PWR that oxidation and hydrogen generation do not violate the safety criteria as long as PCT is under the safe limit. Results from this study show that the calculated value for the PCT is 1229 K well below the acceptance criteria of 1477 K, lower than the DCD value of 1311 K and higher than the TRACE value of 1186 K. © 2016 Elsevier B.V. All rights reserved.


Allison C.,Innovative Systems Software | Hohorst J.,Innovative Systems Software | Venturi F.,University of Pisa | Forgione N.,University of Pisa | Lopez R.,Comision Nacional de Seguridad Nuclear y Salvaguardias
International Congress on Advances in Nuclear Power Plants, ICAPP 2014 | Year: 2014

Immediately after the accident at Fukushima Daiichi, Innovative Systems Software, ISS, and other members of the international SCDAP Development and Training Program started an assessment of the possible core/vessel damage states of representative LWRs during Fukushima-Daiichi-like Station Blackout transients. The initial 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. These initial calculations have been documented in the open literature. Since that initial assessment, additional calculations have been performed using the original Laguna Verde model as well as models with expanded nodalization to (a) better represent the turbine driven cooling system and (b) add a detailed 2D/3D thermal hydraulic nodalization of the containment. The expanded models have been used to evaluate (a) the heatup and potential failure of the RCS piping, and (b) the influence of containment heat up on the rate of core uncovery and heatup. This paper will describe the expanded models and the response of the vessel, RCS piping (including the turbine driven cooling systems), and containment for a range of Fukushima-Daiichi-like transients using the new models and then discuss the influence of the extended nodalization and calculations relative to the initial assessment performed in 2011.


Trivedi A.K.,Indian Institute of Technology Kanpur | Allison C.,Innovative Systems Software | Khanna A.,Indian Institute of Technology Kanpur | Munshi P.,Indian Institute of Technology Kanpur
Canadian Nuclear Society - 34th Annual Conference of the Canadian Nuclear Society and 37th CNS/CNA Student Conference 2013 | Year: 2013

Immediately after the accident at Fukushima Daiichi, Innovative Systems Software (ISS), and other members of the international SCDAP Development and Training Program (SDTP) 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, 2011 to support their emergency response team and later to the SDTP Japanese members to support their Fukushima Daiichi specific analysis and model development. Since the initial calculations were performed and documented in the open literature, a series of related calculations have been performed by ISS and SDTP members. This paper documents the first in a series of assessment calculations performed by the lead author at the Indian Institute of Technology, Kanpur. Specifically, these calculations have looked at the influence of water addition using thermal hydraulic conditions representative of the reactor core isolation cooling (RCIC) system where the water injection is impacted by time and reactor vessel pressure. Although these calculations were extended to the point of likely vessel failure or stable core cooling, this paper focuses on water addition during the initial heating and melting of the core where water addition may be the most effective in limiting the extent of fuel melting. The paper also presents the results of a base case, a station blackout transient without water addition, for comparison purposes. The base case calculations were carried out to 10 hours after reactor scram to a point beyond the point of likely vessel failure.


Trivedi A.K.,Indian Institute of Technology Kanpur | Allison C.,Innovative Systems Software | Khanna A.,Indian Institute of Technology Kanpur | Munshi P.,Indian Institute of Technology Kanpur
Nuclear Engineering and Design | Year: 2014

This work analyses the influence of water addition in a boiling water reactor during a core isolation event in BWR. Injection of water is impacted by time as well as reactor vessel water level and it uses thermal hydraulic conditions representative of the reactor core isolation cooling (RCIC) system. A detailed RELAP/SCDAPSIM model of Laguna Verde BWR vessel and related reactor cooling system (provided by the Mexican Nuclear Regulatory Authority) has been used for this analysis. These calculations have been extended to the point of likely vessel failure or stable core cooling. They focus on initial heating and melting of the core where water addition is found to be most effective in limiting the extent of fuel melting. It also presents the results of a base case, a station blackout transient without water addition. These calculations have been carried out up to 5 h (after reactor scram) beyond the point of likely vessel failure. The maximum core surface temperature of 3042 K and hydrogen production of 367 kg is observed in this case. The importance of timing can be seen from 3500 s and 3700 s injection cases. One case leads to maximum core surface temperature of 1520 K with hydrogen production of 21 kg while the second case leads to temperature of 2940 K with hydrogen production of 193 kg. Temperatures (at the time of start of first injection) in both these cases are1371 K and 1590 K which explains this switching from stable core cooling to very high core surface temperature. © 2014 Elsevier B.V.


Akbas S.,University of Idaho | Akbas S.,Bozok University | Martinez-Quiroga V.,Innovative Systems Software | Aydogan F.,University of Idaho | And 2 more authors.
ASME International Mechanical Engineering Congress and Exposition, Proceedings (IMECE) | Year: 2015

These codes were developed by a variety of institutions (universities, research centers, and laboratories) geographically located away from each other. Each of the research group that developed these coupled code systems used its own combination of initial codes as well as different methods and assumptions in the coupling process. For instance, all these neutron kinetics codes solve the few-groups neutron diffusion equations. However, the data they use may be based on different lattice physics codes. The neutronics solvers may use different methods, ranging from point kinetics method (in some versions of RELAP5) to nodal expansion methods (NEM), to semi-analytic nodal methods, to the analytic nodal method (ANM). Similarly, the thermalhydraulics codes use several different approaches: different number of coolant fields, homogenous equilibrium model, separate flow model, different numbers of conservation equations, etc. Therefore, not only the physical models but also the assumptions of the coupled codes and coupling techniques vary significantly. This paper compares coupled codes qualitatively and quantitatively. The results of this study are being used both to guide the selection of appropriate coupled codes and to identify further developments into coupled codesThe design and the analysis of nuclear power plants (NPPs) require computational codes to predict the behavior of the NPP nuclear components and other systems (i.e., reactor core, primary coolant system, emergency core cooling system, etc.). Coupled calculations are essential to the conduct of deterministic safety assessments. Inasmuch as the physical phenomena that govern the performance of a nuclear reactor are always present simultaneously, ideally computational modeling of a nuclear reactor should include coupled codes that represent all of the active physical phenomena. Such multi-physics codes are under development at several institutions and are expected to become operational in the future. However, in the interim, integrated codes that incorporate modeling capabilities for two to three physical phenomena will remain useful. For example, in the conduct of safety analyses, of paramount importance are codes that couple neutronics and thermal-hydraulics, especially transient codes. Other code systems of importance to safety analyses are those that couple primary system thermal-hydraulics to fission product chemistry, neutronics to fuel performance, containment behavior and structural mechanics to thermal-hydraulics, etc. This paper surveys the methods used traditionally in the coupling of neutronic and thermal-hydraulics codes. The neutron kinetics codes are used for computing the space-time evolution of the neutron flux and, hence, of the power distribution. The thermal-hydraulics codes, which compute mass, momentum and energy transfers, model the coolant flow and the temperature distribution. These codes can be used to compute the neutronic behavior and the thermalhydraulic states separately. However, the need to account with fidelity for the dynamic feedback between the two sets of properties (via temperature and density effects on the cross section inputs into the neutronics codes) and the requirement to model realistically the transient response of nuclear power plants and to assess the associated emergency systems and procedures imply the necessity of modeling the neutronic and thermal-hydraulics simultaneously within a coupled code system. The focus of this paper is a comparison of the methods by which the coupling between neutron kinetics and thermal-hydraulics treatments has been traditionally achieved in various code systems. As discussed in the last section, the modern approaches to multi-physics code development are beyond the scope of this paper. From the field of the most commonly used coupled neutron kinetic-thermal-hydraulics codes, this study selected for comparison the coupled codes RELAP5-3D (NESTLE), TRACE/PARCS, RELAP5/PARCS, ATHLET/DYN3D, RELAP5/SCDAPSIM/MOD4.0/NESTLE. The choice was inspired by how widespread the use of the codes is, but was limited by time availability. Thus, the selection of codes is not to be construed as exhaustive, nor is there any implication of priority about the methods used by the various codes. Copyright © 2015 by ASME.


Martinez-Quiroga V.,Innovative Systems Software | Akbas S.,University of Idaho | Akbas S.,Bozok University | Aydogan F.,University of Idaho | And 2 more authors.
ASME International Mechanical Engineering Congress and Exposition, Proceedings (IMECE) | Year: 2015

High-fidelity and accurate nuclear system codes play a key role in the design and analysis of complex nuclear power plants, which consist of multiple subsystems, such as the reactor core (and its fuel, burnable poisons, control elements, etc.), the reactor internal structures, the vessel, and the energy conversion subsystem and beyond to grid demand. Most commonly the interplay between these various subsystems is modeled using coupled codes, each of which represents one of the subsystems. And the most common direct coupling is that of thermalhydraulics and neutronics codes. The subject of this paper is the coupling of codes that model not only thermal-hydraulics and neutronics, but also structural components damage. Furthermore, the neutronic component is not limited to the sole core solver. The coupled code system encompasses thermal-hydraulics, material performance of the fuel, neutronic solver, and neutronic data preparation. Thus, this paper presents a framework for coupling RELAP5/SCDAPSIM/MOD4.0 with a suite of neutron kinetics codes that includes NESTLE, DRAGON and a version of the ENDF library. The version of the RELAP5/SCDAPSIM/MOD4.0 code used in this work is one developed by Innovate System Software (ISS) as part of the international SCDAP Development and Training Program (SDTP) for best-estimate analysis to model reactor transients including severe accident phenomena. This RELAP5/SCDAPSIM/MOD4.0 code version is also capable of predicting nuclear fuel performance. It uses nodal power distributions to calculate mechanical and thermal parameters such as heat-up, oxidation and meltdown of fuel rods and control rods, the ballooning and rupture of fuel rod cladding, the release of fission products from fuel rods, and the disintegration of fuel rods into porous debris and molten material. On the neutronics side, this work uses the NESTLE and DRAGON codes. NESTLE is a multi-dimensional static and kinetic neutronic code developed at North Carolina State University. It solves up to four energy groups neutron diffusion equations utilizing the Nodal Expansion Method (NEM) in Cartesian or hexagonal geometry. The DRAGON code, developed at Ecole Polytechnique de Montreal, performs lattice physics calculations based on the neutron transport equation and is capable of using very fine energy group structures. In this work, we have developed a coupling approach to exchange data among the various modules. In the coupling process, the generated nuclear data (in fine multigroup energy structure) are collapsed down into two- or four-group energy structures for use in NESTLE. The neutron kinetics and thermal-hydraulics modules are coupled at each time step by using the cross-section data. The power distribution results of the neutronic calculations are transmitted to the thermalhydraulics code. The spatial distribution of coolant density and the fuel-moderator temperature, which result from the thermalhydraulic calculations, are transmitted back to the neutron kinetics codes and then the loop is closed using new neutronics results. Details of the actual data transfers will be described in the full length paper. Copyright © 2015 by ASME.

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